ML20034E681

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Annual Rept for 1992
ML20034E681
Person / Time
Site: Neely Research Reactor
Issue date: 02/22/1993
From: Karam R
Neely Research Reactor, ATLANTA, GA
To:
NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
References
NUDOCS 9303010188
Download: ML20034E681 (38)
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Georgia Institute of Technology 1 / l NEELY NUCLEAR RESEARCH CENTER j (' wy [ 900 ' ATLANTIC ORIVE .] M'a "[ ATLANTA, GEORGIA 30332-0425 l (4o4) 894-3600 USA ') February 22, 1993 U.S.' Nuclear Regulatory. Commission

j Region II g

101-Marietta-Street, N.W. -Atlanta, GA. 30323 1

Reference:

Annual Report Docket 50-160; License R-97 'l Gentlemen: Pursuant to Section 6.7.a of the Technical Specifications for.the-

Georgia Institute of Technology Research Reactor ( License.R-97 ),..

~ the following annual report is submitted. The reporting period is January 1, 1992 through December 31, 19 92. (calendar.. year.19 9 2 )'. j The ' designation of the sections below follow the title and order of A Section 6.7.a of our Technical Specifications. 1. OPERATIONS

SUMMARY

j a. Chances in' Facility Desian j .i There were three facility. design changes'during calendar year 1992:.one. involving the. addition of'an alarm which y activates wher1 the power-level on picoammuter #1 and I picoammeter #2 deviates ' by more. than ' 10% '. from steady - state; another. comprised: replacing the reactor primary t and secondary coolant temperature measuring systems with 'l a new one; and the third,. consisted of. replacing the; temperature probe wells of the primary and secondary coolant. All three design changes are described: in } Appendix A.. b. Performance Characteristics -l 'During the reporting period, the reactor was' operated at j power levels up to 4.0 MW using a 17-element core. ' An 8-c element fuel exchange - to enhance self protection.was ' performed. A weld failure of - fuel-element. B015' was . discovered and reported to'the'NRC and the manufacturer, B&W.- .Otherwise,- fuelcperformance has' continued to.be satisfactory with no'known problems. L 1 n100 @ 9303010188 930222 gDR ADOCK 05000160 YL ~ PDR-1 -L p-l . Tejer: 542507 GTFUOCAATL Fax. 404-853 9325 (Veafy 404-894I3500)[ dunit of the Urnvesty Sys},quf Georgia An $ qual Education and Ernployment Opportunity Inst:tution;.

.c .a-U.S. Nuclear Regulatory Commission - Annual Report February 22, 1993 -Page 2 c. Chances in Operatina Procedures The list of new and/or. revised ' procedures which were approved by the Nuclear Safeguard Committee during calendar year 1992 were as follows: Proc. #. Title 2002 Reactor Operations-Precritical Startup Checklist and Shift Supervisor Approval .i 2006 Reactor Shutdown Checklist t 9150 Operation and Calibration of. Area Radiation Monitors 9312 Sealed Sources Leak Test 9041 Storage Pool Water Sampling and Analysis-9304 Routine Facility Radiation Surveys 9018 Charcoal Cartridge Analysis a 9400 Environmental Monitoring 7204 Floor Fuel Storage Water 9250 raci..ities contamination Surveys 9300 Respiratory Protection 4 6080 Accidental Release of High Levels of Gaseous Activity.to the Atmosphere 6090 Personnel Monitoring After Building Evacuation Emergency Situations 9250 Facilities Contamination Surveys 9057 Calibration Procedure for Eberline Model' E-120 GM Survey Meter 9058 Calibration Procedure for Eberline Model RM-14 Rate Meter

s 4. 'U.S. Nuclear' Regulatory Commission - Annual Report February 22, 1993 Page 3 Proc. # Title 9060 Calibration Procedure for Bicron Model-RSO-5 Survey Meter 9063 Calibration Procedure for Ludlum Model 2 GM Survey Meter 9065 Calibration Procedure'for Bicron Model RSO-500 Survey Meter ~ 9072 Calibration Procedure for Eberline Model RO-2 Survey Meter 7285 Calibration of Reactor Coolant Temperature Measuring System 2002 Reactor Operations-Precritical Startup Checklist and-Shift Supervisor Approval-2006 Reactor Shutdown Checklist-1 7202 Control Rod-Drop-Time 7226 Scram Insertion Delay Time Measurement. 9501 Control & Accountability of Radioactive Sources 7241 Reactor Tank Level Transmitter Maintenance and Calibration Check 'l 4950 Tagging Equipment Out of Service .l A list of procedures which were deleted by the Nuclear, - ) Safeguards Committee were: Proc. # Title 3301 -Sampling Log-D20 3302 Sampling Log H20-3303 Sampling' Log Blanket' Gas 4060 Top Reflector' Dump Time Measurement i

v D 4 U.S. Nuclear Regulatory Commission - Annual Report February 22, 1993 Page 4 d. Results of Surveillance Tests and Inspections The surveillance tests and inspection of the facility required by the Technical Specifications were performed. Documentation of each of the tests'and inspections are available at the site for review. e. Chances, Test and Experiments Aporoved by USNRC There were no changes, tests or experiments that required the approval of the USNRC pursuant to 10 CFR 50.59(a). f. Current Staff and Nuclear Safeguards Committee Membership Dr. R.A. Karam, Director, Nuclear Research Center Dr. Rodney Ice, Manager of the. Office of Radiation Safety Mr. B. D. Statham, Reactor Supervisor and Electronic Engineer l Mr. William Downs, Senior Reactor Operator ~ ~ Mr. Dixon Parker,: Reactor Operator Mr. Jerry Taylor, Senior Safety _-Engineering Assistant Mr. Edgar Jawdeh, Health Physics Mrs. Clara Galleshaw-Mrs. Arlene Robinson Smith Mr. Nazee Chebeir, Health ~ Physics In addition to the full time staff, the NNRC employed the following graduate students on part time basis: Mr. John Hawkinson Ms. Kathleen Klee ~ Ms. Hannah Mitchell Mr. Thomas Evans i Mrs. Hong Ning The current membership of the Nuclear Safeguards Committee is: (1) Mr. Emsley Cobb, Chairman. Disciplines Reactor Operation and Reactor Safety ) (2) Dr. Bernd Kahn Discipline: Radiation ' Protection and j Environmental Measurements ] j

