ML24283A156

Enclosure 5 Idaho State University - Safety Analysis Report (Updated April 2024) - Redacted

ML24283A156
Person / Time
Site:	Idaho State University
Issue date:	02/13/2023
From:	Idaho State University
To:	Office of Nuclear Reactor Regulation
References
Download: ML24283A156 (1)

Text

REDACTED VERSION See ML24137A312 for Full Version

SAFETY ANALYSIS REPORT IDAHO STATE UNIVERSITY AGN-201M RESEARCH REACTOR LICENSE NO. R-110 DOCKET NO. 50-284 UPDATED April 2024

CONTENTS Page No.

1 Introduction 1

2 Location and Site Characteristics 2

2.1 Location and Demography 2

2.2 Meteorology 7

2.2.1 Introduction 7

2.2.2 Temperature 7

2.2.3 Precipitation 9

2.2.4 Wind 10 2.2.5 Other Climate Factors 10 2.2.6 Adverse Weather Effects on the ISU AGN-201 Reactor 11 2.3 Geology and Hydrology 18 2.3.1 General Physiographic Setting 18 2.3.2 Local Geology and Physiography 18 2.3.3 Subsurface Water of Pocatello and the ISU Campus Area 19 2.3.4 Surface Waters 20 2.4 Seismology 20 2.4.1 Introduction 21 2.4.2 Methodology 21 2.4.3 Historical Earthquakes and Earthquake Swarms 21 2.4.4 Surface Faults and Fault Swarms 25 2.4.5 Floating Earthquakes 33 3

LILLIBRIDGE ENGINEERING LABORATORY 35 3.1 General Description 35 3.2 Nuclear Operations Area 38 3.2.1 Reactor laboratory 39 3.2.2 Counting laboratory 40 3.2.3 Subcritical laboratory 41 3.2.4 Reactor observation room, supervisors office 42 4

AGN-201 Reactor 43 4.1 Introduction 43

4.2 AGN-201 Characteristics 46 4.3 Control System Upgrade 59 4.4 Control System 59 4.4.1 Control rods 59 4.4.2 AGN-201 Instrumentation System 63 4.4.3 The Safety Channel 82 4.4.4 The Interlock and Magnet Module 87 5

SAFETY ANALYSIS 92 5.1 General 92 5.2 Reactivity Considerations 93 5.3 Radiation and Shielding 93 5.3.1 Shielding 93 5.3.2 Operational Radiation Levels at Full-Power Operation 94 5.3.3 Radiation damage to the fuel matrix 102 5.3.4 Production and handling of radioisotopes 102 5.4 Production and release of radioactive gases 103 5.4.1 Production of argon-41 103 5.4.2 Release of argon-41 from tank water 106 5.4.3 Production and release of nitrogen-16 106 5.5 Maximum Credible Reactivity Accident 107 5.6 Loss of Water Shield from AGN Tank 116 5.7 Energy Released 117 5.7.1 Operational Containment of Fission Products 117 5.8 Gaseous Radioactive Product Release 119 5.8.1 Water activity 121 5.8.2 Exposure inside the reactor room 121 5.8.3 Exposure outside the building 124 5.9 Emergency Procedures 125 5.10 Safety Devices 126

Fig. 4.4-1 AGN-201 control rod and drive mechanism 62 Fig. 4.4-2 Cross section of reactor showing locations of neutron detectors 64 Fig. 4.4-3 AGN-201 control system 65 Fig. 4.4-4 Right Console Panel 66 Fig. 4.4-5 Center Console Panel 67 Fig. 4.4-6 Left Console Panel 68 Fig. 4.4-7 Block diagram for nuclear safety channel No. 1 69 Fig. 4.4-8 Channel 1 Module 70 Fig. 4.4-9 Block diagram for nuclear safety channel No. 2 73 Fig. 4.4-10 Channel 2 Module 74 Fig. 4.4-11 Period Calibrator 75 Fig. 4.4-12 Block Diagram for nuclear safety channel No. 3 76 Fig. 4.4-13 Channel 3 Integrator Module 77 Fig. 4.4-14 Channel 3 Linear Amp Module 78 Fig. 4.4-15 Simplified circuit diagram of Safety Chassis 81 Fig. 4.4-16 Scram Relay Module 82 Fig. 4.4-17 Block diagram of reactor interlock system 86 Fig. 4.4-18 Interlock & Magnet Module 87 Fig. 5.3-1 External concrete block shielding for AGN-201 reactor. Plan view of shielding.

Elevation A is view facing north. Elevation B is view facing west.

97

Fig. 5.3-2 External concrete block shielding for AGN-201 reactor. Section A-A is a cut-away view facing north. Section B-B is a cut-away view facing west 98 Fig. 5.3-3 External concrete block shielding for AGN-201 reactor. Elevation C is facing south. Top view of shielding 99 Fig. 5.3-4 Radiation levels of the AGN-201 reactor operating at 100 mW 100 Fig. 5.3-5 Radiation levels of the AGN-201 reactor operating at 5 W 101 Fig. 5.5-1 Reactor power during maximun credible accident 101 Fig. 5.5-2 Energy release during maximum credible accident 1012 Fig. 5.5-3 Core temperature rise during maximum credible accident 1013

SAFETY ANALYSIS REPORT LIST OF TABLES Table No.

Title Page No.

Table 2.1-1 Incorporated Cities of more than 1000 population within 75 miles of Pocatello 5

Table 2.2-1 Local and Climatological Data 12 Table 2.2-2 Temperature, Cooling and Heating Days 13 Table 2.2-3 Precipitation 14 Table 2.2-4 Growing Degrees 16 Table 2.2-5 Snowfall 17 Table 2.4-1 Earthquakes Near SE Idaho 1909-1994, with Magnitudes Greater than 4 27 Table 2.4-2 Earthquakes after 1994, with Magnitudes Greater than 5 29 Table 4.2-1 Reactor Characteristics 51 Table 4.2-2 Reactor Control and Safety Specifications, and Trips 55 Table 5.7-1 Activity Contained in the Reactor Core for Various Times after Shutdown 117 Table 5.8-1 Gaseous Fission Products in AGN Fuel at 5-W Operation for 30 days 120

1 1 Introduction This report is a periodic review and update to the Safety Analysis Report (SAR) to the United State Nuclear Regulatory Commission (NRC) by Idaho State University (ISU) at Pocatello, Idaho. The AGN-201 Reactor is a research and training reactor, class 104 license, R-110.

Because of the extensive operating experience (since 1965) of this reactor and several other AGN-201 reactors, the system has a well established database, and hence no research and development activities were required to evaluate the system, from that covered in the previous edition of the SAR, dated January 2003. The reactor is administered and operated by Idaho State University for education, research, and training of students. However, the reactor is deemed a facility that benefits many interested parties, such as the Idaho National Laboratory, within the regional area.

The primary differences between this report and the last report (dated August 2021, are the following:

a) The description of the Channel 2 scrams based on the 2024 License Amendment Request and the drawings for the Lilibridge Engineering Building were updated.

This report contains information on the Pocatello, Idaho location, and site characteristics such as meteorology, geology, seismology, and demography, etc. It also includes a description of the Lillibridge Engineering Laboratory building where the reactor is housed on the ISU campus and a description of the reactor and listing of its characteristics.

The reactor is licensed to be operated at a steady state power level up to 5 watts. The total operational fuel loading provides for a maximum excess reactivity of 0.65% above a delayed critical condition. (The effective delayed neutron fraction for this very small core is 0.745%). The ultimate safety of the AGN-201 reactor design lies in its large negative temperature coefficient, its nuclear instrumentation safety system, and the thermal safety fuse. However, the maximum credible accident analyzed takes no credit for the effective operation of the nuclear instrumentation safety system. Thus, even if a sudden reactivity insertion were made, the reactor power rise would be attenuated and terminated by the negative temperature coefficient, and the reactor operation would be terminated by the melting of the thermal safety fuse which holds the upper and lower halves of the core together.

2 The principal university officer involved responsible for the administration of the AGN-201 reactor license is the Vice President for Research at ISU. Responsibility for the safety and general operation of the reactor has been delegated to the Nuclear Engineering Department in the College of College of Science and Engineering.

2 Location and Site Characteristics Information in this section includes maps, building drawings and data on local population characteristics; meteorology, geology, hydrology and seismology of southeastern Idaho.

2.1 Location and Demography Figure 2.1-1 is a map of southeast Idaho showing county and state demarcations, major rivers, lakes and reservoirs and the principal cities in the region. Pocatello in Bannock County and Idaho Falls in Bonneville County are the largest cities. The main campus of Idaho State University (where the AGN-201 reactor is located) is in Pocatello, and there is a branch campus in Idaho Falls, 50 miles northeast of Pocatello. The U.S. Department of Energy's Idaho National Laboratory is headquartered in Idaho Falls, where many of the research facilities are located, and also has research facilities on a 900 square mile desert site, 30-70 miles east of Idaho Falls (about 40-65 miles north northwest of Pocatello. Hill Air Force Base is about 120 miles south of Pocatello in the vicinity of Salt Lake City, Utah. Idaho State University is located in Pocatello, as shown in Figure 2.1-2.

The major employers of this city of approximately 60,000 people is ON (a semi-conductor chip manufacturing company), a phosphate rock processing company Simplot, the Union Pacific Railroad, and the University. The major north-south highway, Interstate 15, passes to the east of the city and the campus. An intersection of I-15 with Interstate 84 which heads west toward the cities of Twin Falls and Boise is located at the north end of the city. The Pocatello Municipal Airport is about 7 miles northwest of the campus on Interstate 84. The Union Pacific Railroad east-west main line tracks go through the center of the city, about 2 miles west of the campus. The city lies in the Portneuf River valley at a general elevation of 4470 ft bordered on the southwest and northeast by hills of the Bannock mountain range which rises about 3,000-4,000 ft above the valley floor.

