ML20090B382

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Annual Rept for 1991 (CY91) for Ga Institute of Technology Research Reactor
ML20090B382
Person / Time
Site: Neely Research Reactor
Issue date: 02/24/1992
From:
Neely Research Reactor, ATLANTA, GA
To:
NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
References
NUDOCS 9203030358
Download: ML20090B382 (32)
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NECLY NUCLUAA REGEAnCH CENTER g 4a.,' / BOO ATLANTIC OnlVi2 N,o%g,gf' A TL ANh GE CAGIA 30332 048 som oua.. moo February 24, 1992 U.S. Nuclear Regulatory Commission Region II 101 Mariotta Stroot, N.W.

Atlanta, GA 30323 Reforonco: Annual Roport Docket 50-160; Licenso R-97 Gentlement Pursuant to Section 6.7.a of the Technical Specifications for the Georgia Institute of Technology Rosoarch Roactor License R-97),

tho following annual report is submittod. The repo(rting period is January 1, 1991 through December 31, 1991 (calendar year 1991).

The designation of the sections below follow the titio and order of Section 6.7.a of our Technical Specifications.

1. OPP.13ATIRNS

SUMMARY

a. Chances in Facility Desian There were three f acility design changes during calendar year 1991: one involving the annunciator panel flasher, another involving the power level measuring channels, picoanurotor #1 and picoammotor #2, and the third portains to the upgrado of the cooling tower. All throo design changes are described in Appendix A.
b. Performanco Clipractoriptics During the reporting period, the reactor was operated at power levels up to 4.5 MW using a 17-olement core. An 8-l olomont fuel exchango to enhanco self protection was performed during the reporting period. Fuel performanco has continued to be satisfactory with no known problems.
c. Changen in Operaling Progoduren The list of now and/or revised proceduros which woro l approved by the Nuclear Safoguard Committoo during l calendar year 1991 were as follows:

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U.S. Nuclear Regulatory Commission Annual Report February 24, 1992 Page 2 h qq_,_ f Tjtle

. 0004 Safeguards Events Log Entrios 4501 Fuel Element Self Protection Monsurement

!!-300 Training Requiremonto for llot-Coll Operators 3109 Instructions for Experimont Approvals 6090 Personnel Monitoring Af ter Building Evacuation in Emergency Situations 6100 Emergency Notifications 7245 Ronctor Shutdown Margin Dotormination 9037 Tritium Dotormination in Urino 9501 Control and Accountability of Radio-activo Sources 2603 Response to Loss of Electric Power 2604 Response to Inoporablo Control Elt.sont 2605 Response to Loaks in llent Exchangor 2015 Reactor Power Calibration 7245 Reactor Shutdown Margin Dotormination 7250 Complete List of Set Points for Modes 1 & 2 7220 Containment Building Isolation Test 9017 Stack Grab Samplos 2601 Response to Reactor Scram Initiated by a Safety System 2602 Response to an Alarm Annunciator i

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U.S. Nuclear Rogulatory Commission - Annual Report i February 24, 1992 Page 3 EI20. . # Title i

3600 Spoeial Nuclear Material Inventory 6100 Emergoney Modification j 9510 Radioactivo Matorial Shipment 2002 Procritical Chock List and Shift Supervisor Approval Radiation Safety Manual The' list of old proceduros which woro rescinded and rologated to '

Systems Manual by the Nuclear Safeguards Committoo in 1991 woro Proc. # Title i 5000 objectivos and Code for Emergency Proceduros 5001 Power Trip - Scram i

5002 Power Trip No. 2 - Scram 5003 Period Trip - Scram '

5004 Porlod Trip No. 2 - Scram 5005 Magnet Actuator Amplifier - Scram '

5006 Low Ion Chamber Voltago - Scram

  • 5007 Calibrate S Ltches - Scram 5008 Low D 0 Plow - Scram 2

5009  !!igh D 2OVER Tomparaturo - Scram 5010 Reactor Tank Low Level - Scram l 5012 Drain Valvos Opon - Scram 5013 No D O Overflow - Scram 2 _

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U.S. Nuclear Regulatory Commission - Annual Report February 24, 1992 Pago 4 Proc. # Title 5014 Doorn Open - Scram 5015 Reactor. Isolation Valvos Not Open - Scram 5022 High 11/3 Temperaturo-Delay Scram 5023 Low H 0 Plow-Dolay Scram 3