m ) 4 C P U.S. Nuclear Regulatory Commission - Annual Report February 22, 1993 ~ Page 5 (3) .Dr.. Robert Braga Discipline: Chemistry (4) Dr. Prateen-V. Desai, Secretary Discipline: Thermal Hydraulics, Mechanical Systems (5) Dr. Billy R. Livesay, Member Discipline: Material Science, Physics (6) Mr. Jack Vickery, Member Discipline: Security (7) .Dr. Thomas G. Tornabene, Member Discipline: Biology (8) Dr..S. M. Ghiaasiaan, Member Disciplines Nuclear Engineering (9) Mr. Len Gucwa, Member Discipline: Reactor Safety (10) Mr. Steve Ewald, Member . Discipline:- Health Physics (11) Dr. Peggy Girard, Member Discipline: Biology (12) Mr. James O'Hara, Member-Discipline: Health Physics 2. POWER GENERATION For the period January 1,1992 through December ' 31,.1992, the total power generation of.the GTRR was 196 MW hours. The reactor was operated.a total of 158 hours: 8 hours at power levels equal to or less than 100 kW, 119 hours.at power levels 100 kW to 1 MW, and 31 hours at power levels above 1 MW.

3. SHUTDOWNS

' During this. reporting period there were 3 unscheduled-shutdowns. Table 1 gives details.

t U.S. Nuclear Regulatory Commission.- Annual Report February 22, 1993 Page 6 TABLE 1 UNSCHEDULED REACTOR SHUTDOWNS DURING 1992 Report Date Trip Reason for Trip'~ Corrective Initiation Action 92-01 4/29 Low lon Flux Amp #2 Adjusted Flux chamber trip point had Amp #2'R3.to voltage drifted. correct value. 92-02 10/14 Low shield low coolant Added coolant ~ coolant level in tank to TS-1. Also level led to pump raised startup q cavitation. level require-ments from 3 to 6 in. 92-03 12/10 Low H O Flow-Cleaned HX-2 2 flow restriction at impact bars. impact bars in HX-2. 4. UNSCHEDULED MAINTENANCE ON SAFETY RELATED SYSTEMS AND COMPONENTS There were approximately eighteen minor repairs performed on-safety-related systems and components. Records of mainten-ance performed on components are available at NNRC offices for inspection. 5. CHANGES, TESTS AND EXPERIMENTS During 1992, there were 27 approved experiments which used the GTRR. The experiments were evaluated prior to their approval with regard to section 3.4 of the Technical Specifications. 6. RADIOACTIVE EFFLUENT RELEASES a. Technical Specification 6.7.(6)(a) - Gaseous Effluents -Summation of All Releases Via Stack, i.e., ground level release.

U.,S. Nuclear Regulatory Commission - Annual Report February 22, 1993 Page 7 (1) FISSION AND ACTIVATION GASES l Tritium Released (gaseous) Non Measurable Argon-41 Released Total Total Avg. Avg. Released Max. Inst. ~% Tech

  • Release Release over period of Release Specs (Ci)

(pCi/cc) reactor opera-(pCi/sec) tion (pCi/cc) 1" Otr 21.810

1. 857x10-7
3. 3 00x10-5 475 81.19 2"d Qtr 11.324 9.643x10-8
1. 5 08x10-5 228 38.97 3 r* Otr 2.686 2. 2 87 x10-8
1. 000x10-5 95.0 16'.24 4 th Qtr 3.653 3.111x10-8 5.616x10-8 62.7 10.72
  • Computation based on the Maximum. Instantaneous Release' Rate-l as evaluated against a TS release limit of 585 pCi/sec.

(2) IODINES RELEASED None Measurable l Lower Limit of Detection <5.76 x 10-"pCi/cc (3) PARTICULATES 4 None Measurable Lower Limit of Detection <6.3 x 10-5pC1 gross beta / gamma Lower Limit of Detection <8.7 ~ x 10-5pCi b. Liquid Effluents .{ (1) FISSION AND ACTIVATION PRODUCTS Cobalt-60 is 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 Materials License do.1147-1-SNM. No' fission products are released via the liquid effluent pathway. .n i~ f

.U.S.' Nuclear Regulatory Commission - Annual Report February 22, 1993 Pdge 8 6 (1) CO RELEASE i Total Release Avg. Release * % Tech Ci Rate (pC1/cc) Specs 1st QTR 0.000016

8. 00 x 10-"

<1% 2nd QTR 0.000027 1.35 X 10-2 <1% 3rd QTR 0.000048

2. 4 0 x 10-2

<1% 4th QTR 0.000017 8. 5 0 x 10-" <1%

  • Average release rate values are based on a Georgia Tech campus water discharge rate of 2.00 x

+ 10" ml/ quarter. (2) TOTAL GROSS RADIOACTIVITY ( / gamma) Total Release Avg. Release * % Tech Ci Rate (pCi/cc) Specs 1st QTR 2.40 x 104

1. 2 0 x-10-1"

<2% 2nd QTR 3.50 x 104

1. 75 X 10-2

<2% 3rd QTR 5.70 x 104

2. 85 x 10-2"

<2% 4 4th QTR 2.00 x 10

1. 00 x 10-2*

<2%

  • Average release rate values are based on a Georgia Tech campus water discharge rate of 2.00 x 10" ml/ quarter.

