ML20106G428
| ML20106G428 | |
| Person / Time | |
|---|---|
| Site: | Neely Research Reactor |
| Issue date: | 02/27/1996 |
| From: | Karam R Neely Research Reactor, ATLANTA, GA |
| To: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| References | |
| NUDOCS 9603040314 | |
| Download: ML20106G428 (34) | |
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/,u 7 'i Georgia Institute of Technology j4 yd . -
NEELY NUCLEAR RESEARCH CENTER 4 % .,, -# - 900 ATLANTIC DRIVE
's, , ' "" "8 ATLANTA GEORGIA 30332-0425 USA (404) 894-3600 February 27, 1996 U.S. Nuclear Regulatory Commission Region II 101 Marietta Street, N.W.
Atlanta, GA 30323
Reference:
Annual Report Docket 50-160; License R-97 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, 1995, through December 31, 1995 (calendar year 1995).
The designation of the sections below follow the title and order of Section 6.7.a of our Technical Specifications.
- 1. OPERATIONS
SUMMARY
- a. Chances in Facility Desian There were three facility design changes during calendar year 1995. The changes were approved by the Nuclear Safeguards Committee. All design changes are described in Appendix A.
- b. Performance Characteristics During the reporting period, the reactor was operated at power levels up to 5.0 MW using a 18-element core. A five element fuel exchange to enhance self protection was performed. The fuel performance, with regard to its ability to contain and isolate fission products, continues to be satisfactory with no known problems.
Minor repairs were made on some of the equipment (see Table 1).
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rw i U.S. Nuclear Regulatory Commission - Annual Report February 27, 1996 .
Page 2
- c. Chanaes in Operatina Procedures f The list of new and/or revised procedures which were !
I approved by the Nuclear Safeguard Committee during calendar year 1995 were as follows:
4 Proc.'# Title I 4902 Corrective Maintenance 3800 Liquid Waste Disposal 1
7272 Log N Period Amplifier Calibration 7280 MAP-1 Recorder Calibration 7281 Temperature Recorder Calibration -
Thermocouple 9013 Calibration and testing of Moving Air Particulate Monitor 9018 Charcoal Cartridge Analysis 9160 Calibration of the LB5100-W Counting System 1500 Irradiated Fuel Discharge to Storage Pool 1501 Lower Top Shield Plug Removal from Spent Fuel 1505 Preparation and Off-Site Shipment of Irradiated-Fuel 1506 Physical Protection of Irradiated Fuel in l Transit l l
1507 ~ Emergency Threats to Irradiated Fuel in I Transit i 1508 Inspection, Testing and Operating Procedure for 6-M Drums i
1510 BMI-1 Maintenance, Inspections and Tests 1511 BMI-1 Cask Operating Procedure ,
1512 Irradiated Fuel Shipment by NAC-LWT Cask i
l I
I U.S. Nuclear Regulatory Commission - Annual Report l February 27, 1996 Page 3 Proc. # Title 9400 Environmental Monitoring 9501 Control & Accountability of Radioactive Sources There were two procedures canceled:
l 4900 System Work Sheet l
4901 Preventive / Corrective Maintenance on Safety Related Equipment
- 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 Acoroved by USNRC There were no changes, tests or experiments that required I the approval of the USNRC pursuant to 10 CFR 50.59(a). '
- f. Current Staff and Nuclear Safeauards Committee Membershio Dr. R. A. Karam, Director, Nuclear Research Center and Reactor Engineer l Mr. Dixon Parker, Reactor Supervisor Dr. R. D. Ice, Manager of the Office of Radiation Safety Mr. B. D. Statham, Electronic Engineer (approximately half time)
Mr. Neil Copeland, Senior Reactor Operator Mr. Johannes Strydom, Senior Safety Engineering Assistant Mr. Edgar Jawdeh, Health Physics l Ms. Debbie McGeorge l Mrs. Arlene R. Smith In addition, the NNRC employed the following graduate students on part time basis:
Peter Newby, Senior Reactor Operator
! Jeremy Sweezy, Senior Reactor Operator l Dwayne Blaylock, Senior Reactor Operator i i
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U.S. Nuclear Regulatory Commission - Annual Report February 27, 1996 '
Page 4 Chris Comfort, Reactor Operator Trainee Ralph Demeglio, Reactor Operator Nick Jenkins, Reactor Operator l Shane Klima, Reactot Operator Trainee i Katherin Norton, Reactor Operator Trainee Tina Weatherman, Reactor Operator Trainee The current membership of the Nuclear Safeguards Committee is: ,
1 (1) Mr. Emsley Cobb, Chairman Disciplines Reactor Operation and Reactor Safety (2) Dr. Bernd Kahn !
