ML20079P026

Annual Rept of Changes,Tests & Experiments Performed Under Provisions of 10CFR50.59 for Oregon State Univ Triga Reactor for Jul 1990-June 1991

ML20079P026
Person / Time
Site:	Oregon State University
Issue date:	11/04/1991
From:	Andrea Johnson Oregon State University, CORVALLIS, OR
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9111120159
Download: ML20079P026 (39)

Text

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'Icitphonc $03 737 2M1 l'un 503 737 MW November 4,1991 U. S. Nuclear Regulatory Commission ATTENTION: Document Control C'sk Washington, D.C. 20555 Gentlemen:

Subject:

Annual Report of Changes, Tests and Experiments Performed Under the Provisions of 10 CFR 50.59 for the Oregon Stato tlaiversity TRIGA Reactor (OSTR), l.icense No. P-106, Docket No.60-243.

The following report is sub.nitted in accordance with the requirements of 10 CFR 50.59(b) and 10 CFR 50.4. and covers tha OSTR5 annual reporting period of July 1,1990 through June 30,1991. The inf ormation ;n this report is compiled annually and is submitted to the USNRC in this specific 10 CFR 50.59(b) report, as well as in a special section of the OSTR annual report, which was submitted on October 23,1991.

During the specified reporting period there were 17 changes to the reactor facility and 5 changes to the reactor procedures conducted pursuant to 10 CFR 50.59. There were no changes to reactor experiments, no tests, and no new experiments performed under the provisions of 10 CFR 50.59 during the current reporting period.

The individual changes being reported are listed below by category and by title, and crs desedbed in more detail in Attachment A. Regarding this attachment, you will note that it includes a brief description of each change followed by a summary of the safety evaluation conducted for the described change. As required, none of the changes r,erf ormed under the provisions of 10 CFR 50.59 necessitated a change in the OSTR Technical Specifications or involved an unreviewed safety question as defined in 10 CFR 50.59(a)(2).

1. Changes to the Reactor Facility:

a. Replacement of the Reactor Power Chart Recorder

b. Removal of One Fuel Element from the G-Ring

c. Replacement of the Signal Cable Patch Panel for the Safety and Percent Power Channels

d. Movement of the Location of the OSTR Stack Smoke Detector 911112015c 911104 PDR ADocg 05000743 R PDR j/ '

USNRC 2- November 4,1991 4

e. Cleaning the Rotating Rack Irradiation Facility i

f. Replacement of the Emergency Power System inverter Batteries and Battery Enclosute

g. Reactor Console Left Hand Drawer Modifications

h. Modification to the Reactor Console Servo System

i. Temporory Addition of a Water Trap to the Rotating Rack Vent Line

}. Resin Flush / Fill and Water Make-up System

k. Change to the Air intake Dampers in the Reactor Bay Ventilation System

l. Replacement of the Reactor Top / Control Room Phone

m. Reactor Tank Water level Float / Switch Replacement

n. Reactor Console Wiring Cleanup

o. Addition of New Valves and Replacement of Domineralizer System Piping

p. High Voltage Scram Annunciator Lock in

q. Replacement of the Percent Demand Control Knob

2. 10 CFR 50.59 Channes to Reactor Procedures

a. Revisions to the Reactor Operations Committee Charter

b. Revisions to OSTROP 6.0

c. Update of Appendix B to the Radiation Center and OSTR Emergency Response Plan

d. Change of Acceptable pH Range for Reactor Primary Water and Bulk Shield Tank Water

e. Emergency Response Plan Revisions i

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USNRC 3- November 4,1991 We trust that you will find this year's report to be in good order. However, should you require more information or have questions regarding our report, please let me know.

Yours sincerely, A. G. Jo nson Director, Radiation Center AGJ:jrs\rc\5059tept.let Enclosure cc: Regional Administrator, Region V, USNRC, Walnut Creek, California Mr. Leroy Norderhaug, Region V, USNRC, Walnut Creek, California Mr. Phil Qualls, Region V, USNRC, Walnut Creek, California Mr. Al Adams, OSTR Project Manager, USNRC, Washington, D.C.

Mr. David Stewart Smith, Oregon Department of Energy, Salem, Oregon T. V. Anderson, Reactor Supervisor, OSTR S. E. Binney, Chairman, Reactor Operations Committee, OSTR B. Dodd, Reactor Administrator, OSTR J. F. Higginbotham, Senior Health Physicist. OSTR e

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ATTACHMENT A Changes to the OSTR Facility, to Reactor Procedures, and to fLtElor Experiments Performed _hlL6.uanliglQ CFR '50.59 The information contained in this section of the report provides a summary of the changes performed during the repor9ng period under the provisions of 10 CFR 50.59. For each item listed, we have included a brief aescription of the action taken and a summary of the applicable safety evaluation. Although it may not be specifically stated in each of the following safety evaluations, all actions taken under 10 CFR 50.59 were implemented only after it was established by the OSTR Reactor Operations Committee (ROC) that the proposed activity did not require a change in the facility's Technical Specifications and did not introduce or create an unreviewed safety question as defined in 10 CFR 50.59(a)(2).

1. 10 CFR 50 59 Chances to the Reactor Facility There were 17 changes to the reactor facility which were reviewed, approved, and performed under the provisions of 10 CFR 50.59 during the reporting period.

a. REPLACEMENT OF TVE REACTOR POWER CHART RECORDER i

(1) Description l

The previous chart recorder for reactor power was many years old and was in need of replacement. As a result, the reactor staff took out the old i recorder and inserted a new chart recorder in its place. An evaluation by the Reactor Supervisor and the Scientific Instrument Technician resulted in the selection and purchase of a now recorder which is similar to the old recorder -

but has multiple ranges of input. For examp!e, the old chart recorder would only accept input on the range of zero to ten millivolts. However, the new l chart recorder can be operated on a zero to one hundred millivolt range 1

which allows signal outputs from the wide range log power channel, the fuel temperature channel, and the wide range linear power instruments to be used without the need fer voltage dividers.

