ML20063L497: Difference between revisions
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| number = ML20063L497 | | number = ML20063L497 | ||
| issue date = 02/25/1994 | | issue date = 02/25/1994 | ||
| title = Annual Rept for Ga Inst of Technology Research Reactor | | title = Annual Rept for Ga Inst of Technology Research Reactor | ||
| author name = Karam R | | author name = Karam R | ||
| author affiliation = GEORGIA INSTITUTE OF TECHNOLOGY, ATLANTA, GA | | author affiliation = GEORGIA INSTITUTE OF TECHNOLOGY, ATLANTA, GA | ||
Latest revision as of 16:06, 6 January 2021
| ML20063L497 | |
| Person / Time | |
|---|---|
| Site: | Neely Research Reactor |
| Issue date: | 02/25/1994 |
| From: | Karam R Neely Research Reactor, ATLANTA, GA |
| To: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| References | |
| NUDOCS 9403070040 | |
| Download: ML20063L497 (51) | |
Text
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/, s I7 ' t o f,\ Georgia Institute of Technology j
L f-
.s "? j NEELY NUCLEAR RESEARCH CENTER 900 ATLANTIC DRIVE
's, ,/ ATLANTA, G5 JRGIA 30332 0425 USA (4o41804-3000 l
l February 25, 1994 ,
1 U.S. Nuclear Regulatory Commission Region II 101 Marietta Street, N.W.
Atlanta, GA 30323
Reference:
Annual Report Docket 50-160; License R-97 Gentlement Pursuant to Section 6.7.a of the Technical Specifications for the ,
Georgia Institute of Technology Research Reactor (License R-97), l the following annual report is submitted. The reporting period is j January 1, 1993 through December 31, 1993 (calendar year 1993). i The designation of the sections below follow the title and order of Section 6.7.a of our Technical Specifications.
- 1. OPERATIONS
SUMMARY
l a. Chances in Fac1.?ity Desion i There were seven f acility design changes during calendar year 1993. All were approved by the Nuclear Safeguards ]
Committee. All design _ changes are described in !
Appendix A. I l
- b. Performance Characteristics l I
During the reporting period, the reactor was operated at l
power levels up to 3.0 MW using a 17-element core. An 8-element fuel exchange to enhance self protection was )
performed. Fuel performance continued to be satisf actory 4 with no known problems. I
- c. Chances in Operatino Procedures The list of new and/or revised procedures which were approved by the Nuclear Safeguard Committee during calendar year 1993 were as follows:
280000 . (f 940307oo4o 940225 I gDR ADOCK 05000160 \
Telex. 542507 GTRIOCAATL PDR g ,4 % 3 9323 3,,n,y40 u ,,.3eco)
A Unit of the University System rst Georgia An Equal Ectucation and Employment Opportunity Institut;on
, i U.S. Nuclear Regulatory Commission - Annual Report February 25, 1994 Page 2 Proc. # Title 0005 Criticality Alarm Testing 2002 Reactor Operations-Precritical Startup Checklist and Shift Supervisor Approval 2300 Bismuth coolant System-Operation 3105 Bio Medical Facility Operation :
4200 Changes in Facility Design 4400 D2 Analysis in Reactor Cover Gas 6010 General Rules and Guides for Handling Emergencies 6020 Response to Heavy Water Leakage in Containment Building 6040 Response to Fire at NNRC 6100 Emergency Notification 7200 Primary Coolant Sampling for Radionuclide Analysis 7220 Containment Building Isolation Test-7260 Automatic Fire Alarm Testing 7500 Bismuth System Operation Check 7910 Calibration Test'for Keithley Model 485 Picoammeter 9015 Cooling Water Gamma Monitor 9016 Calibration and Testing of Filter Bank Monitor 9053 Basic Portable Neutron Meter Calibration
i
. . ; i U.S. Nuclear Regulatory Commission - Annual Report February 25, 1994 Page 3 Proc. # Title 9250 Facilities Contamination Surveys 9304 Routine Facility Radiation Surveys 9501 Control & Accountability of Radioactive Sources 9502 Control and Accountability of Radiation Generating Devices 9510 Radioactive Material Shipment
- 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 the approval of the USNRC pursuant to 10 CFR 50.59(a).
- f. Current Staf f 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 (approximately half time)
Mr. William Downs, Senior Reactor Operator Mr. Dixon Parker, Senior Reactor Operator l
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, the NNRC employed the following graduate students on part time basis:
Ms. Kathleen Klee i Ms. Hannah Mitchell j Mr. Thomas Evans Mr. Joseph Martin
U.S. Nuclear Regulatory Commission - Annual Report February 25, 1994 Page 4 s Mr. Nick Jenkins (Trainee for Reactor Operator)
Mr. Neil Copeland (Trainee for Reactor Operator)
Mr. Jeremy Sweezy (Trainee for Reactor Operator)
The current membership of the Nuclear Safeguards Committee is: -
(1) Mr. Emsley Cobb, Chairman Discipline: Reactor Operation and Reactor Safety (2) Dr. Bernd Kahn Discipline: Radiation Protection and Environmental Measurements (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 Disciplines Biology (12) Mr. James O'Hara, Member Disciplines Health Physics 1 l
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U.S. Nuclear Regulatory Commission - Annual Report February 25, 1994 Page 5
- 2. POWER GENERATION For the period' January 1,1993 through December 31, 1993, the total power generation of the GTRR was 75.82 MW hours. The reactor was operated a total of 132 hour0.00153 days <br />0.0367 hours <br />2.18254e-4 weeks <br />5.0226e-5 months <br />s: 34.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> at power levels equal.to or less than 100 kW, 90.2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> at power levels 100 kW to 1 MW, and 7.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> at power levels above 1 MW.
- 3. SHUTDOWNS During this reporting period there were 5 unscheduled shutdowns.
Table 1 gives details.
U.S. Nuclear Regulatory Commission - Annual Report February 22, 1993 Page 6 TABLE 1 UNSCHEDULED REACTOR SHUTDOWNS DURING 1993 Report Date Trip Reason for Trip Corrective Initiation Action 93-01 3/8 High TR-2 Initiated Loose terminal Bismuth trip on the TR-2; Coolant although the trip circuit Temp temperature was tightened.
was normal.
93-02 8/11 Low H2O H2O flow Drained Flow measuring moisture from instrument instrument air malfunctioned system and re-due to moisture paired air in instrument dryer system.
air system 93-03 9/28 Period Operator Repaired trip observing a Period nonfunctioning Recorder; Period recorder modified allowed period procedure to exceed trip requiring limit. operators to observe Period meters in addition to the Period recorder.
93-04 10/11 Low Bismuth Pump Informed Bismuth switch was students to be Coolant accidentally aware of how Flow bumped by a they move about student. control room.
93-05 10/11 Low Air in lines Allow 3 minutes-Bismuth caused after starting Coolant temporary drop Bismuth pump to Flow in flow get air out of lines before resetting trip ,
l circuits.