Bannock County population information, including distribution by race, is given in Table

4 Table 2.1-1 Incorporated Cities of more than 1000 population within 75 miles of Pocatello City Name Location from Engineering Building Population 2010 Census Population Estimate 2018*

American Falls 30 miles west 4,457 Ammon East side of Idaho Falls 13,816 16,476 Arco 80 miles NW 995 Blackfoot 25 miles N 11,899 11,946 Chubbuck N side of Pocatello 13,922 15,316 Idaho Falls 49 miles NE 56,813 61,635 (3)

Iona 57 miles NE 1,803 Malad City 60 miles S 2,095 Montpelier 95 miles SE 2,597 Pocatello 54,255 56,268 (1)

Preston 75 miles SE 5,204 Rexburg 74 miles NE 25,484 28,687 (2)

Rigby 66 miles NE 3,945 Ririe 72 miles NE 656 Shelley 39 miles NE 4,409 Soda Springs 55 miles E 3,058 Sugar City 75 miles NE 1,514 Ucon 49 miles NE 1,108

6 Fig. 2.1-3 Campus of Idaho State University

7 2.2 Meteorology 2.2.1 Introduction The climate of Pocatello is semiarid and may be described as a middle latitude steppe climate, where temperatures are relatively high in summer but fall below freezing in the winter and precipitation is sparse and characterized by great variability. Pocatello lies in a valley two to five miles wide (Figure 2.2-1), with mountains on either side rising three thousand to four thousand feet above the valley floor. About three miles north of the city center, the valley broadens and merges into the gently rolling topography of the of the Snake River Plain. The official Weather Bureau station is located at the airport on the southern margin of the Snake River Plain approximately seven miles northwest of town.

Because Pocatello is situated in the Portneuf Valley between two spurs of the Bannock Range, weather data obtained at the Weather Bureau may not reflect the actual weather conditions in the city itself.

Records from a second-order weather station (one at which only temperature and rainfall data were recorded) which existed for about 8 years in the northeastern section of the city indicated that temperatures in Pocatello are two to three degrees higher than those at the official weather station except in winter when they may be more than 10 degrees higher. Except in summer, storms bring rain or snow to the whole area, but the location of Pocatello near to the mountains causes some variation in the amount and distribution of precipitation received at the Weather Bureau station. In summer, precipitation is usually produced by thunderstorms which are extremely localized. Average annual precipitation at the weather Bureau Station is 13 inches.

2.2.2 Temperature The average annual temperature of Pocatello is 47.2°F and monthly temperatures average from 24°F in January to 71.4°F in July. Temperatures in July may reach 100°F or more for short periods, while January temperatures of -30°F have been recorded. The daily range of temperature is also high, reaching 30°F to 40°F in July. This high diurnal range in summer is due to the fact that while daytime temperatures may reach 90°F or more, excessive re-radiation at night under cloudless skies causes the temperature to drop considerably.

8 In winter, moist maritime air masses from the Pacific bring periods of mild weather when winds blow from the southwest, but otherwise temperatures may stay below freezing for several days and the daily minima fall below zero for approximately 4 months. Frost depth to two or three feet is common during the winter season.

Fig. 2.2-1 Physiographic Location In the spring, temperatures gradually rise but freezing temperatures at night are general through most of April. The average first occurrence of 32 degrees Fahrenheit in the fall is September 20 and the average last occurrence in the spring is May 20. The first cold wave may appear during late November but usually not until late December.

9 2.2.3 Precipitation Average annual precipitation for Pocatello as recorded at the U.S. Weather Bureau is 12.32 inches but the precipitation varies greatly in both amount and distribution. In the thirty year period between 1964 and 1993, seven years had less than 10 inches of precipitation, the minimum for any year being 5.34 inches in 1966. Six out of thirty years had precipitation over 14 inches, the maximum for one year being 20.33 inches in 1983.

During the winter, precipitation falling as snow sometimes accumulates to a depth of a foot or more, but snow depth on the valley floor (elevation about 4470 feet) reaches only 5 or 6 inches at most, and the snow usually melts in a weeks time, during a thaw.

The mountains surrounding Pocatello receive more moisture than the town itself and are covered with snow at higher elevations from November to May. Dry land wheat is raised on the hillsides near Pocatello where slopes are not too steep to prevent cultivation but agriculture can be carried on in or near the valley only by irrigation.

Precipitation is distributed unevenly through the year with 82% of the annual precipitation falling during the period of October through June and only 18% during the months of July, August and September. The fact that precipitation minima occur during the season of high temperatures when evaporation rates are also high, results in dry summers. During the summer, precipitation usually falls as local showers accompanied by light to-moderate thunderstorms, and occasionally by hail. Damage by cloudbursts is rare in Pocatello because contour furrowing done by the Civilian Conservation Corps in the 1930's on the hill slopes above the city has prevented excessive runoff. Cloudy and unsettled weather prevails throughout the winter and spring with measurable amounts of precipitation on about one-third of the days.

10 2.2.4 Wind Pocatello lies in the belt of westerlies, consequently, the prevailing wind direction is from the southwest. Average wind speeds are 9 to 10 miles per hour but on rare occasions, during heavy winds, gusts up to 68 miles per hour have been recorded. Winds of 20 to 30 miles per hour may blow continuously for several days in the spring.

Windstorms associated with cyclonic systems and cold fronts do some damage to trees each year, often causing temporary disruption of power and communication facilities; only minor damage is done to structures. Storms of this type may occur from October to June, while during the remaining three months of the year, high winds are almost invariably associated with thunderstorms.

No permanent official wind-recording instruments are located in Pocatello or in the valley adjacent to it. Movement of smoke from the stack of chemical plant located on the northern outskirts of the city indicates that, at times, air may be moving up-valley on the western side of the valley and down-valley during the early morning hours along the eastern side. Winds also appear to blow down-valley during the early morning hours along the eastern side of the valley. These mountain and valley winds are light; velocities are probably not more than 1-3 miles per hour.

2.2.5 Other Climate Factors Relative Humidity -Relative humidity is higher in winter and spring and during these seasons is near 70 to 80 per cent. During the summer months, relative humidity is never greater than 50 per cent during the day, not exceeding 65 per cent at night.

Fog -The Weather Bureau records an average of 10 days of heavy fog per year for Pocatello, and nearly half of these (four) come in January. The valley and mountain winds tend to prevent the formation of fog in the valley and so occurrences of fog in the city itself averages considerably less than 10 days per year.

Sunshine -Sunny skies prevail over Pocatello during the summer when roughly 50 per cent of the days in July, August, and September are cloudless, and possible sunshine rises to 80 per cent. Cloudiness increases in winter and spring. December and January

11 are the cloudiest months, when the sky is more than eight-tenths covered about two-thirds of the time.

2.2.6 Adverse Weather Effects on the ISU AGN-201 Reactor 2.2.6.1 Flood In the very unlikely event of a flood, no special precautions are necessary other than those normally taken in the event of a flood at an industrial site. The reactor will be secured and not operated at this time. The radiological hazard problems are not severe as the reactor is built to withstand a minor flood (one foot of water). In the event of a major flood where the reactor might be overturned or carried away, there is again no serious problem since the self-contained reactor has been designed to withstand such an emergency.

2.2.6.2 Storm It is highly unlikely that a storm could damage the AGN-201 reactor; however, in the event of a severe storm, the reactor will be shut down and secured. It should also be noted that there is no recorded history of tornadoes in Pocatello, Idaho.

Local climatological and meteorological data from the Pocatello Weather Bureau station is given in Tables 2.2-1 through 2.2-5.

12 Table 2.2-1 Local and Climatological Data

18 2.3 Geology and Hydrology 2.3.1 General Physiographic Setting Pocatello is located on the boundary between the northeastern corner of the Great Basin Section of the Basin and Range Physiographic Province and the southeastern edge of the Snake River Plain Section of the Columbia River Plateau Province (as shown in Figure 2.2-1). The area thus has stratigraphic, structural, petrologic and geomorphic characteristics of both areas. The Basin and Range Province, particularly in the Great Basin Section, is characterized by alternating basins and mountain ranges; the basins commonly are partially graded up onto the mountain sides by Bogota (filled) and pediment (cut) surfaces. The Snake River Plain is an arcuate, flat-surfaces basalt plateau, dissected by the Snake River and to a lesser extent by some of its tributaries on the western end. Both physiographic areas are relatively young geologically. Seismic and volcanic activity during "Recent" time and the freshness of tectonic forms and surficial extrusive rocks show that the area is still in the process of evolving.

2.3.2 Local Geology and Physiography The city of Pocatello is in the valley bottom of the Portneuf River, a tributary to the Snake River. The Portneuf drains about 1200 square miles of the Bannock and Portneuf mountain ranges east and southeast of Pocatello. The city is bordered on the southwest and northeast sides by hills generally less than 4000 feet high, with the general terrain becoming more mountainous to the southeast. The river valley is flat floored and at the townsite widens abruptly to the northwest out onto the Snake River Plain.

