ML19325C700

Annual Rept of Changes,Tests & Experiments Performed Under Provisions of 10CFR50.59 for Triga Reactor,For Jul 1988 - June 1989

ML19325C700
Person / Time
Site:	Oregon State University
Issue date:	10/02/1989
From:	Andrea Johnson Oregon State University, CORVALLIS, OR
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 8910170192
Download: ML19325C700 (22)

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4 Octo'ber 2, 1989 I jh[ yn l w n

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? 'U.S; Nuclear' Regulatory Commission '

h ' ATTENTION: Documentf Cont'rol' Desk i

A , Washington, D.C. 20555 i

,. ' Gentlemen: '

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Subject:

' Annual Report' of Changes,iTests and Experiments Performed Under -

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'the: Provisions of 10 CFR 50.59 for the Oregon State University

-TRIGA' Reactor (OSTR).. License No.-R-106, Docket No. 50-243.  ;

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-The following report:is submitted in accordance with the requirements

'of'10 CFR 50.59(b) and 10 CFR 50.4, and covers the OSTR's annualt reporting period of: July- 1,' 1988 through' June. 30, 1989. The information in this report is. compiled. annually and is submitted to the USNRC in this specific ,

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. jl0 3CFR150.59(b)freport, as well as in a special section of the'0STR annual-report, which is submittea in mid-September of each year.  ;

~y' 'During the specified reporting period there were eight changes to the ,

ireactor: facility, four; changes to the reactor procedures, and three changes

, n to' reactor experiments conducted pursuant to 10. CFR 50.59.-. There were

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? , lno' tests, and no new' experiments performed under the provisions of 10 CFR

- 50.59 during the current reporting period.  ;

.TheLindividual changes being reported are listed below by category . J

, , 'and by' title, and are described in more detailln Attachment A. Regarding L this' attachment, you will note that it includes a brief description of

-_each change' followea by a summary of the safety evaluation conducted _ for ithe described change. As required, none of the changes performed under

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the; provisions ~of-10 CFR 50'59 required a change in the OSTR Technical

• o DSpecifications or involved an unreviewed safety question as defined in '

4 .10CFR'50.59(a)(2).

t y 4 1. Changes to the Reactor Facility:

n . :a. Installation of a Cadmium-Lined In-Core, Irradiation Tube t

! b. Changes to the Cadmium-Lined, In-Core, Irradiation Tube

• $] c. Removal of an Experimental Water R 1dioactivity Monitor 4*7 s

d. Addition of a Particulate Filter Down Stream of the Demineralizer Tank

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e. Replacement'of the Water Conductivity Monitoring System.

- f. Monitoring of Reactor Power and Fuel Temperature with a CDAS During Non-Pulsing Operations ,

' g. Monitoring of Reactor Peak Power and Fuel Temperature with

'a CDAS During Pulsing Operations

h. Installation of an Exterior Light on the East Wall of the Reactor Building.

2. Changes to Reactor Procedures: -

i a .~ Revision.of OSTROP 6.0

b. Revision of the Emergency Response Plan

c. Revision of OSTROP 10 4

d. (Additional) Revisions of OSTROP 6.0

3. . Changes to Reactor Experiments:

a.- Revision of OSTR Experiments B-3 and B-11

b. Revision of OSTR Experiment B-31

c. Revision of OSTR Experiments B-3, B-11 and B-12 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, Di k.v.,v w -

A. G. Johnson Dir .or, Radiation Center

AGJ/ef/50.59/031 Enclosure cc: Regional Administrator, Region V, USNRC, Walnut Creek, California 4

OSTR Project Manager, USNRC, Washington, D.C., ATTN: Mr. Al Adams Oregon Department of. Energy, Salem, Oregon, ATTN: Mr. Harry Moomey 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

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I N ATTACHMENT A Ch'anges to the OSTR Facility, to Reactor Procedures and'to Reactor Experiments Performed Pursuant to 10 CFR 50.59 The.information contained in this section of the report provides a' summary of' the . changes performed during the reporting period under the provisions

'of 10 CFR 50.59. For each item listed, we have included a brief description 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 ,

i Technical Specifications and did not-introduce or create an unreviewed safety quescion as defined in 10 CFR 50.59(a)(2).

