ML20149J985
| ML20149J985 | |
| Person / Time | |
|---|---|
| Site: | Texas A&M University |
| Issue date: | 07/22/1997 |
| From: | TEXAS A&M UNIV., COLLEGE STATION, TX |
| To: | |
| Shared Package | |
| ML19317C540 | List: |
| References | |
| NUDOCS 9707290179 | |
| Download: ML20149J985 (121) | |
Text
{{#Wiki_filter:. .._ _ _ _ .. - _ . _ .. _ _ _ _ _ _ . . _ _ _ ... _ _ _ _ ._. . _ !O ! APPENDIX G . i i i i SAFETY ANALYSIS REPORT i { LICENSE RENEWAL APPLICATION l OF j FACILITY LICENSE R-23 iO DOCKET # 50-59 l AGN201M (serial #106) i i i l i TEXAS A&M UNIVERSITY i i i i i 1 I i
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l 9707290179 970722 i PDR ADOCK 05000059 l P PDR . i
SAFETY ANALYSIS REPORT f)i
't TABLE OF CONTENTS Pace No.
1 TIIE FACILITY l-1 1.1 Introduction 1-1 12 Summary and Consideration on Principal Safety Considerations 1-2 1.3 General Description of the Facility 1-3 1.4 Shared Facilities and Equipment 1-7 1.5 Comparison to Similar Facilities 1-8 1.6. Summary Operations 1-8 1.7 Compliance With the Nuclear Waste Policy Act of1982 1-9 1.8 Facility Modifications and History 1-9 f^) 2 SITE CIIARACTERISTICS 2-1 2.1 Geography and Demography 2-1 1 2.1.1 Site Location and Description 2-1 j 2.1.1.1 Specifications and Locations 2-1 2.1.1.2 Boundary and Zone Area Maps 2-1 2.1.2 Population Distribution 2-1 2.2 Nearby Industrial, Transportation and Military 2-7 2.2.1 Location and Routes 2-7 . 2.2.2 Air Traffic 2-7 , 2.2.3 Analysis of Potential Accidents at Facilities 2-7 l 2.3 Meterology 2-8 1 2.4 Hydrology 2-8 2.5 Geology, Seismology, and Geotechnical Engineering 2-9 l 2.5.1 Regional Geology 29 2.5.2 Site Geology 29 . l l
2.5.3 Seismicity 2-10 2.5.4 Maximum Earthquake Potential 2-10 2.5.5 Vibratory Ground Motion 2-10 2.5.6 Surface Faulting - 2-10 , 2.5.7 Liquefaction 2-10 3 DESIGN OF STRUCTURES, SYSTEMS AND COMPONENTS 3-1 3.1 Design Criteria 3-1 3.2 Meteorological Damage 3-2 3.3 Water Damage 3-3 3.4 Seismic Damage 3-3 3.5 System and Components 3-3 4 REACTOR DESCRIPTION 4-1 4.1 Summary Description 4-1
-.4.2 Reactor Core 4-2 4.2.1 Reactor Fuel. 4-2 4.2.2 Control Rods 4-5 4.2.3 Neutron Moderator and Reflector 4-16 4.2.4 Neutron Startup Source 4-16 g
4.3 Reactor Tank or Pool 4-16 4.4 Biological Shield 4-18 4.5 Nuclear Design 4-18
'4.5.1 Normal Operating Conditions 4-19 4.5.2 Reactor Core Physics Parameters 4-19 4.5.3 Operating Limits 4-20 4.6 Thermal-Hydraulic Design 4-23
r 5 REACTOR COOLANT SYSTEMS 5-1 l O O i 6 ENGINEERED SAFETY FEATURES 6-1 J l 7 INSTRUMENTATION AND CONTROL SYSTEM 7-1 i 7.1 Summary Description 7-2 7.2 Design ofInstrumentatios and Control Systems 7-2 7.2.1 Design Criteria 7-2 , 1 7.2.2 Design-Basis Requirements 7-2 l 7.2.3 System Description 7-5 7.3&4 Reactor Control and Protective Systems 7-10 7.6 Control Console and Display Instruments 7-12 7.7 Radiation Monitoring Systems 7-18 8 ELECTRICAL POWER SYSTEMS 8-1
.i O 8.1 Normal Electrical Power Systems 8-1 Q
8.2 Emergency Electrical Power Systems 8.2 , 1 l 9 AUXILIARY SYSTEMS 9-1 9.1 Heating , Ventilation, and Air Conditioning Systems 9-1 9.2 Handling and Storage of Reactor Fuel 9-1 9.3 Fire Protection Systems and Programs 9-1 9.4 Communciation System 9.2 l 1 9.5 Possession and Use of Byproduct, Source and Special Nuclear Material 9-3 1 10 EXPERIMENTAL FACILITIES AND UTILIZATION 10-1 10.1 Summary Description 10-1 f-10.2 Experimental Facilities 10-2 b
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' 10.3 Experiment Review 10.2 11 RADIATION PROTECTION PROGRAM AND WASTE MANAGEMENT I1-1 11.1 Radiation Protection 11-1 11.2 Radiation Sources 11-2 11.1.1 Airborne Radiation Sources Il-3 11.1.2 Liquid Radioactive Sources 11-3 11.1.1.3 Solid Radioactive Source , 11-3 11.1.2 Radiation Protection Program Il-4 I 11.1.3 ALARA Program Il-5 11.1.4 Radiation Monitoring and Surveying 11-5 1
11.1.5 Radiation Exposure Control and Dosimetry 11-6 11.1.6 Contamination Control 11-7 11.1.7 Environmental Montoring 11-8 11.2 RadioactiveWaste Management 11-8 12 CONDUCT OF OPERATIONS 12.1 Organization 12.1 l 12.1.1 Structure & Responsibility 12-1 12.1.3 Staffing 12-4 l 12.1.4 Selection and Staffm' g of Personnel 12-4 l 12.1.5 Radiation Safety 12-5 12.2 Reviews and Audit Activities 12-5 12.2.1 Composition and Qualifications 12-7 12.2.2 Charter and Rules 12-7 l 12.2.3 Review Function 12-7 12.2.4 Audit Function 12-8
- 12.3 Procedures 12-8 12.4 Required Actions 12-8 l
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i 12.5 Reports 12-11 ()
.n 1 12.6 Records 12-13 l
. 12.7 Emergency Planning 12-15 12.8 Security Planning 12-16 12.10 Operator Training and Requalification 12-16 12.12 Environmental Report 12-16 i I ; i 13 ACCIDENT ANALYSIS 13-1 ' 13.1 Accident-Initiating Events and Scenarios 13-1 l 13.1.1 Maximum-Hypothetical Accident 13-1 l . 13.1.2 Insertion of Excess Reactivity 13-2 13.1.3 Loss of Coolant 13-2 13.1.4 Loss of Coolant Flow 13-2 1 13.1.5 Mishandling or Malfunctioning of Fuel 13-3 (^ ( 13.1.6 Experimental Malfunction 13-3 1 l 13.1.7 Loss of Normal Electrical Power . 13-3 l !3.1.8 External Events 13-3
- 13.1.9 Mishandling or Malfunction of Equipment 13-4 i l 13.2 Accident Analysis and Determination ofConsequences 13-5
- 13.3 Summary and Conclusions 13-5 14 TECHNICAL SPECIFICATIONS 14-1 15 FINANCIAL QUALIFICATIONS 15-1 15.2 Financial Ability to Operate a Non-Power Reactor 15-1 15.3 Financial Ability to Decommission the Facility 15-1 16 OTIIER LICENSE CONSIDERATIONS 16-1
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4 17 DECOMMISSIONING AND POSSESSION ONLY LICENSE AMENDENTS 17-1 1 1 l O O
_ ld f3 4 V) Chapter 1 THE FACILITY l Introduction 1.1 This report is intended to be an a part of the application to the United States Nuclear Regulatory Commission (USNRC) by Texas A&M University (TAMU) at I l College Station, Texas for the renewal of the class 104, Facility License R-23, Docket # l 50-59 for the University's AGN-20lM research reactor. The reactor is primarily used for the teaching of classes in the Department of Nuclear Engineering and for the training of reactor operators. The University has operated this reactor for a period of almost forty years and in its present location, the Zachry Engineering Center room 61B for nearly 25
,.s years. l / \ l \ l 1 V This application contains information on the local communities of Bryan and College Station are as well as the reactor facility located inside the Zachry Engineering Center. The local area information includes the meteorology, geology, seismology as well as other pertinent data required by NUREG-1537 Part 1.
This reactor has been rated for a maximum power level of 5 watts thermal since 1972. The AGN-201M reactor design has several inherent safety features, the most significant of these are the negative temperature coefficient. and the design limitation of a maximum of 0.65% excess reactivity above delayed critical. These designed safety features along with the core thermal fuse and the trips and interlocks associated with the nuclear instrumentation safety systems provide for protection of the reactor fuel even under the worst case accident conditions. If a sudden insertion of positive reactivity were LJ
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s-3 FIGURE 1 FIRST FLOOR LEVEL ( f - SITE BOUNDARY EMERGENCY SHOWER
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; . T' to occur the reactor power rise would be attenuated by the negative temperature /
(h) coefficient and then terminated by the separation of the reactor core due to the melting of the core thermal fuse. Other important reactor design considerations include, the type of fuel used, with its low enrichment, the core can, as well as the sealed fueled control rods. These features provide a complete barrier between the reactor fuel and the environment. While the core tank allows for thermal dispensation the heat generated due to the fissioning of the fuel, because of this no outside secondary cooling system is necessary. A unique feature of the TAMU reactor room is the thickness of the walls and ceiling, these components which act as a boundary to the reactor facility, are constructed of concrete. The walls are 3.5 feet thick and the ceiling is 4.0 feet thick. The AGN reactor design has proven to be a safe and effective training device for thousands of future engineers at this as well as other universities. The system has an established data and ( I L corroborated data base, so no research and development activities were required to evaluate the system. 1.2 Summarv and Conclusions on Principal Safety Considerations The the safety criterion associated with the AGN-201M reactor at Texas A&M University is to limit as much a2 possible any adverse effect that reactor operations many have on the general public and students within the reactor facility. It is believed that no adverse consequences have resulted from many years of reactor operations in this current facility. Radiation surveys conducted at 5 watts indicate very low doses outside the reactor room and none outside the areas controlled by the Department of Nuclear Engineering. iA)
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(-L l A complete description of the AGN-20lM reactor is presented in Charter 4 of I j V this report. I 1 All the design bases for the AGN-20lM reactor system and components are presented in the Technical Specifications for this AGN-20lM reactor and are include C o this report The reactor facility in the Zachry Engineering Center is constructed in such a way that the reactor room and accelerator laboratory are in a cube. This cube has 3.5 foot ! l I thick reinforced concrete walls and 4 foot thick ceilings. This area of the building is ! isolated from the general public of as well as most personnel outside of the nuclear engineering students that have laboratories or offices in the area. Because of this design and the design of the reactor and its components, the reactor will remain in a safe, I isolated condition even under the most severe hypothetical accident. I t U This reactor facility has been in its present location since the early 1970's and a l renewal license was issued by the USNRC in August 1977, for this reactor in its present , 1 location. Some small changes have been made to the reactor facility in this time period. 1.3 General Description of the Facility The AGN-201M reactor facility is located on the main campus of Texas A&M University in the Zachry Engineering Center. The location of the center on campus is presented in Figure 1.3-1. The facility is located in areas under the direct control of the Department of Nuclear Engineering on the 1st and ground floors of the center with the reactor room below the ground floor in a specially constructed reactor room. The floor l plans for the facility are presented in Figures 1.3-2 and 1.3-3. l
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lO The location of the facility in the center allows for complete isolation of the (D C/ reactor room , room 61B from the rest of the engineering center. The AGN-20lM reactor is designed to operated in populated areas without undo risk to the general public. This facility produces no effluents to the environment and is located in a room with thick i reinforced concrete walls and below ground. The AGN-20lM reactor system has been in service for many years and the University has had an operating license since 1957.
- Specific site characteristics of the facility are addressed in Chapter 2 of this report.
The site is in an area with a very low threat of earthquakes and flooding and the facility is l i designed to be fail safe. The reactor system has many safety features associated with it, some are passive systems while others are designed into the reactor protective systems. i The reactor systems and safety features are described in greater detail in other chapters of this report. d A unique feature of this reactor facility is the design of the reactor room itself. The control console is shielded by an L-shaped walls to lower the dose to the operators and students and the walls of the room are 3.5 foot thick reinforced concrete and the ceiling is 4 foot thick reinforced concrete. This greatly enhances the ability of the operating staff to operate reactor with confidence that no member of the general public is I being negatively affected. I 1.4 Shared Facilities and Equipment The only facilities that this reactor facility shares with the surrounding building is the electrical power supply and air supply. The reactor room has a ventilation unit inside g and the discharge from the reactor room passes through a grate in the ceiling and into an V
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\4 exit filter unit located inside the accelerator room. A separate fan on the roof provides U,a suction on this filter unit and air is drawn out of the reactor room and discharged on the roof. Power to the AGN-20lM reactor console is supplied from the buildings power supply, but power can be remotely isolated by operation of two large breakers outside of l room 60C. Inside the reactor room is a large graphite pile and a subcritical assemebly M Comparisons to Similar Facilities Two other AGN-201M research reactors are currently licensed by the USNRC, these are both university reactors and have asimilar training and teaching role as does the AGN-20lM at Texas A&M University. But the roles are somewhat different, since TAMU has a much larger TRIGA reactor facility on the campus, this reactor is used for irradiation of materials for medical and commercial use while the AGN-20lM is used as (3 d a teaching and training device only.