1 The zero to one hundred millivolt range on the naw recorder required a t odification to the original linear channel circuit which changed two resistors and one potentiometer in order to provide the proper output from the n-v circuit during pulse operation. No other circuits were affected by this change.

j (2) Safety Evaluation Care was taken to ensure that the new recorder met or exceeded all of the specifications of the recorder it replaced. The installation of the new recorder was supervised by the Scientific Instrume it Technician. Upon completion of the installation, a number of checks wera performed by the Scientific Instrument Technician and the Reactor Supervisor to ensure that the recorder, the wide range log channel, the fuel temperature channel, the wide rar.go linear power channel and the n-v circuit were functioning '

correctly before routine operation of the reactor was resumed. Replacement of the recorder will result in increased safety because the new device is

, much less likely to fail than the old one. Additionally, elimination of the #

voltage dividers will .ncrease the effective input impedance of the recorder channels from ten kohms to five megohms as well as eliminating a prime location for picking up noise in the circuits. Furthermore, the magnitude of any noise picked up will ba greatly reduced with the new recorder enabling it to track the reactor power rnore responsively and steadily than the old .

recorder. This wi:1 also contribute to increased safety.

b. REMOVAL OF ONE FUEL ELEMENT FROM THE G-RING (1) Description As a result of a routine measurement of the shutdown margin (SDM), as defined by the Technical Specifications (TS), the value was found to be approximately $0.85. Or.e fuel element was removed from the G7 core p# tion in order to increase the SDM and create a greater difference between the SDM and the TS limit of 00.57. From previous experience and measurements it was expected that the removal of a G ring element would

increase the SDM to about $1.10. Because tha fuel element removal required recalibration v' the control rods, the change was made just before the scheduled rod calibrations in the beginning of August. The removed element was placed in a convenient locatan in the Y fuel storage rack.

This fuel element movement also necessitated a minor change in OSTROP 10.7.B. including the associated " NOTE," and a change in OSTROP 11.D 2 regarding fuel handling. A summary of those changes follows. Normally, a fuel element in the B-ring (generally B1) is moved to G6 when using the sample-holding dummy fuel element or the CLICIT. As a result of removing G7, it is now better to move the B1 element to G7. Therefore,10.7.B was changed to: "Using standard procedures, move the fuel element from the B1 position to G7," and the NOTE was changed to reflect the intended use of the G7 position. Corresponding changes were also made in OSTROP 11.D.2 regarding use of the G7 podtion.

(2) Safety Evaluation Removing a fuel element increased the already sufficient shutdown margin of the reactor, and thus slightly increased reactor safety. The movement was perforrned using standald, approved fuel handling procedures. The element was stored in a f"el storage rack which is largely empty, but which was analyzed to be t,xticiently sub-critical even when completely full, Removing une element from the core does not result in a change to the OSTR physical security category because the mass of non self protecting 2asU is stillless than 1 kg. Changing OSTROP 10'and 11 as detailed above ensures that all of the FLIP _ fuel elements will remain in a contiguous grouping and will, therefore, give a better core configuration than just putting B1 in G6 and leaving G7 vacant.

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c. REPLACEMENT OF THE SIGNAL CABLE PATCH PANEL FOR THE SAFETY AND PERCENT POWER CHANNELS (1) Description As a result of the reportable event which occurred on July 11,1990, the reactor staff replaced the signal cable patch panel for the safety and percent power channels. The original panel was replaced with a relay and e rotary switched system which eliminates the need to disconnect and reconnect signal cables to the patch panel, and allows switching of the picoammeter  ;

to one of four positions. The positions are:

Position O. Normal; Position 1. Wide-Range Fission Chamber Position 2. Safety Channel Chamber.

Position 3. Percent Power Channel Chamber When the picoammeter is switched to positions 2 or 3, the current coming from the corresponding ion chamber is measurable on the picoammeter.

During this time the rest of the circuit is disconnected from the chamber, and a rod-withdra wal-prohibit interlock is activated. Therefore,it is not possible to operate the reactor while the switch is in positions 2 or 3.

The need for position 1 was not related to the reportable event, but was incorporated ut this time for convenience, its addition restored the capability to use the picoammeter to measure the current from the fission chamber while the fission chamber is being used for normal reactor operation. This was possible in the past, and is useful in determining the linearity of

. response of the linear channel and the chart recorder.

Appropriate procedures were modified to describe the proper operation of the new system.

(2) Safety Evaluation Replacing the signal cable paten panel with the new system definitely enhanced safety by reducing the dependence on procedural controls to

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5-ensure proper connections of signal cables from the safety and percent power channels, Additionally, the new system makes it physically impossible to operate the reactor unless the safety and percent power channels are properly connected. The relay and switch system was carefully designed to enable proper switching of the low current circuits, and to be fail safe, if the mercury-wetted refays should fail, then they will failin a manner which ensures that the circuits wll be properly connected. Such a failure will be noticed during the next startup check when an attempt is made to measure detector current.

Allowing the fission chamber current to be accurately monitored on the picoammeter during reactor operation has proved to be very useful in the past and contributes to the increased assurance of correct operation of that channel. The connection to which the picoammeter is attached is an output specifically designed by the manufacturer for a remote readout such as a recorder or meter. No failure mecnanism can be determined which would result in any feedback to the fission chamber's circuits as a result of attaching the picoammeter.

d. MOVEMENT OF THE LOCATION OF THE OSTR STACK SMOKE DETECTOR (1) Description An on-going investigation of the OSTR stack smoke detector revealed that

- the detector itself was functional, yet its positioning was such that its smoke detecting efficiency would be significantly improved by relocating the detector. A representative of the manufacturer felt that improved operation would be possible due to the pressure differentials and turbulence present at the detector's current location. Therefore, following the advice of the representative, the reactor staff moved the detector to a 'ocation which was upstream of the reactor bay exhaust fan instead of downstream.

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(2) Safety Evaluation This facility change enhanced safety, because the detector's smoke detecting ability has been improved. Af ter being relocated, the detector was again tested to ensure that it functioned properly.- The change of location introduced no safety implications with respect to the reactoi bay ventilation system, which continues to function normally. Relocation of the detector ,

l only required that a small hole be made in the ducting at the new location.

The old hole was appropriately plugged.

e. CLEANING THE ROTATING RACK IRRADIATION LACILITY (1) Description As a result of nearly 25 years of operation, the rotating rack had become increasingly difficult to turn, On several occasions the torque required to-turn the rack had reached a point where the clutch assembly slipped and the rack's rotation either slowed significantly or stopped. Attempts had been

. made tu rectify the situation by lubricating and cleaning the drive mechanism. However, in each case the efforts seemed to provide only a temporary solution to the problem. The reactor operations staff felt that the l-rack's resistance to rotation was due to a buildup of sticky irradiated oil and not due to any mechanical problem. The rationale for this argument was the l: ease with which the rack's rotation could'be temporarilyimproved by further lubrication.

In order to provide a long term solution to'the rotational resistance in the

! rack assembly, the reactor operations staff used a cleaning technique recommended by General Atomics,. the rack's manufacturer. More -

specifically, the staff filled the rotating rack with Simple Green, an over-the-L - counter non-organic and non-corrosive liquid degreaser, and then rotated the rack for five days. This provided adequate time and agitation to dissolve and remove the irradiated oil from the rack's chain, sprocket and bearings.1 The Simple Green was then pumped from the rack, and the rack was flushed three times with water to clean out the-remaining . Simple' Green end l- contaminants.