5024 Control Air Low Prosauro - Delay Scram 5030 Low Shlold Coolant Flow - Delay Scram 5031  !!igh Shield Coolant Temp - Delay Scram 5032 Low Bismuth Coolant Flow . Delay Scram 5033 Iligh Bismuth Coolant Temp.-Delay Scram 5052 Building Radiation fligh-Alarm 5053 Stack Exhaust 111gh Activity - Alarm 5054 Radiation liigh Vent Duct - Alarm 5055 110 3

liigh Radiation - Alarm 5056 Low Ncutron Count Rato - Annunciator 5057 D 3 0 Leak - Alarm

'5058 Procosa Room Doors Open - Alarm 5059 ECCS - Alarm 5060: Outside Sorvo Range - Alarm 5061 Regulating. Rod Low Limit - Alarm 5062 Regulating Rod liigh Limit - Alarm 5063 CW Basin Low Lovel Alarm 5064 Vent System Low Flow - Alarm l

4 l.

U.S. Nuclear Regulatory Commission - Annual Report  ;

February 24, 1992 Pago 5 Proc. # TLt)>l '

5065 Low D 0 Temperature Alarm 2

5066 liigh D 0 conductivity - Alarm 2

i 5067 liigh D 0 Conductivity No. 2 - Annunciator 2

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5066 Low 1I 0 Temperaturo Alarm 2

5010(5069)Roactor Tank Low Levol - Annunciator 5070 Low Ilolium Flow - Alarm 5071 Low Nitrogen Lovel - Alarm .

5072 ~ liigh Nitrogen Levol - Alarm t

5073 Low Recombiner Temperaturo - Alarm 5105 Electric Power Failures 5109 Inoperablo Control Element of Position Indicator '

5120 Moderator Leak in.lleat Exchangor

d. Results of-Surveillance Tests and Inanoctions The surveillance tests and -' inspection of the facility required by the Technical _ Specifications woro performed..

Documentation-of each of the tests and inspections aro +

available at-the site for review.

e. Chances. Test a_nd Experiments Apprqved by USNRC There woro no chhnges, tests or experiments that required the approval _of_the USNRC_ pursuant to 10 CFR 50.59(a).-

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U.S. Nuclear Regulatory Commission - Annual Report February 24, 1992 Pago 6

f. Chancos in Pl_ Ant _ Staff and Committoo Membership

{

Dr. R.A. Karam, Director, Nuclear Roscarch Contor Dr. Botty Rovain, Associate Director _and Acting Manager of the Office of Radiation Safety Mr. D.-D. Statham, Reactor Supervisor and Electronic Enginoor l Mr. William Downs, Senior Reactor Operator Mr. David Cox, Roactor Operator (torninated 10/28/9t; Mr. Dixon Parker, Reactor Operator Mr. Jerry Taylor, Sonior Safety Engineering Assistant Mr. Edgar Jawdoh, lloalth Physica Mrs. Clara Galleshaw i Mrd. Arlano Robinson Smith In addition to the full timo staff, the NNRC omploys the following graduate students on part timo basis: '

Mr. John Hawkinson Ms. Kathloon Kloo Mr. David llustoad Mr. Nazih Chbeir The current membership of the Nuclear Safoguards I Committeo_is:

(1) Mr. Emsloy Cobb, Chairman Disciplino: Reactor Operation and Reactor Safety i (2) Dr. Bernd Kahn Discipline Radiation Protection and 1

Environmental Measuromonts (3) Dr. James Mahaffey, Vice Chairman Discipline: Instrumentation and Control, Nuclear .

Enginooring, Reactor Operations (4) Dr. Pratoon V. Desai, Secretary Discipline Thermal liydraulics , - Mechanical Systems (5) Dr. Billy R. Livosay, Member L. Disciplino: Material Science, Physics p-L

I U.S. Nuclear Regulatory Commission - Annual Roport February 24, 1992 Pago 7 (6) Mr. Jack Vickory, Member Disciplince Security (7) Dr. Kent Barofluid, Member DJeciplino: Organometallic Chemistry (8) Dr. James Gordon, Member Disciplino: Medicino (9) Mr. Lon Gucwa, Member Discipline Reactor Safoty (10) Mr. Stovo Ewald, Member Discipline lloalth Physics (11) Dr. Peggy Girard, Member Discipline liiology (12) Mr. James Ohara, Member Discipline: Health Physics

2. E.0ER GE1LYl%T10N For the period January 1, 1991 through December 31, 1991, the total power generation of the GTRR was 276 MW hours. The reactor was operated a total of 292 hour0.00338 days <br />0.0811 hours <br />4.828042e-4 weeks <br />1.11106e-4 months <br />s: 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at power levels equal to or less than 100 kW, 243 hours0.00281 days <br />0.0675 hours <br />4.017857e-4 weeks <br />9.24615e-5 months <br /> at power lovel 100 kW to 1 MW, and 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> at power levels abovo 1 MW.
3. SHUTDOWNS During this reporting porlod thoro woro 4 unscheduled shutdowns. Table 1 gives details.