1.6 Summarv of Operations The AGN-20lM operations have been limited in the past few years with the downturn in enrollment and more widespread use of the TRIGA reactor. The AGN-201M continues to be used for student reactor startups and for some limited experiments in two different nuclear engineering classes. The reactor has run for maintenance and operator proficiency more than anything else in the last four years, but it continues to be an important learning device for the students. With the presence of two reactosr at one university the students get the unique opportunity to work with and operate both. The reactor should have critical run hours of 25 -35 hours per year at an average power level of 0.75 - 1.0 watts. As part of the ALARA program, the reactor operated at power levels V
I (d which are as low as possible to achieve the desired results in order to keep doses to the D I L
- student low..
In the future the AGN-20lM will be operated a schedule similar to its current l usage. The reactor is used in the first four weeks of both the fall and spring semesters for both graduate and undergraduate reactor laboratory classes. i-I l 1.7 Compliance With the Nuclear Waste Policy Act of 1982 i Texas A&M University is in compliance with the Waste Policy Act. The University has entered into an agreement with Idaho National Engineering Laboratory for I i 1 the final, no-cost disposal of the AGN-20lM reactor fuel. l I 1.8 Facility Modifications and Historv i O This AGN-201 reactor has been licensed by Texas A&M University since 1957. 4 The initial maximum rated power was 100 mW but this was changed in 1973 to the j reactors present 5 watt thermal power limit. The current reactor room was constructed in l 1970 under construction permit CPRR-112 and the reactor moved to this room in 1972. , l In September of 1984, the reactor control console was moved from room 60C to its present location in the reactor room, 61B, behind the installed shielding. This was l done so that the operator could be in the same room as the reactor during experiments. At i i I the same time another modification was made to add a series of concrete blocks around i a the base of the reactor to lower general area radiation levels. Current radiation surveys indicate that the reactor produces very low levels of radiation outside the reactor room , 1 and in the accelerator room. The AGN-201M is used at Texas A&M University to train
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24 Chapter 2 SITE CHARACTERISTICS Q O 2.1 Geography and Demography 1 2.1.1.1 Specification and Location i The AGN-20lM reactor facility is located at the main campus of Texas A&M University, in College Station, Texas in the county of Brazos. The facility is located on [ the first and ground floors of the Zachry Engineering Center and the reactor room is on the ground floor in room 61 B. i A map, Figures 2.1.1.1-1 A&B, is provided with the both Bryan and College
- Station clearly shown and the pin marker on the location of the Zachry Engineering Center. The next Figure 2.1.1.1-2 is a map of the main campus of the University with the Zachry Engineering Building clearly marked.
2.1.1.2 Boundary and Zone Area Maps V , The reactor room and accelerator room areas are easily seen on Figures 2.1.1.1-3 and 2.1.1.1-4, the dark outer line indicates the boundary of the facility. The campus of Texas A&M University is bordered by Texas Highway 6 and 30 2.1.2 Population Distribution it can be seen from the population density on map 2.1.1.1-I A&B that the population density remains constant until you leave the towns of Bryan or College Station. and it falls off to near zero. The 1990 census recorded the population of the Bryan and College Station area at 107,500, many of which are associated with the University. Since the reactor is located on the main campus, up to 55,000 students and employees could be on campus during reactor operations. But since the reactor has O G
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L l-l no effluents and has been proven safe under the most serve conditions this does not pose a safety hazard. 2.2 Nearby Industrial. Transportation and Military Facilities Any potential accident associated with the AGN-20lM reactor will have no effect ' upon the environment nor the surrounding area. No military installations are close the , Bryan / College Station area, industrial development is very limited within the 8 kilometer limit since most of that area is still part of Texas A&M University or a residential area. > 2.2.1 Locations and Routes The only major route of concern is the railroad that cuts across campus,1.8 kilometers southwest from the facility. Texas Highway 30 is located 0. kilometers due north of the facility. But in both cases the facility is located on the far side of the most l l- f, direct point from the routes.
- 2.2.2 ' Air Traffic The reactor facility is located 6.5 kilometers from a small commercial airport.
The flight path for this
- port is some 1.0 kilometers to the northwest of the facility.
Since the facility is located on the first floor and the reactor room is in a steel reinforced t concrete room the danger posed by passing planes is 5 very small. 2.2.3 Analysis of Potential Accidents at Facilities. No heavy industry nor chemical plants are located within 20 kilometers of the reactor facility. The worst possible accident deal with the trains that use the railroad tracks through the middle of campus. If an accident occurred that would require an evacuation of the facility and the surrounding building I still believe that the reactor facility would be in no real danger. If a large explosion occurred the shockwave would
Y tear apart many other structure before getting to the Zachry Engineering Center and since O V the reactor room and accelerator room are located inside a steel reinforced cube the danger to the facility is very low. Even in the event of a highway accident the reactor is well protected from potential harm. 2.3 Meteorology The Bryan-College Station area is located approximately 100 miles inland from the Texas Gulf Coast. The local weather is determined to a great extent by the high pressure areas which are predominant over the Gulf of Mexico. As a direct result of this condition, warm southeasterly winds occur a large majority of the time on an annual basis. It should be noted that in the winter months the winds come out of the north. The average wind speed is 7.34 knots. Hurricanes are very infrequent in this area do to the g distance from the Texas Gulf Coast, in the last 75 years only two hurricanes have caused kj any damage in this area and that was due to the thunderstorms and gusty winds associated with the remnants of the system. Tomadoes are fairly common in Texas, not is in this particular part of the state but not unheard here. Data collected on tornado frequency between 1950 and 1976 indicates that 17 tornadoes were reported within a 25 nautical mile radius of College Station. 2.4 Hydrology The nearest large independent free flowing water the Brazos river located about 10 kilometers from the facility. Several smv! lakes and streams are located within 3 kilometers but pose n, haz nd to the facility. The facility is located 314 feet above sealevel and not in any flood plains. The highest recorded crest of the Brazos river was O recorded at 54 feet above flood stage or 246 feet above sea level. The nearest acquifer is ) O
a ~ 2 i j . the Bryan Sandstone and it is located under other geological foundations at a depth beyond concern for nis study. t' ' 2.5 Geoloev Seismoloev, and Geotechnical Engineerine
- 2.5.1 Regional Geology 1
The facility is located in a geological region known as the Gulf Coastal Plain. This region of Texas has only one fault zone associated with it and the fault lies of the l 3 1 western boundary of the coastal directly plain. ] 2.5.2 Site Geology 1 The local geology is dominant by a formation known as the Easterwood Shale. l The thickness of this formation varies from 10-300 feet. This formation is located
- l
, throughout the local area. s ! j 2.5.3 Seismicity The State of Texas lies in a geological area that has minor seismic activity. Most of the activity that exist is limited to extreme West Texas. This region lies over 600 1 1
- miles west from the local area but is the nearest active belt along the west coast of l
l ', Mexico and the United States. No recorded earthquakes have ever occurred in the local ; ! area, the nearest fault zone is the known as the Balcones Escarpment and is located on the western boundary of the Gulf Coast Plain,100 miles to the west of this facility. i
- i. 2.5.4 Maximum Earthquake Potential i i
The largest earthquake ever recorded in the State of Texas occurred near El Paso 2 and measured 6.4 on the Richter scale. No seismic activity has ever been recorded in the 3 local area and the possibility of any activity occurring is very low. 4 4
,. g , - - -
24 b) i 2.5.5 Vihrotory Ground Motion The area surrounding the facility has no geological activity associated with it. The area has been stable for as long as people have recorded such things and no threat of earthquakes or other seismic actives is concerned a possibility in this local region of Texas. ! 2.5.6 Surface Faulting The local area is considered to have a very low potential for surface faulting. No earthquakes have ever been reported in the local area. 2.5.7 Liquefaction f The local foundation material have no potential for liquefaction. The facility is !A constructed in such a manner that this is not considered to be a potential problem. l l I i i 1
.7
- I j
^ <N (J' Chapter 3 DESIGN OF STRUCTURES, SYSTEMS AND COMPONENTS.
l 3.1 Design Criteria 1 l The AGN-201M reactor control and instrumentation system has two basic design criteria, the first is the ability to safely control the reactor under all normal and foreseeable accident possibilities and the second is the " fail-safe" design mode. The fail-safe design mode is a concept that the reactor will be placed in a safe shutdown condition if any equipment or power failure were to occur. The nuclear l instrumentation system is constructed with a series of trips and interlock systems that will l i accomplish this design criterion. A complete description of these trips and interlocks is presented in Chapter 7 of this SAR. The most critical passive design feature in this " fail-1 . safe" mode, this mode is the present in the form oflarge springs on each of the safety (3 C rods and the coarse controi rod. These springs ensure that the rods will be forcible removed from the core if power is interrupted to the scram magnets no matter the cause of this power interruption. The Technical Specification limit is less than 200 msecs , for the safety system rods to be removed from the core, this limit ensures that the reactor will be safety shutdown even under serve transient conditions. The reactor control and instrumentation system is designed to monitor reactor power all ranges for the AGN-201M reactor system . The construction of the reactor room and adjoining room provides for a safe isolated environment in which to conduct reactor operations. The facility is designed to have very limited interaction with the building systems. The reactor control console D b
1 3* 2 j i power and power to the local ventilation unit and exhaust fan come from the in buildmg l power system. No systems are associated with the reactor facility. 3.2 Meteorological Damage The location of Texas A&M University provides some threat to severe weather conditions. Rare tornadoes have occurred in the Brazos county and have been within 10 miles of the facility they have done little damage to the Universtiy. Only the remnants of one hurricane has been detected in the county over the last 50 years and this lead to some damage in the local community. The reactor facility in the Zachry Engineering Center is designed to be a large cube inside the northwest corner of the building. This cube extends from the first floor down to the reactor room and laboratory adjacent to it. The walls of this cube are 3.5 feet O ! of reinforced concrete and the fail ceilings are four feet of reinforced concrete, the outer l wall is a direct component of the building foundation. This system will stand up to any l l storms that may pass into the area. In the event a tornado is sighted within a 5 mile radius of TAMU, the radio operator at the TAMU Communications Center will notify the first available person listed l on the AGN-20lM Emergency Response phone numbers. The communications center j receives notification of tornadoes from both the TAMU weather radar and the Brazos ; County, Bryan-College Station Disaster Emergency Planning Organization. Tornados are detected by the use of TAMU radar and Doppler radar at a local television station. i A U l j
3-3 3.3 Water Damace () The design of the reactor facility and the Zachry Engineering Center makes the likelihood of water damage very remote. The only route that floodwater could enter the area is through two sets of double doors. In the event of weather severe enough to warrant concern ,these doors have been sandbagged to eliminate any possibility of water entering the facility. A large drain is located in the parking lot adjacent to these double doors to drain off the water. If water entered the facility it would have to rise up to the level of the double leading to room 60C. Then flood the floor in this room in order to reach the double doors leading to the laboratory outside the reactor room. If water were to enter the reactor room a drain system is set up so that the water would drain into a 1000 gallon holding tank ,a located in the basement floor of the facility. This sump would have to be completely (w) filled in order to raise the water level in the reactor room. 3.4 Seismie Damace The possibility of seismic activity is extremely low in this region of Texas. This combined with the fact that the walls of this facility are constructed on 3.5 feet of steel reinforced concrete in the form of a cube give the facility the strength to resist any seismic activity in the area. 3.5 Systems and Components The fail-safe AGN-201M control and instrumentation system and the earthquake switch are designed to place the reactor in a shutdown condition in the event of a natural disaster. The loss of electrical power would also scram the reactor, since this is a likely fm byproduct of such an event this also ensure the safe condition of the reactor. An ( ) v
i 34 Emergency Procedure EA-4, Severe Natural Phenomena, has also been written for this facility in the case of such events and ensures the reactor is in a safe condition if the J likelihood of such events is present or has already occurred.. The scram system for the AGN-201 is of a fail-safe design. Loss ofpwer to the ) electromagennets will allow for the reactor to scram and be placed in a safe condition. j The design basis for the scram system is to remove enough postive reactivity from the i 3 core to lesson the serveiuty of any accident. The design basis for the control and safety g 7 systems for the AGN-201M reactor are presented in section 3.2 of the Technical !
- Specifications.