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l U.S. Nuclear Regulatory Commission - Annual Report February 25, 1994 Page 7
- 4. UNSCJfDULED MAINTENANCE ON SAFETY RELATED SYSTEMS AND COMPONENTS There were approximately twenty-one (21) minor repairs performed on safet.y-related systems and components. Records of maintenance performed on components are available at NNRC offices for inspection.
- 5. CHANGES, TESTS AND EXPERIMENTS During 1993, there were 35 approved experiments which used the GTRR.
The experiments were evaluated prior to their approval with regard to saction 3.4 of the. Technical Specifications.
- 6. EADIOACTIVE EFFLUENT RELEASES
- a. Technical Specification 6.7.(6)(a) - Gaseous Effluents -
Summation of All Releases Via Stack, i.e., ground level release.
(1) FISSION 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 (C1) (pCi/cc)** operation (pCi/cc) (pCi/sec)*
l Otr 3.752 2.98*10-8 9.83*10-6 76 2"d Qtr 1.792 1.42*10-8 6.98*10-6 43.7 3ra Qtr 7.567 6.01*10-8 7.03*10-6 180 4" Otr 6.383 5.07*10-' 5.86*10-6 275.5
- Technical Specifications release limit is 585 pCi/sec.
- Basis = Stack effluent at 34,000 cfm
U.S. Nuclear Regulatory Commission - Annual Report February 25, 1994 Page 8 (2) IODINES RELEASED None Measurable Lower Limit of Detection <1.14*10-" pCi/cc (3) PARTICULATES None Measurable Lower Limit of Detection <6.49*10-5 pCi gross beta / gamma Lower Limit of Detection <7.07*10-* pCi
- 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 No. 147-1-SNM. No fission products are released via the liquid effluent pathway.
(1) CO 6 RELEASE Total Release Avg. Release * % Tech Ci Rate (pCi/cc) Specs 1st OTR 2.45*10-5 1.24*10-2" < 1%
2nd QTR 2.26*10-5 1.30*10-2 < 1%
3rd QTR 2.11*10-5 1.05*10-2 < 1%
4th QTR 1.60*10-5 7.99*10-" < 1%
- Average release rate values are based on a Georgia Tech campus water discharge rate of 2*10" ml/ quarter. ,
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U.S. Nuclear Regulatory Commission - Annual Report February 25, 1994 Page 9 (2) TOTAL GROSS RADIOACTIVITY-(p/ gamma)
Total Release Avg. Release * % Tech Ci Rate (pCi/cc) Specs lat QTR 1.27*104 6.34*10-" < 1%
2nd QTR 4.75*10-7 2.38*10-" < 1%
3rd QTR 1.71*104 8.57*10-" < 1%
4th QTR 1.90*10-7 9.50*10-" < 1%
- Average release rate values are based on a Georg2ia Tech campus water discharge rate of 2*10 ml/ quarter.
(3) TRITIUM Total Release Avg. Release * % Tech Ci Rate (pCi/cc) Specs lot QTR 2.54*10-3 1.27*104 < 1%
2nd QTR 1.76*10-3 8.78*10-' < 1%
3rd QTR 9.47*10-3 4.74*104 < 1%
4th QTR 1.64*10-3 8.22*10-8 < 1%
- Average release rate values are based on a Georgia Tech campus water discharge rate of 2*10" ml/ quarter.
(4) GROSS ALPHA RADIOACTIVITY RELEASED None Measurable Lower Limit of Detection - l
<7.07*104 pCi (5) VOLUME OF WATER RELEASED (ml/Ouarter)
From Reactor Building l
1st QTR . . . 5.45
- 107 ml !
2nd QTR . . . 5.68
- 107 ml 3rd QTR , . . 8.02
- 107 ml 4th QTR . . . 3.97
- 107 ml
U.S. Nuclear Regulatory Commission - Annual Report February 25, 1994 Page 10 (6) VOLUME OF DILUTION WATER UERD DURING EACH OUARTER From Georgia Tech Campus 1st QTR . . . 2.0
- 10" ml 2nd QTR , . . 2.0
- 10" ml 3rd QTR . . . 2.0
- 10" ml 4th QTR . . . 2.0
- 10" ml
- 7. ENVIRONMENTAL MONITORING 1( 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 environmental radiation surveillance table (attached). The letter M was used to designate any reading which was less than the minimum detectable limit.
- 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.
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. Nonetheless, because of its location inside the reactor building stack, it would not be representative of environmental exposures, but rather would represent a worst case exposure.
U.S. Nuclear Regulatory Commission - Annual Report February 25, 1994 Page 11 Several badges indicated radiation exposure above background levels. Badges No. 2, 14 and 15 all gave higher readings in May, June and August. These badges are located on the roof of the Neely Nuclear research Center (NNRC). facility. See the attached figure. The badges were attached to metallic surfaces with duct tape. Generalized increased badge readings were also observed during the months of July and August, two unusually warm months of the year for Atlanta in 1993.
An analysis of the indicated exposures indicates:
(1) The badge (No.1) which would most probably indicate significant gaseous / particulate radioactive reloase showed only "M" minimal exposure all year.
(2) The generalized indicated exposures do not correlate with nuclear operations. See workload at bottom of Table.
(3) Communication with Dr. W. G. Vernetson (TRTR Mtg., Oct.
1993) at the University of Florida nuclear reactor indicate that they also were experiencing anomalous environmental exposuren.
Based upon our analysis, we conclude that the exposures were not correlated with reactor operations and were due to either reflective heat effects or systematic error associated with the badges or badge processing. To prevent reoccurence we 1) installed new dosimetry holders that would allow air circulation around the dosimeters, i.e. prevent reflective heating from attched surfaces, and 2) switched from film badge dosimeters to environmentally specific thermoluminescent dosimeters (TLD's).
The switch to TLD's will also provide a more sensitive (unit millirems versus tens of millirems) assessment plus minimize any
" heating effect" that may have caused our previous anomolous readings.
- d. The highest, lowest and average levels of radiation for the sampling point with the highest average radiation exposure due ;
to reactor operations and location of that point with respect to the site. j i
Average annual level - 27* mrem i Highest annual level - 130* mrom '
Lowest annual level - < 10 mrem.
- There are reasons to question the validity of these numbers (see 7c above).
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U.S. Nuclear Regulatory Commission - Annual Report February 25, 1994 Page 12
- e. Based upon the EPA Comply Code, 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.
- 8. Occupational Personnel Radiation Exposure:
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. All personnel dosimetry data is kept at NNRC. Summary of personnel dosimetry follows.
- a. Summary of exposure for persons under 18 years of age greater than mrem -
None
- b. Summary of occupational exposures greater than 500 mrem -
None '
- c. Person-Rem for the Neely Nuclear Research Center - R-97.
Person-Rem = Sum of occupational workers = 0.16 rem The highest, lowest and average levels of personnel exposure due to reactor and hot cell operations:
Average annual level - 8.4 mrem Highest annual level - 40 mrem Lowest annual level - < 10 mrem.