U.S. Nuclear Regulatory Commission - Annual Report February 24, 1992 Page 8 TAllLE 1 UNSCIIEDULED S110TDOWrJS DURING _1991 4

Report Date Trip Reason for Trip Correctivo Initiation Action i 91-0 2/15 Operator Criticality Nono:

Alarm Accidental trip, pool level setting was being testod when actuation took place.

91 4/29 Air During the Nono nooded.

prosauro procras of This is ,

low unisolating characteristic i containment of the system building, design.

prosauro dropa momentarily.

This happened when air pressure was near low limit causing pressure to go  :

below set point.

91-2 7/48 General Georgia Power None Power

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U.S. Nuclear Regulatory Commission - Annual Report February 24, 1992 Pago 9 91-3 10/11 Roactor IA-D2 pressuro Evaluating Tank D,0 switch measures replacement level low incromontal with moro difforontial sensitivo pressuro in a switch.

range of 12 in.

11, 0 . The switch dead-band is six inchos. Conso-quently it is not sensitivo.

Density changes in D 2 0 duo to temperaturo '

rise often actuates set point.

4. lWSCIIEDULED MAINTENANCE ON SAFETY REIATED SYSTMS AND CQMEQNENTS There woro approximately thirty minor repairs performed on safety-related systems and components. Rocords of-maintenanco performed on components are available at NNRC officos for inspection.
5. CHANGES, TESTS AND EXPERIMENTS During 1991, there woro 39-approved experiments which used the GTRR. The experiments were evaluated prior to tboir approval with regard to section 3.4 of the Technical-Specifications.
6. RADIOACTIVE EPPLUENT RElgASEfq a- Technical Specification 6.7.(6)(a) - Gaseous.

Effluents -Summation of All Roloasos via Stack, i.e.,

ground level-release.

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U.S. Nuclear Regulatory Commission - Annual Report February 24, 1992 Page 10 (1) FISSION AND ACTIVATION GASES Tritium Released (gaseous)

None Measurablo Argon-41 Roloased Total Avg Rolease Max. Instan-  % Tech

  • Releaso Rate (uC1/cc) taneous Specs  !

(C1) Release ist QTR (uci/sec) 5.369 3. 84 x 10" 190 32%

2nd QTR 11.043 7.89 x 104 106 18%

3rd QTR 17.497 1.25 x 104 171 .

29% '

4th QTR 25.304 1,81 x 104 285 49%

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  • Computation based on r:e Maximum Instantaneous Releaso Rate as ovaluated against a TS release limit of 585 uCi/ soc.

(2) IODINES RELRAEED None Measurable ,

Lower Limit of Detection <2.5 x 10*"

(3) PARTICULATES Nono Measurable Lower Limit of Detection gross beta / gamma 6.46 x 10"uci Lower Limit of Detection gross alpha 6.85 x 10**uci

b. Technical Specification 6.7(6)(b) - Liauid Effluent -

Epmmation of all Reactors (R-97) ,

1. FISSION AND ACTIVATION PRODUCTS Cobalt-60 in the only activation product released via the liquid pathway from the reactor facility.

The Co-60 does not result from reactor operations, but is attributable to material stored in the spent fuel storage pool that is part of the-State '

of Georgia Radioactive Materiale License No.-147- .

L 1. .No_ fission producta are released via the 11guld effluent pathway.