The bedrock floor of the valley is buried under some 200 feet of alluvium; whether the original shape and depth of the valley was primarily structural (down-faulted) or erosional is unknown. A much larger river (Bear River) probably occupied the valley prior to the diversion of that river to the south 30,000 years or so ago, possibly accounting for the size of the valley relative to that of the current river. Regardless of Its original configuration, the valley has been altered by several episodes of cutting and filling since its inception. Broad benches slope 300-400 feet per mile toward the valley from each side; these are remnants of a Pleistocene valley fill. Much of this sequence has been removed down the axis of the valley, and the surfaces are currently being dissected by small tributaries to the Portneuf. A basalt flow about the same age as the diversion of the

19 Bear River (about 30,000 years) floors part of the valley upstream from the main part of the city.

Idaho State University buildings are located on terraces 40-60 feet above the valley bottom on the northeast side of the valley. They are built on a combination of the valley-fill alluvium and loess (wind-blown silt deposits), both of which are common in the area.

The southeastern corner of the campus is bordered by a bedrock hill composed of Cambrian (Brigham) Quartzite; the block has been elevated along a high-angle fault which may extend to the northwest under part of the campus or which may terminate at the end of the quartzite block. If the fault does extend under the alluvium beneath the campus, the fault is apparently entirely pre-alluvium in age, as no displacement in the alluvium along strike with the fault has been noted.

2.3.3 Subsurface Water of Pocatello and the ISU Campus Area Gravel lenses and beds near the center of the Portneuf Valley supply water for Pocatello. The non-artesian production is from about 30 to 40 feet in depth for a number of these supply wells in the lower part of the valley, at an elevation of about 4400 ft. The campus is on the eastern slope region, with the campus elevation ranging from about 4450 ft to 4700 ft. The Lillibridge Engineering Building entry doors are at an elevation of about 4500 ft.

Those individuals living on the eastern mountainous slope outside of city limits draw water from deep supply wells, with the producing aquifers typically in the range of 4400 ft or lower. About 1978 the University drilled two wells on campus at extreme ends of the large plot housing the Eli Obler Library (about 150 ft from the Lillibridge Engineering Building). These wells were used to supply heat and cooling through a water source heat pump system, drawing water from one well in the winter and discharging it to the other well. The aquifer depth for water extraction or disposal was in the range of 150 to 200 ft. For cooling in the summer, the direction of flow was reversed, from what was the slightly cooler aquifer area into the slightly warmer aquifer. This system served the library well, until it was abandoned about in the year 2005 because the grid screens through which the water was passed to or from the well to the aquifer become blocked with calcification. (The highly efficient water source heat pump system was replaced

20 with the steam district heating system for the winters, and air-to-air heat pumps for the summer.)

There is a producing well about 0.6 miles south of campus, at an elevation of about 4450 ft., that produced water from about an elevation of 4410 ft., 90 feet below the level of campus. There is also a well about 5 miles north of the campus on the western side of I-15 at an elevation of about 4900 ft that was drilled to a depth of 1100 ft in 1980, and produced 108 F artesian water. (The artesian effect is believed to be due to the temperature/density effect.) If that geothermal aquifer extends to campus, it would be at an elevation of 3800 ft, or at a depth of nominally 700 ft below the campus region where the reactor is sited. Considerations of the elevation (or depth) of the water tables, the elevation of the campus, and the character of the subsurface suggest that no ground water problems will be encountered in the University area 2.3.4 Surface Waters During the early spring of 1962 and 1963, flooding occurred in the low areas of the valley along the river to elevations of about 4460 feet. This flooding was due to unseasonably warm weather and runoff of meltwater into the river over a frozen subsurface in the mountains southeast of Pocatello. A U.S. Corps of Engineers project to straighten and line the banks of the river with concrete has eliminated flooding; even without this correction there would be little chance of flooding in the future to anywhere near the height of the campus. Local drainage is generally away from the Lillibridge Engineering Laboratory so that no surface drainage problems associated with surface runoff are likely.

2.4 Seismology Contributions made:

By James E. Zollweg, Northwest Geophysics, Aug 16, 1995, by Dr. James Mahar, Dept. of Civil and Environmental Engineering, ISU, July 1, 2021, by Dr. Mustafa Mashal, Dept. of Civil and Environmental Engineering, ISU, June 28 2021

21 2.4.1 Introduction Seismic safety is an important engineering consideration in southeastern Idaho because of the proximity of active, late Quaternary/Holocene faults. Careful analysis of potential maximum grown motions caused by earthquakes is required in engineering design and reactor safety. This analysis relies heavily on current expert opinions regarding seismic safety issues.

Pocatello lies along the boundary of two major geologic provinces. The Eastern Snake River Plain (ESRP) is the generally flat plain north of the city. The highland area around the city lies within the Basin and Range province. The Basin and Range province is a region of generally moderate to high seismicity as there are numerous late Cenozoic faults, one of which (Wasatch Fault) is within 30 miles of Pocatello. The Idaho Geologic Survey (IGS 2016) has identified two seismic zones located east and north of the Pocatello area: the north-south Intermountain Seismic Belt and the east-west Central Idaho (Centennial) Seismic Zone (see Figure 2.4-1). Both zones contain active faults.

2.4.2 Methodology A three-step methodology was used in the assessment of seismic risk at Pocatello. The first of these is consideration of the distribution of historical seismicity. The second is consideration of the proximity of potentially active surface faults. The third element, utilized if the first two steps do not indicate a larger magnitude event controls seismic risk, is consideration of a "floating" earthquake that can occur essentially anywhere within the Pocatello region.

2.4.3 Historical Earthquakes and Earthquake Swarms Intermountain Seismic Belt:

The Intermountain Seismic Belt is located along the border between Idaho and Wyoming and extends southward into the northern Utah and northward into western Montana (see Figure 2.4-1). The Idaho Geological Survey (2016) reports that the region has experienced numerous earthquakes a few of which had magnitudes of 6.0 and up to 7.5 since 1884. Two earthquakes of note are the Bear Lake/Paris and Hansel Valley

22 earthquakes south of the campus. The Bear Lake Valley/Paris earthquake (magnitude 6.0) in 1884 caused damage and shaking in the area.

In 1934, the Hansel Valley earthquake (magnitude 6.6) located approximately 20 miles south of the Idaho border produced a 1.6 ft maximum ground surface offset in the 7.1 mile long fault (Hecker 1993). The quake caused structural damage in both southern Idaho and northern Utah. Associate Press (AP) newspaper accounts describe the following 12 March 1934 damage in Pocatello:

Cracked chimneys in Emerson, Lincoln and Jefferson elementary schools Cracked walls in several schools and other buildings Loosened beams in the Reed Hall gymnasium balcony Large cracks in the administration, auditorium and library on the ISU campus And broken residential windows The most severe recorded earthquake in the Intermountain Belt occurred in 1959 with an epicenter located approximately 6.5 miles northwest of West Yellowstone, Montana. The Hebgen Lake of Yellowstone earthquake registered 7.3 on the Richter scale and caused a massive landslide in the Madison River canyon. Several aftershocks ranging in magnitude 5.8 to 6.3 were recorded after the main earthquake event. There are no reports of the earthquake damage in the Pocatello area.

Central Idaho (Centennial) Seismic Zone:

The 1983 Borah Peak earthquake is the largest recorded seismic event in the Central Idaho (Centennial) Seismic Zone. The earthquake registered 7.3 on the Richter scale and occurred along the Lost River Fault. The quake caused severe damage in towns such as Mackay along the Lost River and in Challis, Idaho close to the epicenter. Even though the earthquake was felt in Pocatello, there are no reports of damage in the area.

Earthquake motions occur in two basic distributions: individual events with aftershocks and swarms. Individual events tend to be infrequent with long recurrence intervals but produce the greatest ground motions because much of the energy is released at one or a limited number of events. In contrast, earthquakes in swarms tend to have lower magnitudes but are much more frequent and occur over longer periods of time. Thus, individual events such as the Hansel and Borah Peak earthquakes release much more energy and cause serious damage, whereas earthquake swarms such east of Soda Spring, Idaho in 2017 release the energy over a series of smaller fault displacements.

23 Based on the available technical literature, the greatest threat for seismic damage in the Pocatello area is from earthquakes along the Wasatch Fault south of city. Based on geologic studies, the IGS opines that magnitude 7.0 earthquakes occur along the Wasatch Fault every 300 to 400 years. Hecker (1993) predicts the average regional occurrence of large magnitude earthquakes in the Wasatch Front Region during the Holocene period to be on the order of 125 to 300 years or less, although events have been non-uniformly distributed in time.

POCATELLO DAMAGE, March 12 1934-(AP)-Pocatello was severely shaken twice today by two earth tremors which sent citizens, agog with excitement, scurrying into the streets for safety. The shocks occurred at eight-seven and eleven twenty-one a.m.

Following the second vibration, which was more intense here than the first, public schools were dismissed until a thorough inspection could be made. Cracked chimneys were reported at the Emersion, Lincoln and Jefferson schools and at the general hospital by Fire Chief A. B. Canfield who made an immediate inspection. The walls in several schools and other buildings were also cracked. The balcony at Reed Hall gymnasium where Pocatello and McCammon High school were to play for the district basketball title this afternoon, was condemned as unsafe after the second shock loosened beams. The game was to go on, however. Several homes and one business house reported broken windows.

March 12 1934 (AP)-Two sharp earth shocks of about 15 seconds duration jarred Pocatello residents at 8:07 a.m. today. The Pocatello Tribune building and others in the downtown section were shaken by the quake being accompanied by a heavy rumbling sound. One home reported a window broken by the tremors SALT LAKE CITY, March 13 (AP)-Pending a thorough inspection, schools of Salt Lake City will remain closed today but educational institutions will resume normal functions in most other north Utah and south Idaho cities which were rocked by a series of earthquakes yesterday. Damage to buildings in all instances was confined to cracks and toppled chimneys. The most severe disturbance was centered in an area bounded by Boise, Idaho, on the north; Rock Springs, Wyo.,

on the east; Richfield, Utah, on the south and Elko, Nev., on the west. The quake was less severe in Wyoming and Nevada than in Utah and southern Idaho. On the recommendations of W. L.

