ML20072G642

SAR Update & Review for Univ of Ok AGN-211P Reactor

ML20072G642
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Site:	05000112
Issue date:	05/18/1983
From:	OKLAHOMA, UNIV. OF, NORMAN, OK
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ML20072G622	List: ML20072G617 ML20072G642 ML20072G680
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NUDOCS 8306280581
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.i SAFETY ANALYSIS REPORT UPDATE AND REVIEW for i

University of Oklahoma AGN-211P Reactor i

License R-53 Docket 50-112 i

Revised I

May 18, 1983 i

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o TABLE OF CONTENTS 2 PAGE-I INTRODUCTION ......-..................... 1 II DESCRIPTION OF THE FACILITY . . . . . . . . . . . . . . . . . . . . -2 III SITE DESCRIPTION ......................... 5 IV OPERATION OF THE FACILITY OVER THE RENEWAL PERIOD . . . . . . . . . 6 A. Failure of Control and Safety Rod System . . . . . . . . . . . 7 B. Gamma Compensated Neutron Chamber . . . . . . . . . . . . . . . 9

'C. Corrosion of Pool Tank .................... 9 D. Fuel Deterioration ...................... 10 E. Operation at 100 Watts .................... 11 F. Maximum Hypothetical Accident . . . . . . . . . . . . . . . . . 12 4

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I Introduction The University of Oklahoma, as a cart of the application for renewal of its operating license R-53, has completed a review of its original Safety Analysis Report (SAR). This review has been conducted by the reactor staff, nuclear engineering faculty members and Reactor Safety Committee members. The application for renewal has received the proper internal approvals as attested by the signatures of the Vice President for Administrative Affairs and the Director of the School of Aerospace, Mechanical and Nuclear Engineering, a component of the College of Engineering, on the renewal application.

The Safety review logically comes down to three basic questions:

1. The facility has operated very successfully for 20 years. Can the facility be operated for the renewal period (20 years) with equal safety?

2. Are there Safety questions related to the increase in power from 15 watts to 100 watts?

3. What Safety questions are connected with operating the core in a flux trap configurations?

Our Safety review will focus on ideas associated with these three quution. Other general information about the facility is also included.

It is noted that the AGN-211 reactor series, designed and built by Aerojet General Nucleonics, is a standard class system. A complete description of the reactor system, associated potential hazards and AEC Hazards Evaluation Report, is on file (Docket 50-88). We request that all prior Aerojet General Nucleonics documents submitted, as well as all prior University of Oklahoma documents submitted, be incorporated by reference.

The University of Oklahoma's AGN 211P (pool type) is now the only remaining 211 in operation in the U.S. West Virginia University at Morgantown 1

o operated a 211 for a number of years at 75 watts. The University of Basel

.(Switzerland) operated a 211 at Ikw (for short periods) but has now switched from AGN polyethylene fuel to metal clad 002 elements in order to operate at powers above ikw.

The University of Oklahoma continues to consider its AGN-211P to be a valuable and highly versatile training and low-flux research reactor. The reactor is used as an integral part of the nuclear engineering program (B.S.,

M.S. and Ph.D.).

II Description g the Facility The reactor is located in the Nuclear Engineering Laboratory Building (NEL) and is physically in exactly the same position since the initial installation. The pool is sunk into the floor surrounded by concrete blocks.

Beam tubes energe from the concrete shield into a experimental area, the floor of which is - 10 feet below the reactor room floor. A fuel storaged pit, on the opposite end, allows for storage of spare fuel elements.

The room in which the reactor is located is a restricted area. Only members of the reactor staff have keys. All persons entering the room during i

reactor operations are required to wear a TLD or a PIC.

The pool and location of all components of the reactor system in the labs are as they were in 1958. The reactor room did undergo a major redecoration in 1979.

Outside the reactor room and immediately adjacent to it (see Fig.1) are graduate student offices (Rooms 113,114,115) and nuclear engineering labs (Rooms 108, 109, 124). See Table 1 for other room utilization in the labora-tory. These activities occupy the entire Porth half of the first floor. The south half of the first floor is occupied by Computer and Personnel Department employees.