Disciplines Radiation Protection and Environmental Measurements j (3) Dr. Robert Braga l Disciplines Chemistry 1
(4) Dr. Prateen V. Desai, Secretary I Discipline: Thermal Hydraulics, Mechanical Systems j l
(5)
Dr. Billy R. Livesay, Member Discipline Material Science, Physics (6) Mr. Jack Vickery, Member Discipline Security (7) Dr. Thomas G. Tornabene, Member Disciplines Biology and Biochemistry (8) Dr. S. M. Ghiaasiaan, Member Disciplines Nuclear Engineering (9) Mr. Len Gucwa, Member Disciplines Reactor Safety l l
(10) Mr. Steve Ewald, Member l Discipline Health Physics I (11). Dr. Peggy Girard, Member Disciplines Biology and Biochemistry (12) Mr. James O'Hara, Member l Disciplines Health Physics i
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U.S. Nuclear Regulatory Commission - Annual Report February 27, 1996 Page 5
- 2. POWER GENERATION For the period January 1,1995, through December 31, 1995, the total power generation of the GTRR was 244.98 MW hours. The reactor was operated a total of 151.6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />s: 21.9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> at power levels equal to or less than 100 kW, 81.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> at power levels 100 kW to 1 MW, and 48.7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> at power levels above 1 MW.
- 3. SHUTDOWNS During this reporting period there were 9 unscheduled shutdowns. Table 1 gives details.
TABLE 1 UNSCHEDULED REACTOR SHUTDOWNS DURING 1995 Corrective Number Date Scram Cause Action 95-01 1/18/95 Manual Scram Operators Stop Leak scrammed added to reactor in shield ;
response to a system. I D O leak alarm caused by a light water leak in the shield system.
l 95-02 10/23/95 Negative Operator Instructed Period Trip induced operators to control rod move rods motion caused more slowly negative during period trip shutdown during procedures.
shutdown.
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U.S. Nuclear Regulatory Commission - Annual Report February 27, 1996 Page 6 Corrective Number Date Scram Cause Action 95-03 10/25/95 Power Trip The reactor All operators power was 5 were trained MW. The in high power reactor was operations.
in auto Operators control. The were reminded reg rod was to carefully near high watch all limit. trainee Operator actions, instructed
! trainee to i move shim j blade to i reset reg rod l position.
The trainee i incorrectly l sbut off auto j controller j' and the shim j motion caused trip.
95-04 10/25/95 Control Air During All solenoids Low Pressure calibration were serviced Trip of the Kanne and tested.
chamber, All electrical functioned noise caused properly.
building Problem has isolation. not All building reoccurred.
isolation valves shut, but one air solenoid '
valve stuck and caused air loss and low pressure in the air l system.
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j U.S. Nuclear Regulatory Commission - Annual Report February 27, 1996 j Page 7 i'
Corrective Number Date Scram Cause Action 95-05 10/26/95 Manual Operators Regulating
-l Shutdown shut down Rod Drive the reactor was in response inspected.
j . to an The angle inoperable gear was
, regulating found
- rod. broken. An
, exact t
replacement was found and installed.
The reg rod was tested and functioned properly.
95-06 .11/1/95 Low Bismuth Movement of Operators Coolant Flow the biomed instructed Trip shutter to watch caused bismuth material to tank level block the carefully bismuth during water operation collection of the
, system. biomed facility.
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U.S. Nuclear Regulatory Commission - Annual Report February 27, 1996 l
Page 8 Corrective l Number Date Scram Cause Action 95-07 11/8/95 Low Ion Flux amp #2 Flux amp Chamber trouble reset Voltage light was immediately.
on, Inspection of indicating a the cable and problem with testing of the flux amp the flux amp or the and power cable. supply showed l
no problems.
, The problem
- l. has not reoccurred.
! 95-08 11/9/95 Negative When an Training Period Trip operator was session to showing a emphasize the i trainee the auto use of the controller auto control use and the system the proper reactor response to power the high
- increased. power.
I Operator i
induced control rod motion to bring the power back down caused the trip.
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U.S. Nuclear Regulatory Commission - Annual Report February 27, 1996 Page 9 Corrective Number Date Scram Cause Action 95-09 11/17/95 Door Open While Operator Trip exiting the reprimanded emergency and others airlock, the were informed doors open of error.
trip occurred.