5 l 1 i A v i
- l i
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Chapter 4 REACTOR DESCRIPTION 4.1 Summary Description The AGN-20lM research reactor in operation at Texas A&M University is essentially identical to all the other AGN-201M reactors that have been operated since 1957. Therefore, the reactor description given in the original Aerojet-General Nucleonics AGN-201 reactor report (Hazard Report and Preliminary Design Report by Aerojet-General Nucleonics, Docket F-40) is applicable to this reactor as are the Hazard and i Safety Analysis Reports for the other AGN-201M reactors on file with the USNRC. The characteristics and operating parameters of this reactor have been calculated i and experimentally determined from data from this and other AGN-201M reactors. The rated thermal power level for this AGN-20lM reactor is 5 watts. This power level was approved by the USNRC in Amendment No.10 to License R-23 issued on January 18,1973. The critical mass of the reactor is approximately 665 grams of 235 U. The fuel l itselfis in a plate type configuration and is homogeneously mixed powder of polyethylene 235 3 and UO2 (61 mg U/cm )in the form of 20 micron diameter particles. The core disc have been formed by pressing under high pressure the homogeneously mixed powder. ; The reactor is a free standing unit, the reactor core is surrounded by a graphite reflector, a lead shield and a large water shield tank for fast neutrons. The reactor has a natural convection cooling system through the graphite reflector I and the lead shield to the 1000 gallon tank of water the surrounds the reactor core, this removes any thermal heat due to the fission of the fuel. O
/
Cf4 The reactor has a hermetically core tank surrounding it , this tank is designed to i n s
\J contain fission products gases if any core damage were to occur.
The experimental facilities associated with this AGN-201M are the small 15/16 inch diameter glory hole that runs completely through the reactor and the 4 -inch diameter access ports that pass through the water shield and graphite moderator and lead gamma shield. A side view of the AGN-20lM reactor is presented in Figure 4.1-1. 4.2 Reactor Core 4.2.1 Reactor Fuel The reactor core of the AGN-201M reactor consist of nine separate fuel discs ~ which are stacked into a right circular cylinder as shown in figure 4.2.1-1. This figure clearly depicts the physical relationship between the control rods, the core and which the core tank. The final core configuration is.10.15 inches (25.8 cm)in diameter and 9.34 d,-3 j inches (23.75 cm)in height. The fuelin the core is made up of approximately 665 grams of 235 U in 3330 grams of U for an enrichment of just less than 20%. Figure 4.2.'-l shows the relationship between the core and the rods. The uranium is homogeneously mixed with polyethylene to form a series of fuel plates. A physical description of these fuel plates is included in the reactor characteristics table 4.2-1 The AGN 20lM reactor core also contains four fueled control and safety rods. , Three of these rods, Safety Rods 1&2 and the Coarse Control Rod, are constructed in a 2 similar manner while the Fine Control Rod is much smaller in size. The fuel contained in these rods is exactly of the same type as the rest of the core. The reactor fuel has some limitations associated with it, in particular the breakdown of the fuel at high temperature. O Several design features of the AGN-20lM will prevent this temperature (200 C) from V
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4.1-1 SIDE VIEW OF AGN-20lM REACTOR s
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1 4.2.2 Control Rods 1 The AGN-201M design contains four fueled control and safety rods. Three of 1 l these rods, the two safety rods and the coarse control rod, are identical in design but differ l l in their function. Each of these rods contains about 15 grams of2" U and operate in such a manner that as the rod is driven into the core the reactivity in the core increases. The 1 exact amount of fuel contained in each of the rods is listed on table 4.2-1. Figure 4.2.2-1 shows the design of the rod which fits into a 2 inch (5 cm ) hole in the lower disc in the I reactor core. l
)
The coarse control rod and the two safety rods are designed to contain 1.25% of
- f.s reactivity each, while the smaller line control rod has a reactivity worth of 0.31%. Tne fine control rods is design to allow the reactor operator to make small adjustments in reactor power. The active length of these rods is 5.9 inches (15 cm) and consists of UO2 embedded in radiation-stabilized polyethylene identical to that of the fuel discs in the reactor core. The active fuel material is enclosed in two aluminum containers, the outer containers provide fluid seals between the rods and their drive mechanisms and the shield tank. The innermost aluminum container are used to seal the active fuel material contained in the fuel slugs. These containers provide a double seal between the fuel and the outside environment. These rods are calibrated as part of the AGN maitennace program, the reactivity worths of these control rods must be known to safety operarate the reactor.
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4.2.2-1 CONTROL ROD
.a h ( Table 4.2-1 Reactor Characteristics l A. General Descriptions Type Homogeneous thermal reactor i Principle Uses Education and Training Maximum Rated Power 5 watts (thermal) ! ! Core Design 235 UO2 in the form of < 20 % enriched U ] particles homogeneously distributed in a solid polyethylene moderator, in nine fuel disc, in a cylindrical geometry Reflector Graphite j
- Shielding Lead and water 1
Rods Two safety and two control, all fueled ! i 1 l"f b) Fuel Fuel Material < 20%-enriched UO2 Fuel Dise Right circular cylinder fabricated from a compressed a mixture of fuel material and polyethylene power Specifications: ] i UO2 4 235
- U enrichment < 20%
Particle size - 15 10 m I l Polyethylene Powder ) Particle size- 100 pm ' Purity - commercial grade
.nximate Core Fuel Disc 9 separate fuel disc m,m . aid 23'U contentl0 Four discs 1.57"(4 cm) high ( 97 g 235 U each )
O
._. . . -. . l
Three discs 0.79"(2 cm) high ( 58 g 235 0 each )
) Two discs 0.39"(1 cm) high ( 30 g 235 U each )
Approximate Safety and Coarse Control Rods fuel disc sizes : 4 per rod 1.85 " (4.7 cm) diameter and 1.57" (4 cm) in height Approximate Safety and Coarse Control Rods fuel disc loading : 4 per rod 235 3.6 g U each Approximate Fine Control fuel disc sizes : 4 per rod , 0.9 " (2.3 crn) diameter and 1.57" (4 cm) in height 235 Total Fuel Loading = 665 gram U 235 235 U Density 0.061 g U/cm 3 Core Thermal Fuse Small, right circular cylinder. 0.87"(2.2 cm) in diameter a 235 and 0.35"(0.9 cm)in height enriched with 0.4 g of U. Fabricated from a mixture of fuel materials and polystyrene powder. n k.) c. Reactor Core-containing Tank Gas-tight,0.026"(65 mil) aluminum cylindrical tank, 12.7"(32.2 cm) diamerer and 30* (76 cm) high. The tank has a l' glory hole on the horizontal tank centerline the appropriate control access areas l Core 9 fuel disc, separated at core midpoint by glory hole and thin aluminum baffle. Lower half of the core has appropriate access holes for control and safety rod thimbles and and core thermal fuse. 3 Reflector Heavy density (1.75 g/cm ) graphite, approximately 7.87" (20 cm) on all sides of the core. Top and bottom reflector
-s are contained in the core tank, side reflector surrounds the v)
I
l-~(-(0 core tank and contains four 4"(10 cm) access ports running i tangentially to the core tank Gamma Shield 3.94"(10 cm) oflead shielding completely surrounding the graphite reflector Reactor Tank A 5/16"(80 mm) thick steel tank,37.4"(95 cm) diameter and 58.3"(148 cm) in height tank. This tank contains core tank, reflector and lead shield and the appropriate access holes for the control and safety rods, the glory hole and access ports. This is a gas-tight vessel when all the seal are made. The upper portion of this tanx contains the removable thermal column. Water Shield Tank Steel tank serves as the main structural tank as well as a fast neutron shield. The tank is 6.5 ' (198 cm )in diameter and 7' ( 213 cm) in height. Reactor Dimensions 6.5 ' (198 cm) in diameter and 9.17' ( 280 cm) in height l Reactor Weight 20000 lbs. 11,400 lbs. without shield water Reactor Control Two Safety Rods 235 Safety Rod #1 14.51 grams of U 235 Safety Rod #2 14.51 grams of U Coarse Control Rod 14.49 grams 235 U 235 Fine Control Rod 3.64 grams of U O
L(. tg All rods, with the exception of the FCR is Q. Q mechanically coupled, are magnetically couple to the carriage which is driven into the reactor on a lead screw by a reversible DC motor. Total travel distance is about 25 cm
- d. Nuclear Data (1) Fuel Loading (a) - Approximate Critical Mass 665 grams 235 U (b) Last measured Excess Reactivity corrected to 20 C with glory hole empty 0.137% Ak/k (2) Neutron Flux 8 2 Average Thermal Flux 1.5 x 10 n/cm -sec at 5 watts 8 2 Peak Thermal Flux 2.4 x 10 n/cm -sec at 5 watts (3) Reactivity Worth of Reactor Components (a) Safety and Coarse Rods = 1,25 % Ak/k (each)
(b) Fine Control Rod = 0.31 % Ak/k (c) Cadmium Shutdown Rod ~ 2.35 % Ak/k (d) Polyethylene Rod in Glory Hole New Poly Rod 0.29 % Ak/k Old Poly Rod 0.26 % Ak/k e) Temperature Coefficient of Reactivity
= - 0.024 % Ak/k per degree centigrade
( 4) Important Figures (a) Fine Control Rod Calibration Curve Figure 4.2.2 (b) Inhour Equation Figure 4.2.2-3
. (c) Flux Profile Figure 4.2.2-4 i
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/^T V When the reactor is operated the rods are inserted into the reactor in a specific sequence to ensure that sufficient shutdown margin is maintained at all times. This sequence is controlled by a series oflimit switches associated with each of the safety rods 1
- and the coarse control rod. In order for rod drive motor power to be applied to Safety Rod #2, Safety Rod #1 must be fully inserted in to the core, this feature is tested during the prestartup checkout in the operating procedures. A similar system is associated with Safety Rod #2, the Control Rod will not have rod drive motor power applied until Safety r
Rod #2 is in its fully inserted position in the reactor core. All three of these rods are part of the reactor safety system and will be ejected f from the core anytime a scram signal is received, even .Nring rod insertion. The safety )' system is designed to be a " fail safe" system, this means that in the event of a scram c) b signal the electromagnets holding the rods have their power interrupted. This will allow the rods to be accelerated out of the core downward by gravity and by the force of spring i loading. The spring constant is such that these three rods are ejected from the core with a
- force of 5 g, with this amount of downward acceleration these rods are out of the core in
= 120 milliseconds. The safety rods insert a negative 0.7 % Ak/k during the first 50 milliseconds following the scram signal. The rods are decelerated by dashpots during the
^ last 10 cm of travel to avoid damage to reactor components. The Fine Control Rod due to its limited reactivity is not associated with the reactor safety system. This wazzu rod is driven out of the core at a rate of 0.5 cm/sec. Both the fine and coarse control rods are driven by reversible DC motors through p a lead screw assemblies that are controlled by switches located on the control console. b
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! 1 g i The maximum speed associated with rod movement is 1.0 cm/sec, this speed will result is I a maximum possible reactivity change of 2 x 10" % A k/k per second for the coarse
- control rod. Both the coarse and the fine control rods can move at two speeds,0.5 cm/sec j and 1.0 cm/sec, a switch for each rod is located on the control console to allow for the
$ different rspeeds. i The positions of all rods are indicated on the reactor control console but the safety rod position indications are limited to full out, fully inserted or somewhere in between. The fine and coarse control rods have numerical indicators of their position in respect to distance traveled. The trips and interlocks associated with the scram and interlock systems for the control rods are summarized in the following table. i Table 4.2.2-1 O Trips and Interlocks Leading to a Reactor Scram l 1 l Scram System Interlock Relay , High Level Channel #3 High Level Channel #2, Manual Scram Channel #2 Min. Period, Skirt Monitor High Level Interlock System Shield Water Temperature Switch, Earthquake Switch Shield Water Level Switch, Channel #1 Low Count Rate ! Channel #3 Low Level , Channel #2 Low Level Relay Chassis Interlock, Rod Drive System Plug i I A G -
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bh 4.2.3 Neutron Moderator and Reflector The neutron moderator associated with the AGN-201M is the polyethylene that fuel is part of the reactor fuel. The fuel is homogeneously mixed with 10,900 grams of polyethylene, to form the fuel disc in the reactor core. Since the fuel and polyethylene are part of the same matrix the neutrons are able to be Coarse slowed down to a certain extent right after fission. The reflector associated with the AGN-201M is made-up of high density graphite (1.75 g/cm') and is 20 cm thick. This reflector surrounds the reactor on all sides and is used to move neutrons back into the core after they scatter out ofit. 4.2.4 Neutron Startun Source A 16 gram plutonium-beryllium startup source is currently installed in the AGN-201M. wazzu This source was installed in access port #3 on May 1,1972 after being approved as Amendment No.11 to License R-23 by the USNRC. The relative position of the source as it relates to the nuclear instrumentation is shown in Figure 4.2.4-1. The source is in a position in access port in the access port 11.25 in. from centerline based upon a study conducted by facility staffin 1972. It was determined at this location the source provide sufficient neutron flux to ensure the detectors are in their normal operating range 42 4 (10 to 10 A). 4.3 Reactor Tank or Pool O V
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1 l L) ~l 9 A large 1000 gallon tank of water surrounds the AGN-201 reactor, while the t a V thermal column at the top of the tank, itselfis a separate tank. The bottom of the AGN -20lM is not shielded by this tank to provide access to the control drive mechanisms and the rods. The tank is filled with water and provides a fast neutron shield. scram features associated with this tank include a low-level trip and a low -temperature trip. This tank while providing fast neutron shielding is also the only cooling system associated with the AGN-20lM reactor system. The tank is a solid unit and rest on the base of the AGN-201 system. the reactor core and lead shielding and graphite reflector are in the center of the tank but not directly associated with it. 4.4 Biological Shield The biological shield is designed to limit total gamma and fast and thermal neutron dose at closest approach to the reactor is less than 100 mrem /hr at reactor power i,~) less than 1.0 watts. This dose rate should be as low as possible to allow access to the experimental aresa of the reactor room and to surround the ALARA principle. The other requirement is that the dose rate due total gamma and thermal and fast neutron dose is less 15 mren/hr in the accelerator room. The accelerator room dose rate is easily achieved due to the construction of the floor on the acceleratro room. A complete copy of a radiation survey has been included in the back of this section of the Safety Analysis Report. The biological shielding must also perform under emergency conditions. 4.5 Nuclear Desien ('h q)
hdd The construction and description reactor core and fueled control rods for the I AGN-201M have been described earlier in this chapter. The behavior of the AGN-201 reactor core has been well defined over many years of operations and its previous Safety Analysis Reports and Hazard Safety Report from Aeroject General Nucleonics. During normal and emergency operations no damage to the reactor core or fuel is expected due to the compact and conservative design of the reactor core as well as the presence of many passive safety features associated with this reactor system. 4.5.1 Normal Operatine Conditions During normal operating conditions the AGN-20lM reactor will exhibit little fuel ' burnup or usage. The reactor core has only a single configuration for its control rods and operates in a very low and limited power range. O 4.5.2 Reactor Core Physics Parameters Table 4.2-1 Reactor Characteristics provides a listing of some core physics ! parameters. The reactivity worths for the control and safety rods are determined annually t using maintenance procedure for rod worth. The values for the polyethylene rods can be j i determined in the same manner, as well as the cadmium shutdown rod. This procedure has been used for many years and has proven an effective method for the determination of rod worths. 4.5.3 Operatine Limits O
&N The limiting conditions for operations reactivity limits are designed so that the b
limitations on total core excess reactivity will help to assure reactor periods of sufficient length so that the reactor protection system and/or operator action will be able to shut the reactor down without exceeding safety limits. The shutdown margin and control and safety rod reactivity limitations assure that the reactor can be brought and maintained suberitical if the highest reactivity rod fails to scram ' i and remains in its most reactive position. These reactvity limits must apply to the l l reactivity condition of the reactor and the reactivity worths of control rods and experiments. l If these limit are followed it will ensure that the reactor can be shut down at all times and that the safety limits will not be exceeded.