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. 1 U.S. Nuclear Regulatory Commission - AntaM1 Report February 24, 1992 Page 11 Total Avg Release *  % Tech Est. Total Release Rate (uC1/cc) Specs Error (%)

1 lat OTR 0.000015 7.50 x 1042 < 1% 7.07% ,

2nd QTR 0.000066 3.37 x 1042 < 1% 10.25% i 3rd QTR 0.000100 5.00 x 1042 < 1% 12.52% k 4th QTR 0.000014 7. 00 x 10'" < 1% 21.01%

  • Average release rate values are based on a Georgia Tech campus water discharge rate of 2.09 x 10" '

ml/ quarter '

2. TOTAL GROSS RADIOACTIVITY ( / gamma)
  • Total Avg Release *  % Tech Release (C1) Rate (u/Ci/cc) Spec lat QTR 7.30 x 104' 3.65 x 10 48 < 2%

2nd QTR 1.04 x 1045 5. 2 0 x 10'" < 2%

3rd QTR 3.01 x 1045 1.51 x-104' < 2%

4th QTR 7.55 x 10d' 3.78 x 1042 < 2%

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  • Average release rate values are based on a Georgia Tech campus water discharge rate of 2.0 x 10" ml/ quarter.
3. TRITIUM Total Avg Release *  % Tech Release (Ci) Rate (u/Ci/cc) Spec ist QTR 0.00831 4.16 x 10" < 1%

2nd QTR 0.00515 2.58 x 10d < 1%

3rd QTR 0.00004 4.02 x 104 < 1%

4th QTR 0.01617 8.09 x 10d < 1%

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  • Average release rate values are based on a Georgia l Tech campus water discharge rate of 0.0 x 10" l ml/ quarter.
4. -GROSS-ALPHA RADIOACTIVITY RELEASED None Measurable Lower Limit of Detection -

< 8.7 x 10-6 uCi/ml l

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I U.S. Nuclear Regulatory Commission - Annual Report February 24, 1992 Page 12

5. VOLUME OF WATER R2 LEASED (ml/ Quarter)

From Reactor Building 1st QTR . . . 1. 55 x 10' 2nd QTR . . . 9.27 x 10' 3rd QTR . . . 1. 3 8 x 10' 4th QTR . . . 3. 69 x 10'

6. VOLUME OF DILUTION WATER USED DURING EACH QUARTER From Georgia Tech Campus 1st QTR . . . 2.0 x 10" 2nd QTR , . . 2.0 x 10" 3rd QTR . . . 2.0 x 10" 4th QTR . . . 2.0 x 10" ENVTRONMENTAL MONITORING TECIINICAL SP"CIPICATION 5.7 (7)

(a) and (b) - The environmental parameter monitored for GTRR crerations is that of direct radiation from the facility and from gaseous eftluents via a system of 30 film badges positions around the perimeter fence and other similar locations (see Figure 1, Environmental Monitoring 4tations).

(c) - The itte badge used for environmental monitoring, which is provided by a NVLAP certified vendor, has a lower limit of detection of < 10 mrem.

None of the film badges positioned around the facility showed radiation exposure due to reactor operations.

If radiation exposure due to reactor operations were expected to occur, it would most likely be seen in film badge #1 which is positioned ir aide of the reactor building stack. Therefore, exposure recorded by this film badge would be directly Lttributable to reactor operations. Nonetheless, because of its location inside the reactor building stack, it would not be representative of environmental exposures, but rather would represent worst case exposure.

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U.S. 'Mclear Regulatory Commission - Annual Report Febrcary 24, 1992 Page 13 Several badges showed radiation exposure above background le als, film badges #14 and 15 being the hichest values. Badge #14 is located on the roof of the laboratory building while badge #15 is located on the roof of the hot cell. Exposures registered by these badges as well as badges #2, #9 and #12 are attributable to environmental damage, e.g., rain and excessive heat.

(d) - Highest, lowest and annual average levels of radiation for the sampling point with the highest average radiation exposure due to reactor operations and location cf that point with respect to the site -

All of the film badge locations were similar Average annual level - < 10 mrem Highest annual level - < 10 mrem Lowest annual level - < 10 mrem OCCUPATIONAL PERSONNEL RADIATION EXPOSURE

a. Summary of exposure for persons under 18 years of i age greater than 50 mrom -

None

b. Summary of occupational exposures greater than 500 mrem -

None Should there be any questions concerning this report, please let us know.

Sincerely yours, A- 'W - _

R.A. Karam, Ph.D.,

Director Neely Nuclear Research Center RAK/ccg cc: 1. Dr. Gary W. Poehlein

2. Members Nuclear Safeguards Committee
3. Director, Office of Nuclear reactor Regulation U.S. Nuclear Regulatory Commission Washington, D. C.
4. Document Control Desk U.S. Nuclear Regulatory Commission Washington, D. C.