Payne, Salt Lake City police chief, and Dr. L. E. Viko, city health commissioner, Dr. L. John

24 Nuttal, Jr., superintendent of public instruction, announced that as a precautionary measure all public institutions under his jurisdiction will remain closed until buildings are closely inspected for hidden flaws. Close to Pocatello, 10 public schools were closed following the disturbance. An inspection was started at once and all structures were found safe. The walls of one building were badly cracked. J. R. Nichols, de an of the University of Idaho, southern branch, at Pocatello, called on the state department of public works at Boise to send an expert to examine building ~~

on the campus. Large cracks appeared in several places in the administration building, in the hall which houses the auditorium and, in the library, considerable damage was done and beams were loosened in the gymnasium.

Earthquake Catalog A search was made of available on-line earthquake catalogs to assess the frequency of magnitude 4 or greater events in Eastern Idaho. The search area chosen was 41 to 44.5° north latitude and 110 to 115° west longitude. Seismic monitoring of eastern Idaho is splintered between several organizations and their catalogs are not routinely integrated. Most of the catalogs are not published in any usual sense; they are maintained by their individual organizations as part of routine seismic monitoring efforts.

The level of completeness, accuracy, and ending date differ between the catalogs.

Those catalogs consulted were developed by Woodward-Clyde Federal Services (see Wong et al., 1992 for a map), the University of Utah, the Idaho National Engineering Laboratory, the Montana Bureau of Mines and Geology, Boise State University, and the U.S. Geological Survey. The Woodward-Clyde catalog was derived from the other catalogs and other sources of historic data, but only runs through late 1989. It was the primary data source for events through its completion date. The composite catalog is believed to be essentially complete through mid-1994.

Figure 2.4-1 shows the resulting epicenter map, and Table 2.4-1 gives epicentral data for the events within 130 km (approximately 80 miles) of 42.868° north, 112.435° west (the approximate location of the Pocatello business district and Idaho State University).

Smaller earthquakes have occurred closer to Pocatello, but the distribution of magnitude 4+ events may be a better index to active source areas. No magnitude 4 earthquake has occurred closer than about 60 km (about 35 mi) to Pocatello. The ESRP, which lies to the north of Pocatello, is nearly aseismic at the magnitude 4 level and is not considered to be a likely source of events capable of causing damage in Pocatello. High activity

25 occurs to the south, east, and northeast at greater distances than 35 mi. Many of these earthquakes have been felt in Pocatello, although damaging levels of intensity are not believed to have been achieved historically. The closest magnitude 6+ earthquake was the 1975 Pocatello Valley earthquake (local magnitude 6.0), about 56 mi from Pocatello, and the closest magnitude 7+ earthquake was the 1983 Borah Peak earthquake (surface wave magnitude 7.3), about 105 mi from Pocatello.

It is concluded that the historical earthquake catalog shows Pocatello is in an area of relatively low seismicity, with higher seismicity areas occurring at least 35 mi from the city. This would suggest that Pocatello is at little risk from a nearby source of large earthquakes, but the historic record is short. Geologic evidence discussed in the next section is felt to be of greater importance than the historical seismicity catalog.

2.4.4 Surface Faults and Fault Swarms Faults in the Basin and Range province usually have long recurrence times between large events, typically 1,000 to 100,000 years on the same fault segment. Therefore, the historical catalog is not the only indication of the potential for damaging earthquakes at Pocatello. Figures 2.4-2 and 2.4-3 (taken from Hilt, 1993) show that while mapped late Cenozoic faults occur within a few miles of Pocatello, the nearest fault with probable late Quaternary movement is located about 100 mi southeast of Pocatello (it is the Bear Lake fault). Late Quaternary movement iS known on faults on the north side of the ESRP at about the same distance from Pocatello. Faults with late Cenozoic movement but no evidence for late Quaternary movement are generally considered inactive for seismic hazard evaluation purposes. Most late Quaternary faults have expectable maximum magnitudes between about surface wave magnitude 6.8 and 7.6.

27 Table 2.4-1 Earthquakes Near SE Idaho 1909-1994, with Magnitudes Greater than 4 Year Date and Time (UTC)

Magnitude Lat(N) deg.

Lon(W) deg.

Location from Pocatello 1909 10/06 02:50:00 6.0 41.75 112.65 77.0 mi S of Pocatello 1909 11/17 06:30:00 4.3 41.75 112.15 78.9 mi S of Pocatello 1915 07/30 18:50:00 4.3 41.73 112.15 78.9 mi S of Pocatello 1930 06/12 09:15:00 5.8 42.60 111.00 75.4 mi ESE of Pocatello 1934 03/12 15:05:48 6.6 41.76 112.66 76.8 mi S of Pocatello 1934 04/14 21:26:32 5.4 41.71 112.60 79.0 mi S of Pocatello 1934 05/06 08:09:42 5.5 41.95 112.81 65.7 mi SSW of Pocatello 1946 08/07 02:30:00 4.3 41.71 112.11 80.1 mi SSE of Pocatello 1960 08/30 16:27:16 4.9 42.40 111.50 57.6 mi SE of Pocatello 1962 09/05 13:35:24 5.7 41.91 111.61 77.3 mi SSE of Pocatello 1962 09/19 03:00:00 4.3 42.00 111.70 70.8 mi SSE of Pocatello 1969 09/19 09:31:45 4.4 43.05 111.41 53.2 mi ENE of Pocatello 1969 09/19 13:33:15 4.9 42.98 111.41 51.7 mi E of Pocatello 1969 09/19 19:57:18 4.4 43.00 111.26 59.9 mi E of Pocatello 1969 09/19 23:58:06 4.1 42.95 111.48 48.4 mi E of Pocatello 1973 04/14 06:45:46 4.4 42.03 112.61 57.9 mi S of Pocatello 1975 03/27 04:48:51 4.2 42.05 112.53 55.7 mi S of Pocatello 1975 03/28 02:31:05 6.0 42.05 112.51 55.8 mi S of Pocatello 1975 03/29 13:01:19 4.7 42.01 112.51 57.8 mi S of Pocatello 1975 03/30 6:56:28 4.1 42.01 112.56 58.1 mi S of Pocatello 1976 11/05 02:48:55 4.0 41.80 112.68 74.4 mi S of Pocatello 1978 10/24 20:30:59 4.1 42.55 111.83 37.6 mi SE of Pocatello 1978 11/30 06:53:40 4.7 42.10 112.50 52.6 mi S of Pocatello 1981 12/09 08:15:04 4.1 42.63 111.41 53.5 mi ESE of Pocatello Table 2.4-1 (contd)

28 1982 05/30 11:06:42 4.0 42.68 111.23 62.2 mi E of Pocatello 1982 10/14 04:10:23 4.7 42.60 111.41 54.9 mi ESE of Pocatello 1982 10/14 11:09:29 4.1 42.60 111.43 53.7 mi ESE of Pocatello 1983 02/08 10:54:54 4.2 43.30 111.16 70.7 mi ENE of Pocatello 1985 07/02 03:03:56 4.0 43.25 111.15 70.2 mi ENE of Pocatello 1987 03/18 00:00:42 4.1 42.61 111.31 59.2 mi ESE of Pocatello 1988 11/13 11:53:24 4.4 42.61 111.91 78.4 mi ESE of Pocatello 1988 11/19 20:00:53 4.3 42.00 111.46 77.0 mi SE of Pocatello 1992 11/10 10:46:18 4.8 43.08 111.61 43.6 mi ENE of Pocatello 1992 11/10 10:54:50 4.9 43.00 111.45 50.3 mi ENE of Pocatello 1992 11/11 12:08:07 4.0 43.01 111.48 49.4 mi ENE of Pocatello 1994 02/02 11:04:25 4.0 42.75 111.06 69.9 mi E of Pocatello 1994 02/03 07:14:51 4.5 42.75 111.03 71.0 mi E of Pocatello 1994 02/03 09:05:03 5.8 42.75 110.96 74.3 mi E of Pocatello 1994 02/03 09:47:36 4.0 42.71 111.03 71.7 mi E of Pocatello 1994 02/03 09:58:40 4.2 42.75 111.03 70.9 mi E of Pocatello 1994 02/03 10:25:51 4.0 42.78 111.11 66.9 mi E of Pocatello 1994 02/03 11:19:07 4.7 42.76 111.00 72.6 mi E of Pocatello 1994 02/03 11:46:50 4.0 42.78 111.15 64.9 mi E of Pocatello 1994 02/03 12:04:57 4.4 42.71 111.08 68.7 mi E of Pocatello 1994 02/03 02:42:12 5.2 42.70 111.03 71.8 mi E of Pocatello 1994 02/04 03:10:08 4.0 42.83 111.08 68.1 mi E of Pocatello 1994 02/04 21:49:12 4.0 42.61 111.05 71.7 mi ESE of Pocatello 1994 02/07 06:35:47 4.8 42.65 111.03 72.7 mi E of Pocatello 1994 02/07 12:15:45 4.5 42.66 111.01 73.3 mi E of Pocatello 1994 02/10 00:56:11 4.3 42.88 111.06 69.3 mi E of Pocatello

29 Table 2.4-1 (contd) 1994 02/11 04:24:29 4.0 42.81 111.11 66.3 mi E of Pocatello 1994 02/11 14:59:50 5.3 42.76 111.00 72.8 mi E of Pocatello 1994 02/14 16:55:34 4.0 42.80 111.01 71.5 mi E of Pocatello 1994 03/03 07:13:17 4.1 42.78 111.05 69.8 mi E of Pocatello 1994 04/07 16:16:44 4.8 42.53 111.01 75.5 mi ESE of Pocatello 1994 04/08 07:26:21 4.1 42.60 111.08 71.1 mi ESE of Pocatello 1994 04/10 20:04:09 4.6 42.65 111.11 68.4 mi ESE of Pocatello Table 2.4-2 Earthquakes after 1994, with Magnitudes Greater than 5 Year Depth Magnitude Lat(N) deg.