1. 10 CFR 50.59 Changes to the Reactor Facility There were eight changes to the reactor facility which were reviewed, approved, and performed under the provisions of 10 CFR 50.59 during -

the reporting period,

a. 1riSTALLATION OF A CADMIUM-LINED, IN-CORE, IRRADI ATION- TUBE ,

(1) Description The reactor operations staff built and installed a cadmium-lined irradiation tube which can be-permanently positioned in the reactor core. As shown in Figure 1.0, the facility consists of an air-filled aluminum tube with an offset bend, which is inserted into a convenient B-ring core grid position.

The cadmium is approximately 0.025 inches thick and is perma-nently encased-in aluminum inside and out. The tube is poti-tively secured near the top to a steel beam (the center channel) which extends across the reactor tank, and is equipped with a cap on top to seal it during reactor operations. To eliminate any small pressure increases due to radiolytic gas production,

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END PLUB AND LOCATING PIN Figure 1.0 Cadmium-Lined, In-Core Irradiation Tube

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< hi" Ug a suction is1 drawn on the irradiation tube by connecting hI a.line.from the rotating-rack' vent' system to.a tee-section-on-the irradiation tube.

NOTE: 0STROPfl0 was revised'to detail'the procedures for using the new irradiation facility, and experiment B-3 was

-modified-to allow:1rradiations in the cadmium-lined tube to be performed under this. experiment. These changes, which wereLalso made under 10 CFR 50.59,~ are' discussed later in ~ q this report.

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(2) Safety Evaluation v The safety considerations for this facility were very similar to those previously evaluated for the. cadmium-lined pneumatic l transfer' tube which was installed under ROC approval and

.then recently removed from the reactor core. The reactivity l changes were expected to- be about the same, or slightly more

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' negative'because a similar amount of cadmium was used and

= this facility was placed in a core location with a slightly higher flux. The cadmium-lined tube was therefore estimated' l

.to be Worth-about'-$2.20_and was expected to reduce-the core excess from about $6.50 to about $4.30. As-indicated earlier, the tube.was-secured (bolted)'to the' reactor facility in such a manner that it could not be easily or unintentionally lx removed. This will prevent any sudden, unplanned addition k, of reactivity to the. reactor.

l The tube was constructed in such a manner that the cadmium s was completely sealed within an inner and outer aluminum ,

o tube. Therefore, there will be no potential for cadmium contamination of the samples or the reactor.

t The offset bend in the tube is similar to that of the other r

in-core facilities and effectively eliminates radiation stream-E

-ing from the tube.

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s The procedures'for the facility stipulate that the tube's

' cap is.to be on the top of the tube except when samples'are b" being inserted.or removed'and whenever the reactor is in

, operation. Additionally, under normal circumstances, samples K .will-be removed from the tube only after the reactor has-b, been shut down for a time period adequate to ensureithat

! the bulk of the argon-41 activity has decayed (usually over- '

night). .This~will prevent any unnecessary release of argon-41 to the reactor bay. The tube could be unloaded after a shorter 3

!l decay period with specific health physics approval and appro-priate precautions, but this option is no different than that currently. employed with the rotating rack and therefore introduces no'new safety considerations.

1 The proce'dures and limitations for encapsulation and irradiation-l of samples using the in-core cadmium-lined facility follow e ~

L current requirements, particularly those in 0 STROP 18. There-fore, no new or untried practices were introduced relative  :

.to.the actual use of the new facility, a In order to position the cadmium-lined end of the tube into .

the core's B-ring, a fuel element must first be moved from '!

n grid. location B1 to G6. The cadmium-lined tube can then l l- be. inserted into the vacant B-ring position. The' estimated reactivity.effect at each stage of the move was calculated

'and is given below:

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1. Removal of the element from B1 = -354  :

11. Insertion of this element into G6 = +30&

iii. Insertion of the cadmium-lined tube into B1 = -$2.20 .

1 iv. Overall-reactivity e M nge = -$2.25  !

1 Core excess measurements were made at each step of the tube 1 insertion procedure outlined above. The control rods were recalibrated after the tube was inserted and the core excess was remeasured.

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I b.- CHANGES-TO THE-CADMlUM-LINED, IN-CORE, IRRADIATION TUBE h .