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NEELY NUCLEAR RESEARCH CENTER Minor Changa

.. Proctduro 4200 Number: Revision 00 By: CHANGES IN GTRR DESIGH Approved 04/28/(

Dater / / Page 3 of 4 APPENDIX A 10 CFR 50.59 SAFETY EVALUATION QUESTIONNAIRE FACILITY MODIFICATION NO: T/-00/

TITLE: Ft.-a s H e-< Eoe 44/4/vx/ceA m ,danEL l

1. Will the probability of the occurrence or the consequentes of '

an accident or malfunction of equipment important to sarbty previously evaluated in the safety analysis report be increased? [yes/no) /lo <

2. Will the possibility for an accident or malfunction of a different type than evaluated previously in the safety analysis report be created? [yes/no] Ac /
3. Will the margin of safety as defined in the basis for any technical specification be reduced? [yes/no) 4/O
4. Is the proposed change an unreviewed safety question?

[yes/no] A/0 KOTE: If additional space is needed to justify conclusion (s) please attach extra sheet (s).

DATE:

PREPARED BY: b/LLY hA7 NAM 7- 3 l- 9/

APPROVALS:

Director NNRC: M /[0- h m 8/ /

Nuclear Safeguards Committee: AfDae ned ff///9l

4 support DOCUMENTATION POR FACILITY MODIFICATION NO.91-001 FLASHER FOR ANNUNCIATOR PANEL Flasher description:

The annunciator panels contain one flasher. The function of the flasher is to blink the lights of an annunciator in the alarm condition until acknowledged by an operator. In addition an audible alarm sounds until acknowledged; the audible alarm is independent of the flasher.

Condition:

The Scam model ACSF-1 f1asher has failed ind the 1ights of the annunciator in the alarm condition are on, instead of flashing.

Direct Replacement flasher:

The Scam Instrument Corporation apparently does not exist in the Chicago area, as evidenced by the failure to get a number f rom the phone company. A search of Nuclear Equipment Buyers Guide showed no listing for Scam Instrument Corporation.

ACSF-1 flasher operation:

The ACSF-1 includes a motor and a cam operated switch, the switch has both normal 1y open and normal 1y elosed contacts. 120 VAC H is connected to one side of the motor (connector pin 7) at all times.

When 120 VAC N is connected to the other side of the motor (connector pin 10), the motor will run. The lights of the annunciator in the alarm condition will be on until the normally closed cam switch goes open, the lights then go of f and remain of f until the switch closes. When the acknowledge button is pressed the motor will stop immediately, if the normally open cam switch is open, otherwise the motor will continue to run until the open position is reached.

Electronic flasher operation:

The electronic flasher contains a CMOS 12 bit binary counter (CD4040), a W232D-3-12 solid state relay (SSR) and a 12 VDC power supply power. In the reset condition terminal S Will have approximately 120 VAC H supplied through the audible alarm, this 120 VAC H is converted to + 12 VDC and used to keep the CD4040 in the reset condition. The SSR is turned on while CD4040 is reset.

An annunciator alarm condition will cause terminal S to go to 120 VAC N, this allows the Qp4040 to count at a-60 Hert: rate. After 64 counts, CD4040 pin '2 goes high and the SSR is turned off.

Another 64 counts, CD4040 pin 2 goes low and the SSR is turned on.

When the acknowledge button is pressed. CD4040 is reset.

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NEELY NUCLEAR RESEARCH CENTER

.. Minor Chango Procedurs 4200 Number: Revision 00 By: CIlANGES IN GTRR DESIGN Approved 04/28/l Dates / / Page 4 of 4 FACILITY MODIFICATION DOCUMENTATION CHECKLIST APPENDIX B FACILITY MODIFICATION NO: 9 /- Oc /

TITLE: G_A TH': P 5 :' .ds . i . i. ; ;; - + r: f,;,

DRAWINGS _ :

NUMBER TITLE REVISED BY DATE UCNG $~L A B Mc->2 CD(t & WWirMCtL M,f%f5L 6Z4.EW Q ff

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PROCEDURES: VfDd 4 ffAO M N O b 2.d.,e. cd A w yt s % IMAIJ S4 -

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I - Georgia Institute of Technology

!! Of ;\ NEELY NLJCLEAA RESEARCH CENTER 1 % Y / ]/ 900 ATLANTIC CAIVE

' , o ,, ,, g'

, ATL ANTA. GE ORGIA 30332 0425 gg gg Pacility Modification Request 91-002 l

Power Level Measuring Channels Picoammater #1 and #2 l

The GTRR Power Level Measuring Channels (Picoammeter #1 and #2) are l old GE vacuum tube instruments which are not manuf actured any more.