Lon(W) deg.

Location from Pocatello 1995 1

5.3 41.5 109.6 130 mi ESE of Pocatello 1999 16 5.1 44.8 112.8 160 mi N of Pocatello 2001 5.3 42.9 111.4 65 mi E of Pocatello 2004 3

5.0 43.6 110.8 100 mi ENE of Pocatello 2005 12 5.6 43.4 112.6 145 mi N of Pocatello 2008 7

6.2 41.1 114.9 153 mi SW of Pocatello 2008 5.1 41.1 114.8 150 mi SW of Pocatello 2015 8

5.0 44.5 114.1 150 mi NW of Pocatello 2017 12 5.8 46.9 112.6 250 mi N of Pocatello 2017 6

5.3 42.6 111.4 50 mi E of Pocatello 2020 5.7 40.8 112.0 165 mi S of Pocatello 2020 6.6 44.2 115.2 160 mi NW of Pocatello

30 Anders et al. (1989) believe that the location of active segments of faults around the ESRP is a function of distance from the ESRP axis, and propose a physical model to explain an increase in strain rates with distance from the ESRP. The Anders et al. model would suggest that faults in the immediate vicinity of Pocatello have very low strain rates, and consequently long recurrence times (in excess of 10,000-100,000 years) are probably to be expected. If the Anders et al. model is correct, there is little risk from a large earthquake occurring in the immediate vicinity of Pocatello. Major earthquakes (magnitude 6.8+) are likely to occur no closer to Pocatello than the regions of observed high seismicity, and the Anders et al. model suggests that the more active faults would be located even farther from the city. It is therefore concluded that the known late Quaternary faults are not greater sources of seismic risk to Pocatello than possible blind faults existing near the city, as will be discussed in the next section.

31 Fig. 2.4-2 Faults with late Quaternary motion in Idaho

32 Fig. 2.4-3 Late Cenozoic Faulting in Idaho

33 2.4.5 Floating Earthquakes Blind faults (those not recognized at the earth's surface) exist in southeast Idaho, as is proven by the 1994 local magnitude 5.8 Draney Peak earthquake and the 1975 local magnitude 6.0 Pocatello Valley earthquake. These earthquakes did not occur on known faults. To account for such events, seismologists have used the concept of a "floating earthquake" which is customarily chosen as being 1/2 magnitude unit larger than the largest historical event that is not associated with a known fault. The largest such events are the Draney Peak and the Pocatello Valley earthquakes. Thus, a reasonable estimate of the magnitude of the floating earthquake is 6.5. The floating earthquake is customarily placed at a distance of 25 km from the site in question. Since Pocatello lies at the edge of the Basin and Range province in which the Draney Peak and Pocatello Valley earthquakes occur, the possibility of blind faults near Pocatello cannot be ruled out.

Therefore, a magnitude 6.5 earthquake at a distance of 25 km is chosen to be the event controlling seismic hazard at Pocatello. The choice of this event is a state-of-the-art assessment and may change in the future as a better understanding of southeastern Idaho seismotectonics develops.

Design acceleration for floating earthquake. The horizontal peak ground acceleration for a floating earthquake of magnitude 6.5 at a distance of 25 km from Pocatello can be calculated from relationships presented by Joyner and Boore (1981). Because of the lack of significant historical seismicity within 35 mi of Pocatello, it is felt that use of the Joyner and Boore (1981) 50th percentile formula is sufficiently conservative; a higher percentile formula would be justified if significant seismicity were if located near the city.

The Joyner and Boore (1981) relationship is:

log A = -1.02 + 0.249 Mw -log r - 0.00255 r where A is horizontal peak ground acceleration in g, Mw is earthquake moment magnitude (roughly equivalent to local magnitude at magnitude 6.5), and r is a distance term defined as:

r = (d2 + 7.32)0.5 where d is the closest approach of the seismogenic structure in km. For the floating earthquake, d is 25 km. It should be noted that the Joyner and Boore (1981) curve was developed mainly from California area data and application to southeast Idaho can be expected to be only approximate.

34 The resulting horizontal peak ground acceleration is 0.13 g. Such an acceleration indicates that minor structural damage in poorly-built facilities is about the most that can be expected in the Pocatello area. Since such an event has not occurred historically, this seems to be a conservative estimate suitable for most uses.

Reference:

Anders, M. H., J. W. Geissman, L. A. Piety, and J. T. Sullivan (1989), Parabolic distribution of circumeastern Snake River Plain seismicity and latest Quaternary faulting: Migratory pattern and association of the Yellowstone hotspot, Journal of Geophysical Research 94, 1589-1621.

Hilt, A. P. (1993), Seismic Siting Considerations in Idaho, M. S. Thesis, University of Idaho, Moscow, 136 p.

Joyner, W. B., and D. M. Boore (1981), Peak horizontal acceleration and velocity from strong-motion records including records from the 1979 Imperial Valley, California earth quake, Bulletin of the Seismological Society of America 71, 2011-2038.

Wong, I., K. Coppersmith, W. Silva, R. Youngs, T. Sawyer, M. Hemphill-Haley, C. Stark, P.

Knuepfer, R. Castro, F. Makdisi, D. Wells, S. Chiou, R. Bruhn, and W. Daning (1992),

Earthquake Ground Motion Evaluations for the Proposed New Production Reactor at the Idaho National Engineering Laboratory, Volume I: Deterministic Evaluation, Informal Report EGG-GEO-10304, Woodward-Clyde Consultants, Geomatrix Consultants, and Pacific Engineering and Analysis, Oakland, CA.

48 lead shield. The detectors are connected respectively to a logarithmic picoammeter, a linear picoammeter, and pulse amplifier and count rate meter located in the reactor console. Each indicator is connected through a relay to a scram circuit. Additional safety interlocks provide for reactor shutdown if the level of the shield water drops, if the reactor temperature falls below 15ºC, or if an earthquake occurs. Sequential interlocks are also present to ensure that the proper operational method is followed. Detailed characteristics of the reactor are given in Table 4.2-1.

50

\

Fig. 4.2-3 AGN-201 core tank and contents

54 Table 4.2-1 (contd)

(2) Neutron Flux Average Thermal Flux 1.5 x 108 n/cm2-s at 5 W Peak Thermal Flux 2.5 x 108n/cm2-s at 5 W (3) Reactivity Worth of Reactor Components (a) Safety and Coarse Control Rods 1.25% k/k ($1.68) (each)

(b) Fine Control Rod Fuel-loaded 0.310% k/k ($0.42)

Polyethylene-loaded 0.155% k/k ($0.21)

(c) Standard Core Material in Glory Hole At Core Edge 0.042% k/k gm-1 ($0.06)

At Core Center 0.100% k/k gm-1 ($0.14)

(d) Polyethylene in Glory Hole (completely filled) 0.29% k/k ($0.39)

(e) Access Port Plugs 1 Wood Plug 0.002% k/k ($0.003) 1 Section Pb 0.015% k/k ($0.02) 1 Section Graphite 0.194% k/k ($0.26)

Total Worth of Plugs in one Access Port 0.422% k/k ($0.57)

(f) Temperature Coefficient of Reactivity (Approximate)

-0.035% k/k °C-1 (g) Reactor Sensitivity at Core Center, Measured with 1/v Absorber

-0.14% k/k cm-2 (4) Pertinent Figures (a) Control Rod Reactivity Shape Curve in Figure 4.2-4 (b) Inhour Equation Given in Figure 4.2-5 (c) Flux Plot vs Radius Given in Figure 4.2-6

55 Table 4.2-2 Reactor Control and Safety Specifications, and Trips SAFETY CHANNEL SET POINT FUNCTION Nuclear Safety Channel No. 1 (Startup Count Rate Channel)

Low Power 0.5 counts/second Scram at levels Below the set points Nuclear Safety Channel No. 2 (Log Power Channel)

High Power 6 watts (120% of licensed power)

Scram at power

> 6 watts Reactor Period 5 sec Scram at periods

< 5 sec Nuclear Safety Channel No. 3 (Linear Power Channel)

High Power 6 watts (120% of licensed power)

Scram at power

> 6 watts Nuclear Safety Channel No. 3 (Linear Power Channel)

Low Power 5% Full Scale Scram at levels

< 5% of Full Scale Manual Scram Scram at operator option Area Radiation Monitor

= 10 mR/hr Alarm at or below level set to meet requirements of 10 CFR 20 REACTOR TRIPS Channel No. 1 Trip:

Less than 0.5 counts/second with the source in the reactor at startup Water Tank:

Temperature less than 15° C Seismic Scram:

Indicator Shield Tank:

Water level lower than 10 inches from the top Channel No. 2:

(log n)

Period shorter than 5 seconds Power indication more than 6 watts Channel No. 3:

(linear power level)

Power indication more than 6 watts.