- 4. UNSCHEDULED MAINTENANCE ON SAFETY RELATED SYSTEMS AND COMPONENTS There were approximately thirteen (13) minor repairs performed on safety-related systems and components. Records of maintenance performed on components are available at NNRC offices for inspection.
- 5. CHANGES, TESTS AND EXPERIMENTS During 1995, there were 36 approved experiments which used the GTRR. The experiments were evaluated prior to their approval with regard to section 3.4 of the Technical Specifications.
There were no new experiments which required approval from the Nuclear Safeguards Committee.
I l U.S. Nuclear Regulatory Commission - Annual Report i February 27, 1996 l
l Page 10
- 6. RADIOACTIVE EFFLUENT RELEASES
- a. Technical Specification 6.7. (6)(a) - Gaseous Effluents -
Summation of All Releases Via Stack, i.e., ground level release, j (1) PISSION AND ACTIVATION GASES Tritium Released (gaseous)
Non Measurable Argon-41 Released Total Total Avg. Avg. Released over Max. Inst. %
Release Release period of reactor Release Tech (C1) (pCi/cc)** opercclon (pCi/cc) (pCi/sec)* Specs l'* Otr 7.744 6.15 E-08 1.51 E-05 304 51.97 !
2"8 Otr 1.922 1.53 E-08 2.68 E-06 114 19.49 3ra Otr 5.802 4.60 E-08 1.26 E-05 285 48.72 4th Qtr 27.008 2.14 E-07 1.61 E-05 266 45.47 Annual 42.476 8.43 E-08 1.29 E-05 304 51.97
- Technical Specifications release limit is 585 pCi/sec.
- Basis = Stack effluent at 34,000 cfm ( 1.26 E+14 cc/QTR )
(2) IODINES RELEASED None Measurable Lower LLait of Detection <1.15 E-14 pCi/cc l l
l 4
(3) PARTICULATES j
None Measurable (LB-5100)
Lower Limit of Detection gross beta / gamma = <5.32 E-06 pCi Lower Limit of Detection gross alpha = <3.45 E-06 pCi i
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U.S. Nuclear Regulatory Commission - Annual Report February 27, 1996 Page 11
{
- b. Liquid Effluents (1) TOTAL GROSS RADIOACTIVITY ($/ gamma)
Total Average Maximum % Tech j Release Release Rate
- Conc. Released Specs
! Ci (pCi/cc) (yci/cc) ist QTR 4.65 E-07 1.70 E-12 2.00 E-08 < 1%
2nd QTR 1.41 E-06 S.14 E-12 7.39 E-08 2.5%
! 3rd QTR 7.90 E-07 2.88 E-12 2.31 E-08 < 1%
4th QTR 1.01 E-06 3.70 E-12 5.18 E-08 1.7%
Annual 3.68 E-06 3.35 E-12 7.39 E-08 2.5%
Average release rate values are based on a Georgia Tech campus water discharge rate of 2.743*10" ml/ quarter.
(2) TOTAL GROSS RADIOACTIVI_Tl (Alpha) l Total Average Maximum % Tech Release Release Rateb Conc. Released Specs C1 (pCi/cc) (yCi/cc) ist QTR <MDA* <MDA* <MDA* < 1%
l 2nd QTR 6.54 E-08 2.38 E-13 4.67 E-09 < 1% I 3rd QTR 7.09 E-08 2.58 E-13 5.46 E-09 < 1%
4th QTR 8.70 E-08 3.17 E-13 7.96 E-09 < 1%
Annual 2.23 E-07 2.04 E-13 7'.96 E-09 < 1%
l a. Lower than minimum detectable activity
! b. Average release rate values are based on a Georgia Tech I campus water discharge rate of 2.743*10" ml/ quarter.
1 (3) FISSION AND ACTIVATION PRODUCTS I l
l 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 storage
, pool that is part of the State of Georgia I i Radioactive Materials License No. 147-1-SNM. No ;
l fission products are released via the liquid l effluent pathway.
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U.S. Nuclear Regulatory Commission - Annual Report ;
February 27,.1996 Page 12 l
(i) CO'8 RELEASE i
Total Average Maximum % Tech Release Release Rateb Conc. Specs ,
C1 (pCi/cc) Released r (yCi/cc) l 1st QTR 2.16 E-05 7.88 E-11 5.57 E-07 1.9%
2nd QTR <MDA* <MDA* <MDA* < 1%
3rd QTR <MDA* <MDA* <MDA* < 1% ~i 4th QTR <MDA* < MDA* <MDA* < 1% !