(3) TRITIUM Total Release Avg. Release * % Tech Ci Rate (pCi/cc) Specs 1st QTR 0.00123 6.15 x 104 <1% 2nd QTR 0.00485 2. 4 3 x 10-8 <1% 3rd QTR 0.01841

9. 21 x 10-8

<1% 4th QTR 0.00494' 2.47 x 10-8 <1%

  • Average release rate values are based on a Georgia Tech campus water discharge' rate of 2.00 x 10" ml/ quarter.

i 5

UeS. Nuclear Regulatory Commission - Annual Report February 22, 1993 l Page 9 -(4) GROSS ALPHA RADIOACTIVITY RELEASED None' Measurable Lower Limit of Detection - <7. 07 x 10-6 on 11/18/92 f (5) VOLUME OF WATER RELEASED (ml/Ouarter) From Reactor Building let QTR 3.67 X 107 ml 2nd QTR 5.37 X 107 ml 3rd QTR. 9.61 x 107 ml 4th QTR. 3.71-x 107 ml (6) VOLUME OF DILUTION WATER USED DURING EACH OUARTER From Georgia-Tech Campus 1st QTR 2.0 X 10" ml 2nd QTR. 2.0 X 10" ml 3rd QTR 2.0 x 10" ml 4th QTR'. 2.0 x 10" ml 7. ENVIRONMENTAL MONITORING:( Tech. Spec. 6.7.a(7)) (a) Thirty sites.are monitored for environmental radiation. The parameter monitored-for Georgia Tech.Research Reactor (GTRR) operations is that of direct radiation from the facility and from emitted gaseous effluents (predominantly Ar-41). The_ location of the sites relative to the reactor are shown in Figure 1, " Environmental Monitoring Stations". The. sites are predominantly around the reactor perimeter fence or down-wind from the reactor. - (b) Total assays = 30 sites X 12 months X 2 assays / site = 720 assays. These data are reported in the environ-3 mental radiation surveillance' table-(attached). The letter M was,1 sed to designate any reading which was. less than the minimum detectable limit. 3 (c)_ The film badge used for environmental monitoring, which is provided by a NVLAP certified vendor, has a lower limit of-detection of < 10 mrem. P 4

1 4 UqS'. Nuclear: Regulatory Commission - Annual Report February 22', 1993-Page 10 None of the film badges positioned around the facility showed radiation exposure, due to the 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 inside of the reactor' building stack. Therefore, exposure recorded by this film badge would be directly attributable to reactor operations. None'the less, because-of its location inside the. reactor building stack, it would not be representative of. environmental exposures, but rather would represent worst case exposure. Several badge showed radiation exposure a' cove background levels, film badge # 16 being the highest value, followed by badges # 9 & 11. Badge # 16 exposure is an anomaly of unknown origin possibly attributable to environmental conditions,e.g.-rain & excessive heat. Badges

  1. 9 & 11 are' located around the barn area (radioactive waste-storage; area). The exposure readings

? are probably due'to the presence of;100 mci-of Radium-226 in the aforementioned area. During the months of August through October,-various radium sources were collected from campus users, consolidated for shipment and shielding, and disposed of:as radioactive waste. (d) The highest, lowest and average >1evels of' radiation-for-the sampling point with tho' highest average radiation exposure due to reactor operations and location of that' point with respect.to the. site. All of the film badge locations were similar Average annual level - < 10 K'em Highest annual level - < 10 mrem Lowest annual level - < 10~ mrem.. (e) The maximum cumulative radiation dose above natural background radiation which.could be received by an-individual continuously present in an unrestricted area during reactor's operation would be less than the lower limits of detection (LLD), i.e. < 10 mrem. NEELY' NUCLEAR RESEARCH CENTER: ~ ~ ENVIRONMENTAL RADIATION' SURVEILLANCE

  • 1992-A JAN FEB MIR APR MAY JUN BADGE #

D S D S D S D S D S D S 09801 M M M M M M M M M M M-M 09802 M M M M M M M M M M 'M M 09803 M M M M M M M M M M M-M 09804 M M M M M M M M. M M M M 09805 M M M M M M M M M M M M 09806 M M M M M M M M M M M M 09807 M M M M M M M M M M M M 09808 M M M M M M M M M M M M 09809 M M M M M M M M M M M M 09810 M M M M M M M M M-M M M 09811 M M M M M M h M M M M M 09812 M M M M M M M M M M' M M 09813 M M M M M M M M M M M M 09814 M M M M M M M M M M M M 09815 M M M M M M M M M M M-M 09816 M M M M M M M M M M M M' 09817 M M M M M M M M M M M M 09818 M M M M M M M M M M M M 09819 M M M M M M M M M M M M 09820 M M M M M M M M M M M M 09821 M M M M M M M M M M M M 09822 M M M M M M M M M M M M 09823 M M M M M M M M .M M M M ) 09824 M M M M M M M M M M M M 09825 M M M M M M M M M M M M 09826 M M M M M M M M M M M M 09827 M M M M M M M M M M 09828 M M M M M M M M M M M M. 09829 M M M M M M M M 09830 M M M M M M M. M M M M M Sum of natural radiation, direct radiation from facility, and gaseous -radioactive effluents. Units in millirems (mR). No background or control subtraction has been considered. Detection by film badge donimeters, and' processed by Landauer. Lower limits of detection are 10'mR. Damaged film badge. 'NEELY NUCLEAR RESEARCH CENTER ~ ENVIRONMENTAL RADIATION SURVEILLANCE