(1)? Description.

Following Reactor Operations Committee approval.~of the facility

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-change regarding'the' cadmium-lined..in-core irradiation tube, y 'the staff recommended two changes related to the irradiation tube. ~The first change involved the n,ethod to be used for 1 insertion and removal of the sample. support. seat and the

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outer container holding the encapsulated sample (s) to be-

] 1 irradiated. .This change was suggcsted.because after the .j

-in-core . irradiation facility was constructed it was discovered

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that'the standard TRIGA tube handling device used to insert

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and remove TRIGA tubes for the rotating rack would also easily- j u go down the new irradiation tube provided the handling device's  !

n' outside borated shield was removed. Therefore, instead of a previously proposed, less desirable method for sample handling it war now proposed that samples be placed in the cadmium-lined facility in. aluminum TRIGA tubes modified to have internal U threads so that the-containers could be inserted and removed i using the standard TRIGA tube handling device (fishing pole ,

and grapple) and standard procedures developed for the rotating R

rack. Simil'arly, the: top of the sample support seat-was modified-by adding an aluminum TRIGA tube cap so that this too could be easily inserted and removed using the TRIGA j l, tube handling device. j The second change involved moving the location of the facility's 1 air suction' tube-to the irradiator tube cap, rather than L, having a "T" in the tube itself. It was determined that this change would allow much more flexibility in the angular j positioning of the air tube.

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(2) Safety Evaluation It was judged that the first change improved the operational safety of the facility. The bhsis for this was related to

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Lthe fact that the previously planned use of wire:and cord

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to position and support samples was deleted, and with this-deletion went any potential for-activation'of wire, risk-e(( ,

Fo of: wire brei.kage; and risks' associated with handling wire or cord'when removing samples from thettube. Conversely, 5

g the new method for sample insertion and' removal employs cur-

.rently approved procedures which have been .in practice at

. .the:OSTR for many years. The new design of the sample support- j seat also enables the' staff to easily remove the seat when i

[ not in use, thus preventing unnecessary activation.

e There are no unfavorable safety implications involved with  ;

the.second. change. A slight air suction will still be pulled 1 on the tube when it is in the core.- Repositioning the air f

suction tube to the cap simply. allows this tube to be oriented in different directions and thus eliminates the inconvenience of having'the tube in the way of other work which could be .i  !

going on in the immediate area. j

. .c. REMOVAL OF AN EXPERIMENTAL WATER RADI0 ACTIVITY MONITOR .

'(1) Description.

' As part .of. a graduate research project conducted approxi- ]

mately 8 years ago, an independent water monitoring loop  ;

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l-- was installed in the demineralizer circuit of the primary water system. The water monitor did not perform satisfac-torily, and once the project was finished, the monitor was never used again. The reactor staff removed the water moni-toring loop from the demineralizer circuit. Specifically, i all of the piping.and equipment between valve DV 17 and the tee-section where valve DV 22 was attached to the main water i u- line-for the demineralizer circuit was removed (see Figure 2.0). The tee was plugged and valve DV 17 ~ remained in the l system as a sample collection valve. This change resulted in a revision of OSTROP 7 to remove all mention of this water monitoring loop.

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.- m apr et Figure 2.0 Removal of an Experimental Water Radioactivity Monitor from the Primary Water Purification System

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y-r .(2)1 Safety Evaluation-N *- lThere.were no unfavorable safety implications associated '

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-with this'facil.ity change as this 1000 was never used to

,' -support reactor operations and the original water monitoring  ?

system for the reactor always remained fully. functional. *

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Appropriate health-physics: precautions' were taken during the removal of- the equipment. No significantLcontamination t was found during removal.due to the long elapsed time since  :

,, .the loop was used and because of the low radioactivity,concen-

1. i trationiin the reactor. primary water.

d, ADDITION 0F A PARTICULATE FILTER DOWNSTREAM 0F THE DEMINERALIZER TANK:

L(1) Description The reactor staff installed a particulate filter downstream .

of the demineralizer tank in the reactor primary. cooling

-water. cleanup system. The new filter. prevents resin fines

, from being-introduced into the primary water system. .

t 7e filter assembly is mounted next to the east wall of the heat exchangor room adjacent-to the demineralizer pump' skid.