Annual 2.16 E-05 1.99 E-11 5.57 E-07 1.9% i
- a. Lower than miminum detectable activity
- b. Average release rate _ values are based on a Georgia Tech campus water discharge rate of 2.743*1022 al/ quarter.
Co" Lower Limit of Detection = < 1.44 E-7 uCi/cc.
(ii) TRITIUM Total Average Maximum % Tech Release Release Rate
- Conc. Released Specs Ci (pCi/cc) (yCi/cc) ;
lat QTR 1.06 E-02 3.85 E-08 2.90 E-04 2.9%
2nd QTR 6.91 E-03 2.52 E-08 2.49 E-04 2.5%
3rd QTR 1.30 E-02 4.74 E-08 1.95 E-04 2.0%
4th QTR 2.05 E-03 7.47 E-09 8.17 E-05 < 1%
Annual 3.25 E-02 2.97 E-08 2.90 E-04. 2.9%
Average release rate values are based on a Georgia Teca campus water discharge rate of 2.743*1022 al/ quarter.
(4) TOTAL VOLUME OF LIOUID WASTE RELEASFD let QTR . . . 5.53 E+07 ml 2nd QTR . . . 3.90 E+07 al 3rd QTR . . . 1.19 E+08 ml 4th QTR . . . 4.88 E+07 al ANNUAL . . . 2.62 E+08 ml (5) GEORGIA TECH VOLUME OF DILUTION WATER USED let QTR . . . 2.743 E+11 ml 2nd QTR . . . 2.743 E+11 al 3rd QTR . . . 2.743 E+11 ml 4th QTR . . . 2.743 E+11 ml ANNUAL . . . 1.097 E+12 ml
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U.S. Nuclear Regulatory Commission - Annual Report
- February 27, 1996 Page 13
- 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) operationa 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, I
" Environmental Monitoring Sthtions". The sites are l predominantly around the reactor perimeter fence or down- )
wind from the reactor.
l
- b. Total assays = 30 sites X 4 quarters = 120 assays. The results are reported in the Environmental Radiation i Surveillance table (attached). The letter M was used to designate any reading which was less than the minimum detectable limit.
- c. Monitors are Landauer "X9" aluminum oxide thermoluminescent dosimeters (TLD). The dosimeters meet ;
ANF,I standards. l The dosimeters positioned around the facility showed only-l very low radiation exposure due to the reactor l operations. Radiation exposure due to reactor operations !
is best estimated from TLD #1 positioned inside the !
reactor building stack. Exposure recorded by this film badge is directly attributable to reactor operations.
Because of its location, i.e. inside the reactor building stack, it is not representative of environmental exposures, but rather, represents a worst case exposure.
Thermoluminescent dosimeter (TLD #9) is located on the perimeter fence near the Georgia Tech Short-Term Radioactive Waste storage and preparation facility licensed by the State of Georgia.
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U.S. Nuclear Regulatory Commission - Annual Report February 27, 1996 Page 15 Thermoluminescent dosimeters (TLD's 17 through 24) are closely position to a granite wall. We attribute the majority of exposure to these dosimeters to natural radioactivity in the granite.
Landaurer reports that 8 dosimeters out of 30, averaged over the year, have radiation levels greater than local background. Note: The exposures on the Table assume a person at that site for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> per day, 365 days per year.
- d. The highest, lowest and the annual average levels of radiation for the sampling point (TLD #9) with the highest average radiation exposure and location of that point with respect to the site.
Average Annual Level - 15.3 mrem /yr Highest Level -
9.7 mrem /qtr Lowest Level 4.6 mrem /qtr
- o. The gross dose rate readings for all TLDs from all stations varied between 30 and 60 mrem per quarter. The control TLD station varied between 34 and 60 mrem per quarter. This range of variation produced some net dose rate readings (gross reading minus control or background reading) that are negative. The negative readings are replaced by the letter M in the Table. Statistically no conclusions can be made about the environmental dose attributed to the GTRR operation. It is very small.