Linear rotating switch indicator less than 5% or more than 95% of full scale

56 Fig. 4.2-4 Fine Control Rod Calibration Curve

59 4.3 Control System Upgrade The major modification made to the reactor and completed in 2020 was to install a newly completed control system that was built using all new components, designed similarly to the original control system using analog logic. The new system was installed in an identical frame to the old console, the frame having been obtained from another university, which had decommissioned its AGN-201 reactor.

The primary modification on the new console was to replace the vacuum tubes, that were in the old console with transistors (or equivalent).

To avoid changes to the main licensing document, the Technical Specifications, it was decided to duplicate all of the current circuitry with equivalent solid-state circuitry. This included replacing the 1950-era relays with modern solid-state relays, in addition to replacing the vacuum tubes.

The new console design changes were vetted with the 50.59 review process in house, and any potential design changes that would affect the 50.59 safety decisions were re-examined and revised accordingly.

It was decided to replace the outdated digital encoders feeding into digital displays for rod position indication, thus replacing the gear system readout for the fine and course control rods. The 50.59 evaluation showed that the electronic light position indicators would be at least as reliable as the original system and would make manual time measurement of steady state period increase more precise. Furthermore, the design basis accident analyzed in this report was unaffected by the operator knowing the rod position indication.

4.4 Control System 4.4.1 Control rods The AGN-201 reactor has two safety and two adjustable control rods. Three of these, the two safety rods and the coarse control rod, are identical in design although their

64 Fig. 4.4-2 Cross section of reactor showing locations of neutron detectors

83 INTRODUCTION The AGN201 requires very specific conditions to be met for start-up and operation. If these conditions are not met, the reactor will stay in a scram condition until all operating parameters are within their normal ranges. Normal operation of the reactor to produce power is to have all external interlocks satisfied, all module scams cleared, a specified source rod inserted, the MAGNETS ON switch depressed and rods sequentially inserted into the reactor vessel by four motorized mechanical drives. The four mechanical drives are commanded from switches on the center console control desktop panel for both travel direction and speed. Three of the four drives incorporate electromagnets that when energized, will maintain a magnetic force that will cause the fuel canisters to rise with the drive carriages. Safety Rod 1 (SR1) is raised to its locked position and when it reaches that position it stops and enables Safety Rod 2 (SR2) to be raised. SR2 is driven to locked position and when it reached that position it stops and enables the Course Control Rod (CCR) to be driven up to its position. Finally, the Fine Control Rod (FCR) can be driven up. Reactor power is then controlled by the movement of the course control rod and the fine control rod by driving those two rods up and down. Since SR1 and SR2 are locked in their inserted position they can no longer be moved down except by a scram signal that will cause the magnets to release the fuel canisters and also drive all four carriages to their initial positions out of the reactor.

A scram causes the immediate interruption of magnet holding currents of the two safety control rods and the coarse control rod which will eject the fuel cannisters downward out of the reactor vessel to their initial position. The ejection rate is supplemented by the stored energy of compressed springs in the drive mechanisms during rod insertion.

Dashpots are incorporated to soften the landing of the canisters once ejected. A scram also initiates an Automatic Carriage Return (ACR), condition where the drives are commanded to be driven out of the vessel and returned to their initial position terminating all reactor operations. Depressing the large red SCRAM button on the consoles center control desktop immediately terminates reactor operations.

85 tripped. Contacts 5 and 6 of relays K1, K3, K5, K7, K9 and K11 are series connected, driven by the +12 Vdc NIM bin power supply, and terminate on the coil of the Scram String Relay K13. Relay K13 is energized when all scram conditions have been cleared and their associated trip relays (K1 thru K12) reset. The normally open contacts (5 and

6) of K13 are routed through connector SR2 (pins1&2), to be connected in series with the reactor scram string.

88 (1). The shield water level switch, which consists of a water-tight microswitch and an actuator connected to a float bob, opens the interlock if the shielding water level is less than the minimum allowed level.

(2). The low-temperature switch, which has been calibrated to open at 15o C, is a simple bimetal thermal switch. As the temperature of the switch reaches this 15o C level, the bimetal strip which makes up one side of the switch bends, breaking the interlock circuit.

Bending action takes place because the two different metals used in the strip have different linear coefficients of thermal expansion.

(3). The earthquake switch consists of a steel ball mounted precariously on two terminal strips to maintain electrical continuity. If the reactor receives a physical shock resulting in a lateral displacement, the ball will move and break the electrical contact being made.

Any event of the above three sensors that generate an open in the interlock chain, will require corrective measures by the operator.

An open interlock lights a panel indicator on the consoles left front panel and tells the operator which fault has happened. There are two other indicators: Interlock Open and Interlock OK. Any fault in the interlock string that is not made will cause the Interlock Open indicator to light. When all interlocks are made, the Interlock OK indicator is on.

The Interlock module controls the merger of those signals with the SCRAM module signals, (CHANNEL 1 count rate Lo trip, CHANNEL 2 Hi Level and Period trip, and CHANNEL 3 Hi & Lo Level trip), the ROD INTERLOCK switch on the reactor skirt, the MODE switch on the reactor consoles right hand front panel, and the Reactor Console SCRAM button. The resulting Scram C signal, generated within the INTERLOCK module, controls the application of power to the electromagnets.

When the Interlock OK indicator is on, the Mode switch is in the Operate position with the Operate indicator on, the SCRAM button is up and all module trips are reset and indicators off, +12V is applied to the Magnets On toggle pushbutton switch which energizes the Magnet On relay that sends power to the three control rod magnets which allow the fuel canisters to rise with the rod drives. Only the Magnet On relay can energize the electromagnets. Any broken interlock or any scram event will cause the

92 5 SAFETY ANALYSIS Information in this section includes an analysis of the maximum credible reactivity accident, as well as consideration of radioactive fission product gases, shielding and radioactive effluents, etc.

5.1 General The following section describes the design features of the ISU AGN-201 reactor and Lillibridge Engineering Laboratory building which ensure that the reactor can be operated under the specified conditions with no hazard to the health and safety of the operators, other occupants of the LEL, or to the general public. Also considered are the effects of conceivable accidents due to component malfunction, human error, or force majeure.

The reactor is located in a medium-sized building on the ISU campus with an estimated daytime occupancy of between 100 to 200 people during the academic year. Thus, an attempt has been made to eliminate or reduce as many of the normal potential nuclear hazards as possible. Emergency procedures for the AGN facility are given in the facility Emergency Plan.

The primary hazard is the possible over-exposure of personnel to radiation. Such over-exposure may occur in any of all of the following ways: chronic exposure to relatively low radiation levels; acute exposure to high radiation levels from sealed sources; acute exposure to elevated radiation levels as a result of an inadvertent power excursion: and exposure to, and possible inhalation and/or ingestion of, uncontained radioactive fission products. The purpose of this chapter is to define and evaluate these hazards and to discuss the various safety features of the AGN-201 reactor. The hazards set forth here have been documented and evaluated by personnel from this facility and other AGN facilities, still licensed and decommissioned. An NRC evaluation of the hazards associated with operation of the AGN-201 reactor is given in Docket F-32.

96 Therefore, the AGN-201 reactor is adequately shielded for 5-W operation under such an expanded operating schedule.

For persons working in the reactor room, under surveillance, the ISU occupational dose limit2 is the more restrictive of (1) 1,000 mrem/yr (10 mSv/yr) total effective dose equivalent or (2) 10,000 mrem/yr (100 mSv/yr) for the sum of the deep-dose and committed dose equivalent to any individual organ or tissue, excluding the lens of the eye. A student would have to remain at the concrete shield in the reactor room for about 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> per week for 30 weeks each year to receive an equivalent dose of this magnitude from the reactor operating at 5 W during the entire time.

2 Radiation Safety Policy Manual, Rev. 13.1, Radiation Safety Division, Technical Safety Office, Idaho State University, October 2020.

100 Fig. 5.3-4 Radiation levels of the AGN-201 reactor operating at 100 mW

101 Fig. 5.3-5 Radiation levels of the AGN-201 reactor operating at 5 W

102 5.3.3 Radiation damage to the fuel matrix Low-density polyethylene, a polymeric organic material, can sustain radiation damage when exposed to neutron bombardment and the radiation emissions from the decay of fission products. In tests performed by Aerojet-General Nucleonics, more than fifty small samples of core material were exposed in the CP-5 reactor at Argonne National Laboratory in a flux of approximately 1012 n/cm2-s for periods of up to 1 week. An analysis of these samples indicated that radiation damage manifests itself in reduced density and loss of hydrogen from the polyethylene after exposures of approximately 1 week for a fluence of 6 x 1017 n/cm2. By extrapolating these data, on the assumption that the time-integrated flux (fluence or nvt) is responsible for the radiation damage at an average power of 5 W, the core life is approximately 200 years for continuous operation.

It is a reasonable assumption that a lower flux for a correspondingly longer time would result in no more radiation damage than occurred in the high flux tests conducted in the CP-5. Therefore, the core should have an adequate lifetime if exposed to no more than an average continuous power level of 5 W.

5.3.4 Production and handling of radioisotopes Neutron activation in the AGN-201 reactor can produce only very limited quantities of radioisotopes and any induced radioactivity in reactor structures is negligible.

Subsequent handling of radionuclides is supervised by individuals experienced in the detection and evaluation of radiological hazards. The reactor staff has been trained in such procedures and supervises all handling of radioactive materials within the reactor area. Outside this area, the use of radionuclides comes under the control of the ISU Radiation Safety Committee.

The maximum amount of activity which can be produced by one irradiation is given by the product of the specific activity and the mass of material irradiated. There are three primary limitations on the amount of mass which can be used. The most obvious of these is the amount of space available in the reactor for irradiation. In the AGN-201 reactor, the useful volume of the glory hole is approximately 50 cm3. For those cases in which this volume is insufficient, the 10-cm-diameter access ports may be used instead of the glory hole.