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Table 1 Room utilization in the Nuclear Engineering Laboratory Area Room Usage 103 Mens Restroom and Shower 104 Locked Radioactive Source Storage 105 Womens Restroom and Shower 106 Sample Preparation and Counting Area 107A Reactor Laboratory 1078 Equipment and Records Storage 108 Undergraduate Laboratory 109 Graduate Laboratory and Conference Room 109B Gamma-ray Spectrometer 110 Hallway 111 Reactor Supervisor,s Office 112 Display and Reading Area 124 Classroom 3

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The second floor of the building is also occupied by Personnel Department empl oyees.

A second floor room, which is directly over the reactor pool, is con-trolled by the reactor staff and is used for storage. The purpose of this is simply a precaution. At 15 watts, the measured gamma-ray dose rate is about 0.09 mr/hr in the center of this second floor room. At 100 watts the antici-pated gamma-ray level is 0.09 x 100/15 = 0.6 mr/hr.

No neutron dose is measurable in the room. Operating procedures require the room be locked prior to startup. In the event of a power excursion it would be decidedly better to have this second floor room empty.

Radiation surveys in all areas immediately adjacent to the reactor room have been conducted repeatedly over the years. With the reactor at 15 watts measurable increases over background have been detected in two graduate stu-dent rooms 114 and 115 (the value is about 3 times background) which are occu-pied an estimated 20". of the working day.

Our conclusion is that, in its present location in the NEL Building, the reactor will continue to operate safely and in full compliance with federal regulations regards radiation doses to non-occupational and occupational per-sonnel. We consider all personnel in the building to be non-occupational if they are not associated with the reactor staff or the nuclear labs. We note also that all persons employed in the building know about the reactor and have gone through building evacuation rehersals. Thus, in the event of a possible or actual radiation hazard, we have confidence that our training program and emergency preparedness would protect the building employees.

III Site Description The NEL Building is located on the Norman Campus of the University of Oklahoma. The'NEL Building is constructed of brick and reinforced concrete.

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It has been modernized several times over the past 20 years. The building it bordered by a street on the west, parking areas on the north, east and south.

There are no known plans to introduce any form of construction around the NEL Building.

'The population of Norman has grown to approximately 50,000 with an additional 22,000 students on campus. It is noted that no liquid or gaseous effluents are released as a ccnsequence of the operation of an AGN-211P. No cooling of the pool water is necessary at 100 watts as the temperature rise of the 800 gallons of pool water is neglible.

The Norman area is subject to tornadoes. A city wide tornado alert system (sirens) operates very effectively. The reactor is not operated if a tornado alert is in effect.

There are no known factors which have altered the climatology or meteorology of the region from the descriptions given in the original license submission.

We note two additional points. The NEL Building now has a light and horn warning system on every entrance of the building. Signs clearly state the building is not to be entered if the lights are flashing and the University Police are present to ensure compliance. This is by emergency procedure.

Once each year a full scale simulated emergency building evacuation is done in order to ensure that all personnel in the building are familiar with the evacuation procedure. Evacuation horns sound throughout the entire building.

Evacuation routes are posted in various areas throughout the building.

IV Operation g the Facility over the Renewal Period The University fo Oklahoma's reactor has been in operation since 1959.

The staff has now completed a review of all components of the system. Our review was cast in the framework, is there a possible safety related question 6

relative to the reactor system or one of its component, since it is to func-tion for_ the period of the license renewal (20 years)?

The evaluation concluded that there were four components of the system which had the potential to affect safety or result in license violation.

1. Failure of the control and safety rods system.

2. Failure of the gamma compensated neutron ionization chambers.

3. Corrosion of the pool tank resulting in a loss of pool water thereby, exposing the core.

4. Fuel deterioration.

In addition, we have considered a Maximum Hypothetical Accident (MHA) as the same as the Design Basis Accident (DBA).

A. Failure j(( the Control and Safety Rod System The AGN-211P operates with 2 safety blades worth about 1.5% reactivity and 2 control rods, one worth - 1.5% and the other - 0.45%. The four together constitute a shutdown reactivity of about 5%.

Each year the rod mechanism is disassembled and inspected. Some rubbing of the rods on the rod guides has shown up but has not affected rod drop times which are also measured each year.