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- a. The available excess reactivity with all control and safety rods fully inserted 1 and including the potential reactivity worth of all experiments shall not exceed 0.65% Ak/k referenced to 20 C.
- b. The shutdown margin with the most reactive safety or control rod fully inserted shall be at least 1%A k/k.
1
- c. The reactivity worth of the control and safety rods shall ensure suberiticality on l l
the withdrawal of the coarse control rod or any one safety rod. 1 The core excess reactivity limits are designed to ensure that the reactor will have reactor periods of sufficient length so as not exceed safety limits if the reactor protective O%/ system and/or operator action will be able to shutdown the reactor. These operating .
QD limits are verified during annual preventive maintenance procedure ROEX-3, Calculation p O of Shutdown Margin and Total Excess Reactivity. When setting operating limits the maximum steady state power level and
- maximum core temperature during steady state and transient opersting conditons one
- must always consider the integrity of the fuel matrix so that no fission products are i
allowed to escape the core matrix. The polyethylene core material does not melt below a temperature of 200 C and is expected to maintain its integrity and retain essentially all of the fission products at I temperatures below that point. The Hazards Summary Report dated February 1962 and submitted on Docket F-15 by Aerojet-General Nucleonics (AGN) calculated a steady l l state core average temperature rise of 0.44 C/ watt. Therefore, a steady state power level I 0 of 100 watts would result in an average core temperature rise of 44 C. The corresponding maximum core temperature would be below 200 C thus assuring integrity of the core and i .i i retention of fission products. 1 The safety limits shall be set to ensure the fuel matrix inaintains it integrity at these levels
- a. The reactor power level shall not exceed 100 watts,
- b. The maximum core temperature shall not exceed 200 C during either steady state or transient operation.
Now that the limits exist reactor safety system setpoints must be determined which will limit maximum power and core temperature. These shall be the Limiting r3 Safety System Settings, at these setting an autimatic protective action will be intiated to V
~ - - - . . .. _ - - . . - . - . . . - - - - - - . - . - - .
a L( '2 3 scram ii.e reactor and prevent a safety limit from being exceeded. j (d Based on instrumentation response times and scram tests, the AGN Hazards Report concluded that reactor periods in excess of 30-50 milli-seconds would be i adequately arrested by the scram system. Since the maximum available excess reactivity l in the reactor is less than one dollar the reactor cannot become prompt critical and the . \
- corresp - ling shortest possible period is greater than 200 milli-seconds. The high power i j LSSS ,1J watts in conjunction with automatic safety systems and/or manual scram 1
l capabilities will assure that the safety limits will not be exceeded during steady state or as f a result of the most severe credible transient. I' In the event of failure of the reactor to scram, the self-limiting characteristics
- 6e to the high negative temperature coefficient, and the melting of the thermal fuse at a j i
tanperature below 120 C will assure safe shutdown without exceeding a core 1-O iemperature of 200 C. So the setting have to be 510 watts l , Channel Condition LSSS i Nuclear Safety #2 High Power 510 watts
! Nuclear Safety #3 High Power $ 10 watts 1-The core thermal fuse shall melt when heated to a temperature of 120 C or lessresulting in core separation and a reactivity loss greater than 5% A k.
< 4.6 Thermal-Hydraulic Desien The thermal-hydraulic design of the AGN-201M reactor system is limited to the O
%.)
transfer of heat to the shield tank. The shield tank is so large,1000 gallons and the heat
i. l~ ' i 3 I j' load so low from the AGN-201M reactor that the reactors effect on the tank is not i notice'd . But the reverse is not true, the tank must be filled to within 10 inches of the top l in order to meet shilding requirements. The shield tank is the primary fast neutron shield in the AGN-201M reactor design. ! I i a' ' I I l i O
I i Procedure for AGN Reactor Radiation Survey J 1. Fill out cover page including instruments to be used in the course of the survey.
- 2. Perform a background radiation survey in the AGN Room. Record the results as being performed at "0" (zero) power level. Measurements made in the AGN Room are performed with the detector placed on top of a meter stick and the bottom of the stick on the "X" on the floor.
- 3. Perform surveys at various power levels. Although rated power is 5 watts, the reactor may not be able to achieve full power due to temperature constraints. At the highest power level achievable. survey in the Accelerator Room. These measurements are made with the detector placed on the floor.
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7 i S'd . l i ' p g !- Chapter 5 REACTOR COOLANT SYSTEMS j _ The AGN-20lM reactor is operated at a very low power levels for short periods of 1 l time. The 1000-gallons of water in the shield tank, whose design function is that of a ] fast neutron shield, will by natural convection flow dissipate any heat that is rejected from 3 the core. The amount of heat tranferred to the shield tank has a little effect on the e i' temperature of the tank. i i i l i
- 1 i-i 4
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- . . - - . . . - - - . ~ . - . . - . _ . . - . . - . - . - . . . - . . - . - . . - , - . . _ . .
6-1 ! Chapter 6 ENGINEERED SAFETY FEATURES I
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( Engineered safety features (ESFs) are systems provided to mitigate the radiological consequences of designed-basis accidents. Since the AGN-20lM reactor i 1 l system operates at such a low maximum power level,5 watts thermal, the fission product
- inventory is very low. Evaluation of the maximum hypothetical accident has indicated no i
significant radiological releases. Therefore, no ESF systems are required or needed at an
- l
- , AGN-20lM reactor facility. l 3-
! l i j 1 I J O 5 b l } O
0 w/ 1 i Chapter 7: INSTRUMENTATION AND CONTROL SYSTEMS (V 7.1 Summary Description The instrumentation and control (I&C) system of an AGN-201 reactor consists mostly of three nuclear instrument drawers that allow for operation to of the reactor and, if necessary, send a variety of scram signals to the reactor control safety systems. The design basis for these trips will be discussed in detail later in this chapter. The reactor control system is designed, through a series ofinterlocks and relays, to ~ control movement of the safety and control rods. This is done to ensures that a reactor startep cannot commence unless both safety rods are fully withdrawn from the core. The interlocks ensure that only one safety rod can be inserted at a time and that the coarse control rod cannot be inserted unless both safety rods are fully inserted. The rods are controlled at the control console by a series of four switches, one for l,_) each rod, and position indication is provided on the console. Indicator lights will illuminate to give the operator indication of the status of each rod. These lighst are illuminated when the rod and the carriage reach certain positions in their travel. A series of eight relays (K10 -17) actually provide the controlling features to this system. The reactor control safety systems receive inputs from the nuclear instruments and many other detectors and sensing systems and voltage detectors on the interlock relay system itself. The console and at its features are clearly presented in a series of five figures presented in section 7.6 of this chapter. The layout of the control console is presented in these diagrams and all detectors and instrumentation is clearly labeled. V)
72 i
- The only area monitor device associated with the AGN-20lM reactor is the skirt
' b d monitor under the reactor. This system is intended to scram the reactor if radiation levels l greater than 2 times the last skirt monitor indication at 5 watts..
- 7.2 Desien ofInstrumentations and Control Systems i
4 7.2.1 Design Criteria All systems and components associated with this system are located in the reactor console, in the shield tank or under it. The reactor control system diagram is presented in Figure 7.2.1-1, this figure shows both the interlock system as well as the scram system for the AGN-201M. In Figure 7 2.1-2 a block diagram of the nuclear instrumentation j systems for the AGN-201M is presented. A complete description of each of the nuclear j instrumentation channels and skirt monitor is presented in section 7.2.3. g The design and operations of these detection systems is such that the Limiting
- U Safety Systems Settings will not be exceeded due to a reactor scram initiated by these systems. The control and safety setpoints are presented in Table 3.1 in the Technical e
4 .; Specifications for the AGN-20lM reactor and will be presented later in this chapter. 1 7.2.2 Desien - Basis Requirements i The design basis of this system is to monitor the the entire range of reactor power i and to provide protective features that will safely shutdown the reactor in the event of an equipment failure, operator error, or accident. The three nuclear instrumentation channels constantly monitor the reactor at all times. The scram and interlock reactor control safety systems are directly associated with the nuclear instruments since many of the signals to q these systems come directly from this instrumentation. V
M. e 2 8v P.S. (comes on when powe r e<ll O r r SCBAP H"T. VUl.1 A2 3 l-------1 ROD CAlllTS HONil"9 RELAY g pop vingvg g l_ _ _INTERI DCM_-_ _ _, ANN. 6 SYS. PLAc R EIAY I I I I I l 1 Ann. 13 f SillEtti W41ER l CHANNEL 3 p ANN. 4 l ( TEMPT. SWITCII l HIGl LEVEL g
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SCRAM REl>Y SCRAM REl.AY 7.2.1 -1 REACTOR CONTROL SAFETY SYSTEM ANN.16 - _ ANN.22 ANN.14 - - ANN.20 At: 1.15 - - ANN.21 l MAGNET l l MAGNET l l MA(. NET l e O *
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2.r 7.2.3 System Description (3 O The nuclear instrumentation system consists of three channels and the skirt monitor. Channel No.1 The components that consprise nuclear instrument Channel No. I are presented in block format in Figure 7.2.1-2. The detector is a BF 3 - filled proportional 6 counter with a range of 10 -10 counts per second. The detector is mounted in the shield tank and the relative position of all detector is presented in Figure 7.2.3-1. The channel drawer for this detector is located directly below channel No. 2 in the left-most side of the control console. Channel No.1 is a low-range detector and its primary function is to ensure a neutron source is present in the reactor during startup. The detector system is made up of g a high voltage power supply, a preamplifier and amplifier. The linear amplifier and
'") power supply are located in the back of the control console in a NIM bin, the preamplifier is located on the top of the reactor in the cabling runs.
The output of this detector could be displayed on one of the chart recorders but normally this channel is not recorded. The high voltage for this detection system is determined annually as part of the preventive maintenance program 5 . The system is also checked prior to reactor startup during the prestartup checklist. The ' push to calibrate" button is depressed and the meter should have a 60 counts per second input signal, if not the system can be adjusted with the calibrate to 60 CPS knob. Once this adjustment is completed the push io calibrate button is pressed a second time and the system comes i back on line. Another check performed as part of the pre-startup checklist is the low O count rate channel #1 interlock. This check is performed to ensure that if the count-rate O
M drops below 10 counts per second the reactor will scram. An annual maintenance f) V procedure also test this interlock to ensure that it is functioning correctly. Channel No. 2 The block diagram for Channel No.2 is presented in Figure 7.2.3-1. The function of this channel is to provide both high and low-level trips as well as an input l signal to the period meter. This system present output in a logarithmic fashion on a 4 ammeter that reads from 103 - 10 amps allowing the power to be monitored over the complete range of reactor operations without switching scales. l l The detector associated with Channel No. 2 is a BFrfilled ionization chamber. The detector is physically located in the shield tank as shown in Figure 7.2.3-1 in a waterproof housing. The high voltage power supply for the detector is located in a NIM i bin in the back of the control console. Once the signal leaves the detector it goes to a Keithley Model 420 Log n O Amplifier and Period Meter for processing This processed signal is fed into the ammeter I on the face of the Channel No.2 drawer for a logarithmic indication of power. A five - 1 position switch is located on the front of the channel No.2 drawer that allows testing of l the drawer and the associated trips and interlock relays. These tests are performed during the prestartup checkout procedure and are required to be done prior to a reactor startup. This proc, 'are test the operation of Channel No.2 drawer and the period meter as well. The first test of the drawer is to ensure that it will operate correctly the range of oreactor power. In order accomplish this, a series of test signals are feed into the system depending upon the switch position. Test signals of 103,10~" and 10 4are selected and A h
. . . . .- - . - . . - . - . . - . - - - - . - - . . - . - ~ - _ - - . - . . . - . - - .