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The system and procedure used for pumping the Simple Green and water in and out of the rotating rack was approved after much discussion by the l Reactor Operations' Committee. All valves, hoses and drums were positioned, leak tested, and shielded, as appropriate and practical, to minimize personnel exposure. '

Great care was taken in drying and decontaminsting the rack and sample loading tube. Overall, the cleaning was very successful with the rack now turning freely.

(2) Safety Evaluation The rotating rack assembly is an air filled system which is isolated from the reactor primary coolant by its aluminum housing. Therefore, no interaction .

between the Simple Green and the reactor water was possible. Simple Green is a commercial, household product which is an aqueous-based cleaning and degreasing compound, and it was concluded that no probloms would be created even if it did contact the primary water. The Material Safety Data Sheet (MSDS) indicated that the only hazardous material component was 2% by volume butyl cellosolve. Since the volume used was small (10 gallons),it was determined that it would not create a mixed waste disposal problem, and it was further determined that the absorbed liquid -

material would be accepted by Allied Ecology Services Inc. for disposal at their Richland, Washington low-level : radioactive. waste disposal site.

Discussions with the manufacturer. indicated that Simple Green istused extensively in the nuclear industry as a decontaminating agent. Tests were performed to provide assurance that Simple Green would not attack metals oi plastics, and neutron activation analysis indicated that the only significant '

activation products wereisotopes of sodium and potassium. The reason for .

leaving the.Simplo Green in-the rack for five days was based-on tests-performed at-the OSTR which. indicated that it required several hours for Simple Green to adequately soften and dissolve the irradiated ' oil.

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8-The pumping, air filtration and radiation monitoring systems were very carefully designed to minimize any potential for personnel exposure or >

radioactive material release during the cleaning process. As much as reasonably possible, drums and hoses which contained radioactive material were snielded, and all potentially contaminated air was filtered many times as well as monitored at least twice prior to discharge. Radiation dose rates were directly monitored at all times during the operation by OSTR staff, and appropriate protective clothing and dosimetry were worn.

The system components which were on the reactor top were positioned so that the heavy shielding was directly over the solid concrete reactor tank pedestal and not on the cantilevered portion of the reactor top.

f. REPLACEMENT OF THE EMERGENCY POWER SYSTEM INVERTER BATTER!ES AND BATTERY ENCLOSURE (1) Description The odginal 6-volt battery system for the emergency power inverter did not seem to have sufficient reserve capacity to provide the required degree of rapid response to electrical power variations. Furthermore, the batteries were of a non standard type which were costly and required special ordering each time they needed replacing. Therefore,- to increase reliability and reduce the cost of annual upkeep, the reactor operations staff replaced the emergency power system's batteries. The original enclosure with twenty 6-voit batteries connected in series was replaced with a new, specially designed enclosure containing ten 12 volt deep cycle batteries connected in series.

(2) Safety Evaluation The change made the emergency power system more reliable by reducing the number of interconnecting cables and terminals, and by replacing the original batteries with batteries having a much greater ampere-hour capacity.

Although the emergency batteries do not directly contribute to reactor

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9 safety, it is clear that this change prov. des greater asswance of desired emergency power system operation when needed, and it will therefore increase the ability to monitor various parameters immediately after a commercial power f ailure. In addition, the new batteries are available locally, olirninating the three-week lead time that was previously required to replace a defective battery.

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g. REACTOR CONSOLE LEFT-HAND DRAWER MODIFICATIONS (1) Description The reactor staff modified the left-hand drawer of the reactor control console by moving the two test buttons for the high voltage scrams from the interior of the drawer to the front panel. At the same time, the tog power / fuel temperature meter in the drawer was replaced to eliminato non-linearity problems associated with these channels which had been traced to the meter.

(2) Safety Evaluation InstcIling the two high voltage scram test buttons on the front panel of the '

left-hand drawer eliminated the daily opening and closing of the drawer, and thus increased safety by reducing the possibility of premature failure of the cables which entered through the rear of the drawer due to repeated flexing.

Additionally, the change eliminated the need for the operator to reach into an energized drawer and, therefore, eliminated any possibility of electrical shock.

The new meter movement was checked for linearhy and smooth : 93 ration prior to installation. Therefore, replacing the log power / fuel temperature '

meter movement with a new identical meter increased safety by eliminating the need for non-linearity corrections.

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h. MODIFICATION TO THE REACTOR CONSOLE SERVO SYSTEM (1) Description When the new reactor console chart recorder is insta3ed, a number of tests were performed to ensure that the recorder was f unctioning correctly.

During these tests there appeared to be a small shif t in the indicated reactor power level when the mode switch was moved from the square wave to the automatic position. It appears that this situation was always present, but that it was recently more noticeable due to a more sensitive and responsive recorder.

An investigation of the servo system circuitry by the Scientific Instrument Technician revealed that the period amplifier was sending a positive 0.75 volt signal to the servo limiter when the reactor period was infinite. The servo system interpreted this voltage as a positive period and, as a result, regulated the actual reactor power to a level about five percent below the demand setting when the mode switch was in the automatic position. When the mode switch was in the square wave position there was no input from the period amplifier and, therefore, in this configuration the actual power was regulated to be the same as the percent power demand setting. It should be recognized that in each case the indicated power was correct, but that in the automatic mode the power was slightly lower than the percent I power demand setting.

1 in response to the above situation, the Scientific Instrument Technician constructed and installed an amplifier circuit with unit gain and an adjustable offset of the input signal. This amp;ifier receives the period amplifier output from the left-hand drawer and then provides a corrected input to the servo limiter. The araplifier was constructed on a printed circuit board and

. installed in an existing open slot in the servo card box.

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(2) Safety Evaluation The change was discussed with electronics personnel at General Atomics.

They concurred with our approach and could see no reason not to proceed.

The change to the circuit has enhanced safety by eliminating the shift in reactor power which occurred when the mode switch was moved between the square wave and the automatic position. The high input resistance of this circuit climinates the possibility of a failure in this circuit which would in any wayinhibit the operation of the reactor period scram. Should a failure occur that would cause the output of this circuit to increase, then the servo system would respond by oriving the regulating rod in and lowering power.

If the circuit failed with the output low, the servo system would respond by withdrawing the regulating rod and raising power. In this case, the operator would manually scram the reactor, or without operator action, the power increase would be terminated by either a high power scram or a reactor period scram depending on the initial power level. This situation would be no different than if, for example, the console power supply providing a-positive voltage to the servo system should fail. Therefore,it is clear that t this failure modo does not present an unreviewed safety question.