103 The second limitation, pertaining to the total mass of material that may be irradiated is the effect of this material on the criticality of the reactor. Since the material must absorb neutrons to become activated, its insertion into or near the core decreases the reactivity of the system. This effect can be compensated by the use of the control rods up to the limit of the CCR and FCR being fully inserted.

The limitation on the amount of material arises from loss of the neutron absorption efficiency because of self-absorption whereby the outer portion of the material shields the inner portion from the neutrons. This effect is present primarily in strongly absorbing materials, such as indium and gold and involves the resonance integral in particular.

The risk of release of radioactivity by breakage of an irradiated sample is reduced by careful design of handling and encapsulating practices and attention to details of irradiation, such as the effect of radiation on the sample. Such criteria are examined closely with respect to experiments. The use of non-porous paintwork in the Reactor Laboratory is an aid in preventing long-term contamination. Emergency decontamination supplies are on hand at all times, as well as contamination survey instruments.

5.4 Production and release of radioactive gases Radioactive argon-41 and nitrogen-16 are produced by neutron reactions with air and water in the vicinity of the core of the reactor. Air may be contained in experimental facilities (glory hole & access ports) and is in solution in the tank water.

5.4.1 Production of argon-41 Experience with AGN reactors operating at higher power has shown that no significant release of Ar-41 (half-life = 1.8 hr) occurs from the glory hole during power operations at 5 W or less. This conclusion was the result of a test at the Naval Post Graduate School where Ar-41 activity was measured by irradiating a sample of air at atmospheric pressure in a closed tubular container just filling the AGN glory hole to the boundaries of

104 the core. The irradiated air was transferred to a chamber counter with thin-walled glass G-M tube. Decay was followed over approximately one half-life and was consistent with the decay of Ar-41. The measured activities agreed with those estimated from a calculated efficiency of the counter.

The next most likely location to produce Ar-41 is below the reactor skirt. On the basis of Naval Post Graduate School operating experience, Ar-41 will not be formed in measurable concentrations under the skirt at operation at 5 W. Since the resulting peak Ar-41 activity for the air volume in a sealed empty glory hole is only 45 times greater than the MPC value for Ar-41, release of the Ar-41 in the glory hole into the reactor room will result in natural diffusion and mixing of this irradiated air volume throughout the room will easily reduce the average air activity in the vicinity of the reactor to less than 1% of MPC values for uncontrolled areas. Also, the reactor area is presently, and will continue to be, a control area with limited access. Thus, no hazard from Ar-41 is anticipated, as shown below.

The maximum equilibrium concentration of Ar-41 produced can be easily calculated. At 5 W, the average thermal neutron flux is 1.75 x 108 n/cm2-s. The saturation reaction rate,, for Ar-41 production is given by

= !

= !

(1) where

= microscopic cross section for 40Ar(n,)41Ar[cm2],

= natural abundance of Ar-40 [dimensionless],

m

= the mass of Ar-40 contained within the volume of the glory hole fully contained within the reactor core [grams],

NA

= Avogadros number [mol-1],

A

= atomic mass of Ar-40 [grams mol-1], and

= average thermal neutron flux [n cm-2 s-1].

The mass of Ar-40 is calculated based on the assumption that the air entrapped within the glory hole is a dry, ideal gas at standard temperature and pressure with argon comprising 1.3% of air by mass. Thus, there are 6.2 mg of Ar-40 contained within the

108 assumptions are made regarding the 2% step increase in reactivity at time zero, negligible energy accumulation at the start of the excursion, and the excursion is modeled as an adiabatic process. With these assumptions, the applicable kinetics equations are then

$& =

'()*(+

+

-/0 (3)

$1!

$& =

+!

--, = 1,2,...,6 (4)

= 2 (5) and

$& =

3 451" (6) where n = space-averaged neutron density in reactor [n/cm3],

= core reactivity[dimensionless],

= temperature coefficient of reactivity [ºC-1],

= neutron generation time [s],

= fraction of delayed neutrons [dimensionless],

i = delayed neutron fraction for the ith group of delayed neutron precursors [dimensionless],

i = decay constant for the ith group of delayed neutron precursors [s-1],

Ci = space-averaged density of the ith group of delayed neutron precursors [cm-3],

P = reactor power [W],

f = macroscopic fission cross section [cm-1],

v = average thermal neutron speed [cm/s],

Y = recoverable energy per fission [J/fission],

V = volume of reactor [cm3],

= density of the core material [gm/cm3],

CP = specific heat at constant pressure [J/gm-ºC], and T = core temperature [ºC].

110 rises to very high values, passes through a maximum of about 170 MW and then rapidly decreases to a power level of about 800 kW which then appears to slowly decay, all within a time window of about 300 milliseconds.

The height of the maxima is observed to be essentially independent of the initial power level, but as might be expected, the time at which the maxima occur increases with decreasing initial power. Following the excursion, the power decreases slowly until after several minutes the thermal power output attains a steady level of about 200 watts in which the reactivity addition is balanced by the compensating reactivity at the consequent elevated core temperature as a result of the large negative temperature coefficient. However, it is reasonable to assume that the thermal fuse has functioned as designed, thereby separating the core so that this equilibrium power is not maintained. The excursion simulation shown in Figures 5.5-1, 5.5-2, and 5.5-3 does not model core separation which

111

112

113

114 would occur when the temperature of the thermal fuse located near the center of the core exceeds about 100ºC, as discussed in Section 4.2.

The final power level is obtained by modifying Eq. (4) to include Newtons law of cooling and then equating the time derivatives with zero in Eqs. (1), (2) and (4). Equation (4) is then solved for the final power, Pf, which becomes

$& =

0 451" ([2 6])

(7)

Where U = overall heat-transfer coefficient [W cm-2}, and A = external core surface area [cm-2].

The overall heat-transfer coefficient U is assumed not to be a function of temperature.

The numerical value of the product UA was obtained from steady-state operation at 0.1 W. According to the manufacturer, the average temperature rise of the core at this power is 0.05ºC, so that

=

3 7* = 2 /

(8)

The value of UA has little influence on the total energy released during the accident, but it does determine the final power level which the reactor ultimately attains. Setting the time derivatives to zero gives 2 = B

) 6C (9) where the substitution 2 = / has been made.

The energy released in the excursion is shown as a function of time in Figure 5.5-2.

These curves are simply the time-integrated curves shown in Figure 5.5-1. The average core temperature is shown as a function of time in Figure 5.5-3. Assuming that the initial excursion is an adiabatic process, the maximum temperature is greater than 170ºC for all three initial conditions. The maximum temperature, however, will be less than 150ºC because of core separation once the thermal safety fuse deploys as designed. Core separation will result in an approximately 5% decrease in reactivity,

116 5.6 Loss of Water Shield from AGN Tank If the reactor is operated without water in the shielding tank at 0.1 watt power, the radiation level just outside the reactor tank will be about 10 mrem/hr (0.1 mSv/hr) of gamma rays, and about 50 mrem/hr (0.5 mSv/hr) of fast neutrons. These levels are six and eleven times as much, respectively, as they would be through the water shield. At the outside surface of the concrete shield, radiation levels would be about 0.14 mrem/hr (1.4 Sv/hr) of gamma rays and 3.4 mrem/hr (34 Sv/hr) of fast neutrons at 0.1 watt without the water shield. At 5 watts the radiation levels would be about 7 mrem/hr (0.07 mSv/hr) for gamma rays and 170 mrem/hr (1.7 mSv/hr) for fast neutrons. This radiation level would trip the high-level radiation alarm mounted on the reactor console and initiate laboratory evacuation.

While it is extremely unlikely for an excursion to occur without the water shield in place, the maximum acute dosage a person might receive at the surface of the concrete shield would be 18 rem (0.18 Sv) of gamma rays and fast neutrons. This would of course be a high amount of radiation but is within the guidelines for emergency doses.

Another potentially hazardous condition which can be envisioned is the case where the control and safety rods are fully inserted and the scram mechanisms are made to be inoperative. Under such circumstances the reactor power would continue to rise until the negative temperature coefficient reduced the reactor to a delayed-critical state at some high power. Under this circumstance, equilibrium is determined by the condition that the rate of energy conducted away from the core be exactly equal to the fission energy generation rate. Since the temperature coefficient of reactivity is approximately -

1.66 x 10-4 ºC-1, and the heat conductivity rate from the core may be estimated, it may be readily calculated that with 0.2% excess reactivity the equilibrium temperature is approximately 10ºC above ambient. This corresponds to a fission rate of approximately 10 watts. Postulating these conditions of the reactor operating at a continuous power level of 10 watts, the radiation received by a person outside the concrete shield would be approximately 12 mrem/hr (0.12 mSv/hr).

The above postulated exposure, although constituting a slight hazard, is considered improbable since it is doubtful that anyone would stay in such a position under reasonable administrative control for more than a few hours. It is interesting to try to

117 predict whether or not operation at this high power level would cause the fuse to melt, and, accordingly, shut down the reactor. Unfortunately this is a very difficult heat transfer calculation, due to the complicated geometries and, although it is believed there is a reasonable expectation that the fuse would function, no claim is made to this effect.