! 70 l 4 .! an individual potentiometers adjused for each test signal, this process is repeated until all 4 (N e t v) three input signals are read on the meter face, 4 A low level Channel No.2 interlock is also associated with the meter on the front of the drawer, the interlock needle is moved to contact the meter needle and the interlock f should activate. A high-level trip is also checked and the switch on the front of the drawer is positioned to do this. This setting is calculated annually during reactor l maintenance and is used to ensure the reactor will automatically scram at about 150% of the five watt reading. Operation of the period meter is also checked during this procedure and a test signal is delivered through the check = switch on the front of the drawer. Once this is check a test signal is delivered to the period meter since it responds directly to this input. Once this check in complete the switch on the Channel No.2 drawer is moved to the 10 4 O input position and the operation of full the ininimum period scram is verified with the trip needle positioned at the 5-second position. An annual maintenance requirement also is in place to check the proper operation of the period meter. The output signal from the Channel No.2 ammeter has a number of very important ftmetions , it provides a signal to both the interlock system and scram system directly as well as to the period meter which also has a direct input to the scram system. The signal also is fed into one of the recorders so a record of Channel No.2 output is permanently recorded.
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l l Channel No.3 The block diagram for Channel No.3 is presented in Figure 7.2.1-2. The primary function of this channel is to provide high power level scram protection and j l to ensure that the LSSS and LCO are not exceeded. This channel is the primary indication of reactor power over all ranges of operation. The Channel No. 3 detector is located inside the shield tank as shown is Figure 7.2.3-1. This detector is fitted into a waterproof container and is a 13F3 -filled ionization chamber. The Channel No.13 drawer has a Keithly 410A picoammeter installed in it with reads the output signal from the detector directly into watts. The lowest scale of detection of the selector switch is 0 -3 mircowatts and the highest scale reads from 0 - 10 watts. i The automatic trips associated with this drawer come directly from the picoammeter and are set by annual maintenance procedure to trip at about 5% and 95% of full scale. The 95% of scale trip will activate a scram system relay and the 5% of scale will activate an interlock relay, both will lead to reactor scrams. These scram features are
. tested prior to a reactor startup and the picoammeter is also tested on the 0 -10 watt scale.
The output of this drawer is also directed to a recorder which records Channel No.3 output on a 0 -10 scale. Recorders Two recorders are present in the control console and in operation during all reactor operations. These recorders are checked for proper operation during the prestartup checklist. One recorder is always recording Channel No.3 while the other maybe selected between the remaining channels. As per the operating procedures, Channel No.2 is always selected and this is verified by the prestartup checklist. Each chart is stamped with the time and date of both reactor startup and shutdown to maintain a permanent record. (] V
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}(O The Technical Specifications clearly state all the required testing for all channels 'V of the nuclear instrumentation system. The LSSS and LCO are clearly described in this document.
7.3&st Reactor Control and Protective Systems The AGN-20lM reactor system only has a manual control system for control of the safety and fine and coarse control rods. The first part of this section will discuss a system design to limit rod movement and the second part will discuss the reactor protective system. The only interlocks associated with this system are directly related to rod movement and the maintenance of adequate shutdown margin. The rods in the AGN-201M reactor system which are directly related to reactor safety, the two safety rods and coarse control rod, have limitations on their respective movement and cannot be moved unless certain conditions are meet. ! \ V The first rod to be movd is Safety Rod #1, this rod must be driven into its fully inserted position prior to the K-10 relay activating, cutting power to Safety Rod #1 and allowing movement of Safety Rod #2. The K-relays are designed to activate only when the appropriate safety system rod is in its fully inserted position or in the fully withdrawn position depending upon their function. Once Safety Rod #2 is fully inserted and the K-12 relay operates and then the Coarse Control Rod can be moved. A set of these K relays also are activated when the carriages are retumed to the down position so that rod movement maybe started again after a reactor scram. , l The reactor protective system, known as the Reactor Control Safety Systems consists of the Interlock System as well as the Scram System. The Interlock System was I designed as an add-on to the original Scram System when it was found over the years of \q~.s
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( *, L/ none of the LSSS and LCO were exceeded as the reactor power levels were raised. I The Reactor Control Safety Systems rely upon signals from the nuclear instruments as well as other detection systems to perform their designated function, that is, to scram the reactor. The features directly associated with the nuclea-instrumentation system were discussed in previous section, so the other parts of the Interlock Relay and Scram System will be addressed in this section. 4 The manual scram switch is located in the middle of the control console and 4 maybe activated in the event of any emergency condition that exists in or around the 1 reactor facility. This switch is also used to scram the reactor after completion of nonaal reactor operations. A diagram of the Scram System is provided in Figure 7.2.1-1. J , The Interlock Relay in the Scram System will activate anytime one of the O kJ Interlock System trips is activated,just as any trip circuit in the Scram System. 1 The Interlock System consists of some trips that are associated with the shield l water tank, a low-temperature trip that activates at 20 !_0.5 C and a low level trip that activates when the level in the shield tank falls below 10 inches from the highest point. { i These are both important to reactivity control in the reactor as well as shielding I requirements in the reactor room. Both of these trips are checked as part of the annual maintenance program and as directed by the Technical Specifications. The other trips will active if the scram bus voltage is sensed to be low a trip will occur or if the rod drive system plug is removed from any of the safety system rods. ( v i
7-/2 s 7.6 Control Console and Displav Instruments The reactor control console is presented in Figures 7.6-1-5. These figures provide a clear idea of the particular console design far the AGN-201M reactor. The console is designed so that the operator has within easy reach all control switches for the nuclear instrument channels as wd! ".s a clear view af all indications including the alarm panel. Figure 7.6-1 depicts the alarm panel as well as the skirt monitor. The indicator lights on the panel are different colors and give the operator a visual as well as audio indication of
. the reactor status.
The rod energized lights are green and the scram indicator lights are red, these are just two examples of how the alarm panel was designed with the operator in mind. The rod position indicator lights are different colors which makes the identification very O simple as well as providing a visual indication of rod position. Figure 7.6-2 displays th: l center of the control console and the position of one the recorders and the rod position indicators and well as the manual scram button. Figure 7.6-3 displays the other recorder, the one selected to Channel No. 3, as well as the drawer for Channel No. 3. Figure 7.6-4, the last of the upper section of the control console indicates the relative positions of nuclear instrument Channels No.1&2. The location of the control panel for the source drive is also shown in this figure. These figures clearly show the case with which the reactor operator can observe and monitor reactor o,oerations. In Figure 7.6-5, the switches that control the rods are shown, this horizontal panel is located directly below the recorder for Channel No. 2. O
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Tle 7,7 Radiation Monitorine Systems / The only radiation monitoring system associated with the AGN-20lM reactor is the skirt monitor. A G.M. tube, inside a lead shield, is in place under the reactor inside the skirt. The skirt monitor panel on the control console provides an indication of the radiation level under the reactor. As part of the annual maintenance program the radiation levels at a reactor power of 5 watts is recorded and a voltage plateau performed on the detector. The skirt monitor is set to alarm at two times this 5-watt reading. It is assumed that, if the radiation levels reach these levels in the local area, some accident or loss of shielding has taken place. When this monitor reads its setpoint the reactor will scram and the evacuation hom will sound in the areas outside and above the reactor room. Should this occur Emergency Procedures will be initiated, and the cause of the problem will be investigated, and corrective actions will be taken. p(). o
di Chapter 4 REACTOR DESCRIPTION j O 4.2 Summan' Description l j The AGN-20lM research reactor in operation at Texas A&M University is essentially identical to all the other AGN-20lM reactors that have been operated since 1957. Therefore, the reactor description given in the original Aerojet-General Nucleonics l AGN-201 reactor report (Hazard Report and Preliminary Design Report by Aerojet-General Nucleonics, Docket F-40) is applicable to this reactor as are the Hazard and Safety Analysis Reports for the other AGN-20lM reactors on file with the USNRC. The characteristics and operating parameters of this reactor have been calculated and experimentally determined from data from this and other AGN-20lM reactors. The rated thermal power level for this AGN-201M reactor is 5 watts. This power level was approved by the USNRC in Amendment No.10 to License R-23 issued on O January 18,1973. The critical mass of the reactor is approximately 665 grams of 235 U. The fuel 1 itselfis in a plate type configuration and is homogeneously mixed powder of polyethylene ; and UO2 (61 mg 235 U/cm3)in the form of 20 micron diameter particles. The core disc have been formed by pressing under high pressure the homogeneously mixed powder. The reactor is a free standing unit, the reactor core is surrounded by a graphite reflector, a lead shield and a large water shield tank for fast neutrons. The reactor has a natural convection cooling system through the graphite reflector and the lead shield to the 1000 gallon tank of water the surrounds the reactor core, this removes any thermal heat due to the fission of the fuel. O
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- . The reactor has a hermetically core tank surrounding it , this tank is designed to iO ld L~
contain fission products gases if any core damage were to occur. ! i The experimental facilities associated with this AGN-201M are the small 15/16 i inch diameter glory hole that runs completely through the reactor and the 4 -inch diameter , t access ports that pass through the water shield and graphite moderator and lead gamma i shield. A side view of the AGN-201M reactor is presented in Figure 4.1-1. ; l i 4.2 Reactor Core I l 4.2.1 Reactor Fuel i
- The reactor core of the AGN-201M reactor consist of nine separate fuel discs 4 l
/ . which are stacked into a right circular cylinder as shown in figure 4.2.1-1. This figure i > )- clearly depicts the physical relationship between the control rods, the core and which the ' ~ core tank. The final core configuration is.10.15 inches (25.8 cm)in diameter and 9.34 I inches (23.75 cm) in height. The fuel in the core is made up of approximately 665 grams l l \ of *U in 3330 grams of U for an enrichment of just less than 20%. Figure 4.2.1-1 2
- shows the relationship between the core and the rods. The uranium is homogeneously mixed with polyethylene to form a series of fuel plates. A physical description of these fuel plates is included in the reactor characteristics table 4.2 1 The AGN-20lM reactor core also contains four fueled control and safety rods.