These instruments are 25 years old and frequently break down. l Spare parts are nearly impossible to obtain. Consequently, it is desired to replace the instruments with Keithly picoammeters model

  1. 485/4853.

Recently, we received funding from the Department of Energy to i I

purchase and install the two picoammeters. The attached document gives more details.

TQie r 542507 GTRIOCAATL f da 404 053 9325 ' *dV40d *4 3M A Uac c' tae Universay System of Georpa An Eaue Ecutanen and Empiovment ocoortunev m:e~

. NEELY NUCLEAR RESEARCH CENTER

. Miner Cheng] Proccdura 4200 Number: Revicion 00  !

By: CHANGES IN GTRR-DESIGN Approved 04/28/89 l Dates / / Page 4 of 4 '

FACILITY MODIFICATION DOCUMENTATION CHECKLIST APPENDIX B FACILITY MODIFICATION NO: 9/- 007.

TITLE: hicD A v1A4GYEtt $P PL.fic.E k/En)7 DRAWINGS:

NUMBER TITLE REVISED BY DATE G45- 62 -cm L Js vg o g e g r n rfn ti d d,ctrot.

15/r 2 G-ara /Ane s PROCEDURES:

NUMBER TITLE REVISED BY DATE 7 co2. E EACXoC, C#ER ArronJr hcet>rrwAL.

GrM*,na c.nescusr Ap3O Satcr SGR0 t r co2 ADP p nVA e .

i 97-74 %en Aa4Mer:1 CAL f(M Art Od .

Reviewed By: I/ i4 . '4dQL^ ~ - - ~ Date: '![ 9 f

NEELY NUCLEAR RESEARCH CENTER

, Minor Changa Procedure 4200 Ntimbert Revicion 00 By: CHANGES IN GTRR DESIGN Approved 04/28/89 Dates / / Page 3 of 4 APPENDIX A 10 CFR 50.59 SAFETY EVALUATION QUESTIONNAIRE FACILITY MODIFICATION NO: 9l- do 2 TITLE: krtoAu4ETM EPL4 c w/ar/T

1. Will the probability of 'he occurrence or the consequences of an accident or malfunction cf equipment important to safety previously evaluated in the safety analysis report be l

increased? [yes/no) A/o l

2. Will the possibility for an accident or malfunction of a different type than evaluated previously in the safety analysis report be created? [yes/no) Me l

1

3. Will the margin of safety as defined in the basis for any technical specification be reduced? [yes/no) Mn  ;
4. Is the proposed change an unreviewed safety question?

[yes/no) Alo NOTE: If additional space is needed to justify conclusion (s) please attach extra sheet (s).

DATE:

PREPARED BY: /$. S ATH a vl / 2-/ 7-9 /

APPROVALS:

Director NNRC: ,p, g , M % /a / /9 /9 f

}

Nuclear Safeguards Cnmmittee: k / L//9N '

Picoammeter Replacement 1.0 PURPOSE The purpose of this facility modification is to replace the Reactor Instrumentation Picoammeters with new modern technology picoammeters.

2.0 SCOPE The proposal is to replace the existing General Electric Micromicroammeters with Keithley model 485 Aut orangin;;

Picoammeters.

3.0 RESPONSIBILITY The approval f or this modification lies with the NNRC director with the concurrence of the Nuclear Safeguards Committee.

4.0 REFERENCES

4.1 Keithley-Model 485 Autoranging Picoammeter Instruction Mantal 4.2 General Electric Stable Micromicroammeter Catalog #

534E745G3,G4 4.3 Related Procedures 4.3.l~ Procedure 7274 Picoammeter Calibration 4.3.2 -Procedure 2002 Reactor Operations - Precritical Startup Checklist and Shift-Supervisor Approval SiO SYSTEM DESCRIPTION 5.1 Existing Picoammeters The_ existing picoammeters are General Electric

-Micromicroammeters equipped with a manually operated range

. switch. Each range decade has a x1 and a x2.5 position.

There is a contact that is closed when the range switch is on the 2.5E-10 amp position. The 2.5E-10 amp position is used for reactor startup and the closed contact is a startup permissive. The Micromicroammeter has a 0 to 10 millivolt output- used to drive the Power Level Recorder. There are no reactor scram signals from the Micromicroammeter.

5.2 Proposed New-Picoammeters The proposed new picoammeters are Keithley model 485 autoranging picoammeters. A 485 range change is 1 decade; in the auto mode the most sensitive on scale range is automatically selected.