Installing the amplifier in the servo card box provides . a protected environment fcr this circuit and helps ensure a long trouble-free existence,

i. TEMPORARY ADDITION OF A WATER TRAP TO THE ROTATING RACK VENT LINE .

(1) Description

.in order to speed up the drying of the. rotating rack subsequent to the cleaning and flushing, the reactor staff inserted a temporary water trap between the loading tube and the first particulate and charcoal filter. At the time, it appeared as though drying was being hindered by the presence of the filters due to the moisture dripping back down the loading tube. By putting a simple water trap before the filters it was reasoned that this moisture could be removed more efficiently. The rotating rack vent system was kept running continuously until the rack dried out.

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(2) Safety Evaluation There were no safety implications associated with this small chango in the rotating rack vent system. The trap was simply placed in the vent line with quick disconnect fittings to facilitate insertion and removal as necessary. All of the filters remained in line, and any radioactive material collected in the trap was routinely monitored, with the material being disposed of as radioactive waste,if necessary. The same security locking device was used for the loading tube as was used during the cleaning process.

j. RESIN FLUSH / FILL AND WATER MAKE UP SYSTEM (1) Description The reactor operations and health physics staff modified the primary water demineralizer/ purification loop by adding a systern which is used te flush spent resins and add fresh resins to the ion exchange column, and to provide make-up water to the reactor primary cooling system or bulk shield tank.

One of the reasons this system was installed was to virtually eliminate the already small quantity of radioactivity which was previously drained to a hold up tank during the resin flush and fill process. In addition, the system provides the option to use distilled water for the reactor and bulk shield tank make-up in order to prolong resin life and further reduce water discharges to the hold up tank.

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Two new tanks (a make-up water tank and a resin dewatering tub) are located along the east wall of the heat exchanger room behind the heat exchanger, while the pump and the plumbing for the new system were placed at convenient locations in that same area.

For resin flush, water is pumped from the rnake-up tank through the demineralizer to a sack in a plastic crate. This crate is raised sufficiently high in the resin dewatering tub to enable the water.to drain from the resin without any further handling. The water in the dewatering tub can then be easily pumped back into the make-up tank. For resin fill, water is pumped-

4 from the make-up tank through the new resin funnelinto the domineralizer and back to the make-up tank. Make-up water for reactor systems is added by pumping water from the make up tank through the demineralizer to the reactor tank or bulk shield tank, as necessary.

(2) Safety Evaluation These facility changes have no effect on reactor safety in .:ny way. In essence, there is little difference in the resin change process and the water make-up procedures except that there is a plumbed, closed system with the option to use distilled water instead of a once through city water flush and fill. There may be a slight improvement in reactor safety by adding distilled make-up water, but the effect is very small because the volume of water added is small compared to the reactor tank volume.

The major benefit of this change is enhancement of the Center's ALARA commitment in that there will be a reduction in the radioactivity which will be drained to the hold up tank. This will in turn reduce the already yaty small amount of radioactivity discharged to the sanitary sewer to levels which will probably be below the limits of detection. This is consistent with the University's policy and the principles of ALARA.

(3) Conforming Procedural Changes OSTROP 7 was revised to add a new section on changing the domineralizer resin using the new system. Appropriate health physics precautions were also added. In addition, section'7.8.A 4 on water make-up was changed so

that it confc ms to the new hardware.

l k. CHANGE TO THE AIR INTAKE DAMPERS IN THE REACTOR BAY VENTILATION I SYSTEM (1) Description The pair of air intake dampers for the reactor bay ventilation system were-operated by two pneumatic motors. One motor was located on each side of

the unit which housed the two air intake dampers, and the motors were coupled together through the dampers so that both motors attempted to open and close both of the intake dampers. This did not seem to be the best design. If tha speed of operetion of each motor was not exactly equal (i.e.,

if the motors attempted to open the dampers at different rates of epeed), the motors worked against each other. This uneven force increased the stress on the mechanical components of the dampers and reduced the overall efficiency of the damper chsing process, As a result of the above situation, the reactor staff made a small modification which allows two alternate methods of operating the pair of intake dampers. Approval for both methods was obtained which allows operational flexibility without the need for further analysis and approval. The first alternative arrangement results in the dampers being coupled and operated by one pneumatic motor while the other motor is disconnected.

The second arrangement involves having each pneumatic motor operate only the damper on its side of the damper unit in this second arrangement, the dampers are decoupled and the motors each independently work one of the dampers.

(2) Safety Evaluation 11 should first be noted that the damper system is of a fail-safe design. The pneumatic motors keep the dampers open by the action of compressed air.

l Any conceivable failure would result in air leakage which in turn would cause the motors to close the dampers in response to the action of a large spring.

In addition, the dampers are designed so that if they are disconnected from the motor for any reason they will clnse under the action of gravity.

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t. Both of the alternative modes of operation involve separating the actien of the two pneumatic motors. This is considered a design improvement since it eliminates the possibility that the two motors will work against each other, which in turn improves the reliability of the mechanical aspects of the intake dampers and results in a more consistent and smooth closing action.

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Both of the dampers can easily be operated by one pneumatic motor, due to the fact that the dampers op9 tate very freely and there is very little resistance to overcome. Having the ability to use only one motor to operate the dampers willincrease operational flexibility, especially should one motor require maintenance. Tests have shown that one motor can easily close all of the dampers well within the specified 5-second response time.

The pacond alternative has each pneumatic motor operating one of the dampers. Due to the fail safe nature of the damper design,it can be seen that this is clearly an improvement over the old system. With this arrangement, the damper motors do not work against each other, and there is an allowance for variations in the individual closing rate of each of the dampers. However, each motor is still required to close the damper it controls within the 5 seconds stated in the Safety Analysis Report.

l. REPLACEMENT OF THE REACTOR TOP / CONTROL ROOM PHONE (1) Description The phone between the reactor top and the control room was not functioning as reliably as desired. it was determined that there was an intermittent problem with the phone wire. Consequently, the reactor staff completely replaced the wiring and power supply. On the old system the wires were routed to room D106 where the phone's power supply was located. The new system uses the +25 volt power supply in the reactor console and new wires were rua directly from the control room to the reactor top in addition, a new buzzer system was designed and installed.

(2) Safety Evaluation The only aspect of this minor facility change which needed evaluating was the plan to use the +25 volt power supply from the reactor console. The consolo 25 volt DC power supply is rated at 1.5 amps. Before adding the phone, the current loadings were about 500 milliamps for the +25 voit supply and 300 milliamps for the -25 volt supply. The new circuit adds

about 300 milliamps of load to the +25 volt power supply when in use.

Clearly this extra load is very small and still well within the range of the reactor console power supply.