F w r .- -r-- - --- - r __ _ - _- __ *_m - -_m__
h U.S. Nuclear Regulatory Commission - Annual Report February 25, 1994 Page 13
- d. Category of exposure NNRC Radiation Workers Annual exposure # Radiation workers
< 10 mrem -11 10 mrem - 49 mrem 8 50 mrem - 99 mrem 0 t 100 mrem - 149 mrem 0 150 mrem - 199 mrem 0 .,
> 200 mrem 0 y
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.. .- ~ .. ~. _ . _ . - - . - . . _ . .. -. - .. ..
NEELY NUCLEAR RESEARCH CENTER ENVIRONMENTAL RADIATION SURVEILLANCE, 1993 JAN FFB MAR APR MAY JUN BADCE f D S D S D ',
. D S D S D S 09801 M M M M M A H M M M M M 09802 M M M M M M M M 20 20 40 40 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 I
09805 M M M M M M M M M M M M 09806 H H 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 M 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 10 10 50 50 09815 M M M M M M M M 10 10 50 50 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 MISSING *** M M M M M M 10 10 09820 M M MISSING *** M M M M M M M M 09821 M M M M M M M M M M 10 10 09822 M M M M M M M M M M 10 10 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 MISSING *** H 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 MISSING *** M M M M 09829 M M M M M M M M M M M M 09830 M M M M M M M M M M M M Workload MW Hrs 10.90 7.44 2.06 .014 7.42 5.21 Sum of natural radiation, direct radiation from facility, and gaseous radioactive affluents. units in millirems (mr). No background or control substraction has been considered. Detection by flim badge dosimeters, and processed by Landauer. Lower limit of det. action is 10mR.
- Damaged fil.a badge
- Represent lost flim badges for unknown reasons.
NEELY NUCLEAR RESEARCH CENTER
. . ENVIRONMENTAL RADIATION SURVEILIANCE' 1993 JUL AUG SEP OCT NOV DEC YEAR BADCE# D 8 D 8 D 8 D 8 D 8 D S D 8 09801 M M MISSING *** M M M M M M M M M M 09802 M M 30 30 20 20 M M M M M M 110 110 09803 M M 10 10 M M M M M M M M 10 10 09804 10 10 10 10 M M M M M M M M 20 20 09805 M M M M M M L M M M M M M M i
09806 10 10 10 10 M M M M M M M M 20 20 09807 M M 10 10 20 20 M M M M M M 30 30 09808 M M M M M M M M M M M M M M 09809 10 10 20 20 10 10 M M M M *
- 40 40 09810 M M M M M M M M 10 10 M M 10 10 09811 10 10 10 10 M M M M M M ** 20 20 09812 M M 10 10 M M M M M M M M 10 10 09813 10 10 20 20 M M M M M M M M 30 30 09814 ** 40 40 20 20 M M M M M M 120 120 09815 M M 40 40 30 30 M M M M M M 130 130 1
09816 10 10 10 10 M M M M M M M M 20 20 09817 20 20 10 10 M M M M M M M 30 M 30 ,
09818 M M 10 10 M M M M M M M 1
M 10 10 '
09819 M M MISSING *** M M M M M M ** 10 10 09820 10 10 20 20 M M M M M M M M 30 30
_ 09821 10 10 MISSING *** M M M M M M M M 20 20 09822 10 10 10 10 M M M M M M M M 30 30 g 09823 10 10 10 10 M M M M M M M M 20 20 09824 10 10 10 10 M M M M M M M M 20 20 09825 M M M M M M M M M M M M M M 09826 M M 10 10 M M M M M M M M 10 10 09827 20 20 M M M M M M M M M M 20 20 -l 09828 M M 10 10 M M M M M M M M 10 10 09829 M M M M M M M M MISSING *** M M M M 09830 20 20 10 10 M M M M M M M M 30 30 Wrkload MW Hrs .02 9.91 1.82 22.65 5.10 3.26 Sum of natural radiation, direct radiation from facility, and gaseous radioactive effluents. units in millirema (mr). No background or control substraction has been considered.
of detection Detection is 10mR. by film badge dosimeters, and processed by Landauer. Lower limits
- Damaged film badge Represent lost film badges for unknown causes.
l U.S. Nuclear Regulatory Commission - Annual Report i February 25, 1994 Page 16 ;
Should there be any questions concerning this report, please let us know. -
S i.ncerely,
@ /7 e h ,
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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APPENDIX A 1
Facility Modifications _
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- * - ' NEELY NUCLEAR RESEARCH CENTER
, Mfnor Change Procedura 4200 R*, vision 00 Number:
By: CHANGES IN GTRR DESIGN Approved 04/28/89 !
Date: / /
Page 3 of 4 l i
APPENDIX A l 1
10 CFR 50.59 SAFETY EVALUATION QUESTIONNAIRE l FACILITY MODIFICATION NO: 3- oo i TITLE: b W GM Flow Ldhdor
- 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) Ao
- 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] AD i
- 3. Will the margin of safety as defined in the basis for any technical specification be reduced? [yes/no) no
- 4. Is the proposed 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: h on k. Orker 2- 9 -13 APPROVALS:
Director NNRC: @ 4 Mh. M /'2 z /D Nuclear Safeguards Committee: // 5 c A,0,n3gstd J/75/97
NEELY NUCLEAR RESEARCH-CEh..:R
'. Minor Change
- Procedure 4200 Revision 00 10anber: Approved 04/28/89 By:- CHANGES IN GTRR DESIGJ Date: / /
Page 4 of 4 FACILITY MODIFICATION uSCUME4YATION CHECKLIST APPENDIX D FACILITY MODIFICATION NO: Il 3 00 l _ _.
TITLE: CwLM Ftow T n h i r <j or _ ._ _
DRAWINGS: No chao36 . n c)(we.j$ fle s- b "
NUMBER TITLE REVISED BY _DATE N Wa'**#* "' = - W E.R.OCEDURES: MD th630 k p C20**'=
REVISED BY_ DATE NUMBER TITc
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Reviewed By: s .,,_
Date:
+ R uby.,
h5 ken IdcNIb grvpe Aq
A Facility Modification 93-001 CWGM Flow Indicator 1.0 PURPOSE The purpose.of this facility modification is to replace the existing non-operational flow indicator in the secondary I- coolant system line to the Cooling Water Gamma Monitor (CWGM) with a new one.
2.0 SCOPE This modification applies only to the one time replacement of the flow indicator in the CWGM water line.
3.0 RESPONSIBILITY The responsibility for the approval of this facility modification lies with the NNRC director with the concurrence of the Nuclear Safeguards Committee. The installation of the flow indicator will be by the NNRC Staff, i
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4.0 REFERENCES
l l Procedure 2002 - Precritical Startup Checklist and shift Supervisor Approval 5.0 SYSTEM DESCRIPTION 5.1 Existing Flow Indicator A brass rotary sight flow indicator with 3/4" female NPT fittings manufactured by Ernst W. C. & G. Company is currently j installed in the 1" copper line. The copper line is reduced j to 3/4" for the flow indicator (see diagram). The flow 'j indicator functions by giving visual indication of flow by verifying that the vanes inside the indicator are rotating.