i Three of these rods, Safety Rods 1&2 and the Coarse Control Rod, are constructed in a similar manner while the Fine Control Rod is much srnaller in size. The fuel contained in these rods is exactly of the same type as the rest of the core. The reactor fuel has some limitations associated with it, in particular the breakdown of the fuel at high temperature. Several design features of the AGN-20lM will prevent this temperature (200 C) from
l l g-1 8 Electrical Power Systems 8.1 Normal Electrical Power Systems The AGN-201 reactor system has an electrical requirements of 2kw of 110 V AC. This power is supplied directly to the reactor room from the Zachry Engineering Center power network. The actual power to the reactor room is both 220 and 110 volt AC. Both of these power systems can be remotely disconnected using two breakers located in the hallway outside of room 60C. The AGN-20lM reactor system has no emergency power requirements. If a loss of power occurs during operations the rods are forced from the core and the reactor is considered shutdown. The cadmium rod is added to the core to ensure the reactor is shutdown since without instrumentation this is hard to verify. A radiation survey can also be performed to ensure the reactor is shutdown. The AGN-201M reactor system uses a variety of voltages and currents in normal A U operations. It would be impossible to present all that information even in this context. The major power distribution electrical block diagram for the Texas A&M University AGN-201M reactor is Drawing Number 2-C-100631. This diagram breaks down the regulated i10 volt AC to all the detector high voltage power supplies and to each of the , channel drawers. This regulated AC passed through two sola type voltage transformers, these transformer have been replaced by modern line conditioners to better handle the power from the Zachry Engineering Center. 8.2 Emercenev Electrical Power Systems The AGN-20lM reactor does not require an emergency power system due to its i design and lack of external systems. The reactor facility has an emergency lighting i i O I 'd ! 1
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I I 9.0 AUXILIARY SYSTEMS AGN-20lM has few auxiliary systems associated with it and none have any safety significance expect for the ventilation system. 9.1 Heatine. Ventilation, and Air Conditioning Systems The ventilation system for the AGN-20lM reactor room consists of one circulating unit by the reactor console, a supply area in the accelerator laboratory and an exit fan on the roof of the 7.achry Engineering Center. The inlet plenum for the fan suction is located in room 135, the accelerator laboratory and the air flow is designed to pass from the reactor room up through the grate in the ceiling and into this inlet plenum. Measurements have confirmed that air passes from outside both the accelerator laboratory and the reactor room in this manner. The supply header in the accelerator consist of three outlets supplying air at 500 cfm each and the exhaust fan drawing off air at 1500 cfm. The unit in the reactor room recirculates air at 1100 cfm. 9.2 Handline and Storane of Reactor Fuel The AGN-20lM reactor has such a low fuel bumup rate that it will never have to be *efueled in its useful lifetime. Some approved experiments can require the removal of fuel from the core but these experiments are no longer performed at Texas A&M \ University. 9.3 Fire Protection Systems and Programs j i No permanent of fixed fire suppression system exists in the Zachry Engineering ! Center. Several CO2fire extinguishers are present in the reactor facility including one by O i I
9-2 the reactor control console and another outside the reactor room in room 61B. The design of the facility limits the effects of fire on the facility. Thc rcactor console is in the comer of the room a fire starting in the area it could be put out before spreading. The reactor room itselfis large and without much equipment in it so everything is fairly spread out. If a fire were to occur outside the facility, access to the facility would be limited and the fire would have to pass the thick concrete walls and ceiling of the accelerator laboratory and the reactor room and adjoining laboratory. The doors into both the reactor room and accelerator room are metal so the fire would have little chance of getting through. Chemicals used by the department are stored inside explosion- prooflockers. also At the end of the biannual reactor emergency drill involving outside agencies one of the local fire department companies reponds at the scene. After the drill is over, fireman are given a tour of the facility and different aspects of the facility are pointed out to them.. V) t One of the Emergency Procedures directly addresses a fire in the facility and the i Reactor Emergency Procedure (RE-1) addresses a fire that could be in a nuclear system. J With the core surrounded by 1000 gallons of water and the isolated nature of the reactor room in the Zachry Engineering Center the chances of any damage to the reactor or any reactor components is very small. 9.4 Communication Systems An public address system is used by the reactor operator to make important announcements in the facility. The microphone is located on the reactor console and the speakers for the system are located in the hallway outside of room 60C and room 135. n s
9-3 9.5 Possession and Use of Bvoroduct, Source and Special Nuclear Material
- The only special nuclear material directly associated with the AGN-20lM reactor i
] is the startup source located in access port #3. This is a 1 Ci Pu-Be sealed source and is 1 i used to provide neutrons in the reactor. Several other sealed source - 4 J within the reactor room but these are just stored in the source locker and a e' - control of Radiological Safety. Two sealed reactor fuel plates are also stored it. me source locker, one from an AGN-211 and one from an AGN-201. These plates will not be used and will be returned with the rest of the AGN-201M fuel when the reactor is decommissioned. O O i
(O'I 10 EXPERIMENTAL FACILITIES AN AND UTILIZATION The experimental facilities associated with the AGN-20lM reactor system consist of the glory hole, which passes directly through the center of the reactor and the access ports which pass outside the reactor in the graphite reflector the lead and water shields. This particular facility has very little experimental use. The TRIGA reactor is primarily used by researchers at the University for experimental research. The AGN-201M reactor is used for experiments in nuclear engineering classes. The only experiments run using the reactor during these classes are from the set of approved i experiments for the AGN-20lM. Section 3.3 of the Technical Specifications clearly listed all the limitations on experiments at the AGN-201M reactor facility. ; i If an experiment were proposed to be performed using the AGN-20lM, the i l experimental procedure first would have to be approved by the reactor supervisor. Then, \ this approved proposed experiment would be presented to the Head of the Department of l I Nuclear Engineering. Once reviewed and approved by the Department Head, the proposed experiment must be presented to the Reactor Safety Board, if a majority on the Board approve the experiment only then can it be performed It may only be perfomied under the direct supervision of the reactor supervisor and wi'h the approval of the Department Head. 10.1 Summary Description The AGN-201M reactor design has both the in-core facility, the glory hole, and in-reflector facility with the access ports. A thermal column may be installed in the AGN-201M. However, this feature has not been used and this area is filled with water. O
lb' 10.2 Experimental Facilities I I 'v The experimental facilities in an AGN-20lM reactor design are limited to the 1 glory hole and the access ports. While the center of the glory hole provides a relatively l large neutron flux, the remaining area of the glory hole will be filled by either an experiment holder or a poly rod so the generation of radioactive gases is very limited. 1 10.3 Experiment Review The experiment review process is a three-level system at the AGN-20lM reactor facilay. The proposed experiment must be presented to the reactor supervisor for review l I and consideration. His responsibilities include ensuring all requirements are meet for 1 limitations on experiments and reactivity as per the Technical Specifications. This l review includes a verification of all reactivity values of the materials and experimental equipment that are exposed to the reactor neutron flux. Special care must be and taken to / ,\ \
'v ensure that if any material becomes activated during the experiment that proper radiological controls are taken and that personnel from Radiological Safety are directly involved in evaluating the experiment. The reactor supervisor must ensure that no part of the experiment can lead to damage to reactor systems or injury of personnel in the event of a failure. If the reactor supervisor is satisfied that the experiment can be conducted safely and poses na danger to the general public, he must present the proposed experiment to tne Head of the Department of Nuclear Engineering.
The Department Head must then complete an evaluation of the proposed experiment and determine ifit is in fact worth the resources to conduct and has no unresolved safety considerations. All the aspects of the experiment will be discussed
,q with the reactor supervisor to ensure no questions remain. The Department Head must
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10 3 then present the proposed experiment to the Reactor Safety Board for consideration. If a majority of the Board approves then the experiment can be conducted. The experiment can only be conducted with the direct approval of the Department l l Head and under the direct supervision of the reactor supervisor. l l 1 l n v
i 1 I -l F 11 RADIATION PROTECTION PROGRAM AND WASTE MANAGEMENT 11.1 Radiation Protection The radiation protection program at Texas A&M University is the responsibility of the Radiological Safety personnel which is part of the University's Environmental Health and Safety organization. The AGN-20lM reactor room is part of the nuclear i engineering but all of these laboratories lie within the physical area of the reactor facility as defined by the Security Plan. The personnel directly involved in the radiation 4 protection program for the AGN-201 are staff of the facility. The reactor operators f control access to and control the actions of all personnel in the reactor room. Radiological Safety personnel conduct an annual radiation survey in association
; with the AGN's preventive maintenance program. This survey verifies the extremely low i
levels of radiation present in uncontrolled areas outside the reactor room. The main concern of the program is to limit exposure to both the students and the general public. The design of the reactor facility and its location in the Zachry Engineering Center isolate the reactor room from unrestricted access by the general public. The Physical Security Plan does not allow unrestricted access to the reactor facility by the general public. The operators do not allow any personnel in the reactor during reactor operations unless they have some form of dosimetery. The reactor operating procedures do not allow for a reactor startup without verification of proper personal dosimetry. An important design feature of the facility is the presence of the C-2 warning device on the door to the reactor room. This device will sound a horn if anyone opens the b o
(p 2 room door whenever the reactor console is energized. As a further limitation on access, the door to laboratory 61 A is always locked closed and only personnel assigned to the facility or doing research in that laboratory have a key to the door. 11.2 Radiation Sources , The only radioactive source that is directly associated with license R-23 is the I curie Pu-Be startup neutron source located inside access port #3. This sealed source is swiped quarterly and these records are maintained by Radiological Safety personnel. . A radioactive source storage locker is located in the comer of the reactor room 61B. All sources contained in this storage locker are under the control of Radiological Safety personnel. Three 2-Ci and two 1-Ci sealed Pu-Be sources are stored in this locker, along with one 2 mci Co-60 source. These sealed sources are stored inside containers in the . m ( ) V source locker. Inside the source locker are two fuel plates one from an AGN-211, and the other from an AGN-201, these plates are inside sealed aluminum storage holders and will remain inside these holders until a time when the AGN fuel is removed. Another source locker is located inside the counting laboratory, this locker and , laboratory are secured at all times unless in use these sources are used as check souses in 4 nuclear engineering laboratories and consist mostly of source sets. All of these sources d remain inside the locked storage cabinet, inside a locker storage container unless they are required for a laboratory. The teaching assistant or professor must obtain a key to the source locker from the laboratory manager and be briefed on precautions reguarding their use. b
it'3 1 l D\ (,/ 11.1.1 Airborne Radiation Sources The only possible airborne radioactive source associated with the AGN-201 could be due to the generation of Ar-41 from a neutron interaction with Ar-40. This is discussed in detail in section 10.88. This analysis shows a minimum danger at a this particular reactor facility. l 11.1.1.2 Liauld Radioactive Sources The design of this reactor facility and the experiments conducted do not lead to the production of any liquid radioactive sources. The only possible source is the 1000 gallons of water that surround the reactor in a sealed tank. This water will be analyzed and level of possible contamination determined as part of a decommissioning plan for the \ facility. 11.1.1.3 Solid Radioactive Sources No solid radioactive sources exist at this facility other than those described in section 11.1 on this chapter. No solid radioactive waste is present at the facility other than small amounts laboratory waste generated by students during experiments. This radioactive reactor waste is removed by Radiological Safety personnel, but by its very nature is a very small about of waste. No waste other than laboratory trash-type waste is generated by experiments directly associated with AGN-20lM reactor operations, and maintenance. O
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(> 11.1.2 Radiation Protection Pracram The radiation protection program for the AGN-201M is cooperative program between Radiological Safety personnel and the facility staff. The Radiological Safety Officer directs the radiological safety personnel in the workings of the program. Under emergency conditions , the Head of the Department of Nuclear Engineering or his designated alternate is the emergency director for the emergency response personnel at the AGN but gets direct input from radiological safety personnel. The main function of this program to protect the general public under normal and emergency conditions from radiation exposure and ensure that staff personnel and students use the ALARA principle to fullest extend possible to keep exposures low. The dose to the students is directly controlled and limited by the actions of the (")N
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facility staff. Any student that is working with the reactor will be issued a dosimeter. If they are classified as radiation workers in the department, Radiological Safety will issue them a suitable personnel monitoring device. If these individuals are nuclear engineering students, the reactor operator will issue pocket ion chambers, obtained from Radiological Safety personnel, prior to reactor operations. When a nuclear engineering class is scheduled to use the AGN-201M for experiments in the laboratory , the class is lectured about reactor operations and construction by a qualified operator. During these lecture, a tour of the reactor room is
- performed and throught this effort the student become familiar with the reactor room layout. It is during this tour that the reactor operator points out the location of specific 1
m items (i.e. the source locker). This leemre will point that avoidance of these areas will
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Li-5 l lower the radiation exposure of the students. When the student is to perform their reactor p V startup, one of the first steps of the pre-startup checklist is to conduct a radiation survey of the room. it is during this time that the student is trained on the location of all survey 1 points and on the practice of ALARA during reactor operations. Reactor operations during laboratory classes is limited to the lowest power level practical for the experiment. Student access to the reactor room is limited to only personnel necessary to successfully complete the experiment and to operate the reactor. All other students involved in the experiment, are allowed access only to room 61B, this I way the students level of exposure is greatly limited. l l 11.1.3 & ARA Program The ALARA program in place at Texas A&M University comes directly from ) procedures created by the Of6ce of Radiological Safety. All of these procedures as well detailed instructions are presented in " PROCEDURE MANUAL FOR TIIE USE OF RADIOlSOTOPES AND RADIATION PRODUCING DEVICES" which was last revised in July,1990. The ALARA Program is addressed in section 4 of the manual 11.1.4 Radiation Monitorine and Survevine At the University personnel from Radiological Safety perform the required radiation surveys and instrument calibrations. The basis for the radiation survey is detailed in the Technical Speci6 cations section 3.4. and speci6 cations and basis for radiation monitoring and control are contained in section 4.4 of the Technical q Speci6 cations. Section 4.4 also details the requirements for calibration of all ponable V
a (4-6 radiation survey instruments assigned to the reactor facility by personnel for Radiological
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() safety. 11.1.5 Radiation Exposure Control and Dosimetrv Radiation exposure control is a very important concern at the AGN-20lM facility. The reactor room is considered to be a radiation area whenever the reactor is operating at a power level ofless than 0.9 watts and a high radiation area whenever the reactor is operated at power levels greater than 0.9 watts. These radiation levels raise many ALARA concerns but the reactor is only operated at power levels greater than 0.9 watts when required by maintenance procedures. When students operate the reactor for laboratory reactor startups, the power level rarely exceeds 50 milliwatts. Ifirradiations are required for a laboratory only the reactor operator and one O b student will be in the reactor room and they are shielded by the shielding around the control console. If the activation foils are to be taken to the counting laboratory they are moved in a lead container and handled with large tweezers and remain in the custody of the reactor operator. When the laboratory is completed. the foils are placed into the source locker inside the reactor room for decay. All students performing reactor startups are issued pocket ion chambers and students performing experiments in the area around the reactor are also issued pocket dosimeters. These dosimetesr are provided by Radiological Safety personnel and issued by the reactor operator just prior to reactor startup. With the reactor operatingonly for about 30-40 hours per year mostly for reactor
- f. maintenance exposure, levels in the reactor room over the course of a year due to reactor G
//- 7 operations are relatively low. The reactor room has also been used for calibration of
- portable survey instruments by Radiological Safety personnel. These exposures have
, added to the dose measured by the local area badges. During years of consistent reactor usage and instrument calibration, the area badges read between 180 and 830 mrem per i
- year of gamma and neutron radiation. Since the calibration ofinstruments has been moved to another location, the area badge inside the reactor room still reads about 750 mrem per year but the area badge outside the reactor room read only 10 mrem.