No existing reactor wiring was interfered with when the new system was installed.

m. REACTOR TANK WATER LEVEL FLOAT / SWITCH REPLACEMENT (1) Description A new reactor tank water level float and switch assembly was designed,-

fabricated and installed in place of the old float and switches. The new float and switch assembly has two floats with very sensitive high and low water level microswitches for each float. The circuit is arranged such that .

actuation of- only one microswitch' is needed to cause control room annunciation and a green light alarm.

(2) Safety Evaluation The new float system is engineered better than the previous system, it has a sturdy, welded bracket which is attached to the reactor tank in an out-of-the-wey location. The new system has duplicate floats and more sensitive micro-switches which will increcse safety by providing redundancy and .

increased sensitivity. In addition, the new system has an indicator gauge marked on the front which makes it easier to observe changes in the reactor tank's water level due to evaporation.

The new float system was thoroughly tested after installation to ensure that it was functioning correctly. This test included actually radng and lowering U the water level to initiate.the respective alarms.

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n. REACTOR CONSOLE WIRING CLEANUP (1) Description Over the years, the numerous modifications anri ed.iitions to the reactor l console resulted in a maze of wiring which was untidy and inefficient. There were several wires which were no longer in use, but their presence made it difficult to trace wires. These wires also may have been potential sources of electrical noise. The Scientific Instrument Technician cleaned up the console wiring generally, and wired several systems more tidily and efficiently by routing them to their destination, rather than to several i locations across the console, in addition, the specific changes detailed I below were made. These changes have made it being much easier to trace wiring and have reduced potential sources of electrical noise.

Specific changes were:

(a) The console power, demineralizer pump power, and fluke temperature .

indicator power were hardwired to a dedicated terminal board.

Previously they were plugged into a power strip.

1 (b) The wiring and parts for the claxon horn which were no longer used were removed.

(c) The wiring to the switches / annunciators on the reactor console were l

completely replaced.

(d) The cabling and wiring associated with the old fission chamber and its drive were tidied up.

l l (2) Safety Evaluation There were no changes associated with the reactor console wiring cleanup.

Therefore, there were no unreviewed safety questions involved. However, the nature of this work was such that the- staff felt that it should be reviewed by the ROC prior to implementation. The wiring cleanup resulted

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in a tidy console with old unused wires removed, and with existing wires properly connected and bundled. This will make it much easier to traco connections when problems are discovered, and it will also minimize electrical noise.

As each part of the sviring cleanup was accomplished, the systems associated with that part were tested to ensure that they were fully

• functional.

o. ADDITION OF NEW VALVES AND REPLACEMENT OF ")hMINEPAL!ZER SYSTEM PIPING (1) Description As one of the follow-up actions associated with an occurrence involvirig leakage of water from the demineralizer system, two new valves were placed in the system so that it can be isc. lated nearer the reactor tank. The new two-way valves are made of stainless steel and replaced the original plastic three-way valves DV1 and DV18. In conjunction with the valve replacements, the old Schedule 40 plastic piping between the reactor tank and the demineralizer skid was replaced with new Schedule 80 plastic piping. New pipe supports were installed at various locations in the reactor bay pipe trench.

(2) Safety Evalu: tion i

The valve additions and the new piping will reduce the already small probability of a demineralizer system leak, and will thereby enhance safety.

The original piping was of a !ighter gauge plastic and it k ad been in use, and exposed to fluorescent lights, for many years. In addtion,it had relatively few supporting hangers on some of the horizontal runs in the trench.

Therefore, replacing the piping with newer, stronger pipe with additional supports has reduced the probability of failure. It made sense to proceed with the pipe replacement at the same time the new valves were being installed.

t

. - _ _ - ~ . . . - _ _ _ _ _ . . - _ . -

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I Replacing DV1 and DV18 with new stainless steel valves will enable them to be opened and closed each day with less wear and greater reliability.

l This, in turn, will ensure that the demineralizer system is isolated closer to l

the reactor tank than was previously possible, further reducing the probability of water leakage. There is no need for three-way valves any more due to earlier piping changes. All the usual water flow lineups are possible with the new valves. To get better pipe runs, the new valves were located on the south side of the reactor shield on the second level rather than on the second level railing where DV1 and DV18 were originally positioned.

p. HIGH VOLTAGE SCRAM ANNUNCIATOR LOCK-IN (1) Description The three reactor power detectors (two ion chambers and a fis? ion chamber) l each require a high voltage power supply. If any of the high voltages to these detectors drops significantly, then a so-called "high voltage" scram will l result. (The scram is actually due to a voltage from the high voltage supply l

which is too low.) As the system was originally configured, only one of the three high voltage scrams (HV2) would lock in an annunciator. Therefore, if a scram occurred via HV1 or HV3 the operator would nct know which of these channels had caused the scram. Th9 Scientific Instrument Technician designed and incorporated a change which now enables each of the high vcitage scrams to lock in and annunciate. HV1 in the left-hand drawer was

!. modified to be consistent with HV2, and HV3 uses a new relay as a Icck-in feature.

(2) Safety Evaluation The changes enable any high voltage scram to be identified by a light which remains on even after the transient conditions. -This in turn allows more rapid identification of potential causes of the trip.- The circuit changes are all designed to be fail safe and are independent of the functional requirements of the fission chamber and ion chambers, in particular, the

catectors' high voltage circuits are ur'affected by the change. This means that a detector's high voltage could return and the detector could resume normal operation regardless of whether or not its high voltage scram annunciator is reset,

q. REPLACEMENT OF THE PERCENT-DEMAND CONTROL KNOB (1) Description The original Wdemand control knob had a range from zero to 100% and could not be turned above 100% When the reactor operates at 1 MW in automatic mode, there is a need to be able to adjust the knob up or down a small amount on either side of the 100% power setting. For this reason, the console was set up so that 100% of the power for any given power range (selected via the powcr range switch) was equivalent to about the '

95% setting ca the Wdemand control knob. This meant the operator could not set the %-demand knob exactly at the desired power level, but instead had to set the knob about 5% low.

To remedy this situation and to enable the desired power level to be dialed in, the Scientific Instrument Technician installed a different style knob. The new knob necessitated a small change in the circuitry to enable the 10 turn potentiometer to cover the voltage range needed in addition, to provide better voltage stability, the 1 3 kO rasistor was replaced by an active voltage regulator. The voltage range covered by the potentiometer was shif ted to enable the new knob to vary from 10% to 110% of the power indicated on the range switch.

(2) Safety Evaluation The changes which were made will result in increased safety. Prior to the changes,it appeared that any changing load demands on the +25 V power supply resulted in the need for slight changes in the %-demand potentiometer setting. Replacing the resistor with the voltage regulator has eliminated any variation in the voltage seen by the potentiometer and has resulted in more stable operation.