. .- - .. . - - - . ~ . . - - . - , , . - . , . - , , . -

I 1

o The 485 autoranging feature eliminates the need for the  ;

startup permissive and this permissive will be deleted. i The 485 analog output is 1 millivolt per count; therefore a Power Level Recorder resistive divider network is necessary. 3 The schematic of this purposed network is included. The network contains an x1, x3 switch; this feature allows the reactor operator to increase the Fower Level Recorder reading ,

by a factor of 3. The x3 position could be used when the i Power Level Recorder is < 20% full scale on the x1 position i and autocontroller operation is to be used. i 1

b( & O LD m y i

i 1

l l

I i

f l

g, .

NEELY NUCLEAR RESEARCH CENTER Minor Change Procedure 4200 Number: Revision 00 By: CHANGES IN GTRR DESIGN Approved 04/28/89 Date / / Page 3 of 4 p,g, APPENDIX A d 10 CFR 50.59 SAFETY EVALUATION QUESTIONNAIRE FACILITY MODIFICATION NO: J O O P #DS g. *

[jjg TITLE: _ REPAIR and UPGRADE COOLING TOWER

1. Will the probability of the occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety' Analysis report be increased? [yes/no] NO v/
2. Will the possibility for an accident or malfunction of a different type than evaluated previously in the satsty analysis report be created? [yes/no] NO t/

The new materials used are the same or better capability than the present materials.

3. Wi.11 the margin of safety as defined in the basis for any '

technical specification be reduced? [yes/no] NO e/

Change does not alter the margin of safety defined in the technical specification.

4. Is the proposed change an unreviewed safety question?

[yes/no] __NO v' NOTE: If additional space is needed to justify conclusion (s) please attach extra sheet (s).

This does not present an unreviewed safety question.

DATE:

PREPARED BY William H. Downs September 10, 1990 APPROVALS:

pirector NNRC: / l' /9t!! ) h kNO v Nuclear Safeguards Committee:

49 NEELY NUCLEAR RESEARCll CENTER Minor Change Procedure 4200 Number: Revision 00 By: QIANGES IN GTRR DESIGN Approved 04/28/89 Dates / / Page 4 of 4 FACILITY MODIFICATION DOCUMENTATION CHECKLIST APPENDIX B FACILITY MODIFICATION NO: 90-005 TITLE: REPAIR and UPGRADE COOLING TOWER DRAWINGS:

NUMBER TITLE REVISED BY DATE None PROCEDURES:

NUMBER TITLE REVISED BY DATE None Reviewed By: C J-d Il Dater C) /> h()

/

l l

l

s' ,

Facility Modification 90-005 Cooling Tower Repair and Upgrade Approvals k lhttn 9 I/6/) 'i c:doOirector NNRC Dste

~

Nuclear Safeguards Committee Date

  • e iO Cooling Tower Repair and Upgrade 1.0 PURPOSE The purpose of this facility change is to rejuvinate the cooling tower and bring it up to original operating specifications.

2.0 SCOPE This proposal is to replace the existing wooden lath fill, plywood top deck and cold water basins with newer ' state of the art' components.

3.0 RESPONSIBILITY '

The approval for this modification lies with the Neely Nuclear l Research Center management with the concurrence of the Nuclear  !

Safeguards Committee. l l

4.0 REFERENCES

4.1 GTRR 5 Mw Safety Analysis Report section 4.4.8.2, page 78 4.2 GTRR Technical Specifications Table 4.2 section 4.4.d 4.3 Related Procedures 4.3.1 Procedure 5063 CW Basin Low Level - Alarm 4.3.2 Procedure 9019 Cooling Tower Tritium Analysis 5.0 SYSTEM DESCRIPTION 5.1 EXISTING COOLING TOWER

-The existing cooling tower has a wooden lath fill. It has a plywood top deck and cold water basins The louvres are asbestos.

5.2 _ PROPOSED UPGRADE $b The proposed upgrade will install a new-3/ " top deck and cold water basirs. The cold water basins will have removable to eliminate algae growth. New target nozzles will be installed also.

The wooden lath fill and drift eliminators will be removed and replaced with Munters PVC. The support lumber will be redwood.

,- )

  • . 1 Remove & replace the asbestos louvres with 12 oz. fiberglass.

Plywood will be Douglas Fir. Support lumber will be redwood. All lumber will be pressure treated. Hardware will be stainless steel.

5.3 POSSIBLE FAILURE MODES None.

6.0 RECORDS The changes in the system will be incorporated in the facility drawings listed in section 4.0.

1.7 '

TED MARSDEN, INC.