The only anticontamination clothing wom at this facility are gloves that are wom to perform swipe surveys on samples, ne respir.itory protective equipment is used at the AGN-201M reactor. l The area badges and personnel dosimetry issued to facility personnel are evaluated monthly. If problems were to occur in this facility personnel from Radiological Safety A would be called in to assist facility personnel in accessing the problem and help in the i restoration efforts.
- Doses at this facility have been very low even to personnel assigned to the facility.
l In calendar year 1996 only one person assigned to the facility had any recorded radiation exposure.. Great care is taken by the facility staff to ensure that all personnel associated with the AGN-201M reactor receive doses which are as low as possible. 4 11.1.6 Contamination Control The reactor facility is swiped monthly by personnel from Radiological Safety. A radiation and contamination surveys are performed on all samples after they are removed from the reactor. With a sealed reactor system. no loose contamination is expected to
, found in the AGN-20lM reactor facility during normal operations. When the control NJ
l-a ! ((-8 L rods are removed from the for reactor maintenance or repair, they are surveyed to ensure no loose contamination is present. All the sources stored in the reactor room are sealed. l
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j '11.1.7 Environmental Monitorine No environmental monitoring is conducted at thia facility ! 11.2 Radioactive Waste Management f l l .. The only waste generated directly from reactor operations is simple laboratory 1 l trash. Efforts are made by facility staff to limit the amount of waste generated. The total amount is very small and Radiological Safety personnel dispose of this laboratory waste. i I i l t l l'
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- 12 CONDUCT OF OPERATIONS 12.1 Ornanization i
The AGN-20M reactor is owned by Texas A&M University and has been l operated by the University since 1957 under USNRC license R-23, Docket 50-59. l Since the reactor is at a university the organization surrounding it is quite unique to this l particular university and will be discussed in the following chapter. A detailed l l description of all duties responsibilities and requirements are presented in the Technical , Specification, section 6.0 12.1.1 Structure & Responsibility l The organizational structure at Texas A&M University is intended to fulfill the ! ! i requirement setforth by the USNRC. The organizational structure is presented in l flowchart format in Figure 12.1.1-1. l l
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l The President of the University is considered to be the chief administrative officer I l for the facility. The application for licensing of the reactor is made in his name. The l Dean of the College of Engineering is administrative officer responsible for the day to l day operations of the College of Engineering and, since the Department of Nuclear ; i < l Engineering is among those departments, he has an administrative role in the operationa l of the AGN-201M reactor. The Head of the Department of Nuclear Engineering is the person administractively responsible for the department. This includes the AGN-201M reactor. The Department Head shall have final authority and ultimate responsibility for the { operation, maintenance and safety of the reactor facility. The Department Head shall I appoint personnel to the position of Reactor Supervisor and shall select qualified O l
h" candidates as Reactor Operators. He shall work with members of the Reactor and i Radiological Safety Boards to ensure that no unresolved safety questions are present. He shall present new experiments and procedures and facility modifications to the Boards for their review and approval. The Reactor Supervisor shall be a licensed SRO and shall be responsible for the l l preparation, promulgation and enforcement of administrative controls including all the l rules, regulations, instructions and operating procedures to ensure that the reactor facility I is operated in a safecompetent and authorized manner. He shall direct the activities of operators and technicians in the daily operation and maintenance of the reactor; schedule reactor operations maintenance; be responsible for the preparation, authentication and storage of all prescribed logs and operating records: authorize all experiment and procedures and changes thereto which received the approval of the Reactor Safety Board and/or Radiological Safety Board and the Head of the Department of Nuclear Engineering. He shall be responsible for the preparation of experimental procedures involving the use of the AGN-201M reactor. Reactor Operators shall be responsible for the safe manipulation of the reactor controls, monitoring ofinstrumentation, operation of reactor related equipment, and maintenance of complete and current records during the operation of the facility. The Reactor Safety Board is respensible for, but not limited to, reviewing and approving safety standards associated with the use of the reactor facility. The Board shall review and approve of all proposed experiments and procedure and changes to either. O
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/2 g The Board must also approve all reactor facility modifications which might affect its safe O
V operation. A complete listing of the Reactor Safety Boards duties and responsibilities is 1 i 1 listed in Section 6.1.6 of the Technical Specifications. j i The Radiological Safety Board shall advise the University administration and the Radiologic;i Safety Officer on all matters concerning radiological safety at the university facilities. , 1 I
; 12.1.3 Staffine l
The staffing requirements for the AGN-201M are as follows
- a. The minimum operating staff during any time in which the reactor is not shutdown shall consist of:
- 1. One licensed Reactor Operator in the reactor control room. l
- 2. One other person in the reactor room or reactor , control room certified by the
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/ Reactor Supervisor as qualified to activate the manual scram and initiate j emergency procedures.
- 3. One licensed Senior Reactor Operator readily available on call. This requirement can be satisfied by having a licensed Senior Reactor Operator performing that duties stated in paragraph 1 o 2 above or be designating a
;icensed Senior Reactor Operator who can be readily contacted by telephone and who can arrive at the reactor facility within 30 minutes.
- b. A licensed Senior Reactor Operator shall supersise all reactor maintenance or modifications which could affect the reactivity of the reactor.
12.1.4 Selection and Trainine of Personnel n
k The selection of personnel to operate the AGN-20lM reactor typically has been
, (Aj restricted to former nuclear navy personnel or graduate students at Texas A&M University. These individuals require less basic traming as operators than would other less-qualified personnel. All personnel that operate the AGN-201M reactor, even students, undergo some form of training. The students are lectured on the construction and operations of the reactor pr ior '.o swdent startups.
12.1.5 Radiation Safety The Radiological Safety Officer at the University is responsible for all aspects of radiological safety. In this facility, the staff will train the student operators on the importance of the ALARA program and what particular aspects of that program can be used at the facility. As started above, all questions and reviews concerning radiological i safety are presented directly to the Radiological Safety Board.. At Texas A&M V University, Radiological Safety personnel have the authority to stop any experiment or c activity they believe to be unsafe. 12.2. Review and Audit Activities The Reactor Safety Board is responsible for reviews and audits at this facility. The review and audit activities of the Board are clearly outlined is Sections 6.4.2 and 6.4.3 of the AGN-201M Technical Specifications. The authority of the Reactor Safety Board comes directly from the office of the President of the University. The Reactor Safety Board shall review:
- a. Safety evaluations for changes to procedures, equipment or systems and tests or g experiments conducted without Nuclear Regulatory Commission approval under V
/M provision of paragraph 50.59 to verify that such actions do not constitute an unreviewed
() V safety questions.
- b. Proposed changes to procedures, equipment or systems that change the original intent or use, and are non-conservative, or those that involve any unreviewed safety question as defm' ed in 10 CFR 50 paragraph 50.59.
- c. Proposed test or experiments which are significantly different from previously l
approved tests of experiments, or those that involve an unreviewed safety question as defined in 10 CFR 50 paragraph 50.59.
- d. Proposed changes in Technical Specifications or licenses.
- e. Violations of applicable statutes, codes, regulations, orders, Technical Specifications, license requirements, or ofinternal procedures or instructions having nuclear safety significance.
/"\ > 1 V f. Significant operating abnormalities or deviations from normal and expected I performance of facility equipment that effect nuclear safety.
- g. Reportable occurrences.
- h. Audit reports.
l Audits of facility activities shall be performed at least quarterly under the cognizance of the Reactor Safety Board but in no case shall personnel responsible for the item conduct the audit. These audits shall examine the operating records and encompass, but shall not be limited to the following:
- a. The conformance of the facility operation to the Technical Specifications and applicable license conditions, at least annually.
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, l 2 ~l) j b. The Facility Emergency Plan and implementing procedures, at least every two years. c The Facility Security Plan and implementing procedures, at least every two years. l 12.2.1 Composition and Oualifications l- , j Reactor Safety Board members shall have a minimum of five (5) years experience in their profession or a baccalaureate degree and two (2) years of professional experience. Members will generally be University faculty with considerable experience l
- in their own field of expertise.
j 12.2.2 Charter and Rules i i The Reactor Safety Broad shall meet at least once per calendar year, or more often { as deemed necessary by the Chairman. A quorum for the conduct of official business 1 I
- shall be the chairman, or his designated alternate, and two other regular members. At no l \ time shall the operating organization comprise a voting majority of the members at any 1 1
- Reactor Safety Board meeting.
!. 12.2.3 Review Function i 2 Reviews are performed when either the facility staff has a proposed change to a a standing item or when some sort of violation has occurred. The review of standing procedures, Technical Specifications, security and emergency plans, etc. is not required but encouraged since same minor changes and updates occur have all the time.
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! /M t 12.2.4 Audit Function 4 The audit function at Texas A&M University is limited to members of the Reactor l Safety Board, In special cases, audits may be conducted by others, e.g. qualified 1 individuals from three University, consultants ,etc. 1 12.3 Procedures I ;
- There shall be written procedures for the followmg; ;
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- a. Startup, operation and shutdown of the reactor, l
- b. Fuel movement and changes to the core and experiments that affect reactivity. I i
, c. Conduct ofirradiations and experiments that could affect the operation or safety l of the reactor. ]. 4 d Preventive or corrective maintenance which could affect the safety of the reactor. l
- e. Surveillance, testing and calibration ofinstruments. components, and systems as I \
specified in section 4.0 of the Technical Specifications. i i
- f. Implementation of the Security and the Emergency Plans.
i } This list is taken directly from section 6.6 of the Technical Specifications for the AGN-l 20lM. 5 Any procedure must be submitted to the reactor supervisor for approval and then ! 4 j to the Head of the Department of Nuclear Engineering . Finally must be submitted to
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! the Reactor Safety Board for approval and signature of the Chairman.- 1 12.4 Reauired Actions Section 6.9.2 of the facility Technical Specifications clearly defines what are considered I to be reportable occurrences. In all cases of reportable occurrences, the reactor shall be shutdown.
/2 4 Reportable occurrences, including causes, probable consequences, corrective actions and (A) measures to prevent recurrence, shall be reported to the USNRC. Supplemental reports may be required to fully describe final resolution of the occurrence. In case of corrected or supplemental reports, an amended license event report shall be completed and reference shall be made to the original report date.
- a. Prompt Notification With Written Follow-up.
The types of events listed below shall be reported as expeditiously as possible by telephone and confirmed by mailgram, or facsimile transmission to the Director of the appropriate NRC Regional Office, or his designated representative no later than the first work day following the event, with a written follow-up report within two weeks. Information provided shall contain narrative material to provide a complete explanation of the circumstances surrounding the event. 1
- 1. Failure of the reactor protection system or other systems subject to limiting safety system settings to initiate the required protective function by the time a l monitored parameter has reached the set point specified as the limiting safety system setting in the technical specifications or failure to complete the required protective function.
- 2. Operation of the reactor or affected systems when any parameter or operation subject to a limiting condition is less conservative than the limiting condition for operation established in the technical summary specifications without taking permitted remedial action.
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- 3. Abnormal degradation discovered in a fission product barrier.
a (d 4. Reactivity balance anomalies involving: (a) Disagreement between expected and actual critical rod positions of approximately 0.3% Ak/k. (b) Exceeding excess reactivity limit. (c) Shutdown margin less conservative than specified in technical specifications. (d)lf sub-critical, an unplanned reactivity insertion of more than approximately 0.5 % Ak/k or any unplanned criticality.
- 5. Failure or malfunction of one (or more) component (s) which prevents or could prevent, by itself, the fulfillment of the functional requirements of system (s)
- used to cope with accidents analyzed in the Safety Analysis Report.
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- 6. Personnel error or procedural inadequacy which prevents, or could prevent, by itself, the fulfillment of the functional requirements of systems required to cope with accidents analyzed in Safety Analysis Repert.
l , 7. Unscheduled conditions arising from natural or man-made events that, as a j direct result of the event require reactor shutdown, operation of safety systems, or other protective measures required by Technical Specifications.
- 8. Errors discovered in the transient or accident analyses or in the methods used
^ for such analyses as described in the Safety Analysis Report or in the bases for , the Technical Specifications that have or could have permitted reactor ) operation in a manner less conservative than assumed in the analyses. i
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- 9. Performance of structures, systems, or components that requires remedial action or corrective measures to prevent operation in a manner less
{ l conservative than assumed in the accident analysis in the Safety Analysis ) Report or technical specifications bases; or discovery during plant life of conditions not specifically considered in the Safety Analysis Report or Technical Specifications that require remedial action or corrective measures to prevent the existence or development of an unsafe condition. 12.5 Reports Routine annual operating reports shall be submitted no later than ninety (90) days following the end of the operating year. The annual operating reports made by licensees shall provide a comprehensive summary of the operating experience havir'g safety 3 ' (O significance that was gained during the year, even though some repetitia af previously reported information may be involved. References in the annual operating report to previously submitted reports shall be clear. Each annual operating report shall include:
- 1. A brief narrative summary of (a) Changes in facility design, performance characteristics, and operating procedures related to reactor safety that occurred during the reporting period.
(b) Results of major surveillance tests and inspections.
. A tabulation showing the hours the reactor was operated and the energy produced by the reactor in watt-hours.
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- 3. List of the unscheduled shutdowns, including the reasons for the shutdowns and I corrective actions taken, if any.