In 1980, we were fortunate to procure from West Virginia University a complete duplicate of the rod system. We thus have a spare system which could

-replace the present system or components of the spare system may be used to affect repairs. We consider the presence of this complete spare control and safety system as a totally adequate assurance of the ability of the system to operate safely over the renewal period.

A schematic of the scram and interlock system is given in figure 2.

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Control and Safety Rod Magnets D

1. Manual SCRAM CIRCUIT

' " Scrar h a a Series Interlocks dbl 0-7 watts-2' 150 watts 120 watts dL5 second period Log and Period Linear Low Water uu amp pu amp Temperature h h n Ion Chamber Ion Chamber Low Water Level e

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High Voltage High Voltage Power Failure A

Loss of Positive High Voltage Loss of Positive High Voltage Figure 2. Scram and Interlock Syste'm.

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B. Gamma Compensated Neutron Chamber The reactor operates with two conventional gamma compensated neutron ionization chambers. These drive standard linear and log calibrated power channels, the output being displayed on recorders. Both of these chambers have been replaced recently due to failure of their seals (moisture in the connectors). Replacement units are readily available. Thus, should the two chambers 1equire replacement (which is likely),..we are in a position to accom-plish the replacement with little difficulty.

We note that our license requires us to confirm the power calibration once each year.

In order to achieve a power reading of 100 watts, the log N chamber wil,1 need to be physically moved back from the surface of the graphite reflector.

This will be done opon receipt of NRC approval to go to 100 watts. We see no safety related questions associated with replacement of the ion chambers or the power calibration.

C. Corrosion of the Pool Tank The AGN-211P reactor core sits on an grid plate in the bottom of an 800 gallon pool. Approximately five feet of water covers the top of the core.

Each fuel element has (permanently attached) five inches of graphite and two inches of lead on the top end. At 15 watts the gamma level at the surface of the pool water is 15 mr/hr. At 100 watts we expect

[x15=100mr/hr A person standing, looking down into the pool, would be in a gamma field of about 20 mr/hr at 100 watts.

We have considered the question of corrosion of the aluminum tank. Should corrosion-failure of the tank wall occur below the top of the core, the water 9

would drain, exposing the top of the core.

If the reactor were operating when the leak occurred, it would scram because of the low pool water interlock. The radiation level will depend on the duration of the last run and the length of the decay time. Such a leak exposing the core, would not result in a hazardous radiation condition. The reason is quite clear. Operation at 100 watts results in a small fission pro-duct inventory which rapidly decays to very low levels making it possible to physically handle a fuel element one day after shutdown from a 15 watt run.

A measurement was made by taking a gamma-ray reading at the surface of

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the water. The reactor had been shutdown 15 days. This reading was 0.03 mr/hr (normal background). The pool water was then emptied until the top of the core was exposed. The reading measured - 0.09 mr/hr (three times back-cround). Had this measurement been made around one day after shutdown, the reading from the exposed core would be approximately 10-50 mr/hr. We, there-fore, conclude that should the aluminum tank leak, exposing the core (loss of the five feet of water normally over the core), no serious radiation hazard would exist, even if the leak occurred a short time after shutdown from a 100 watt run.

D. Fuel Deterioration The core of the AGN-211P consists of homogeneous ' fuel arranged in a series of elements. The fuel consists of polyethylene clad UO2 in the form of fine particles. Each fuel element appears as a homogeneous, unciad, solid polyethylene block.

The pool water is continuously monitored by circulating the water past a shielded 2 x 2 W (TA) detector. This highly sensitive system allows us to observe the very slow growth of gaseous fission products in the water. Fis-sion products appears about an hour af ter reaching 15 watts. A slow growth is 10

observed over the balance of the run. Water analysis has established the pre-sence of Kr-87, Kr-88, Xe 137 and Xe 138 and several daughter products in the water. At 100 watts we anticipate that fission produc levels in the water will be higher by the ratio about 100/15. This seems reasonable since no heating of the fuel will occur. The heat capacity of the fuel-pool system is quite large. This water fission product activity will be monitored carefully after first reaching 100 watts.

To ensure the fission product gases are not leaving the water in danger-ous quantities, entering the air in the reactor room, a new air monitor will be made operational. This sampling system will take air directly from the surface of the pool water into a sensitive air-ionization chamber. At 15 watts the present system does not normally record the presence of any gaseous fission products in the air drawn from over the water. We anticipate that the new air monitoring system may be capable of detecting the presence of gaseous fission products at a power level of 100 watts. Through calibration, the con-centration of any measured activity can be determined.