The: pipe and valving system (see Figure 3.0) is primarily' 1" plastic'p'pe with three.valvea (two shutoff valves and 7

a-bypass valve), two pressure gauges (to' measure the pressure i drop across the filter) and a drainable housing for the 25 micron filter. The filter housing can be shielded by concrete blocks,' if needed.

(2) ~ Safety Evaluation c

There are no unfavorable safety implications related to the addition of this new filter. All of the materials used are of good quality, and the system passed all preoperational tests. The new filter is essentially no different than the existing particulate filter for the primary water which has i

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j p $ _ ADDE noal w se r arm m Apr et Figure 3.0 . Addition of a Particulate Filter Downstream of the Demineralizer '

in the Primary Water' Purification System

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been'inisuccessful operation since-the OSTR was-built. Proce-

~-dures used-to change the new filter cartridge are the~same e) .

The new filter contrib-as- those'used for the current filter. _;

l .utes to increased. safety by removing any resin fines which l

might pass fromctheidemineralizer tank intoithe primary water system.>  !

e. REDLACEMENT: OF THE. WATER CONDUCTIVITY MONITORING SYSTEM L(1) Description ,

The _retctor staff upgraded the reactor water conductivity monitoring system to a digital readout instrument with auto-matic temperature cc.apensation.

While the original conductivity monitoring system was still-

< quite reliable, it had one disadvantage related to the. fact that the= " cat eye" indicating tube in the conductivity system had to be changed periodically. These old-style " cat eye" electronic tubes were becoming increasingly expensive and  :

could possibly becomc vnavailable in a few years.

In order to install the new system the-following. changes were made:

L - 1. Two new signal cables were pulled from the heat exchanger room to=the reactor console. The original cable did not have enough wire pairs to accommodate the temperature L compensating feature of the new instrument.

.11. New temperature compensating conductivity probes were installed.

iii. Adapters were installed so that the ucw probes would fit into the water system piping. <

$ iv. Modifications to the console side cabinet were made i

t:> physically cecommodate the installation of the new digital readout instrument.

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( '(2) Safety'Evaulation, .

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There is a. great deal of historical data documenting the.- ,

normal range of conductivity for the OSTR primary water.

Any significant. discrepancy observed between established -l wb' values and results obtained with the new system could,and -l m A would be immediately investigated using other conductivity- ,

1 ' instruments.

e Failure'of the new system also does not create any immediate safety implications; since.the conductivity of-the reactor l primary water changes very slowly with time, and thus allows

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. plenty of, time for detection of abnormalities and repair i- of conductivity. equipment. In addition, the OSTR primary water is kept at an extremely. low level of conductivity, which gives-an even greater margin of protection.

l Because=this is new equipment, it is expected that the potential 1

for failure will actually decrease and that this installation

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L - will provide more accurate conductivity readings. Hence, the new device will actually increase reliability and safety.

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f. MONITORING OF REACTOR POWER AND FUEL TEMPERATURE WITH A CDAS DURING ~

NON-PULSING OPERATIONS l l

(1) Description l 6 On an as-needed basis, a computer-based data acquisition J system (CDAS) can be connected to test terminals on the OSTR j console to passively measure reactor power and fuel temperature l signals curing non-pulsing operations. Terminals TP2 and j

, TP3 would be used to measure fuel temperature from the fuel thermocouple amplifier board (card XA16) in the left-hand console drawer. These terminals ware extended to the rear

, of the console for easier and safer access. The LINEAR and )

1 LOG terminals already on the rear of the console would be l used to measure the linear and log reactor power, respectively. ]

No active circuitry will exist between the test terminals l and the CDAS. j i

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- .The referenced terminals are designed to be used as indi-gv, cated above;- however, additional measures will be taken to s . .

f .7 ensure the safe use of the CDAS while measuring reactor power ,

, and. fuel temperature during non-pulsing operations. First, a 100-ohm resistor was permanently installed between console- ,

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terminals TP3 and TP4 for the Nyt circuit, thus ensuring }

H proper connection'of the CDAS or other external recording [

j' devices to-these terminals. Second, the CDAS system will. f be operationally tested to assure its proper functioning before it-is connected to the reactor console. This will w

<' ,* eliminate the possibility that the CDAS will be connec'ced to the console with an improperly functioning or failed compo-C nent.' In addition, cables used to connect the CDAS to console terminals were labeled and color-coded to minimize the possi-bility of an incorrect connection.- It is important to note, i i however, that an incorrect connection would create no safety m problems and would not affect console electronics. The impact would be an incorrect result on the external recording device.