This indication is sufficient, but the existing indicator no longer functions.
i 5.2 Proposed Flow Indicator l The proposed flow indicator is very similar to the existing one. A brass rotary sight flew indicator with 3/4" female NPT fittings has been selected from McMaster Carr Company. The indicator has s maximum operating temperature of 200 F and maximum operating pressure of 125 psig. The operating temperature and pressure are well below these values.
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6O IMPLEMENTATION
- 1. Drain water lines around the CWGM.
~
- 2. Cut the 3/4" pipe on one side of the flow indicator.
- 3. Remove the current flow indicator.
- 4. Unsolder the cut 3/4" pipe from the 1"-3/4" bushing.
- 5. Make a new 3/4" pipe section with 3/4" male NPT fitting.
- 6. Screv the 3/4" pipe section onto the new flow indicator.
- 7. Screw the new flow indicator onto the uncut 3/4" pipe. >
- 8. Solder the 3/4" pipe into the existing 1"-3/4" bushing.
1" Copper Pipe t
M. .h 1" - 3/ 4" Bushing 3/4" Copper Pipe r > 3/ 4" Male NPT ,
W I I Flow Indicator I l M
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NEELY NUCLEAR RESEARCH CENTER '
dinor Changa Procedure 4200 Number: Revision 00
'By: CHANGES IN GTRR DESIGN Approved 04/28/89 Dates / / Pihge3of4 i i
i APPENDIX A 10 CFR 50.59 SAFETY EVALUATION QUESTIONNAI1ltB FACILITY MODIFICATION NO: 43-002 L TITLE: kep!4ce k M of 4 4 -})2 hre550te b li th
- 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
- 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] AD
- 3. Will the margin of safety as defined in the basis for any technical specification be reduced? [yes/no] o0
- 4. Is the proposed 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: .h ar e4- .3-t$-43 APPROVALS:
Director NNRC: M . 7 4 . W a ]Le - /M /f3 Nuclear Safeguards Committee: A/Sc 4. ppurfg4 2 /2-j'/f2
NEELY NUCLEAR RESEARCH CENTER Minor Change Procedure 4200
' Number: 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: h3' 802 Db Lt4 -D2 h/ eS5 s(f 5 ;4ch TJ.TLE: efkA LC Wfl DRAWINGS:
NUMBER TITLE REVISED BY DATE 04c-60-00% ft; mam (eks %Oem s
(2 kn f v/
8/crh3 PROCEDURES: A) Ort neded TITLE REVISED BY DATE NUMBER TED 2 6 596 Reviewed By: \ -
Date:
l Facility Modification 93-002 Replacement of LA-D2 Pressure Switch i
1.0 PURPOSE The purpose of this facility modification is to replace the-old pressure switch that provides the input for the scram from ;
the reactor tank level (LA-D2) with a new one.
2.0 SCOPE This modification applies only to the one time replacement of the LA-D2 pressure switch.
3.0 RESPONSIBILITY The responsibility for the approval of this facility modification lies with the NNRC director with the concurrence of the Nuclear Safeguards Committee. The installation of the pressure switch will be by the NNRC Staff.
4.0 REFERENCES
Procedure 2006 - Reactor Shutdown Checklist Drawing 045-50-008 - Primary Cooling Water System (D 20) 5.0 SYS'.?EM DESCRIPTION 5,1 Existing Pressure Switch The existing pressure switch was manufactured by Meletron Corp. The switch has a range of 0.75-20 psig and a proof pressure of 65 psig. The switch contacts are rated for 10 Amps at 125, volts. The switch is located next to the pipe chase and approximately 13.5 feet below the dump level of the reactor tank. The pressure switch has been a continual source-of problems and causes spurious reactor scrams for two reasons. The first is that the smallest . division for the setpoint adjustment is one psig, not. sensitive enough. The second is that the switch is not.a differential pressure switch; it is affected by the pressure of the cover ' gas system.
5.2 Proposed Pressure Switch The proposed pressure switch is manufactured by Omega Engineering. The switch is model PSW-364. It has a range of 23" to 150" of water and a proof pressure of 21.6 psig. The electrical contacts are rated for 10 amps at 125 volts. The setpoint repeatability is 1% of full scale or 1.5" of water, and will provide adequate sensitivity. The switch is a differential pressure switch, and will be connected to the cover gas system to negate the pressure supplied by the cover
gas. The proof pressure is still well above the maximum pressure the switch will be subjected to. The new switch will be located in approximately the same place, but will be a maxim u of 12 feet below the dump level of the reactor.
6.0 IMPLEMENTATION
- 1. Drain primary water lines around the pressure switch.
- 2. Mount the new switch on the wall.
- 3. Remove the D,0 line from the existing switch, and connect it to the high pressure side of the new switch.
- 4. Install a line from the cover gas t- the low pressure side of the new switch.
- 5. Remove the alarm contacts from the old swj tch and connect them to the new switch.
- 6. Check the new switch for proper function.
- 7. Remove the old switch from the wall.
- NEELY NUCLEAR RESEARCH CENTER Minor Change Procedere 4200 I Revision 00 Number: Approved ~ 04/28/89 By: CHANGES IN GTRR DESIGN Page 3 of 4 Date: / /
APPENDIX A 10 CFR 50.59 SAFETY EVALUATION QUESTIONNAIRE FACILITY MODIFICATION NO: @ - 003 TITLE:
kePl4remuif ob fitl bl htmole [coel InM m b $yd e
- 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
- 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
- 3. Will the margin of safety as defined in the basis for any technical specification be reduced? [yes/no] Ao
- 4. Is the proposed change an unreviewed safety question?
[yes/no] nn NOTE: If additional space is needed to justify conclusion (s) please attach extra sheet (s).
DATE:
PREPARED BY:
- r N 2 13-43 APPROVALS:
Director NNRC: h .O . /rdJtew- [k) [J Nuclear Safeguards Committee: A/ 5 e A.90m34 ~2/2f70v i
NEELY. NUCLEAR RESEARCH CENTER l Minor Change Procedure 4200 l Number: Revision 00 J By: CHANGES IN GTRR DESIGN Approved 04/28/89 i Date: '/ /
Page 4 of 4 l FACILITY MODIFICATION DOCUMENTATION CHECKLIST APPENDIX B FACILITY MODIFICATION NO: $3'003 _
TITLE: ke0!4te mint o f L I A - D 1 k w d l e.ve \ I n t)i u.fio n $ g { e m DRAWINGS:
NUMBER TITLE REVISED BY DATE 004i-62.Gon Loolrnl (l.c.a m Mc4 rurdea FE" i e M'
PROCEDURES:
NUMBER TITLE REVISED BY DATE 7241 dealhe Tank Icwei TrMsMil 'r /hin k 3 F 9*4d 'IISIM
/' TES 2 g '.'-
Reviewed By: % '
Date:
s %
1 l
Facility Modification 93-003 Replacement of LIA-D1 Remote Level Indication System 1.0 PURPOSE The purpose of this facility modification is to replace the old . remote level indication system .(LIA-D1) with a new one -
that provides readout in the control room of=the reactor tank level and input for a reactor tank level scram.