,,l Epula[ising in Coofixg Jowet bessieu & Sepaits 956 Strap Hinee Trall Stone Mountain, GA 30083 8

July-25,-1988 Georgia Institute Of Techno)ogy-Neely Nuclear Research Center 900 Atlantic Drive Atlanta, Ga. 30332 Attention: Mr. Dean Mc Dowell

Subject:

Marley Cooling Tower, Model #68-102 Dear Mr. McDowella This week, we conducted an inspection of this tower. Heroin are our recommendations for getting the tower back into condi-tion to operate as originally designed (1200 GpM -116-87-79 degrees).

The structure'is-sound except for the top deck area.

We propose:

1. Install new 3/4" plywood top deck and cold water basins; the cold water basins will have removable covers to eliminate algae growth. Install.new target nocales.
2. Remove all wooden lath fill and drift eliminators.

Install-Munters pVC fill and drift eliminators with redwood' supports. Hunters fill is ' state of the art' in cellu;.ar film cooling. Their drift eliminator has the lowest pressure drop and is the-moct efficient in today's market.

3. Replace all louvers with 12 oc. fiberglass.

Contract price -

$ .27,800.00 g , c h Our contract price-includes-supervision,~1 abor, materials, freight, and disposal of all debris. plywood will be Douglas Firs

-cupport lumber will be redwood;- all lumber will be pressure treated. Hardware will be stainless steel.

The asbestos louver blades will be disposed of at a certified landfill;-documentation of this'will be forwarded to the cus-tomer.

We look forward to working with you on this project.

C cen/&y,

'py h F. oden, President TED MARSDEN, INC.

-Enclosure PM/jm1

Reactor Coolino Tower Test Date 0 ??kl-.

Time lb') l b/6 / 4- 4 0- l(}2 Reactor Power '@@I 7#O7 7 '7 /,f V - W13 'd

  • T o f7,. , as m 45.=~ fo Relative Humidity 42 n. , 3 $' ,79.f ,.76, 769 .g f g f, ,3 Outside Temperature 6/ C 8f Sl/ 85 6f 6g ff6 ,

Primary Flow / 7/ O /74(- / 7 9d / Tf)d ,

Primary Inlet Temp. N,T T& C TT,9 / (;g. $

Primary Outlet Temp. T / , () Tc O /6 6, / / ) C, p Secondary Flow #/I8 (/ 7 (, C/?c) /8 C Secondary Hot Temp, 7#/. .C /[9,( /[(, / / /2, [

Secondary Cold Temp. 67< f 8[,7 7S, f 96,f Shield Flow 34- 84- 3 f- 3 9-Shield Hot Temp, '7 2 72. T 77,0 Tr 3. '/'

Shield Cold Temp. '72 7 / , <;s '7 7. C Ts 3,C Fan Condition M/ M/f 2 h' / (Z /d-/ 6 i Bypass Valve $5k i f,Es */d, [$'Ti Operator < 6A Reviewed / m

,, ', Reactor Cooling Tower Test Date /2 Pet'fd Time l9fVT //6? I3 ?3 /39'f \

Reaetor Power AM / c25u $N'6 3 874.7 Relative Humidity 62D 7. 4-07o kDN s35 % ,

Outside Temperature 50 M sc/ (o I Primary Flow nio /7b) (*7 9 *, /f/6 10%.3 Primary Inlet Temp. 79. ~7 8.F Mr'M /db

'I 7. b' Primary Outlet Temp. 82.[2 77. / 46 r31 _ /s/, 3 Secondary Flow 9 70 fW 9 55 9 d r; _

Secondary Hot Temp, fa 'f ,fe //77 t, //c/

Secondary Cold Temp. 7t_ - 7/./ N . 6 ,_ _[J. [ _

Shield Flow J[. J4 s3(, Jg Shield Hot Temp. 77. / ></,s'~ 8 f. , O d'J' Shield Cold Temp. 7/. 3 77.2 ' 74 cpg f--

Fan Condition /.rff f ,

(P

[/$w d, ri QJ f,//m Bypass Valve 70k gyf TC T ;77 y[

Operator 2/w.e

/' A />

Reviewed, ,, M erm

N. b -

    • HYGRO-THER MOGR APH t NYGRO.TNERMOGRAPH

, - CHART NO. 5 207.W . CHART NO. 5 207.W

, LVOftT INSTRUMENT COMPANY

.. - . ....4,..

gtVONT INSTRUMENT COMPANY

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