- 4. Discussion of the major safety related corrective maintenance performed during the period, including the effects, if any, on the safe operation of the reactor, and i
the reasons for the corrective maintenance required.
- 5. A brief description of: l (a) Each change to the facility to the extent that it changes a description of the facility in the application for license and amendments thereto.
(b) Changes to the procedures as described in Facility Technical Specifications. (c) Any new or untried experiments or tests performed during the reporting period.
- 6. A summary of the safety evaluation made for each change, test, or experiment not (V submitted for NRC approval pursuant to 10 CFR 50, paragraph 50.59 which clearly shows the reason leading to the conclusion that no unreviewed safety question existed and that no technical specification change was required.
- 7. A summary of the nature and amount of radioactise effluents released or discharged to the environs beyond the effective control of the licensee as determined at or prior to the point of such release or discharge.
(a) Liauid Waste (summarized for each release) (1) Total estimated quantity of radioactivity released (in curies) and total volume (in liters) of effluent water (including diluent) released.
/2'IL (b) Solid Waste (summarized for each release) /
()\ (1) Total amount of solid waste packaged (in cubic meters) (2) Total activity in solid waste (in curies) (3) The dates of shipments and disposition (if shipped off site).
- 8. A description of the results of any environmental radiological . surveys performed outside the facility.
- 9. Radiation Exposure - A summary of radiation exposures greater than 100 mrem (50 mrem for persons under 18 years of age) received during the reporting period by facility personnel or visitors.
If a Safety Limit is violated then the appropriate USNRC Regional Office of l '~'\ (O Inspection and Enforcement must be contacted, the Director of the USNRC, and the Reactor Safety Board not later than the next working day. In addition a Safety Limit Violation Report shall be prepared for review by the Reactor Safety Board. This report shall be submitted to the USNRC and the Reactor Safety Board within 14 days of the violation. 12.6 Records Generally record retention falls into two broad categories, records that need to be retained for a period of at least five years and records that are required to be retained for the life of the facility. The records that are retained in this facility are contained in a series of file cabinets, one located inside the reactor room and one next to the reactor 7 supervisors desk. I (
12 - 0 The following is a list of records to be retained for a period of at least five years, (D V taken ia . 2 the Technical Specifications section 6.10 Records to be Retained for a Period of at Least Five Years:
- a. Operating logs or data which shall identify:
- 1. Completion of pre-startup checkout, startup, power changes, and shutdown of the reactor.
- 2. Installation or removal of fuel elements, control rods or experiments that could !
affect core reactivity.
- 3. Installation or removal ofjumpers, special tags or notices, or other temporary changes to reactor safety circuitry.
I 4. Rod worth measurements and other reactivity measurements. i
- b. Principal maintenance operations.
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\ c. Reportable occurrences.
- d. Surveillance activities required by technical specifications.
- e. Facility radiation and contarr.ination surveys.
1
- f. Experiments performed with the reactor. ;
. 1 This requirement may be satisfied by the normal operations log book plus,
- 1. Records of radioactive material transferred from the facility as required by the license.
- 2. Records required by the Reactor Safety Board for the performance of new or special experiments.
- g. Changes to operating procedures. !
O-/f (3 V Records to be Retained for the Life of the Facility:
- a. Records ofliquid and solid radioactive effluents released to the environs.
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- b. Appropriate off-site environmental monitoring surveys.
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- c. Fuel inventories and fuel transfers. !
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- d. Radiation exposures for all personnel. '
- e. Updated as-built drawings of the facility.
l f. Records of transient or operational cycles for those components designed for a j l limited number of transients or cycles. . 1
- g. Records of training and qualification for members of the facility staff. '
- h. Records of reviews performed for changes made to procedures or equipment or reviews of tests and experiments pursuant to 10 CFR 50.59.
- x. i. Records of meetings of the Reactor Safety Board. l 12.7 Emercency Plannine The main function of the Emergency Plan is to provide an guide by which the facility staffis able to contain and overcome emergency situations. The plan clearly states the role and response of area emergency services and with the procedures in place the severity ofincidents associated with the AGN-20lM will be limited .
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- l. 12.8 Security Plannine 3
(V The Security Plan associated with the AGN-20lM is intended to take full advantage of the isolated nature of the facility inside the Zachry Engineering. The location of the facility in the building is over two floors but with restricted access to the outer doors, access to the reactor room itselfis extremely limited. 12.10 Operator Trainine and Reaualification A copy of the Requalification Program for Licensed Reactor Operators and Senior Reactor Operators for the AGN-201M has been included in this application for license renewal as Appendix C. 12.12 Environmental Reports An environmental rpeort has been submitted with this license renewal application and is included as Appendix E. I l 3-
__ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ . _ _.~. _ _ _ _ j] 1 i Chapter 13 Accident Analyses 13.1 Accident-Initiating Events and Scenarios 13.1.1 Maximum Hypothetical Accident Studies have demonstrated the Maximum Hypothetical Accident (MHA) involving an AGN-20lM reactor system is a large reactivity addition. This analysis is based upon the assumption of an instantenous increase of 2%Ak/k reactivity above delayed critical with the reactor at a power level 5 5 watts. This situation would not t occur under normal situations but for the MHA it is being considered. This instantaneous 2% positive reactivity addition is considered to be the maximum creditable accident for the AGN-201M reactor system. In the past, this scenario has been analyzed using one group theory with one group of delayed neutrons. This analysis was first performed by Aerojet General Nucleonics and O was published in August 1957. This simple analysis assumed that:(1) at time zero a 2% 1 l step increase in reactivity was inserted with the reactor at a power level of 100 milliwatt ' and (2) the energy in the core at time zero was negligible in comparison with the energy liberated during the ensuing excursion. Further, this analysis assumed that there was no energy transfer from the core during the excursion. A complete description of the analysis can be found in the Hazard Summary Report. the results are wazzu summarized below. ! Numerical solution of the governing equations by finite-difference methods yielded a total energy release of 1,7 MJ and raised the average core temperature by y 71 C. This excursion produced a peak power excursion of 54.4 MW at 204 msee after the reactivity insertion. These results demonstrated that the core material would not sustain
_ __ _ _ _ _ _ _ _ . . _ _ _ . - . _ . . . _ _ . _ _ . . _ . _ ~ . _ _ _ . _ _ . - . _ _ . ._. l3- 2 catastrophic damage nor release any fission products because the average core temperature remained well below the melting point of the polyethylene matrix. A study completed by the University of New Mexico using the same parameters resulted in a total energy release of 2.41 MJ causing a temperature rise of 100.7'C with the temperature at the center of the core rising to a level of 156 C, total dose to a person < next to the reactor would be about I rem. Another study completed by the Idaho State University using the same parameters 0 resulted in a total energy release of 5.8 MJ causing a temperature rise of 170 C with the temperature at the center of the core only rising to a level of 120 C. Since this is the point the core thermal fuse would separate and insert a large amount of negative reactivity. Total dose to a person next to the reactor would be about 3.2 rem. Based upon these data and using the highest area radiation level next to the reactor at 5 watts. the dose would be 18.14 rem using the New Mexico data and 7.54 rem using the Idaho State data. These exposures are quite high but these conditions are not I realistic and will never occur. l I 13.1.2 Insertion of Excess Reactivity Consider in 13.1.1. I 13.1.3 Loss of Coolant Not considered in the AGN-20lM system 13.1.4 Loss of Coolant Flow i Not considered in the AGN-20lM system O
.g (33 , 13.1.5 Mishandline ff f alfunction of Fuel
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V Not considered in the AGN-201M system 13.16 Experiment Malfunction No explosives or corrosive are permitted in the AGN-20lM. 13.1.7 Loss of Normal Electrical Power t In the event of a loss of normal electrical power the reactor will scram and be l placed in a shutdown condition. i 13.1.8 External Events l A large scale meteorological disturbance would certainly interrupt reactor ' i ! operations. The reactor would be shutdown and placed in a safe condition, if an !
- unforeseen event occurred the construction of the reactor facility would ensure the safety
' i of personnel and the reactor would be shutdown in a controlled manner. l l In the event of a large seismic event, the earthquake switch would activate and ! shutdown the reactor. l The walls of the facility at ground level are constructed of 3.5 feet of steel l reinforced Type 11 concrete. The rest of the structure in the reactor facility around the reactor room and accelerator laboratory is not, if any large impact were to occur the weaker structure would bear the bumt of the damage. If a large explosion were to occur the construction of the facility would mitigate most of the damage. O
(J' : 13.1.9 Mishandline of Malfunction of Equipment. Operator error is a very important consideration during reactor operation. Since the design of the AGN-201M reactor system with all the scrams and associated interlock the likely result of operator error would be a scram. The goal of proper operations is not to let an engineered system scram three reactor because the operator has made a mistake. The most dangerous error could occur if the operator exceeded 5 watts, on a low ,
'l period. A scram would occur at 95% of scale on Channel No. 3 so no core damage would i
result. Since the only 5 watt operations of the AGN-201M reactor occur during reactor i maintenance procedures which require controlled reactor operation this event is unlikely to occur. If reactor period were to go to less than 5 seconds, the reactor interlock system O would scram at 95% scale on Channel No. 3, unless the operator changed scales but to keep pace but eventually the reactor would scram would trip on the 0-10 watt scale. When students operate the reactor, the operation of the Channel No. 3 scale selector switch and how to correctly read the picoammeter seem to be one of the most l challenging aspects of reactor operations. Some students have inadvertently scrammed the reactor due to misoperation of the scale selector switch. i If an equipment malfunction were to occur, such as failure of a nuclear instrument the reactor would scram on the interlock relay or the scram system.. l O
(J5 13.2 Accident Analysis and Determination of Conscouences A d (1) In the case of the MilA, initial reactor conditions would be, reactor critical, operatingjust below 5 watts, shield tank temperature at 21 C. Glory hole empty J (2) The cause would be an instantaneous addition of 2% reactivity to the center of the reactor in the glory hole (3) Reactor would scram on high power,95% of 10 watt scale, Channels No.1 & 2 would be off scale high, all the scram lights would be on but with the release of so much energy in such a short amount of time it is possible the core thermal fuse would melt. If this happens the core will separate adding a negative 5% reactivity l
- to the core which will control the reactivity addition (4) It has been should that the reactor fuel would not be damaged but since the core I separated some major repairs would be required.
A t V; (5) With the presence of the core tank to trap any of the gases released from the fuel and the separation of the core no releases to the environment would occur. The radiation levels in the reactor room would be very high, but will decrease rapidly. 13h Summary and Conclusions Many studies have demonstrated with USNRC approval that the MilA for the AGN-20lM reactor will not melt the polyethylene which surround the fuel. Consequently an insignificant amount of fission products would be released and the direct radiation would only be a small fraction of the 10 CFR 20 limits. O
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, 14-/ , i I ! 3 , e i Chapter 14 Technical Specifications i 1 . i
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i 1 The Texas A&M University AGN 20lM Technical Specification in their present ' 1 j ' format were approved by the NRC on April 25,1979 as Amendment 12 to NRC License j l l R-23. version submitted with this license renewal is thesame set of Technical i i l' Specification with a few minor revisions. At this time. these revisions are being a I j submitted to the NRC for approval. !
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\S-l Chapter 15 FINANCI AL OUALIFICATION 15.2 Financial Ability to Operate a Non-Power Reactor Texas A&M University for the last 40 years has demonstrated the financial ability j 1
to own and operate an AGN-20lM reactor. The overhead cost of operation of this facility j is small. The reactor supervisor is only supported,1/2 time in this capacity and the other operating cost of the reactor are quite minimal. Based upon Department of Nuclear d Engineering records the operating cost for the AGN-20lM in 1996 was about fifteen l , thousand dollars. ) These cost include salary, repair parts and some new equipment for the facility. In ! ) the next five years it is estimated that this figure will vary between fifteen and twenty thousand dollars. All operating capital associated with the AGN-20lM reactor is j provided by the Department of Nuclear Engineering. 4 The reactor is not used for any commercial work and no plans have been made to change the operating schedule of the reactor. ] 15.3 Financial Ability to Decommission the Facility l Texas A&M University is a state supported educational institution that receives I the majority ofits financial support directly from the state. Texas A&M University has l pledged to apply to the State of Texas for whatever financial resources are necessary to decommission the AGN-201M. Texas A&M University has filled a letter ofintent, dated July 19,1990, with the USNRC concern the decommission funding. O
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i f IV , i Chapter 16 OTHER LICENSE CONSIDERATIONS l The AGN-201 reactor at Texas A&M University is not used for medical research and no future considerations are being made for this type of work. The components that presently made up all the systems associated with the AGN-201
-contain no prior use components from any other reactors since the license was last renewed in Auguest of 1977.
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m._- ___ - _ ... . _ ._ . 4 r7 -/ Chapter 17 DECOMMISSIONING AND POSSESSION ONLY LICENSE AMENDMENTS 4 At the current time Texas A&M University has no plans to decommission to the AGN-201M reactor. When necessary, facility staff will prepare a preliminary decommissioning plan which will be followed by a complete decommissioning plan for the facility. The University responded to the NRC INFORMATION NOTICE NO. 90-16 ( { with a letter to the Document Control Desk stating that since the University is a state ) institution that, at the time a decommissioning decision is made, the proper funds would I be requested from the State of Texas.
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