2.0 SCOPE This modification applies only to the one time replacement of the LIA-D1 remote level indication system.
3.0 RESPONSIBILITY The responsibility for the approval of this facility modification lies with the NNRC director with the concurrence of the Nuclear Safeguards Committee. The installation of the pressure switch will be by the NNRC Staff.
4.0 REFERENCES
Procedure 2002 - Precritical Startup Checklist and Shift Supervisor Approval Procedure 7241 - Reactor Tank Level Transmitter Maintenance and Calibration Check.
Drawing 045-50-008 - Primary Cooling Water System (D 20)
Drawing 045-62-001 - Control Room 5.0 SYSTEM DESCRIPTION 5.1 Existing Remote Level Indication System The existing remote level indication system consists of three instruments and associated piping and wiring. The three instruments are:
- 1. A Mercury Differential Pressure Transmitter (LI-DlT) manufactured by Hcneywell,
- 2. a pressure gauge denoted as a Pneumatic Indicating Receiver (LI-DlI) manufactured by J. P. Marsh, and
- 3. a Pressure Switch manuf actured by Meletron Corp (LA-D1) .
The specifications for all three instruments are attached.
9Cv w'
a 4
The pressure transmitter (LI-D1T) is connected to the Reactor Tank Level Column (LC-D1) and senses the pressure differential of the reactor tank water level and the cover gas pressure. This differential pressure is converted to an wir signal and is transmitted to the control room via air lines. In the control room this air signal is connected to the pneumatic indicator (LI-D1I) to provide a readout and to the pressure switch (LIA-D1) to provide a scram.
The current system is no longer able to maintain an acceptable accuracy, and requires excessive maintenance.
5.2 Proposed Remote Level Indicating System The proposed level indicating system will consist of two instruments, both manuf actured by Omega Engineering, and associated piping and wiring. The instruments are:
- 1. A Heavy Duty Industrial Differential Pressure Transmitter,
- 2. a Digital Panel Indicator with Excitation and Dual Alarm Relays.
The specifications for these instruments are attached.
The new pressure transmitter will be connected to the Reactor Tank Level Column in the same manner as the existing transmitter.
The differential pressure is converted into an electrical signal that is transmitted to the control room. In the control room this electrical signal will be connected to the Digital Panel Meter to provide readout and a scram signal.
The proposed system has much better accuracy than the original and will be more reliable and provide digital indication of the reactor tank level in the control room. The new system is also compatible with a proposed change to allow the reactor tank level to be viewed in the Emergency Command Center during emergencies.
6.0 IMPLEMENTATION
- 1. Drain primary water lines around the old pressure transmitter.
- 2. Remove the existing transmitter.
- 3. Mount the new transmitter.
- 4. Install the water and gas lines from Reactor Tank Level Column to the new transmitter.
- 5. Remove the pressure gauge and pressure switch from the control room.
- 6. Install the. panel meter in the control room.
- 7. Run wires from the new pressure transmitter to the panel meter.
- 8. Connect the panel meter to the scram input.
- 9. Calibrate the system per the manufacturer's instructions.
- 10. Ensure the system is functioning properly.
- 11. Remove air lines.
Mercurvless Pressure Transmitter Brown Instrument Division Minneapolis-Honey well Regulator Company Wayne & Windrim Avenues Philadelphia, Pennsylvania Model #228N4C - As defined by the following specifications Conditions for Operation:
Fluid Heavy Water Temperature: 100 F Pressure: 6 psig Range: 0 to 100 inches of D 2O differential, u Construction:
Case: Black baked enamel, vapor proof, with yoke for 2 inch pipe or mounting bracket.
Manometer Bellows type with pulsation damper.
Materials: Body - stainless steel Bellows - stainless steel Bolting - alloy steel Connections: Air - 1/4" NPT Manometer - 1/2" NPT flanged process connection to connect to three valve manifold.
Accessories: Transmitter shall be complete with integrally mounted combination air filter
& reducing valve.
I Pneumatic Indicatina Receiver J. P. Marsh Company Skokie, Illinois Model -Master Movement Dial Indicating Refinery Pressure (12") as defined by the following specification.
Construction:
Case: Black baked enamel, smooth finish, flush mount.
Dial: 12" Scale: 0-100 uniform scale.
Input Air: 3-15 psi from transmitter.
Accuracy: 1/2%
Other:
Remarks: Standard accessories & legend plate required.
Use: Panel mounted receiver. for Brown Instrument Model 228N4C non-indication bellows type differential, transmitter with pulsation damper.
Pressure Switch Meletron Corp.
Model: 420E ,
Catalogue #: 420E 6SS10A Range: 0.5-10.0 psig Proof Pressure: 25 psig
.~
lleavy Duty Industrial Differential Pressure Transmitter Omega Engineering Inc.
Stamford, CT Model PX761-150WDI-B3-FL-FL as defined by the following specifications.
Conditions for Operation:
Fluid: Water Temperature: -22 to 212 F Proof Pressuro: -2000 psig Range: Adjustable 0-25 to 0-150 inches of HO 2 differential.
Construction:
Electrical Housing: Epoxy painted low copper aluminum.
Gauge Type: Capacitance sensor, using 316SS isolation diaphragm with silicone oil fill.
Materials: Body - stainless steel Diaphragm - stainless steel Bolting - alloy steel Connections: Electrical - 1/2" NPT Gauge - 1/4" NPTF Accuracy: 0.25%
Output 4-20mA two wire Excitation: 12-45Vdc
?
l Dicital Panel Meter Omega Engineering, Inc.
Stamford, CT j Model DP205 panel meter as defined by the following specification.
Construction:
Case: Aluminum case.
Displays 4 digit, red LED, 0.56" high Range: -1999 to 9999 counts.
Input: 4-20mA, 0-100mV, 0-10V Accuracy: 0.02%
Qther Operating Temp: 32 to 122 F.
Analog Output: Scalable 0-10V or 4-20mA.
Excitation: 12V 0 100mA or 24V 0 50mA Relays: Dual 250 vac, 6 amp, SPDT Power: 115 Vac, 6 watts maximum 1
" - - n--- - -p-.. <- -
y .
NEELY NUCLEAR RESEARCH CENTER
. . ' " ~ Nino_r Change Procedure 4200 Number: Rovision 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: Y3 - OC Y TITLE: EMe%EI)cH 00MofMD 0sWE/L MO AJ /TDR / Ald)
- 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) dd
- 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) Ao
- 3. Will the margin of safety as defined in the basis f,pr any technical specification be reduced? [yes/no) ND
- 4. Is the propospd change an unreviewed safety question?