.Furthermore, extending terminals TP2 and TP3 from the thermo-couple amplifier board (card XA16) to the rear of the reactor console will increase safety by making these terminals more accessible, by eliminating the need to directly access elec-o tronics in the left-hand' drawer of the console, and by making itieasier to confirm that the connections are correct.

An evaluation of the worst-case failure of the CDAS while connected to the reactor console to record reactor power

. and fuel temperature indiceted that the consequences are no worse than those created by the failure of an existing

, y console component in the same circuit, and such consequences would be immediately obvious to the reactor operator so that appropriate action could be taken.

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. A new OSTROP 26 was written and approved.. It details operating:  ;

procedures-for the CDAS when it'is being used to measure reactor' power level and fuel temperature, y # - g;. MONITORING 0F REACTOR PEAK POWER AND FUEL TEMPERATURE WITH A CDAS t

b DURING PULSING.0PERATIONS

[ -(1) Description The '0STR staff installed an operational amplifer inside the o reactor console cabinet.- The amplifier provides a gain of g about 100 to ' amplify the peak power (Nvt) signal taken frem i console ~ terminals TP3 and TP4. Signal amplification at these

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terminals-is fr% 7 bout 60 mV-to about 6 V. Although the camplifier circuit nas been permanently installed inside the reactor console cabinet, it will be" connected to Nyt circuit ,

1 terminals TP3 and TP4 only as needed during pulsing. Also, as~needed during pulsing operations, a computer-based data acquisition system (CDAS) will be connected to the output from the operational amplifier to measure the peak reactor

. power during' a pulse.

. -To measure fuel temperature during pulsing, the CDAS will

.also be connected to fuel temperature output terminals TP2 <

and TP3-on the back of the OSTR console in order to passively ,

measure fuel temperature signals from the thermocouple amplifier board (card XA16) in the left-hand console drawer. No active circuitry will exist between the fuel temperature terminals ,

and the CDAS.

< (2) Safety Evaluation L The referenced terminals are designed to be used as indi-cated above; however, additional measures have been taken to ensure the safe use of the CDAS and amplifier during pulsing operations. First, a 100-ohm resistor has been permanently installed between console Nyt terminals TP3 and TP4, thus

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. _other external recording devices to these terminals. Pende nentlyLinstalling a 100-ohm resistor across terminals TP3 p ,

and TP4 increases the- system's reliability by eliminating W*- the need .for a' jumper cable across TP3 and TP4. Second,

.' the amplifier and the CDAS system will be operationally tested-to_ assure ~ proper functioning before they are connected to the reactor console. This will eliminate the possibility j 7

.that these devices will be connected to the console with

-an improperly functioning or failed component, which will thereby eliminate the chance that a signal from the amplifier .

-or CDAS will affect console electronics. In addition, cables used to connect the amplifier to the console and the CDAS l

• Lto the amplifier were labeled and color-coded to minimize the possibility of an incorrect connection. It is impor-tant to-note, however, that an incorrect connection would create no safety problems and would not affect console elec-LP tronics. The impact would be an incorrect result on the I external recording device. Furthermore,'only cables from F input channels of the covered distribution board for the CDAS will be present, and thus no output signal cables will

• be available for connection between the CDAS and the reactor console. As en added feature, protective diodes were added to the amplifier circuit to isolate the CDAS and the amplifier .

from the reactor console. This action will prevent the CDAS or amplifier from introducing a measurable charge to the .

Nyt circuit capacitor.

An evaluation of the worst-cose failure of the CDAS and amplifier while connected to the reactor console to record peak pulse power and fuel temperature indicates that the consequences

, are no worse than those created by the failure of an existing console component in the same circuit, and such consequences would be immediately obvious to the reactor operator so that appropriate action could be taken.

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, 'InLaddition to the.above safety considerations, fuel temperature:

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monitoring during pulsing.is described and fully evaluated by the 10 CFR'50.59 evaluation entitled ',' Monitoring of Reactor.  !