Changing the % demand control knob so tLt the % power indicated correlates much clocer with the actual reactor power enhance = safety by avoiding the need for the operator to make a mental adjustment. Having a

%-demand control knob covering from 10% to 110% of the power indicated on the range switch is suitable for all operational needs. This range fits well with the chart recorder, as the paper covers a range up to 110% of the range switch setting.

In addition, while the normal maximum operating power is 1 MW, the reactor is licensed to 1.1 MW so that the maximum setting of the % demand control knob on the maximum power range will not cause the licensed power level to be exceeded.

l

2. 10 CFR 50.59 Chanaes to Reactor Procedures There were five changes to reactor procedures which were reviewed, approved, and performed under the provisions of 10 CFR 50.59 during the reporting period,

a. REVISIONS TO THE REACTOR OPERATIONS COMMITTEE CHARTER (1) Description On an annual basis, a standing Reactor Operations Committee (ROC) subcommittee reviews the Committee's charter. The 1990 review indicated that several changes should be made in order to keep the charter current.

Several other changes were made in order to clarify activities specified in the charter. None of the changes involved major changes in ROC operating policies.

The- changes to the ROC charter included several very minor editorial clarifications which were- of no possible significance from a safety standpoint. Other changes which were of a minor nature, but perhaps not totally editorial, are listed below.

22- -l

'i Section Chanae ll.2 (1) The section on ROC membership was changed to delete the statement that the! chairman must hold at least a Reactor Operator license valid for the OSTR. The statement was replaced '

by the requirement that at least one of the ROC members must possess a Senior Reactor Operator's license valid for the OSTR.

The former statement is not required by the _OSTR license or Technical Specifications and is replaced by a statement which

. brings more technical corapetence to the Committee. ,

(2) Statements f and g under _ membership were changed to .

include the words "at leam one" person whose field of expertise -

is ... instead of-just "a person".

(3) in the last paragraph of section 11.2, the abbreviation ROC was substituted for the word Committee in several places.

111. 1 (1) Under the Audit section, item a was modified by adding _the words "and the fuel inventory and transfers" to the list of items specified for auditing in item a.

(2) ' Under the Audit section, a new item e was added to clarify that facility drawings are an item which will be audited as part of the quarterly ROC audit process.

Ill.1 - Under the Audit section, there exists an additional list of iterns to -

- be' audited on a quarterly basis which includes radiation surveys-(item d) _ The radiation survey statement was expanded. by _ _

' ~

adding the parenthetical statement." including routine surveys L such as daily, weekly, and monthly; special surveys; R AM receipt -

surveys; and environmental monitoring surveys".

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Under the Audit section, tbcra exists an additionallist of items to j be audited on an annual basis which includes the Radiation l Center Health Physics Procedures (item f)s The statement listing the Radiation Center Health Physics Procedures was expanded by l adding the following clarifying sentences: "The ROC reviews the - l Radiation Center Health Phywes Frocedures (RCHPPs) annually on a rotating basis (i.e., a certain fraction of -the. RCHPPs are- e reviewed nach quarter). However, during the interim period, the Senior Health Physicist has the authority to review and revise RCHPPs as necessary without ROC approval.

_.c (1) The first sentence of item c in the Review and Approval section was changed to remove the statement "within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />" ' -

since all events which are required to be reported to the NRC will be reviewed and, as appro'priate,' approved by the ROC. The-reporting time required by the NRC will have no bearing -on ,

whether or not such events are reviewed by the ROC.

(2) The last sentence-of item c in the Review and Approval section was deleted because the requirement for ROC review and approval of.. written reports to be submitted to the NRC before the reactor may be brought back to an operational state was overly testrictive, is ~ not required by the .. OSTR -license - or-Technical Specifications, nor by ANSI Research Reactor Standard -

, .15.1. This standard deals with the development of technical' specifications for research reactors and is endorsed by and used _

by the NRC.

(3) -- A final sentence was added to item c in the Review and Approval section which reads as follows: -"This review and' approval will occur before formal submission.of reports to the NBC."

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24-j Section Chance lil.2.d&e items d and e in the Review and Approval section had a final sentence added to each of them stating "This review and l

appr .~al will occur before formal submission of reports to the involved regulatory agency or agencies.

l lit.2.f item f in the Review and Approval section had the statement "as the need arises" added at the end of the existing statement.

(2) Safety Evaluation Based on the nature and scope of the changes to the ROC Charter, it is evident that most cf the changes were editorial, in cases where the changes added or deleted statements, these additions or deletioris arose from the need to ensure that the charter is consistent with and responsive to the requirements of the OSTR license and Technical Specifications, and clearly reflects the operating policies and procedures of the ROC. Therefore, all

! changes can be categorized as being either editorial, clarifying or updates to the charter. Minor editorial changes, by definition, have no safety implications, while additions or deletions resulting in updates and clarifications are clearly beneficial to safety since they minimize or eliminate l

the possibility for misunderstanding. Updates to the. charter to ensure compliance and/or consistency with the OSTR license and Technical l

Specifications . also enhance safety by ensuring that- the required memberships, audits and reports are strictly in accordance with NRC requirements,

b. REVISIONS TO OSTROP 6.0

0) Description The OSTR staff decided to perform a complete review of OSTROP G and to revise it where necessary to bring it up to date, in addition; the staff recognized that there were changes which could be made to improve the efficiency of various processes. All of the changes, other than typographical or grammatical changes, are listed below.

' 25-Section Chanae 6.2.A.3 The parenthetical references to which personnel were or were not licensed reactor operators were deleted, as they were not relevant here, 6.4 The recently approved changes to the Reactor Operations Committee Charter were incorporated into this OSTROP.

These changes were discussed in detailin a 10 CFR 50.59 safety evaluation dated October 17,1990.

6.5 A new paragraph was addsd to the beginning of this section stating that normailines of succession have been established for certain specified positions.

6.5.B.7 Designation of the Reactor Administrator as the Principal Security Officer in'tho absence of the Radiation Cer.ter Director is now included as one of the items in the lin(s of succession for key reactor positions. Therefore, this rnatter was deleted from the new OSTROP 6. Other Reactor Administrator responsibilities were renumbered.

6.5.D.1 This section was revised to allow the person filling the l Reactor Operator position to hold an NRC reactos operator or senior reactor operator license.

6.5.G.p _ Reference to the Student Health Center was deleted in order to be consistent with the latest revision of the OST3 Emergency Plan.

6.5.H.8 A minor change was made here to recognize that there are now two cobalt-60 irradiators.