  • 1992 JUL AUG SEP OCT-NOV DEC YEAR BADGE #

D S D S D S D S D S D S D S 09801 M M M M M M M M M M M M M M 09802 M M M M M M M M M M M M M M 09803 M M M M M M M M M M M M M M 09804 M M M M M M M M M M M M M M 09805 M M M M M M M M M M M M M M 09806 M M M M M M M M M M M M M M 09807 M M M M M M M M M M M M M M 09808 M M M M M M M M M M M M M M 09809 M M 10 10 20 20 M M M M M M 30 30 09810 M M M M M M M M M M M M M M 09811 M M M M 20 20 10 10 M M M M 20 30 09812 M M M M M M M M M M M M M M 09813 M M M M M M M M M M M M M M 09814 M M M M M M M M M M M M 10 10 09815 M M M M M M M M M M M M M M 09816 M M M M 40 40 M M M M M M 40 40 09817 M M M M M M M M M M M M M M 09818 M M M M M M M M M M M M M M 09819 M M M M M M M M M M M M M M 09820 M M M M M M M M M M M M M M 09821 M M M M M M M M M M M M M M 09822 M M M M M M M M M M M M M M 09823 M M M M M M M M M M M M M M 09824 M M M M M M M M M M M M M M 09825 M M M M M M M M M M M M M M 09826 M M M M M M M M M M M M M M 09827 M M M M M M M M M M M M 09828 M M M M M M M M M M M M M M ^ 09829 M M M M M M M M M M 09830 M M M M M M M M M M M M M M Sum of natural radiation, direct radiation from facility, and gaseous radioactive effluents. Units in millirems (mR). No background or control subtraction has been considered. Detection by film badge dosimeters, and processed by Landauer. Lower limits of detection are 10 mR. Damaged film badge.

j i .o U.S. Nuclear Regulatory Commission - Annual Report February 22, 1993 1 Page 14 -j I 8. Occupational Personnel Radiation Exposure: Radiation workers of Georgia Institute of Technology are' l monitored through the use of film badges which are provided by a j NVLAP certified. vendor and have a lower limit of detection of i <10 mrem. A monthly radiation dosimetry report is issued for the d I personnel of the Neely Nuclear Research Reactor. All personnel l dosimetry data is kept at NNRC. Summary of personnel dosimetry follows. a. Summary of exposure for persons under 18-years of age greater then mrem - None b. Summary of occupational exposures greater than-'500 mrem-None c. Person-Rem for the Neely Nuclear Research Center - R-97. q t Person-Rem = Sum of occupational workers = 0.54 rem The highest, lowest and-average-levels of personnel exposure due to reactor and hot cell cperations: Average annual level - 49 mrem Highest annual level - 150 mrem Lowest annual level - 10 mrem. d. Person-Rem for Ga Tech campus users. Person-Rem 2.04 rem = The highest, lowest and average' levels of personnel exposure due to' experimental use of radionuclide primarily P-32, I-131, and S-35 and/or the use of x-ray machines. Average annual level - 68 mrem Highest annual level - 250 mrem Lowest annual level -- < 10 mrem. _-_______a____-__ _ _ _ =

4 9 4 U.S. Nuclear Regulatory Commission - Annual-Report February 22, 1993-Page 15-e. Category of exposure NNRC Radiation Workers Annual exposure

  1. Radiation workers

< 10 mrem 7' 10 mrem - 49 mrem 7 50 mrem - 99 mrem 2 100 mrem - 149 mrem l' 150 mrem - 199 mrem 1 - 1 200 mrem 0 T Ga Tech On-campus users Annual exposure

  1. ' Radiation workers

< 10 mrem 32. 10 mrem - 49 mrem 17 50 mrem - 99 mrem 5 i 100 mrem - 149 mrem 0 t 150 mrem - 199 mrem 7 1 200 mrem 1 A F i

3 I, F L U.S. Nuclear-Regulatory Commission - Annual Report-February 22, 1993'- 'Page 16 Should there be any questions concerning this report, please let us know. Sincerely,

k. /4. (M 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. 1 4 J

t .t APPENDXX A Facility Modifications

.f. M ( FACILITY MODIFICATION REQUEST 92-001 PICOAMMETER MONITORS FOR PICOAMMETER #1 AND 12 The Nuclear Safeguards Committee approved a change in instruments for power level measuring channels #1 and #2 from the.old GE vacuum tube picoammeter to Keithleys' Model #485/4853. It is desirable to have an alarm added to the Keithleys' such that once the power level deviates by more than 10% from steady state operation, the alarm activates with sonalert. The device is intended as an additional aid to the operator to maintain " Cognizance" of' reactor power level. The attached document gives more details. b r v

NEELY NUCLEAR RESEARCH CENTER -r Minor Changa Procedura 4200 Number: Revision 00 By: CHANGES IN GTRR DESIGN Approved 04/28/89-Date: / / Page 3 of 4 APPENDIX A 10 CFR 50.59 SAFETY EVALUATION QUESTIONNAIRE F ILITY MODIFICATION NO: 9Z-OOI TITLE: I C_O 4 A fME72EC cA// To d 1. Will the probability of the occurrence or the consequences-of an accident or malfunction of equipment important to safety previously evaluated in the gafety analysis report be increased? [yes/no] Alo 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) Alm 1 f Will the margin of safety as defined in the basis fo,r,O 3. any technical specification be reduced? [yes/no] N 4. Is the propgsed change an unreviewed safety question? [yes/no] NO NOTE: If additional space is needed to justify conclusion (s) please attach extra sheet (s). DATE: PREPARED BY: /- 24-72_ J APPROVALS: Director NNRC: . /d MG - - I[d9-/9 A / Nuclear Safeguards Committee:

NEELY, NUCLEAR RESEARCH' CENTER < - [ . Minor.Changa. Proc dura 14200-Number

Revision' 00..