[yes/no) /JD NOTE: If additional space is needed to justify conclusion (s) please attach extra sheet (s).
DATE:
PREPARED BY: TW77hA/ 2 -/ 8_ '
APPROVALS:
I Director NNRC: @ 4 - kil1L% SI/'Jf[fy Nuclear Safeguards Committee: 4/ fC ftpSVEd zg I
. , , NEELY NUCLEAR RESEARCH CENTER
, Minor Change Procedure 4200 Number: 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: @3-COY TITLE: EF/CdCSEdC Y HN/A A/D b-/Y7~C/C kod/TOD 9 DRAWINGS:
NUMBER TITLE REVISED BY DATE 04 5 - (,2 - co l TMsr4du/6drArfod h doMT72.0L 'D. Fart.v'
/41kr
- 25 -
SH'Et=T 2. ScHeH nr1 c3 PROCEDURES:
NUMBER TITLE REVISED BY DATE
~2D C 'L R eta cro L 6 p etLhr1 o ed 5. - ik A < < '
h?Ecnt7-/ cat- STA crUR G466 KLt ST AM13 54IFr 50PGaVi 20R AP4/loV% . l l
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FACILITY MODIFICATION 93-004- '
EMERGENCY COMMAND' CENTER MONITORING 1.0 PURPOSE _
The purpose of this facility modification is to. increase the GTRR parameters that can be monitored in the ECC (Emergency Command Center).
2.0 SCOPE At present, the reactor tank level can be monitored in _ the ECC via Closed Circuit Television (CCTV). Both the camera and-monitor require utility power to function. This proposal will add three (3) GTRR fuel element temperatures to the parameters that can be monitored in the ECC. In addition, battery backup system will be provided.
3.0 RESPONSIBILITY The approval for this modification lies with the NNRC director with the concurrence of the NSC (Nuclear Safeguards Committee).
4.0 REFERENCES
4.1 Omega DPS3200-TC Pcogrammable Process Scanner Manual 4.2 Related Procedures 4.2.1 Procedure 2002 Reactor Operations - Precritical Startup Checklist and Shift Supervisor Approval 5.0 SYSTEM DESCRIPTION 5.1 Existing System Tho existing system is comprised of a CCTV camera in the GTRR cortrol room, a monitor in the ECC and a connecting coaxial cable. The camera and monitor are powered by utility power.
The camera is energized during reactor operation at power levels greater than 1 MW (per procedure 2002). The camera views the pneumatic instrument that indicates the D 20 level in
-the GTRR vessel.
5.2 Proposed New System for the Control Room 5.2.1 Add an Omega DPS3200-TC Programmable Process Scanner to the system. The Scanner is capable of scanning up to seven (7) inputs with mix of signals from four (4)
1 different type thermocouples, 4-20 mA and 0-10 V. The. ,
scanner has an alpha numeric LED (Light Emitting Diode) '
display.
5.2.2 Replace the camera with a low' light level type camera.- )
This camera will have the capability of functioning with only the light from the scanner display.
5.2.3 Add a battery backup system that is automatically energized if utility power is lost- during reactor operation at a power level greater than 1 MW.
5.2.4 Connect three (3) of the fuel element thermocouples .tx>-
the Scanner.
5.2.5 The instrumentation for the measuring the D/) level in the GTRR vessel is being changed; reference - f acility modification 93-003. Connect this instrumentation into one (1) of the Scanner's 4-20 mA inputs.
5.3 Proposed New System for the ECC 5.3.1 Add a battery backup system for the existing monitor that can be manually activated.
6.0 OBJECTIVE To enable the Emergency Director to make a more informed add water to the ECCS (Emergency Core Coolant System) decision.
l
.l oc , ., .
s b,rogrammable Process Scanners
'hChannel input a:
i l'.
n q L:fg e f 911 1 in 9Wh a
. A je D gption
%ptt.113G Load Cell. 369 D60GV Pressure Transducer, $175 q
%500 LVDT Sensor.5150 p i, '.gypt _ .
TMS-125G-6 Therrnoccuple $24
- . +
i T PX181100G5V Pressure ik. ,
I . Transducer. $169 t !
{ , 7tTC 50 Load Cell. 3350 -
O I O! " S '," gl gpj,,
{ - 74110 500SI Pressure
,,, ~ mducer. 5295 g
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< Independent Scaling From
- can
- ( Thermocouple, Voltage, Current input 2ndo (4 e
Rate of Change and Min or Max Display Olgit Display Plus 3 Alphanumeric EE.ecifyModel
'Model No. Price input Typ (s) -
)asse.
- 20 t rhe DPS3000 series scanners accept seven individua DPS3100-(*) $595 see
- below l Bputs The DPS3200 can accept different input types DPS3200-TC 695 J, K, T, E,4-20 mA,010 V l-
, [f am m each channel, while the DPS3100 is limited to one DPS3200-R 695 R,4-20 mA,0-10 V l 5
, [put type on all seven channels.
DPS3200 S 695 S,4-20 mA,0-10 V l N ess km the DPS3100 and DPS3200 meters can be setup DPS3200 8 695 B,4-20 mA,0-10 V l )
mm) msplay in any of six operating modes. The unit can 0-100 mV,4-20 mA,0-10 V 4
the $an each channel; display elapsed time; display the DPS32. -' 695 l p
- annel with highest or lowest reading; scan each DPS320 d 695 400 series thermistor, f
-anr^# deviation from a predetermined setpoint; -22 to 212*F (-30 to 100 C)
(use.Kant difference between a ' master' channel and the
- Specifyinput type:J. K. T E R S B T' er t ;nputs, ordenng Example: DPS3100.J, $$95. DPS3100 7.channelscanner, .?
.t oedicated to type J thermocouple input.
specifications 2 guracy: 0.5% Rd (J K, T E; 1 C; S 2 C; R,B 3 C) Options-DPS3100, DPS3200 and DP3400 I
. u n1 / F(01 C/ forthermistorinput);
- her
- 120 Vac,60 Hz; optional -1 $5 240 Vac power l S.
ton WC. 50 Hz; optional 7 to 12 Vde,900 mA -2 25 7 to 12 Vdc power l $*
n Pate: fixed,2 channels /s
- . garn Display Time: 1 to 999 s per channel -3 110 l 6 alarms,4 SPDT,2 dc drivers l garm optiont.1): mechanical relay: 1 A @ 28 Vdc or -4 110 l 6 alarms, all dc drivers g 4 (120 Vac, SPDT; de driver: 5 Vdc @ 50 mA. One l 200' 125 Analog output,1 mV/ C (DP3400
~
Der channel, (channel 7 is monitonng only) -5 hetadband: 2 to full scale. adlustable only) tout: 45 x 92 mm (1.772" x 3.662"); 's DIN: 9 0" depth
' Tl IL-?? ' Q3 10: 25 iD SCS PROC-tUCLEA TEL NO:205-870-6197 #027 P02 t ACCElvED 7m _00: 18 1993 4T 1734567 PAGE 2 (PRINTED PAGE Z) )
NEELY NUCLEAR RESEARCH CENTER
~. Procedure 4200 L- Minor Chango Revision 00 Numbert CHANGES IN GTRR DESIGN Approved 04/28/09 gy .Paga 3 of 4 Dates /' /
APPENDIX A 10 CFR 50.59 SAFETT EVALUATION QUESTIONtukIRE FACILITY MODIFICATION ND #/3- 003 _
TITLE [G8 LAM.A/EMT d# MdR TAl d6Cd/r/7y' 8YITEv/
- 1. Will the probability of the occurrence or the consequencess of an accidont or malfunction of equipment important to anfaty previously evaluated in the gafety analymis report be 1.ncreased? [yes/no] A/O Will the possibility for an accident or malfunction of a 2.
different type than evaluated previously it; the safoty analysis report be created? (yos/no] No _
7 .