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NEELY NUCLEAR RESEARCH CENTER ENVIRONMENTAL RADIATION SURVEILLANCE
- 1995 A B C D E f jan i . Apr 1 July 1 Oct 1 1995 loial Comments Mar 31 lune 30 Sept 30 Dec 30 Laudauer I 1 5.7 M 6.5 8.8 5.0 in stack J
l 2 M M M M M 3 4.5 M M 4.3 M 4 5.6 M M 4.5 M 5 3.3 M 0.2 1.7 M 6 8.1 M M 2.5 M l
7 40 M 1.2 2.7 M l I
8 9.0 M 3.2 5.4 11.3 1
9 9.7 M 46 8.3 15.3 Rad Waste Bam 1
10 8.4 M 1.6 0.8 1.5 l
11 5.2 M 0.1 3.8 0.7 l l
12 6.6 M M 5.6 M 13 9.2 M 6.7 3.4 11.2 14 0,3 M M M M 15 M M M M M 16 M M 2.5 4.8 M 17 10.8 0.4 4.4 4.9 20.4 18 5.0 M M 2.6 M 19 5.2 M 2.0 4.5 M 20 6.0 M M 1.3 M 21 6.6 M 1.7 2.6 M 22 6.6 M 0. 7 Ab M 23 6.0 M M 2.7 M 24 8.4 M 1.9 6.8 9.9 Gramte Wall 25 2.7 M M M M 26 1.2 M M M M 27 M M M 14 M 28 6.0 M M M M 29 M M M M M 30 1.6 M Ab M M WoAload MW+tR5 38.84 12.08 34.56 159.50 244.98
- 5um of natural radwiean, direct radiaium from facility and gascum radioactive effluents momfored with Al,0, TLD's less control badge kept at G1 Police Dept. Badges procened by Landauer.
The lower limd of detection is 0.1 meem. All negative readings are indicated by M. Atnent Ab.
U.S.-Nuclear Regulatory Commission - Annual Report ,
February 27, 1996 Page 17 i
- 8. OCCUPATIONAL PERSONNEL RADIATION EXPOSURE 1995:
Radiation workers of Georgia Institute of Technology are monitored through the use of film badges which are provided by a NVLAP certified vendor and have a lower limit of detection of < 10 mrem. A monthly radiation dosimetry report is issued for the personnel of the Neely Nuclear Research Reactor, a-summary shown in Table 1. i
- a. Summary of exposure for persons under 18 years of age greater than urem -
None
- . b. Summary of occupational exposures greater than 500 mren - t None i
- c. Person-Rem for the Neely Nuclear Research Center - R-97.
l Person-Rem = Sum of occupational workers = 0.490 rem The highest, lowest and average levels of personnel l exposure due to reactor and hot cell operations: j l'
Average annual level - 20 mrem l Highest annual level - 100 mrem l Lowest annual level - < 10 mrem. 4 I
- d. Category of exposure '
NNRC Radiation Workers l
Annual Deep Dose # Radiation workers
< 10 arem 11 10 mrem - 49 mrem 8 50 mrem - 99 mrem 5 1 100 mrem - 149 mram 0
- 150 mrem - 199 mrem 0 i
> 200 mrem 0 i,
Total Workers 24/490 mrem total i
i I
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, , l U.S. Nuclear Regulatory Commission - Annual Report February 27, 1996 Page 18 Should there be any questions concerning this report, please let us know.
Sincerely, A-~ AA R. A. Karam, Ph.D., Director Neely Nuclear Research Center RAK/dmcg cc 1. Dr. Jean-Lou Chameau
- 2. Dr. John White
- 3. Members Nuclear Safeguards Committee
- 4. Director, Office of Nuclear Reactor Regulation l U.S. Nuclear Regulatory Commission ,
Wgshington, D. C. 20555 i
- 5. ptfocument Control Desk {
U.S. Nuclear Regulatory Commission j Washington, D. C. 20555 '
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a 5 t e APPENDIX A f
. . - . -- . . ~ . . . - . -. . . - . . .-.
NEELY NUCLEAR RESEARCH CENTER Minor Change Procedure 4200 1
Number: Revision 00 i
By: CHANGES IN GTRR DESIGN Approved 04/28/89 Page 3 of 4 Date / / ( _
APPENDIX A 10 CFR 50.59 SAFETY EVALUATION QUESTIONNAIRE FACILITY MODIFICATION NO: 6-00t TITLE: St0.cllIdid m /,M Afveu 1
- 1. Will the probability of the occurrence or the consequences of an accident or malfunction of equipment important to safety j previously evaluated in the safety analysis report be increased? [yes/no] A/d i
- 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) 4/ ,,
- 3. Will the margin of safety as defined in the basis for any technical specification be reduced? [yes/no) A1.
1 l
l
- 4. Is the proposed change an unreviewed safety question? -
(yes/no) da NOTE: If additional space is needed to justify conclusion (s) please attach extra sheet (s).