Power. and Fuel . Temperature with a CDAS. During Non-Pulsing 3 g , Operations," (see Attachment' A, section 1.f). .

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Operating-procedures-for the CDAS and amplifier when used ,

h during' pulsing are~ included in the new OSTROP 26, " Procedures l

.for the Use of External Monitoring and Recording Devices."

p h '. INSTALLATION 0F AN. EXTERIOR LIGHT ON THE EAST WALL OF THE REACTOR BUILDING-(1) Description

,In order to provide additional lighting for the northern half of- the Radiation Center's parking lot, and to simulta- ,

neously enhance exterior lighting on the east side of the reactor building, an exterior light was installed.at the north end of the east wall of the reactor building.. Electrical ,

power was supplied by extending a conduit from an interior east' wall electrical outlet located adjacent to the east  ;

fire exit door.

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-(2) Safety Evaluation The conduit penetration in the east reactor bay wall, as

- described above, was filled with an electrical conduit which was appropriately caulked inside and out to prevent air leakage.

Therefore, this facility change does not affect the ability of the reactor building to maintain the originally designed L containment integrity .and does not alter the probability of minimal radiological releases as described in the facility SAR. Consequently, the change introduces no increase in the probability or consequences of occurrences evaluated in the facility SAR. Furthermore, no new types of occurrences are introduced and no margin of safety is reduced by the proposed change.

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'2. 10 CFR 50.59 Changes to Reactor Procedures There were'.four chang'es to reactor procedures' which were reviewed - .

g.  : approved, and performed under the' provisions of 10 CFR 50.59 during ithe. reporting _ period.- .

a.-' REVISION OF 0 STROP 6 (1)-'D$scription OSTROP 6.0 was modified to-incorporate revisions necessi-

-tated by recent organizational and procedural changes. In particular job descriptions were modified for the Radiation '

Center Director, the Reactor Administrator, and most of the health physics; group. The revised ROC charter was also incor- I porated into the procedure.: In addition, the access control procedure for the reactor bay was changed to incorporate

, the daytime use of. signs when the security alarms are set during the day.

n (2) Safety Evaluation i

There'are no unfavorable safety implications associated with 1 the job description changes. All of the necessary responsibil- j ities are well-covered, and the organizational changes have n

already been approved by the NRC and incorporated into the OSTR Technical Specifications. The changes to the ROC charter were also previously approved.by the ROC, and these changes were merely being incorporated into OSTROP 6.0. The slight change to the reactor bay access procedure will not affect the. security plan for the reactor. The change only affects those people with reactor bay keys and these people are the a

most responsible members of the Radiation Center security ,

staff. Even if the procedure is not followed, security is O' not compromised, only an alarm is sounded unnecessarily.

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>, b. REVISION OF THE EMERGENCY RESPONSE PLAN

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(1)~ Description

' A' number'of changes to the;0STR; emergency response plan were made as a-result of the annual review of the plan by the

1. 1. standing' subcommittee of the ROC. This review'was conducted

~on October 26, 1988. Most of the changes were required.as a result of the revised Radiation Center organization'recently approved by the NRC. Remaining changes were merely updates of such things as telephone numbers,.first-aid qualifications, 4 etc.

4

, (2)L Safety Evaluation None of the changes involve revisions of the actual er.ergency response described'in the plan. Instead, the changes simply

= update the plan to incorporate the current titles and organiza-tional. structure of the Radiation Center, and involve appropriate-modifications to designated lines of succession in the emergency plan. As a result, none of the changes have any impact on safety. Changes to the plan were also reported to the USNRC

-under-the provisions-of 10 CFR 50.54(q). i

c. REVISIONu0F OSTROP 10  ;

(1) Description-y

' The reactor staff amended section 10.7 of OSTROP 10 to specify the procedures for moving the cadmium-lined in-core irradiation o

tube from its storage' location in the S-rack to the in-core position, and for returning the tube to the storage location.

In addition to amending section 10.7, the staff changed the title of OSTROP 10 to " Operating Procedures for OSTR Irradiation Facilities."