L:- = By:- CHANGES IN.GTRR DESIGN . Approved 04/28/85. . pate: / / Page. '4 l of ~;4 f FACILITY MODIFICATION DO..4_"' a ?ATION. CHECKLIST d APPENDiL J; FACILITY. MODIFICATION NO: $2 - Oc / TIfLE: 8 c.c A M44'ETSA oA// tor i DRAWINgg: + NUMBER TITLE REVISED'BY-DAIR

1

.i !q . PROCEDURES: EHEER TITLE REVISED BY DATE

L AlcAls j

q J ] 1 'k . Reviewed By: Date: I 14, My + i W. y i t l r n. e n

i Picoammeter Monitor 1.0 PURPOSE The purpose of this facility modification is to add a device that will give an audible alarm when the one of the Reactor Picoammeters deviates from a preset point. 2.0 SCOPE The proposal is to add the Picoammeter Monitor that will be connected to two Keithley model 485 Picoammeters. 3.0 RESPONSIBILITY The approval for this modification lies with the NNRC director - with the concurrence of the Nuclear Safeguards Committee.

4.0 REFERENCES

4.1 Schematic for the Picoammeter Monitor 4.3 Related Procedures 4.3.1 None 5.0 SYSTEM DESCRIPTION 5.1 Picoammeter Monitor The Picoammeter Monitor (PM) will be connected to tne Picoammcters analog output. The PM will be energized only when Reactor power is stablized. When energized, the PM comes on in the alarm condition. The operator.will adjust.the #1-set point potentiometer until the #1 set point LEDys a its - brightest level; this action will be repeated f or #2 set polne potentiometer. The operator will then press the reset switch to turn the~ alarm condition off and silence the Sonalert. @l Should either Picoammeter indication increase or decrease.byOE* approximately 10% of full scale value (for what ever scale it may be on) the PM will go to alarm mode. This condition requires operator action to reset the PM. A PM alarm could be. caused by a change in Reactor power level (autocontroller l problem) or malfunction in the one of the Picoammeter measuring circuits. During power level changes the operator would turn the PM off until a new' level is reached. At the new level the set points l would be adjusted.

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NEELY NUCLEAR RESEARCH CENTER Procedure 4200-Minor Change Numbe2: R; vision 00 By: CHANGES IN GTRR DESIGN Approved 04/28/89 Date: / / Page 3 of 4 APPENDIX A 10 CFR 50.59 SAFETY EVALUATION QUESTIONNAIRE FACILITY MODIFICATION NO: 92-002 TITLE: kEAL ACCMh/T" Of $2AcfaC'Diuf4eY AWD $n.co^1D A2 'l Cmt.MT' twf&t2Ardite VIGASO2td6 SYsr&MT 1. Will the probability of the occurrence or the consequences of an accident or malfunction of equipment'important to safety previously evaluated in the safety analysis report be increased? [yes/no) Alo 2. Will the possibility for an accident or malfunction of a different type than evaluated previously ip the' safety analysis report be created? [yes/no)- A/o 3. Will the margin of safety'as defined in the basis for any technical specification be reduced? [yes/no) 4)e 4. Is the proposed change an unreviewed safety question? [yes/no)- A4 NOTE: If additional space is needed to' justify conclusion (s) please attach extra sheet (s). DATE:- PREPARED'BY: /$. $ hare /44' M-/ o - 9 t- ~ APPROVALS:. Director NNRC: M.M./v6>i w 3 /O 9d Nuclear Safeguards Committee:

NEELY INCLEAR.~RESEARCH CENTER Numbdr: Revision 00 By: CHANGES IN GTRR DESIGN Approved 04/28/89 Date: / / Page 4 of 4 FACILITY MODIFICATION DOCUMENTATION CHECKLIST APPENDIX B FACILITY MODIFICATION NO: 92-00t TITLE: IhPLAcr4Gdf Of SEALTod fkoaAK.1 Ano %eeacaL1 CooLAIJT ' $GMAGAA702G YGA S d2l Al& bYSTEA45 DRAWINGS: NUMBER TITLE REVISED BY DATE o4 s - c 2 - oct h sre vuerJr4 rend ( Cearroc_ sa sr t Sc4s.u4v,cs f' ) ) PROCEDURES: NUMBE13 TITLE REVISED BY DATE 2-CO L E0CtOA 00tAbrin41-AECPJYo CAL &Atv dP C+1catt.rsr A WD.54 r W SUPtMesot. Afreets. 72 76 ~re uDcAArde e C = = ma Aces a C AL t 8 /t#T/od RTb / A)Ad7-72R3 R 7-D C_'A L it3 A A rf n d Reviewed By: - Dates.

REPLACEMENT OF REACTOR PRIMARY AND SECONDARY COOLANT TEMPERATURE MEASURING SYSTEMS FACILITY MODIFICATION 92-002 1.0 ' PURPOSE The purpose of this facility modification is.to replace the existing primary and secondary coolant temperature. measurement, alarm and scram systems with modern. equipment. 2.0 SCOPE The proposal is to replace. the resistance temperature detectors (RTD) and temperature recorders with new RTDs, digital panel meters (DPM) and new recorders. 3.0 RESPONSIBILITY The approval for this modification lies with _the NNRC director ' with the concurrence of the Nuclear Safeguards Committee.