I DATE: ;
PREPARED BY: ' heu'%
~
24 05 l
APPROVALS:
Director NNRC: 4, M k[
Nuclear Safeguards Committee: g/ /G(
s, NEELY NUCLEAR RESEARCH CENTER Minor Change Procedure 4200 Number: Revision 00 By: CHANGES IN GTRR DESIGN Approved 04/28/89 Dater / / Page 4 of 4 FACILITY MODIFICATION DOCUMENTATION CHECKLIST APPENDIX B FACILITY MODIFICATION NO:
TITLE: l4.1 (.A ,(l d .' ~ d o ~ btM d{de DRAWINGS:
NUMBER TITLE REVISED BY DATE l
I PROCEDURES:
NUMBER TITLE REVISED BY DATE l
l j
Reviewed By: Dates
-_ == .
Facility Modification 95-001 Hot Cell Window Level Alarm Description j l
The hot cell windows are two zine bromide filled viewing windows necessary for the !
operation of the hot cell. The zine bromide acts as a radiation shield during hot cell experiments.
If the level of either window were to drop below the upper steel shielding of the window assembly while sources were present in the hot cell, a beam of radiation would escape through !
the window possibly endangering the operators or the public. The current window level alarm is connected to the criticality alarm system. The new level alarm will be connected to the criticality alarm system in the same manner. This modification improves the window level detection system and does not change the intended function of the system.
Current Design !
i The current hot cell window level alarms consist of a float switch assembly in each i window. An acrylic float connected to a rod and plate activates a microswitch if the zine bromide level in either window drops (Diagram 1). The trip level is adjusted by moving the plate up or down the metal rod. The current system has several problems. First, the sensitivity of the current system is not sufficient to ensure that the alann is activated if only a small drop in window level occurs. Second, the acrylic floats are seriously degraded by the window solution and need frequent replacement. Third, the microswitch is deteriorating due to corrosion caused by the zine bromide solution. Finally, the float rod guide occasionally sticks in the window plug and could permit a leak of several inches before breaking free and activating the alarm.
New Design The new system senses the window level by monitoring the conductivity between two copper probes dipping down into the zine bromide (Diagram 2). Since the walls of the windows are lined in copper, deterioration of the probes is not anticipated. The sensing circuit (see Circuit Diagram) supplies around 25 millivolts and one microamp. This current is not sufficient l to deteriorate the zine bromide solution. The height of the two probes is easily adjusted and the
alarm is activated as soon as the solution breaks contact with either probe. Each window will have its own set'of probes but uses a common circuit.
The probes will be threaded so that the trip level can be adjusted by tightening or loosening a nut on the probe and observing the probe. The system will be tested by slowly
', , extracting one of the probes from the liquid until an alarm sounds. Operators can check to see if
) the system is operating by observing LED's on the circuit board. This system will allow for more reliable operation of the window level alarm. Also, this design allows for more accurate control over the trip point.
Failure Modes The window level alarm is designed to be fail safe. The relay that activates the alarm is normally opened and must be energized to deactivate the alarm. If the power to the circuit or the continuity of any wire is lost the alarm will activate. Also, the probes have been designed so that neither touches the same surface except for the zine bromide. This will prevent any spilled solutien from keeping the circuit closed if the window level should drop. The design also prevents the probes from contacting each other and defeating the alarm.
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i MEMORANDUM i
DATE: 3/6/95 i
Jerry Taylor, Manager of Hot Cell Operations i TO:
L
- i. FROM: Dixon F. Parker, Reactor Supervisor M
SUBJECT:
- Hot Cell Window Level Alarm l
l 1-As part of the recent modification of the hot cell window level alarm system the Nuclear 1 Safeguards Committee stipulated that formal testing of the new sensor be performed. Also, you l l
' must verify the operability and you familiarization with the system in writing prior to l commencing any operation in the hot cell. The system is described in the facility modification package.
l
! Several points to keep in mind while testing the system are:
- 1. Do not touch the metal part of the middle probe with bare skin as this will ground out the system and prevent the test from working properly,
- 2. The center probe must not extend below the steel plates on the upper portion of the !
window. If a window leak occurred this would cause a slit beam to appear prior to the alarm being activated.
- 3. The side probe will not necessarily give an alarm ifit becomes uncovered.
- 4. The probe level can be easily verified by visually inspecting liquid level in the 1.5 inch hole where liquid is added to the windows. The probes can be seen projecting below the surface of the liquid.
l I suggest that you test the system several times to gain familiarity with the sensitivity and
! adjustment capability of the new prebes. Dr. Karam has requested that he be present when you do so. After testing send a memo to file describing what testing actions you have taken, and l
L confirming your acceptance of the system. Also I have attached a training sheet for you to sign. i j
1 have given verbal instruction on the system operation to you, Dr. Ice, Peter Newby, and Billy '
Statham. Please ensure that all of them sign the sheet. Also, I will train any additional personnel that you feel need to be familiar with the system.