- 3. Will the margin of safoty as defined in the basis for Nn $ny technical apuciiication be reduced? Iyas/no)
- 4. Is the propos9d change an unreviewed safety question?
[yes/no) No p_gn; If additional space in needed to justify conclusion (s) please attach extra shcot(s).
DATE:
PREPARED DY: .,
I 7^b~93 APPROVALS:
Director NNRC: /h.4
/ NOAA,u.t Nuclear Safeguards Committoct P -
NEELY NUCLEAR RESEARCH CENTER
- kinor Change Procedure 4200 Number: Rsvision 00 By: CHANGES IN GTRR DESIGN Approved 04/28/89 Date: / / Page 4 of 4 FACILITY MODIFICATION DOCUMENTATION CHECKLIST APPENDIX B FACILITY MODIFICATION NO: k3- M TITLE: WEPLACEAEAT OF THER 1h) Sease ir'/ SYsred DRAWINGS:
NUMBER TITLE REVISED BY DATE PROCEDURES:
NUMBER TITLE REVISED BY DATE Reviewed'By: M AL MS- Date: 97
~
FACILITY MODIFICATION 93-005 REPLACEMENT OF TIMER IN SECURITY SYSTEM ,
t 1.0 PURPOSE The purpose of this facility modification is to replace the timer in the security system, i
2.0 SCOPE The proposal is to replace the timer.
I 3.0 RESPONSIBILITY The approval f or this modification lies with the NURC directc r with the concurrence of the Nuclear Safeguards Committee.
4.0 REFERENCES
4.1 Included sheet that shows the current model and the proposed replacement model.
5.0 SYSTEM DESCRIPTION NOTE: The timer in the security system activates the card lock at the entrance to reactor control l
- one during business hours- .
3 NOTE: The timer must be capable of- continued i operation during a power outage and contain a minimum of two Single Pole Single Throw (SPST) j Normally Open (NO) contacts. The size of 1.he '
i timer must be such that it will fit in the security box. .l 5.1 Existing timer is a Paragon model- 7618-56. The unit losses time at the rate of approximately 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> per week. It is an obsolete model with no ceplacement- ' parts available from H Paragon. A recent attempt- to correct the- problem was unsuccessful.
5.2 Proposed replacement timer is an Omron model H5L-A. This unit I meets the requirements of the security system having tuc SFST No contacts and battery backup to keep the unit opera t.iona l during a power outage. The model HSL-A timer will fit in .
R security box.
__ ~- - .- -- - - . _ . . . . . _ . _ _ _ _ _ ._
4
'T (, / 8 - 56 COKRENT /VoDEL G #- d 7600 SERIES $) @ ,
t, Three Hour Minimum ON Or OFF Time ,
Four Pole, Quartz Carry-Over 4 ~
e
- 7. day catencar permits different ON'OFF schedules on diff erent days of the y .
week. Program up to tour ON/OFF operations per day. 28 per week. Three h[ .g a g
hour minimum and 21 hour2.430556e-4 days <br />0.00583 hours <br />3.472222e-5 weeks <br />7.9905e-6 months <br /> maximum ON or OFF time per day. Independent tour pole desgn anows for SPST. DPST or SPDT switching. Manual
' hy M[hp!- >g g 1;.a l overnde lever temporanty reverses switch operation without permanently ,
gj r
-g ,
disturbing preset schedute. Switen shdor bar assures positive switching. ,
Heavy duty industrial type synchronous motor. Heavy duty terminals ac- '
2 commodate up to AWG #8 wire. Standa d NEMA 3 indoor / outdoor metal .
- j. _,
enclosure side hinged with combination 1/2-3/4 incn knockouts in bottom l
and side. Hasp tot padlock or seal. Quartz carry over to take over during power outage. Quartz carry-over keeps control on time for a maximum of l 7 days. l How To Specify ]
l installer shall fumish and install Paragon (Model No.) 7 Day General '
Purpose Time Control with (4PST-4 NO,2 NO-2 NC) switch. Contacts to be rated at 40 amperes tungsten or 40 amperes noninductive per pole up to 240 volts. Control shall have NEMA 3 indoor / outdoor metal enclosure.
I Switch Hatinq AC lane i rictosure 'ihipping EDP No. Time Control - Switch Winng Otag.
Type Wqt j 4988 Model No Type See pg. 49 Per Pole Hz ! b. Kg .
Figure: Amp Amp 1 VA Ho Volts l l
27 l 40 1 40 l 690 -
l 120 l 50/60 ' 1 1 10 l 4.5
' 71670 7617-56 l 4PST-4 NO i l76172l 7617-62 l 4PST-4 NO l 27 l 40 l 40' l ' 690' 'l "r# f208-240 I 50/60 l 1 l 10 lc 4.5 !
)' I 76182! 7618-62 l 2 NO-2 NC l 76180 7618 56 i 2 NO 2 NC 28 28 l 40 l 40 i -
I 690 l 690 l 1
1
! 120 ; 50/60 l 208-240 l 50/60 l 1
1 l 10 l 10 4.5 4.5 _
SPDT OPST or OPDT operation possete ey usmg entra surncers (turnesneci For enclosur e cimensions see page 39 Furnisreo vuin ' sets at ON OFF toppers for addmonal seis see pace 41 Ho rating tone pole onry; i . _ . _ . - - . _ _ _
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i 24 HOUR AND 7-DAY PROGRAMMABLE TIMERS 4
ELECTRONIC, 24 HOUR, PROGRAMMABLE, 2-CHANNEL TIMER
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',3. * % ':c ,, ~.P e 2 channels with independant program. e 10. year memory protection by tm C T, .+ ..i 3 % ff
.yi ,j.. ming for each circuit battery
' E32800' '
l
- Weekly timerr 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> x 7 days using e E asy.to read,0.5" high LCD
~" % c.[yU I five programming keys display
~ - '. t t.
l ,j p' 't e Surface, track (DIN type mountirn l ,.j e Manual On.Off switching for each can' see page 367), and flush mountm
- q. trol output without changmg program 1/4 DIN panel cutout
- C d h, glM 1}.; .g\ e 4 adjustable cycle models e oimensions: 3%H x 3%W = 2%"D l LR2231,0 i ppcf.1 . 4.