4.0 REFERENCES

4.1 Operator's Manual, Omegarometer series # 10294ML-99 (manual for the DPM) 4.2 Operator's Manual, Omegarometer series # 10262ML-93 (manual for DPM dual setpoint control) 4.3 Operator's Manual, Omegaline Recorder model 620-2V 4 5.0 SYSTEM DESCRIPTION 5.1 Existing system description j Equipment: i d. 2 each RTDs for: primary coolant b. 2 each RTDs for secondary coolant c. 1 each temperature recorder for primary coolant d. 1 each temperature. recorder for secondary' coolant. There is an RTD in the primary coolant heat. exchanger inlet and outlet; also an RTD in the secondary ~ coolant heat exchanger inlet and outlet. One temperature recorder will e accommodate 2 RTDs. There are relays and mechanically actuated switches in each recorder that make up the scram, high temperature annunciator and low temperature. annunciator contacts. I

m. m_A s .5.2-Primary Problem-With Existing System The temperature recorder is a single input type recorder with the two.RTD inputs. sequentially zwitched with.~a mechanically driven stepper switch. This arrangement is satisfactory for an input device that has high resistance but the resistance of the RTD is 100 ohms near the mid range. of the. measured temperature. A resistance change of 0.2 ohm is approximately equal to l'F. The stepper switches require frequent cleaning and still their performance is poor. 5.3 Other Problem With Existing System The recorder scale is 11 inches wide for a span of 100 F; this makes resolution of 0.l*F very difficult. Replacement parts are difficult to impossible to locate. 5.4 Proposed system description Equipment: 4 each Omega model PR-11-2-100-1/4-6-A platinum RTDs 4 each Omegarometer model DP2101R2 Digital Panel Meters 4 each Omegarometer dual setpoint_ controls (in DPM) 2 each Omega 11ne model 620-2V recorders D O measurement, alarm and scram circuit is shown in figure 1. The low temperature annunciator requires a closed contact-across TC10 and TC11 to enable reset. -Both DPM low set points. will be at 53 F, should either DPM indication be less than this point the Low D,0 Temp annunciator will alarm. This low temperature alarm is only f'or the reactor operator information. The High D,0 Temp scram requires a closed contact across TC6 and~TC7 to enable reset. The.high set point for TRA-D1-1 DPM. is 137 F and the high set point for TRA-DI-2 DPM is 123 F. For a closed contact to be present at TC6 and TC7 there must be power to TRA-D1-1 DPM and the indicated temperatures less than the high set points. The High D O Temp annunciator requires a closed contact across~ TC8 and Td9 to enable reset. This is furnished by contacts of Kl. TRA-D1-1 RTD is connected through' normally closed contacts of. a spring. loaded switch. Weekly pre-critical startup checklist requires the operator to. test the High D 0 Temp scram point. 2 The operator will~ press and hold the spring loaded ~ switch while adjusting the 100 ohm 10 turn pot to test the scram set point. Once the switch is released the pot and resistor are shorted across. f V

H O measurement, alarm and scram circuits'are shown in figure. 2 2. TRA-H1-2 DPM high set point is 134 F. Circuit description is similar ' to the D O primary system.- 2 Figure 3 shows figure 1 and figure 2 connection into the existing wiring. 5.5 Technical Specifications and-Procedure Requirements-Appendix A contains reprints of pages 7, 8 ' and 11_ of the Technical Specifications and page 6 of procedure-7250. These are requirements that are applicable to this' facility. modification. t i b c y r

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k ff.:.llL !): h ~ 2.1.2 SAFETY LIMITS IN THE NATURAL CONVECTION MODE APPLICABILIIT This specification applies to the interrelated variables asso-ciated with the core thermal and hydraulic performance in the natural convection mede of operation. SPECIFICATION The reactor thermal power shall not exceed two (2) W. BASIS Experience with the GTRR has shown that no damage to the. core and no boiling occurs without forced convection coolant flow at power levels up to two W. 2.2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 LIMITING SAFETY SYSTEM SETTINGS IN THE FORCED CONVECTION } ODE APPLICABILITY Applies to the settings of those instruments monitoring the safety limits. OBJECTIVE To assure automatic protective action is initiated before a safety limit is exceeded. SPECIFICATION The safety system trip settings shall be as follows: Thermal Power 5.5MW - Reactor Coolant Flow 1625 GPM Peactor Outlet Temperature 139 *F t

q . v BASIS -The trip settings are chosen so that the reactor'is operated with no incipient boiling. An cualysis was made showing that at 1800 gallons per minute total' coolant flow, five MW thermal power an inlet reactor coolant temperature of ll4*F and the application of all the engineering uncertainty factors, a maximum fuel surf ace. temperature 8*F less than the local D 0 saturation temperature 2 might occur.(1) REFERENCE (1) Letter, R. S. Kirkland to USAEC, October 22, 1971, Response No'. 10. 2.2.2 LIMITING SAFETY SYSTEM SETTINGS IN NATURAL CONVECTION MODE APPLICABILITY Applies to the values of safety system settings when operating in the natural convection mode. OBJECTIVE To assure the reactor is not operated at a power level sufficient to cause fuel damage. i SPECIFI MTION The reactor thermal power safety system setting shall not' exceed 1.1 kW when operating in the natural convectit.: mode. BASIS In the natural convection mode of reactor operation the main coolant pumps are not operating. The reactor. isolation valves may be closed so that only internal, natural convection is available for cooling. Experience with the GTRR has shown that the reactor can be operated at one kW indefinitely. without exceeding a bulk reactor temperature of 123*F. j