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NEELY NUCLEAR RESEARCH CENTER Minor Change Procedure 4200 Number:
Revision 00 By: CHANGES IN GTRR DESIGN Approved 04/28/E Date: / /
Page 3 of 4 APPENDIX A 10 CFR 50.59 SAFETY EVALUATION QUESTIONNAIRE FACILITY MODIFICATION NO: 4T /5 - 00Z.
TITLE: [GPLALEPfEff- Oc ff)E FigG ALARM'
~TfAd.sHisstor) 0 NIT
- 1. Will the probability of the occurrence or the consequences of an accident or malfunction of equipment important to safety previouslyevaluatedinthe/Dsafety increased? [yes/no] A analysis report be
- 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] AlO
- 3. Will the margin of safety as defined in the basis foyNOany technical specification be reduced? [yes/no]
- 4. Is the propos,ed change an unreviewed safety question?
[yes/no] MO ,
l NOTE: If additional space is needed to justify l conclusion (s) please attach extra sheet (s). l f
f DATE:
PREPARED BY:
3 ~2/~ 9 6
.)
APPROVALS:
Director NNRC: 8.AL h 3/2.3 /9C Nuclear Safeguards Committee: [ / /
NEELY NUCLEAR RESEARCH CENTER Minor Change Procedure 4200 Number:
Revision 00 l'HANGES IN GTRR DESIGN Approved 04/28/8)
By: Page 4 of 4 l Date: / / 1 I
FACILITY MODIFICATION DOCUMENTATION CHECKLIST APPENDIX B FACILITY MODIFICATION NO: 96-COZ TITLE: [EPLACE M'E dT OF 7M6 bRG h_ ARM
~I~EAd SkhSS/ od 0NI7~
DRAWINGS:
NUMBER TITLE REVISED BY DATX PROCEDURES:
TITLE REVISED BY DATE NUMBER ,/
726o PdfoWAYI& Flitc ALMM Tesrds nh No Peac.eDoeA L C+4A NGrE rJeEDErs. 7W6 F:UAlcT/ e A) D/-
T+4 G 2EPL4cr Mtdr di22 ALARM TEA Als Wh_Lt to d l)Al 17~ T~S 34WG AS l Of_T% L) U IT~.
Reviewed By: Date:
1
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FAC LITY MODIFICATION 95-002 REPLACEMENT OF THE FIRE ALARM TRANSMISSION UNIT 1.0 PURPOSE The purpose of this facility modification is to replace the fire alarm transmission unit.
2.0 SCOPE The proposal is to replace the fire alarm transmission unit.
3.0 RESPONSIBILITY The approval f or this modification lies with the NNRC director with the concurrance of the Nuclear Safeguards Committee.
4.0 REFERENCES
4.1 Procedure 7260, Automatic Fire Alarm Testing 5.0 SYSTEM DESCRIPTION The fire alarm transmission unit sends a NOTE:
signal to the Georgia Tech Police Station (GTPS) indicating the condition of the fire alarm at the Neely Nuclear Research Center 4 (NNRC).
5.1 The old fire alarm transmission unit developed a problem was and upon careful inspection of the unit, it determined that it was not practical to repair this unit. l 5.2 The proposed replacement unit consists of one (1) relay, i two (2) diodes, four (4) resistors and two (2) LEDs l making it very straight f orward to repair (if necessary) .
The relay is a time delay relay, set for two (2) second delay on pul1 in; this reduces the possib.ility of sending ,
a false fire alarm to the GTPS during a momentary power interruption. The unit contains a green LED to indicate a safe condition and a red LED to indicate an alarm condition. The power f or the signal send to the GTPS is The battery taken from an existing battery supply.
supply consists of two (2) 12 volt lead acid batteries and a constant trickle charger (supply is also used to power the PA system in case of loss of utility power).
Using the battery power for the GTPS signal prevents a false fire alarm f rom being transmitted during loss of utility power at the NNRC.
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- NEELY NUCLEAR RESEARCH CENTER Minor Changa Procedurs 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 FACILITY MODIFICATION NO: @@03 TITLE: REPLN.2&f.htr OF TdC [GAC.702 AMAR'l (2x>(Adr IwW IrJDcAf/ Akr Adn I?wenide .D4Wcs
- 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) M.