26-Section Chance 6.6.A.4 A clarification was made stating that all reactivity changes will be made by, or in the presence and under the direction of, the licensed operator of record at the time the reactivity changes are made.

6.6.A.5 Personnel other than the reactor operator can coniplete the startup checklist. Therefore, this section was changed to reflect that fact.

6.6.B.2.c The range within which the Reactor Supervisor must remain while carrying a radio was reduced from 15 to 10 miles.

6.6.B.4 For clarity, this section was split into two sections, and subsequent sections renumbered. In addition, the Reactor Administrator was added to the list of personnel who may suspend operations if necessary. The Reactor Administrator was also added as a person whc needed to give permission for restart.

6.6.C A new section with this number was added and subsequent sections renumbered and revised for conforming consistency. The new section addresses the fitness for duty of licensed operators and the goals of a drug-free environment.

6.6.D 4 A parenthetical note was added to clarify the intent of the requirement that the reactor operator of record not be the instructor of an ongoing class.

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Section Chanoo 6.7.B.7 A new section was added to specify the physical security requirements when staff observe or suspect unauthorized ,

entry into a security area. -This section is consistent with the OSTR Physical Security Plan.

6.8 This section was completely revised except for the cart relating to the reactor status board. The section vrar split into four new sections (6.8 6.11). One of the major intents of these revisions was to remo'(e the current need to have several pieces of papn Act .ating which refer to the-same procedural or facP.ty change. 'a the revised procedure, normally one document wW be used for the requisite review, approval, and informational purposes.

In section 6.8, the list of procedures which require a 10 CFR 50.59 safety evaluation before they are changed was expanded to include OSTROPs 12 and 26.

The new procedure requires all 50.59s to be circulated to all of the ROC (rather than to just a quorum) for. their .

approval,'and to reactor operators for information. The only exception to this is when an ROC member is, w . ble to sign the 50.59 form due to absence or illness then he or -

she may be omitted from the circulation. Provisions are also g ven for.an ROC member who feels he/she cannot give approval without further discussion.

6.9 . This section addresses the new Reactor Operations Committee Approval Sheet (ROCAS). To a certain extent, this sheet replaces the old OlBs and FCBs. The sheet is used for anything except 50.59 evaluations which requires ROC approval between meetings. As with 50.59s, all ROC

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members sign off on the ROOAS'to signify review and approval, and. reactor operators sign to signify review, ROC signatures also--indicate agreement with the conclusion that a 50.59 safety evaluationis not needed for the item- or action under consideration. Provisions- are .

given if a member feels he/she; cannot give approval withou'. furthcr discussion.

6.10 This section adds a new information Bulletin used for communicating informational items only. . It has no approval -

function.

Flowchart The flowchart an'd examples of the standard forms and Forms have been revised to reflect the changes descr; bed above.

(2) Safety Evaluation This safety evaluation addresses only the substantivo changes, in addition,

the changes due to the revision of the ROC charter are not covered here because they have already been addressed in the 10 CFR 50.59 evaluation dated October 17,1990.

Section Chanae.

6.5 - Adding a formalized line of succession for all of the key positions- with respect to reactor operations enhances-reactor safety.' It clarifies who has certain responsibilities,

- in the absence of. key people. Lines of succession for emergency - situations- are already specified L in ' the' '

+

emergency plan and implementing procedures.

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29-Section Chano2 6.5.D.1 The position of Reactor Operator can clearly be fulfilled by a licensed senior reactor operator as competently as c licensed reactor operator. Therefore, this change does not lead to any reduction in reactor safety.

6.6.A.5 Stating that the Reactor Operator must complete the startup checklist is unnecessarily restrictive. Any licensed reactor operator or senior reactor operator, or a reactor operator trainee or nuclear engineering student in the presence of, and under the direction of, a licensed operator -

is allowed by the NRC's regulations to perform such a function.

6.6.B.2.c Reducing the range of travel for- a Reactor Supervisor ensures that he/she can return more expeditiously to the Radiation Centerif necessary. Therefore,if this change has any effect at all it is beneficial.

6.6.B.4 Adding the Reactor Administrator to these lists increases the number of people who may suspend reactor operations j and authorize subsequent restart, thereby giving further opportunity for safety review.

6.6.C This nsw section addresses objectives and requirements of the USNRC, the state of Oregon, and Oregon State l University with respect to maintaining an environment free from the effects of drugs and alcohol. In addition, the l fitness for duty requirements f or licensed reactor operators are clearly stated. Each of these changes can only lead to greater reactor safety, by alerting staff and operators to the high standards expected of them.

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Section Chacan f 0.0.0.4 The initialintent of this requirement was to ensure that the person operating the reactor was not distracted by 1

students in a class. Thus, it was meant for larger classes and not for situations where there woro just one or two poopio with whom the operator was interacting. This -

chango clarifies this original intent.

l 6.7.B.7 This section providos a more accessible reiteration of statements already included in the OSTR Physical Security Plan and therefore, does not affect roactor safety.

l 6.8 6.10 These changes will be addressed together as they aro Inte: rotated. Roducirsg the number of piccos of paper relating to a particular change miniinizou the chances of confusion, and reduces duplication. Even though thor < is j now only ono document related to each chango, there are, in f act, more people reviewing each document. Under the now process, normally every member of the ROC and overy reactor operator reviews, and as appropriato, approves tho l changes. In the past, most of the responsibility for assessing and approving a 50.59 safety evaluation was carried out by a quorum of the ROC The now process gives the ROC great 6r initial involvement in changes associ-ated with the reactor and, agaln, contributos to increased >

safety by reducing the chances that some important issue t

l will be overlooked.

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c. UPDATE OF APPENDIX B TO THE RADIATION CENTER AND OSTR EMERGENCY RESPONSE PLAN (1) Description Appendix 83 of the Radiation Centor and OSTR Emergency Response Plan contains representative listings of inventories of emergency equipment availabio at various onsite and offsite locations. A now Appendix B was issued in order to koop thoso equipment inventories up-to date.

(2) Safety Evaluation Updating Appendix B has no effect on reactor safety and doos not reduce the offectiveness of the OSTR Emergency Responso Plan, in some small measure it increases the reliability of the plan t y keeping it current, and thus -

enhancos safety.

A review by the reactor operations and health physics groups determined that the indicated distribution of emergency equipment was more than adequato for any emergency addressed in the Emergency Responso Plan. In

)

the unlikely event that extra equipment is needed, more is available in the Radiation Center. This extra equipment is used for teaching purposeh in the Nuclear Engineering Department's radiation protection classes,

d. CHANGE OF ACCEPTABLE pH RANGE FOR REACTOR PRIMARY WATER AND BULK SHIELD TANK WATER (1) Description The acceptable pH range for the reactor primary v'ater was shown in OSTROPs 7 and 13 as 5.0 7.5. The acceptable ranga Sr the bulk shield tank water was inconsistent, being 6.0 6.5 in OS* WOP 10 and 5.0-8,5 in OSTROP 13.