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The~ reactivity effects of-inserting and' removing the cadmium-lined in-core irradiation tube-have already been addressed -

lin a previous 10 CFR 50.59 safety' evaluation. The required use of; a. specific set of control rod calibrations depending on whether the tube-is in or'out of the core will ensure -

that accurate-measurements of core excess and shutdown margin

, are made. .A= 2( limit for comparing control rod worths is i also included in revised section 10.7 of.0 STROP 10 because i this is the estimated error.for a rod calibration.

'With respect to movement of the tube in the reactor tank, j aluminum TRIGA tubes-filled with. lead shot'will be placed L' inside the ' cadmium-lined in-core irradiation tube and the .

tube's top cap will be sealed to ensure that the tube is' ,

neutrally bouyant. Therefore, the impact of accidentally.

releasing th'e tube'will be minimal. In addition, the tube will not be over the core when it is passed under the center L channel, and:as a~ result there will be no possibility of

. dropping.the tube on the core,

d. (ADDITIONAL) REVISIONS OF OSTROP 6.0

.(1) Description

' Three (additional) amendments were made to 0 STROP 6.0, " Admin-istrative and Personne1' Procedures." The first amendment states that no external measuring or recording device will be connected to reactor measuring channels or safety channels without ROC approval of a 10 CFR 50.59 safety evaluation and any needed operating procedures. However, it was not intended that ROC approval be required for normal use of standard diagnostic equipment by the Scientific Instrument Technician or a designated replacement. A second amendment to OSTROP 6.0 states that when classes are in the reactor control room, the operator of record will not be the instructor

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L Finally, a third amendment was added to ensure

of the class.

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that any connection of an external' system to reactor measuring.

or safety channels will be checked by.the Reactor Supervisor, p' , Again,.this requirement was not intended to apply to the D

e normal use of standard diagnostic equipment by the Scientific U Instrument Technician or a designated replacement.

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All'of the three revisions.were made to increase safety, and'therefore the' safety implications are all positive. 1 The first amendment' noted above will. prevent. the addition

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E of systems which might' affect the reactor measuring-or safety 4 channels. The second amendment will decrease the possibility k of the reactor operator being distracted by also having to  ;

instruct a class. The third amendment will help to ensure -l that any. connections mada to important console systems are lmade correctly.

3, 10 CFR 50.59 Changes to Reactor Experiments

.There-were three changes to reactor experiments which were reviewed.

approved, and performed under the provisions of 10 CFR 50.59 during

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ithe reporting _' period.-

a.

REVISION' 0F OSTR EXPERIMENTS B-3 AND B-11

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- (1) Description  ;

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' Experiments B-3 and B-11 were revised to add the new cadmium-lined in-core irradiation tube as one of the standard OSTR irradiation facilities.

(2) Safety Evaluation The installation of the cadmium-lined tube in the core provides  !

a new standard irradiation facility. From an operational x and health physics standpoint, irradiation of samples in

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ithis facility is no differentithantirradiating samples in

~ cadmium cups in the (dummy) _ sample-holding _ fuel element, i ~

.in the rotating rack or.-in the_ pneumatic transfer' facility.

> a' -In_ fact,lusing the new tube will enhance safety by reducing j fradiation doses associated with the use-and handling of cadmium i

, cups, which will not 'now be needed for many experiments.

All irradiations will be controlled following the usual proce -

dures associated =with irradiation requests.- ,

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b. - REVISION OF OSTR. EXPERIMENT- B-31 (1) Description Experiment B-31 was revised to expand the types of niaterials ,

that can be.activ'ated in OSTR facilities for flux-mapping

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1. purposes. All' OSTR 1rradiation facilities can be used including

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.the_ reactor' tank and associated in-core locations. since i

.c , ' this was the original intent of Experiment B-31.

-.(2) Safety Evaluation-t No reduction'in safety effectiveness results from these minor- 1 revisions. Reactivity values and radioactivity-limits have

.not changed.

c. REVISION OF 0STR EXPERIMENTS B-3, B-11 AND B-12 =

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.(1) Description-

, Experiments B-3, B-11 and B-12 were revised to prohibit the irradiation of elemental mercury or substances where mercury  :

is a major constituent, j (2) Safety Evaluation

-The change increases safety by preventing the introduction of mercury into the reactor. Mercury reacts with aluminum

'and could therefore cause undesirable corrosion of reactor components.

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