NEELY NUCLEAR RESEARCH CENTER + Minor Change Procedure 7250 Number: Revision 01 By: COMPLETE LIST OF SETPOINTS FOR. Approved 08/01/: ,Date: / / MODES 1 AND 2 Page 6 of 6 APPENDIX A TRIP (SCRAM) CONDITIONS AND SetpointS TYPE OF TRIP MODE 1 MODE 2 Power Trip 1.25 MW 5.5 MW Period Trip 15 second same -10 second Beactor Tank Low Level > 66" same Low D O Flow 1000 gpm 1625 gpm 2 High D 0 Temp 125* F 139* F 2 Low Ion Chamber Voltage 650 VDC same Calibrate Switches Not in Operate same Position Magnetic Actuator Amp Not Energized same Reactor Isolation Valves Valves Leave same Not Open Open Position Drain Valves Open Valves Open same No D 0 Overflow No Reactor Vessel same 2 overflow Doors Open 20 psig on Truck, same Personnel, Emergency door (s) High H O Temp 134* F same 2 Low H O Flow 900 gpm 900 gpm 2 Control Air Pressure 60 psig same i Low Shield Coolant Flow 30 gpm same High Shield Coolant Temp 120* F same Low Bismuth Coolant Flow 0.75 gpm same i High Bismuth Coolant Temp 120 F same.

r_ 'I TABLE ' 3.1'. REQUIRED SAFETY CHANNELS Minimum No. Channel. Setpoint . Requi red Function 1* Minimum countrate permissive .2' cps. Start up rod. withdrawal interlock <10 see (pos or'neg) 2(c)-~ Scram: -Period trip .2 "I Scram I 5.5 MW ' Power. trip -Low D 0' flow 1625 gpm 2 Scram 2 139*F-2 . Scram High D 0 Temperature 2 d- -Isolate reactor vessel ,' Low D 0 Level 2 Scram <12" below over flow- . Initiate ECCS 2 1 Scram No D 0 Overflow 2 1 Scram Manual scram Reflector drain 1 Backup scram Containment doors open 1 per airlock ' Scram - Reactor. isolation valves closed 2 " per valve Scram '(a) Required during startup and for operation with less than 1 decade overlap between the startup' channel and the pico-ansmeter channel. (b) Hot required.for natural convection: operation-One 'of 'the twelve requiredf safety. channels may be bypassed. for a period not -to-exceed' 8' hours g - (# ~'C foritest, repair,"or-calibration? 'e

e NEELY NUCLEAR RESEARCH CENTER Minor Change . Procedure 4200 Number: Revision 00 i By: CHANGES'IN GTRR DESIGN Approved 04/28/89 Date: / / Page 3 of 4 APPENDIX A I 10 CFR 50.59 SAFETY EVALUATION QUESTIONNAIRE i FACILITY MODIFICATION NO: T2-003 TITLE: [EPL.Ac EWEd7 OG EGAC.TOK- ?tttMAzy Asa SCCodbAR.Y C.nouAUT ieHPe RATur E FR.osE kJE LLS 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] MO 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) A/o 3. Will tim margin of safety as defined in the basis fo. y any technical specification be reduced? [yes/no) No 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).. l DATE: PREPARED BY: [p -/O - 9 2 j APPROVALS: Director NNRC: MA $ dA.% $ [/o /fd Nuclear Safeguards Committee: 49/.Atvock 6 /'25/92 M4 /c 5

NEELT' NUC'5 EAR RESEARCH CENTER ~ Ri.nor Change Procedure 4200 Number Revision 00 py: cnANsEs in GTRR DESIGN Approved 04/28/89.. Date: / / Page 4 of 4 FACILITY MODIFICATION DOCUMENTATION CHECKLIST APPENDIX B FACILITY MODIFICATION NO: 92 -CO 3 TITLE: [EpLAC.E.&&NT t%- EGAc7nE ?RtHAKY $ND $coss/Z)ACY COOLANT' % MPeRArd/W PROBE U]EU S DRAWINGS: NUMBER TITLE REVISED BY DATE 045-si-cq D10 Pipes T> <a s pA sg 2 cc 2 Aas LAyour J' PROCEDURES: NUMBER TITLE REVISED BY-DATE M M S-Reviewed By: Date:

1 \\ i f L REPLACEMENT OF REACTOR PRIMARY AND SECONDARY COOLANT TEMPERATURE PROBE WELLS FACILITY MODIFICATION 92-003 1.0 PURPOSE The purpose of this facility modification is to replace the existing primary and secondary coolant temperature probe _ wells with wells sized f or the new resistance temperature detectors (RTDs). Installation of the new RTDs was' approved by Facility: Modification 92-002. 2.0 SCOPE The proposal is to replace the temperature probe wells. 3.0 RESPONSIBILITY The approval for this modification lies with the NNRC director with the concurrence of the Nuclear Safeguards Committee.

4.0 REFERENCES

4.1 Omega Temperature Handbook pages B-11 and B-28 5.0 SYSTEM DESCRIPTION 5.1 Existing temperature probe wells Equipment: -i a. 2 each temperature probe wells for primary. coolant b. 2 each temperature probe wells for secondary coolant 5.2 Problem with existing temperature probe wells The existing temperature probe wells inside ' diameter.-is. approximately 0.375 inches and the new RTD outside diameter is ' 0.25 inches. Initially the plan-was install a brass sleeve over the new RTDs to fill the space and use.the = existing j wells. More ef ficient ' heat - transf er will occur. by using wells sized for the new RTDs. 1 i 5.4 Proposed temperature probe wells I I Equipment: H 4 each Omega model.3/4-260S-U4 1/2-304SS lI l