Changes in pH instrumentation and in the method of measuring pH resulted in some of the pH readings for the reactor water being close to, and slightly in excess of the upper limit of 7.5. Investigations over a few months led to n  ;

. - .- - - ~ - - - - - - - ~ _ - - . . - - - . -- ._ .

d 32-b the conclusion that there was really no change in the reactor water pH, but that the method of measuromont was what caused the chango in readings.

In the past, the procedure required waiting for a stable reading before recording the pH. Howevo.', under the new procedure the reading is take,'

very quickly. Taking the reading immediately provides a more representative vatuo by avoiding the change in pH due to CO, absorption from the air.

Discussions with General Atomics and an OSU chemistry professor led us to the conclusion that the rango of acceptablo pH values could easily be expanded with no increase in the potential for corrosion. The reactor staff, therefc,ro, revised the procedure to allow the acceptable pH rango for both the reector water and the bulk shloid tank water to be 5.0-8,5.

(2) Sofoty Evaluation Investigations of the pH issuo have revoated the following information. The Nuclear Safety information Contor in Oak Ridge made the statomont in 1969 that it was impossible to maintain the pH of swimming pool type reactors between 6.5 7.0 (as many facilities were trying to do) due to the carbon j dioxide from the air dissolving into the water. They also stated that pH measutomonts and pH control were difficult with water conductivity below 5 mho/cm (ours is about C.5 niho/cm). General Atomics agreed with this and further pointed out that pH readings obtained at low conductivity are very much a function of such things as the technique used and the electrode  ;

history of r.he pH instrumentation.

It should also be noted that there was not any real change in the pH'of tho water. The recorded results changed simply because the readings were taken immediately af ter taking the sample rather than waiting for the valuo to stabilize. pH measurements similar to those obtained on the old pH instrument can be read on the new pH instrument if the water is allowed to stand as before.

0 .

J 33-The key to minimizing reactor component corrosion is maintaining a low conductivity, and not trying to control pH. It did not seem to be prudent to i

add chemicals to the reactor water or the bulk shield tank water in an attempt to control or odjust pH. In particular,it was felt that the addition of potassium chromate to the bulk shield tank (as was written in OSTROP 10.4) should not be allowed due to the hazardous nature of that material.

A member of the ROC whois a professor of chemistry was involved in these discussions on pH and recommended that the acceptable range for both the reactor tank water and the bulk shield tank water be 5.0 8,5. He concluded that there will be no chango in corrosion ratos for wetted components within this range.

e. EMERGENCY RESPONSE PLAN REVISIONS (1) Description Several changes necessitated minor revisions to the OSU Radiation Conter and TRIGA Reactor Emergency Response Plan. These changes are grouped below under the appropriate heading, and are referenced to the page on which th6 change or changes occur.

Increase in the Licensed Power Level of the OSTR The maximum licensed steady state power lovel of the OSTR was raised from 1000 kW to 1100 kW. This necessitated one change.

11 Chango 1000 kW to 1100 kW.

Chance in Resoonsibilities of the Emeraency Director and Emergency Coordinator it was decided that the Public Information Of ficer (PlO) should report to the -

Emergency Director rather than the Emergency Coordinator. Thic i

necessitated several changes in the emergency plan and many changes in the emergency implementing procedures. However, only the changes in the plan need to be addressed here.

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• 34-37 "vi) Notify the Public Information Officer and koop him/ hor informed regarding the emergency situation" was added to the list of Emergency Director duties. '

37 "and the Public Information Officer" was doloted from item il) of the Emergency Coordinator duties.

3 10 in the list of PIO duties, "li) Coordinate with the Emergency Coordinator (and Recovery Operations Coordinator)...", was changed to "ii) Coordinate with the Emergoney Director...",

3-11 "and Public Information Officor" was doloted from item iv)in the list of Recovery Operations Coordinator duties.

3 14 The chart was revised to show the line to the PlO going from the Emorgency Director rather than the Emergency Coordinator. The Radiation Contor Director box was removed.

Removal of the Gamewell Fire Annunciator Syjilem The Gamewoll fire annunciator system is no longer used by the City of Corvallis Fire Department. This means that fire alarms from the Radiation Contor now will no loriger sound in the City 911 Dispatch Offico. However,

they will sound inside the OSU Polico and Security Dispatch of fico. The OSU l

dispatcher will then notify the Fire Dopartment.

7-6 The last sentence in 7.2.1.d was changed to road "For examplo, OSU Police and Security will roccivo physical security and firo alarms directly."

85 The first two sentences of 8.2.5 Fire Doctectors, paragreph b), '

woro changed to road "If one of the fire sensors detects a fire, l an alarm will be automatically ac:Nated, not only in the building, but also at OSU Police and Security Dispatch. The OSU dispatcher will then call the Corvallis City Dispatch who will then alert the nearest available fire engine crow."

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s 35-(2) Safety Evaluation The safety considerations related to the increase in licensod reactor power level have been previously evaluated and the change was approved by the USNRC. Thorofore, this change moroly brought the plan up to-dato, and is not discussed further in this document.

Discussions betwoon the Radiation Con'or Director and the Reactor Administrator led to the decision to have the PlO report to the Emergoney Director rather than the Emergency Coordinator.11 was rocognized that the Emergency Coordinator would be very t,usy coordinating actions betwoon many groups, while the Emergency Director was relatively uninvolved in the emergency. It is a logical assignment for the Emergency Director because the press releases havo to be approvod by this individual, and also because the Emergoncy Director coordinatos with the University administration. The change of responsibility should froo up the Emergency Ccordinator without unduly burdening the Emergency D:rtetor, and thereby will result in an increase in safety.

The fact that the fire alarm signal does not directly annunciato at the City's 011 Dispatch Contor does not reduce reactor safety. OSU Police and Security will immediately call the Fire Department who will dispatch the nearest fire crew. This will normally be a team from the local sub-station which is located a few hundred yards from the Radiation Contor. The total response time will not be appreciably changed. It should be noted that all of the other alarms which are annunciated outside of the Radiation Contor also go directly to the OSU Police and Security Dispatch, and therefore, this chango does not present a new situation with rcspect to emergency responsa.

(3) Conf yming Procedural Changes Conforming procoeural changes to the Emergency Responso implementing Procedures (EHIPs) were also mado.

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3. 10 CFR 50.59 Channes to Reactor Evoeriments There were no changes to reactor experiments during this reportin0 period.

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