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1 59 Min On OtT periods
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 APPENDIX A j 10 CFR 50.59 SAFETY EVALUATION QUESTIONNAIRE i FACILITY MODIFICATION NO: i3' cob i TITLE: k<ph & Nid of $vmp ano bmp [rr A /3/9nJA (osl;ny tudu GIlghin SyNn1
- 1. Will the probability of the occurrence or the consequences of l
' an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report be increased? [yes/no] /Jo
- 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] do
- 3. Will the margin of safety as defined in the basis for any technical specification be reduced? [yes/no] #0 l
- 4. Is the proposed change an unreviewed safety question?
[yes/no] vo NOTE: If additional space is needed to justify conclusion (s) please attach extra sheet (s).
DATE:
PREPARED BY: DJ.harker ll/43 j I
APPROVALS:
Director NNRC: @-/9- 44>Le N3NS Nuclear Safeguards Committe9: k 1 5 1
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ - _ _ _ ___- _______ _-_ - __ - --_ O
NEELY NUCLEAR RESEARCH CENTER Minor Change Procedure 4200 Number: Revision 00 By: CHANGES IN GTRR DESIGN Approved 04/28/89 Date: / / Page 4 of 4 FACILITY MODIFICATION DOCUMENTATION CIIECKLIST APPENDIX B FACILITY MODIFICATION NO: k3 -0Cb TITLE: 5 plPs,, R,piu g na t 6;sma co;cs DRAWINGE:
NUMBER TITLE REVISED BY DATE 645 -5J-004 h dismo A loel;o) b3.ke# (ellehen Sv; tem ))ai utet4 D@ es 2 f, m.
PROCEDURES:
NUMBER TITLE REVISED BY DATE 73ec Alyndh (cohnn Syde C#okon WAer *Z5S r
Reviewed By: ( % .
Date: O 25 a l
l
Facility Modification 93-006 Replacernent of Sump and Sump Pump for the Bismuth Collection System 1.0 PURPOSE The purpose of this facility modification is to replace the sump and sump pump for the bismuth cooling water collection system.
2.0 SCOPE The modification applies only to the one time replacement of the sump and the sump pw 1p for the bismuth collection system.
3.0 RESPONSIBILITY The responsibility for approval of this facility modification lies with the NNRC director with the concurrence of the Nuclear Safeguards Committee. The implementation wm be done by the NNRC staff.
4.0 REFERENCES
1 Procedure 2300 - Bismuth Cooling System Operation Drawing 045-53 0004 - sheet 2 of 2 -- Bismuth Cooling Water Collection System 5.0 SYSTEM DESCRIPTION 5.1 Existing Eump anci Sump Pump The current sump and sump pump is an integral unit with a maximum sump capacity of approximately one gallon and an original pumping capacity of one ha!f gallon per minute. The pump has a built in level switch to control pumping.
Recently the pump has become clogged several times from silt like material that is getting into the collection system. The pump has also been damaged by thN silt like material and will no longer pump a su'ficient amount of water.
5.2 Prorosed Sump and Sump Pump ,
A new sump with capacity of 15 gallons will be used The sump will be a polyethylene drum. A new sump pump with a 1.5 gallon per minute pumping 4 capacity and the ability to pass solid materiah is proposed. The pump has a l built in level switch to control pumping. Thc pump is manufactured by Flotec. i The current piping for the system can be .r4ed. l l
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t-4 6.0 IMPLEMENTATION ,
- 1. Unplug the current pump.
- 2. Disconnect the plumbing fittings.
. 3. Remcve old sump and pump.
- 4. Place new sump and pump.
- 5. Connect plumbing fittings. '
- 6. Plug in pump.
- 7. Test for proper operation.
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NBELY NUCLEAR RESEARCH CENTER Minor Change ~~~
Procedure 4200 Number: Revision 00 By: CHANGES TN GTRR DESIGN Approved 04/28/89 Date: / / Page 3 of 4 APPENDIX A 10 CFR 50.59 SAFETY EVP.UATION QUESTIONNAIRE FACILITY MODIFICATION NO: N'0O TITLE: [eorl dIum on Sismo/h (A>ofo foNe kon (ys
- 1. Will t.he 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
- 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] 00
- 3. Will the margin ci entoty era defined in the basis for any l technical specificuton be reduced? [yes/no] ^! O
- 4. Is the proposed change on unreviewed safety question?
[yes/no] No NOTE: If additional space is needed to justify conclusion (s) please attach extra sheet (s).
DATE:
PREPARED BY: ad's 4 2d63 APPROVALS:
Director NNRC: O.A. h 7/ I523 NJ N? ear Safeguards Committee: C 2k '/2 93
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NEELY NUCLEAR RESEARCH CENTER Minor Change Procedure 4200 Number: 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: N'OU3 TITLE: LeveI fare en ilt ISi5nu,lh loch / C oll<cf 5ysL m DRAWINGS: NUMBER TITLE REVISED BY DATE 045 -5 3 -004 th. 6invih blic$ bach, fellnlion S 5kem M PROCEDURES: NUMBER TITLE REVISED BY DATE 23e o 6isndh (.colina Sskew Oe(chor 1 (o.t\. on \ki5 mh6 W - mod. Nb udWi g [ m .4 FT.B 2 5a- ~ ] , 3PO l 1 l Reviewed By: Date: l
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, , aq, Facility Modification 93-007 Installation of a Level Alarm on the Bismuth Water Collection System 1.0 PURPOSE The purpose of this facility modification is to Install a level alarm with 'an annunciator in the control room on the bismuth water collection system.
2.0 SCOPE The modification applies only to the one time installation of the alarm on the - bismuth water collection system 3.0 RESPONSIBILITY The responsibility for approval of this facility modification lies with the NNRC director with the concurrence of the Nuclear Safeguards Committee. The ' implementation will be done by the NNRC staff.
4.0 REFERENCES
1 Procedure 2300 -- Bismuth Cooling System Operation Drawing 045-53-0004 - sheet 2 of 2 -- Bismuth Cooling Water Collection System 5.0 SYSTEM DESCRIPTION < 5.1 Existing Alarm None. 5.2 Proposed Alarm A float type level switch from McMaster Carr will be installed on the new sump . for the bismuth cooling water collection system. The switch will be connected to an audible and visual alarm in an existing NIM bin in the control room. The audible alarm can be silenced, but the visual alarm will remain lit until the level in. the sump drops below the alarm level. The alarm level will be approximately half way full on the 15 gallon sump. The level will give approximately 30 minutes of , response time before water overflows the sump. 6.0 IMPLEMENTATION
- 1. Installlevel switch on sump.
- 2. Run wire to control room.
- 3. Connect wire to alarm Nim module.
- 4. Test system to ensure proper operation, i
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