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%"-PTBS-12 12" $28 W" %"-PTBS-12 12" $25 l %"-PTBS-18 18" 31 '/2" %"-PTBS-18 18" 28 .? I %"-PTBS-24 24" 33 W" % "-PTBS-24 24" 30 ga 5 l 2 Nominal %" Schedule 40-304 SS Pipe Nominal W" Schedule 40-304 SS Pipe l ! (.8241.D. x 1.050 0.D.) (.6221.D. x.840 0.D.) a I t S THREAD CAT. NO. LENGTH PRICE THREAD CAT. NO. LENGTH PRICE } } 7 %"-PTSS-12 12" S43 v2" %"-PTSS-12 12" S40 l %"-PTSS-18 18" 52 W" %"-PTSS-18 18" 48 l -I %"-PTSS-24 24" 70 W" % "-PTSS-24 24" 67 l 1 ..e. - f i {.7 Techn:calDATA l UEf Male NPT thread on one end. spun closed on the other end. l O',) r. Max. Continuous Mech. Strength Material Operating Temp. Cold Hot Applications Remarks g - W.![. g .. ~ " 304SS 1650*F Excellent Fair Food processing. Dairy Embrittles in 800*F to l' "*{Q ^ prod., mild acids, alkalies. products. Petroleum 1400*F range. i ... gciej i Black 1200*F Excellent Good Molten Babbit. Tin. Lead Low cost j Steel and Magnesium e p 6 M*WfWSJ@g ' Sts-2'wu25fS S i f, Z P j W @ F_" TM: t i Honto;ordermpg y Discount Schedule: i e-m. and i 1 10 units . Net Quantity discount applie5 to N.PJ.thrend 11 24 units. .10% like tubes. Other sizes 46ConsultSalesforPhcf materials and configur'ations 25 100 units. . 20% - [ N-i -b W /p 101 and up. nsult by quotation. ..._ _ _ BAIL _ - --- -------

y Series 260S 1 G Prottetive Coatings For Thermowells- - Standard Threaded Well for %" Diameter Elements

  • ResistCorrosion
  • ForChemicalBaths e
  • Color-CodedSensorsforProcessControl Application:
  • Available in PFA Teflon *, Epoxy and other materials Standard Length, %" Stem, Bimetal therm 0m,eters; #20 Please consult our Sales Department for complete information.

gage thermocouple elements; unarmored liquid-in-glass Materials-test thermometers. Other temperature sensing elements Brass (ash M B-16); Carbon Steel (C-1018); Stainless Steel having.252 in. maximum diameter. A.I.S.I. 304 & A.LS.I. - 316; Monel. Wells are available also in Connection Size: special materials, prices on request. W". %" and 1" NPT are standard. Other thread sizes are Cap and Chain Options: For Brass cap, add $4 to price and add availablo upon request. suffix-CC(Brass)totheendof thecatalognumber. For 304SS cap, adJ $7 to price and add suffix-CC (304SS) to the I endof thecatalognumbar. Saries 260S-General Use .e 1% U INSERTONLENOTH All ditnenstons arein mehes. +- 2% [4 W THREAD ALLOWANCE S' I l OLA. m 4 i - M .n (I7D q n _ w oi4. W NPSM k to02 f P NPT .2e0 SORE j 'NPSM internal psps throod will { accept both NFT and NPS male l c A STEW LENOTH ~3 =e.% threads. f .go When ordering probes udth NPT Fittings specify this stem length. 1 %" 260s -U 2% C i 4. 2% i 5 22.00 l $ 30.00 ' $16.50 s16.50 ji i U 4% C ' 6 4% 26.50 36.00 20.00 20.00 l MostPopularslaes $ U 7%0 g 9 7% 39.50 5100 31.50 31.50 e %" NPT l -U10% D 12 10 % 46.00 65.00 41.00 41.00 ~ -U13% D 15 13 % 69.00 93.00 $3.00 5100 i: Ut6% C 16 16 % 82.50 111.00 63.50 63.50 +U22% 0 _ '[4 l ??w 10600 143 00 74 00 74 00 'i ~ %"-260S U 2%D 4 2% 22.00 30 00 16.50 1650 Most Popular sizes $ *u 7% 0 U 4%0 6 4% 26.50 36.00 20.00 20.00 9 7% 39.50 5100 31.50 31.50 %"NPT -U10% D 12 10% 46.00 65.00 41.00 41.00 -U13% D 15 13 % 69.00 93 00 53.00 $300 U16% D 18 16 % 62.50 111.00 63.50 63 50 -022% D 24 22 % 106.00 143 00 74 00 74 00 1"-260s.U 2% D 4 2% 29.00 35.50 22.00 22.00 Most Poputat sl es $ +U 4% 0 6 4% 36.50 49 00 26.50 26.50 t -O7%Q 9 7% 48 00 65.00 34.50 34 50 1" NPT .U10% a 12 10 % 60.00 80.00 46.00 46,00 -U13% D 15 13 % 6100 111.50 61.00 61.00 -U16% D 16 16 % 97.00 130.50 71.50 71.50 3 -U22% D 24 22 % 132.50 164 00 83 00 83 00 ~ U l Matenal. 304 S.S.. Carbon Steet, etc. #1G#US#TED MODE '~ l Insertson tengtn dimension Bore size-6nch EfUONU[UkN ' Estomal tnrend. NPT DBlVERY. iessets Termperaturerating-Ibs.persq. inch These wells are compatible with OMEGA

  • NB1, NB2 Brass 5000 4200 1000 B 3 & B-4, PR12, NP% M) )and caroon steet 5200 5000 4800 4600 3500 1500 14 W E-5 A t.s I 304 7000 6200 5600 5400 5200 4500 1650 A t s.1 316 7000.

7000 64to 6200 6100 5100 2500 . B 4), as WeX as ' uonei e500-e000 5400 5300 5200 1500 b Page B-27 for Maximum Fluid Vetoetry emp* 0 j r To order, piesse specify: 1. Complete Type Nurnber

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1-10......... Net Use OT 201 Conouctive Sillcon Paste P*8' 11-24...,,.. 10% Discounts apply to similar thermowell types 25 100....,,. 20% 101 and up . Consult - - ~ ~

.s- + 4 s t-s 5.5 The new temperature probe well material.is 304SS (stainless steel) which same as existing wells-(ref dwg # 045-51-001 sh. 2 of 2). W 5.6 Included are copies Omega Temperature Handbook pages B-11 and.' B-28. F t f --g 9}}

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