The reliability of the Digital Panel Meter should be greater than the 30 year old flow recorder _.
- 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) No The proposed system will provide the same functions with greater sensitivity. ,
- 3. Will the margin of safety as defined in the basis fgr any technical specification be reduced? [yes/no) A/o The margin of safety should be increased with the proposed system because of greater reliability and increased accuracy in the ability to set the scram point.
- 4. Is the proposed change an unreviewed safety question?
[yes/no) A n Safety questions will be the same for both systems. The proposed syctem, providing the same functions, with greater reliability and increased . sensitivity has no unreviewed safety questions.
DATE:
PREPARED BY: /d/U X N T'4T//AA[ 7-/f-9 3 APPROVALS:
Director NNRC: h4- Mtu 7hO!7(
Nuclear Safeguards Committees - 9-10 I
,' ., *i. NEELY NUCLEAR RESEARCH CENTER Procndura 4200 Minor Changa Number: Rsvicion 00 By: CHANGES IN GTRR DESIGN Approved 04/28/89 Date: / / Page 4 of 4 FACILITY MODIFICATION DOCUMENTATION CHECKLIST APPENDIX B FACILITY MODIFICATION NO: h5-OC)3 TITLE: EdLAccMENT Or %c Eu M t%AKY 0oot Aar ku)
TA h tc A n orJ A,JD & ncoin he.sc.e DRAWINGS:
NUMBER TITLE REVISED BY DATE D 4 5 - la 2 - 00I $STRt)H'sdrAT/oni hab(hdreoG S*4enT I os 4- &HsMArics c4 5 -6 2-oc t C .5 Ana )
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PROCEDURES:
TITLE EgyISED BY DATE T27 D2O be Escoe.De.R. (.At.igit.pTsod
?R07 EtACToR. C39G A ATs o Al$ ~ % ud.t TICAL STM.TU! CAaCKU5T AslO 514 r-1~
& Pa.a? ViscK A P/Ro/AL 2.006 KeAcroe MarboM Caecxusr Reviewed By: 'b A Date: 7 80 3
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FACILITY MODIFICATION 95-003 .
REPLACEMENT OF THE REACTOR PRIMARY [
COOLANT FLOW INDICATING AND RECORDING DEVICE 1.0 PURPOSE :
! The purpose of this facility modification is to replace the Reactor primary coolant flow recorder with a Digital Panel .
Meter and a strip chart recorder.
i 2.0 SCOPE
! The proposal is to replace the Reactor primary coolant flow :
indicating and recorder system. [
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 Omega DPF700 Operator's Manual I
4.2 omega Operator's Manual for Model 620 Strip Chart :
Recorder 1
5.0 SYSTEM DESCRIPTION 5.1 The existing system has a Potter flowmeter which generates a signal whose frequency is proportional to the primary coolant flow rate. An Acromag frequency to voltage converter changes the frequency signal to a DC 1 millivolt signal. A zero (0) to ten (10) millivolt recorder is used to indicate and record thiscam DCactuated millivolt signal. The recorder has a two (2) switches, one for generating a reactor' trip signal and 3' another for generating the Low D0 Flow annunciator 3 signal. The recorder has a r el a'y that generates a reactor trip signal should the power to the recorder be turned off. A 60 Hertz line frequency signal can be l applied to test the system (providing
~
440 GPM flow indication).
5.2 The replacement system will utilize the existing Potter flowmeter. An Omega digital panel meter (DPF700) equipped with a dual relay board and analog output board l will be used as the indicating device. One relay on the I
dual relay board will generate both the reactor trip signals; the second relay will generate the annunciator l signal. The analog output board will generate a sional for one (1) channel of a dual channel flow recotaer (Omega Model 620 strip chart recorder). A test signal that is near the operating range (~1790 GPM flow) will be
. ~ . .- . . - . . . . . . . . - _ . . ~ - - - - - . - . - . - . . . - - . - . . _ _ - . . . .- .- .
- o provided. The second recorder channel is for future secondary coolant flow recording; this will be a separate Facility Modification.
Existing system l l
Potter Frequency Primary Reactor Trip l Flowmeter- To Coolant Signals !
Voltage Flow Annunciator Converter Recorder Signal 60 Hertz Test Signal !
i Proposed system [
I Potter Primary Reactor Trip Signals l Flowmeter Coolant !
Flow Annunciator Signal l DPF700 Dual Channel Flow Recorder ,
Omega Model 620 l Test Signal I
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