5.7 Energy Released At 5 watts, the total fission rate is about 1.6 x 1011 fissions/sec. Each fission produces approximately 0.6 neutrons that may leak out, 5 MeV of prompt gamma rays, 6 MeV of delayed gamma rays, and a small number of delayed neutrons. By far the largest source of radiation is due to the radioactive fission products. If the reactor has been operated at this level for a long time, the activity in MeV-curies equivalent at a time t seconds after shutdown is given by 0.4t -0.2, 7: the activity which produces 3.7 x 1010 MeV of ionizing radiation is defined as one MeV-curie equivalent. Table 5.7-1 shows these values for various times after shutdown. In the event that the reactor is operated on an eight-hour-per-day schedule, the figures for one day and one month may be reduced by a factor of one-fourth. Five-watt operation leads to 50 times the radiation fields generated at a power level of 100 mW, as shown in Table 5.7-1.

5.7.1 Operational Containment of Fission Products 7 A.T. Biehl, R.P. Geckler, S. Kahn, and R. Mainhardt, Elementary Reactor Experimentation, Aerojet-General Nucleonics, San Ramon, CA, October 1957.

118 The one significant difference resulting from 5 watt operation is the increased fission product inventory. Concerning the levels of activity in the core following 5 watt operation, reference is made to Biehl, Geckler, Kahn, and Mainhardt, Elementary Reactor Experimentation, Aerojet-General Nucleonics, San Ramon, Calif., October 1957, pp. 19-21. Further, it is noted that operation of Aerojet-General Nucleonics AGN-201 at 5 watt levels resulted in no detectable release of radioactive effluents due to the retention of fission products by the fuel matrix material. Even if there were gaseous effluents released they would be contained in the core tank. If, following recent 5-watt operation, it becomes necessary to open the core tank, samples of the gas within the core tank will first be taken and analyzed to assure that there has been no hazardous release of radioactive effluents from the fuel material. If any significant levels do exist, appropriate radiological safety procedures will be followed prior to and during subsequent opening of the core tank. These procedures are under the direct supervision of the Reactor Supervisor as authorized by the Idaho State University Radiation Safety Officer.

The core tank is vented to purge radiolytic hydrogen and noble gas fission products as part of a biennial surveillance procedure. Only small quantities of radioxenon and radiokrypton have been detected in the performance of these surveillances. Thus, significant fission product leakage from the fuel core is assumed not to occur. At power levels of 5 watts or less, leakage is insignificant and measurable amounts have not been found in other similar facilities. If leakage is experienced under any conditions of operation within the scope or the authorization requested, procedures for the safe and authorized disposal of fission gas that may be formed together with procedures to assure containment will be formulated.

Two assumptions are used as a basis for calculating the power generated in an accident:

119

a.

At time zero, a 2% step increase in reactivity is inserted with the reactor at low power ( 5 watts).

b.

At time zero, the thermal energy in the core is negligible in the comparison with the energy liberated during the accident and there is no heat removed from the core during the excursion.

The time-dependent behavior of the neutron density, including one average group of delayed neutrons is considered. A numerical finite-difference solution of the three coupled nonlinear differential equations yields a value of 54.4MW for the peak power at t=204 milliseconds and a total energy release of 2.61 MJ, resulting in a temperature rise of 71.3ºC. There will be about 1.45 x 1017 Mev of prompt gamma radiation produced in this excursion. Table 5.2 presents the residual activity formed in the core as a function of time after this excursion.

5.8 Gaseous Radioactive Product Release For the purpose of analysis, the gaseous fission products have been divided into two groups as shown in Table 5.8-1. The first group comprises those radionuclides that will remain in the tank water should the release occur when the tank is filled with water. This group includes the bromines and iodines. In the incredible event that no tank water is present, those isotopes would be added to the radioactive cloud and add to the hazard.

120 The second group comprises the insoluble volatiles, the krypton and xenon isotopes.

They are the major source of potential radioactivity in the room (and outside) if tank water is present.

Table 5.8-1 Gaseous Fission Products in AGN Fuel at 5-W Operation for 30 days

123 For a long time, t, and using Te=0.693/e, Eq (11) becomes:

> =

(?.3& x '8$)%%:&@)

(rads)

(12)

x 5 ABB

(?.3& x '8$):&0

%++@)+

)

C B

(rads/sec)

(14)

For the standard man,9 fa = 0.23, m= 20 gm, R= 10 m3/8 hr = 3.47 x 10-4 m3/s, and the value of the constant is thus 3.41 x 10-3.

The data necessary for the summation are contained in Health Physics, 3, June 1960, and the necessary activity concentrations are calculated from data given in Table 5.8-1.

The summation for I-131 through 1-136 yields a value of about D = 1.4 mrem/sec (14.4 Sv/sec).

Assuming that 300 rem (3 Sv) to the thyroid is a limiting dose,10 a person will have approximately 60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br /> to evacuate from the reactor room. Actually the time will be longer than this since the room exhaust will be in operation and will reduce the dose.

Further the leakage of the soluble iodine through the moderator and two containment 9 Report of ICRP Committee II on Permissible Dose for Internal Radiation (1959), Health Physics 3, June 1960.

10 J. J. DiNunno et al., loc. cit.

125 1 =

0F, G8 1

8

?

E

) @

(16)

= 8 70E1 ?

@G If the exposure is for an infinite time, 1 =

0F,)

G8E (17) and, comparing with the previous calculation, the total dose 1H1EI =

J(*EKL/L<N))(#-)

G8E(#-/L<N)

(18)

= 180 (1.8 )

Thus we may conclude that for any location outside of the building, the maximum possible dose to the thyroid from a fission product release will be considerably less than 0.07% of a maximum permissible dose, even when no credit is taken for radioactive decay.

5.9 Emergency Procedures The postulated Maximum Credible Accident assumes that the maximum available reactivity in the laboratory is inserted into the core, that as the control rods are inserted, the high-level trips on channels two and three do not work, and the operator is oblivious of the fact that the trips are not working. The excursion turns around after 14 seconds, and at about 20 seconds the temperature of the core no long rises. By that time the health physics alarms would have activated and the operator should then be alerted to the fact that an incident had occurred.

If the operator decided that an accident might have occurred, the operator would evacuate the reactor area and initiate building evacuation. The operator would then inform the appropriate administrative personnel, who would see that reentry to the area was not made until conditions were tolerable in regard to both radioactivity from reactor

126 as well as any airborne radioactivity that might have escaped. Complete emergency procedures are given in the facility Emergency Plan.

Any person close to, or exposed to radiation from the reactor excursion would be placed under observation at the Bannock Regional Medical Center which is prepared to handle and treat such persons. Similarly, an early attempt would be made to evaluate exposures by pocket dosimeters, film badges, etc., and to determine if any fission product leakage has occurred so that appropriate action could be taken.

After reentry, an estimation of the energy release would be made by gamma-ray flux measurements. When the radioactive hazards had been sufficiently reduced to permit working near the reactor another effort would be made to detect any radioactive gas leakage and the reactor would be secured by inserting cadmium poison rods into the glory hole and access ports. After a period of about a week, the control and safety rods would be permanently removed. At this time the core activity should be down to the millicurie range where the normal radiological handling procedures could be used. To place the reactor back in operation, the core would be assembled as a new critical assembly with a new safety fuse.

5.10 Safety Devices Every effort has been made to make the reactor safe against all foreseeable circumstances. In the event of an electric power failure, the control system is designed to fail safe and scram the reactor. The terminology of fail safe for a power failure means that energy from the power system is used to hold the control rods in a critical position in the reactor, i.e., spring and gravity forces acting on the safety rods are held in check with an electromagnet. Loss of power de-energizes the holding magnet and the rods containing fuel are accelerated out of the reactor to their safe, stable positions.

Although every effort has been made to make the standard instruments used in the reactor fail safe, the very nature of their electronic operation makes this quality intrinsically imperfect. However, instrument failure as a potential danger in the reactor operation is decreased by having three independent systems.

128 housed within a compartment that is enclosed by the main water shield tank, will not be pushed into the core. Finally the concrete shielding walls should prevent severe tilting of the reactor.

Some additional accidents or events that might conceivably result in the release of radioactive materials from the reactor are considered below. In the event of fire, damage to the reactor is not considered likely, since the AGN-201 reactor superstructure is intrinsically fireproof, as is the reactor room. The reactor may be sprayed with water, CO2, or any other fire-quenching material without damage to the reactor tank or fear of inadvertent criticality. The reactor will be shut down and the reactor room locked in the case of fire, and the Reactor Supervisor or designated alternate will be notified. If a fire involving the reactor or laboratory does occur, the reactor will be thoroughly inspected for damage before operation resumes.

In the extremely improbable event of a flood, no special precautions are necessary other than those normally taken in the event of a flood at an industrial site. The reactor will be secured and not operated at this time. The radiological hazards are not severe as the reactor is built to withstand minor flooding (one-foot of water). In the event of a major flood where the reactor might be overturned or carried away, there is no serious problem since the self-contained reactor has been designed to withstand such an emergency.

Further, because the core resides within a water-tight shield tank, there is no risk of inadvertent criticality in the event of complete immersion of the reactor.

It is extremely unlikely that a storm could damage the AGN-201 reactor. However, in the event of a severe storm, the reactor will be shut down and secured. In the event of civil disturbance such as a strike or riot, the reactor will be shut down and secured and guards will be posted at the entrance of the reactor laboratory to prevent unauthorized entry.

In addition to all the above safety features and administrative controls, there exists a negative temperature coefficient in the reactor core. The temperature of the reactor will vary during normal operating conditions as well as during an excursion. In both cases, the change in temperature will cause a change in reactivity. The temperature equilibrium rise of the core can be shown to be on the order of 2ºC when the reactor is

129 operated at 5 watts. Thus, under this condition, the steady state temperature is essentially ambient temperature.