Further, the fuel has been used-for twenty years and shows no deteriora-tion. It is important to bear in mind that clad corrosion or clad rupture is not a problem. The AGN fuel is unciad.

E. Operation at.t 100 Watts Operation of the reactor at 100 watts has been discussed and considered throughout the various sections of the SAR. In remains, perhaps, to be stated that we find no impediment to increasing the power from 15 to 100 watts. The UWV reactor operated at 75 watts for a period of years without problems to our knowledge. In order to keep the gamma-ray level at or below 0.1 mr/hr at the operator's console, a gamma shield equivalent to 1 inch iron plate will be layed across the pool at power levels above 50 watts. We have always used a 11

very conservative exposure policy because of the presence of students.

Additionally, neutron and gamma radiation surveys around the reactor clearly establish that the concrete shielding is fully adequate for 100 watt operation. The power increase will be undertaken in several steps with a period of operation at each power step (25, 50, 75,100) to allow time to observe any unforseen effects.

F. Maximum Hypothetical Accident (MHA)

The University of Oklahoma AGN-211P is a homogeneous, thermal reactor, using 20 w/o U-235 oxide. The maximum permitted positive excess reactivity is limited to 0.65%. The effective beta for the reactor is estimated to be around 0.75%. Control of the excess is through a fuel loading limitation.

We will define a design basis accident (DBA) for an AGN-211 reactor to be the same as the MHA.

We will define the maximum hypothetical accident (MHA). Our license renewal asks for approval to operate in more than one core configuration.

This is important in the training of students in that several different criti-cal experiments may be run. Changes in rod worths among different cores could also be demonstrated.

We have asked for approval to operate in two basic core configurations.

These are shown in Figure 3. The standard core is a 12 element parallelepiped with a graphite reflector. The flux trap core will be either a 15, 16, 17 or 18 element core depending on the excess reactivity needed. The missing element at the center produces a neutron pileup which results in a thermal flux of about a factor of two over that in the standard core. (Both ANISN and EXTERMINATOR computer codes have been used to compute the flux in the hole.)

Such flexibility in core configuration is highly beneficial to the nuclear engineering program, both research and teaching.

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g K- n A A g R A R R A R R .R F F F A Figure 3 R A F F F F F = fuel g element F F F F R = graphite

& R R R R lx element I

& R R R R R R

A. Standard Core, Configuration R 'R R R it R R R A R g

R F F F F F F F W F F 8

E F F b E W = water g R A. R A R R R R &

B. Typical Flux Trap Core Configuration 13

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Studies of the flux trap configuration have, however, pointed to what we feel defines the MHA.

Consider: The flux trap configuration is in place at a temperature of 20*C. The excess is measured to be the maximum allowed, 0.65%. The 4 rods, fully inserted, are worth an estimated 4%. This means the reactor is shutdown by about 4-0.65 = 3.35%. A possible accident senario may now be postulated.

A new operator takes over and is told to change to a standard core. In absolute violation of written procedure he loads a fuel element into the empty flux trap, slot W. That element is estimated to be worth 4.65% (three group EXTERMINATOR CALCULATIONS). The reactor, with all four rods in_, is now prompt critical, having a 1.3% excess. This is less than the 2% reactivity insertion postulated by AEROJET-GENERAL. ,

The result of a 2% step input was calculated by AEROJECT General in the original hazards analysis for the AGN. In suinmary, a 2% step would result in a period of 5-10 milliseconds. The excursion would last about 100 milliseconds. The average core temperature would have reached approximately.

92*C terminating the reaction due to the negative reactivity coefficient. The total energy release is estimated to be about 4 megajoules. Since the maximum temperature reached would be approximately 130 C and 150 C is required to soften the polythylene, no large fission product release would occur. The excursion would result in a dose of 1 to 2 R should a person be standing looking down into the pool at the time of'the excursion. We note also that the 2% step would produce about 100 millicuries of I-131. The energy released in the accident would raise the pool water temperature approximately 0.3 C.

The ability of the ' water to retain all gaseous fission products and ' iodine would be undiminished due to the large solubility of these fission products in water.

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