ML20197D548
| ML20197D548 | |
| Person / Time | |
|---|---|
| Site: | 05000131 |
| Issue date: | 12/17/1997 |
| From: | DEPT. OF VETERANS AFFAIRS MEDICAL CENTER, OMAHA |
| To: | |
| Shared Package | |
| ML20197D519 | List: |
| References | |
| NUDOCS 9712290082 | |
| Download: ML20197D548 (163) | |
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![ SAFETY ANALYSIS REPORT i
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Revised December 17,1997 ,
to include " Reply to Additional Information" dated Dec. 17,1997 I
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- Table of Contents - -
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d 1 . Safety Analysis Report j c Omaha VA Medical Center-TRIGA c , _
GbagleL1 .Page INTRODUCTION AND SU WMARY .........................................................-.. 1 1.1 PRINPIPAL DE SIGN CRITERIA ....................................................... 1-1 C t 1.2 L DESIGN HIGHLIGHTS ......... .............. .L....... .. .. . ................ .............. .. 1-1 11.3 : CONCLUSIONS .. . .. .. . .. . . . . .. . . .. .. . ... . .. . . . . .. . . . ... . .... . . . . ... .. . . . . . . . . ... . . ... . . .. . .. ... ],
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t p :: SITE DE SCRIPTION . . . .. . . . . ... .. . .. .. . ... . . . . . . . .. . . ... . . . . . . . . . . . . .. . .. .. . . . . . . . . . . . . . . .. . . . .. . . . . . . . 2-1 !
t 0 i 2.1 GE NE RAL LOCATION . . . .. . . . . . . . . . . .. . . . .. . .. . . . . . . . .. . ... .. . . . . . . . . . . . . . . .. . . . . . . . . . . . . .. 2-1; i 2.2 ? POPUl.ATION DENSITY . . . . . . . . . . . . .. . .. .. . . .. . . . . . . . . . . . . . . .... . . . . .. . . . . .. . . . . .. . . . . . . . . 1 1
- 2. 3 M ETE OR OL OG Y . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . 21 ;
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- 2. 4 ! GE OLOG Y . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ; . . 2-8
. # 2. 5 - EARTHQUAKE S . ... .. . . . . . .. . . . ;.. . . . . . . ... . . . . . . . . . . . .. . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-13_ 'i 4
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- 2. 6 TOR NADOE S ; . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
2 2.7 ? ATMOSPHE R IC STABILITY . . . . . . . ... . .. . . . . .. . . . . . . . . ... . . . . . ... . . . . . . . .. . . . . . . . . . . .. .2-19 .. l 1
Chanter 3 l .
i j - FACILITY DESIGN STRUCTURES, SYSTEM AND COMPONENTS........... 3-1 :
. 3.1 - REACTOR LABORATORY . . . . . . . . ... . . . . . .. . . . . . . . . . . . . . . . .. . . . . . . . . . .. . ... . . . . . . . . . . . . . . 3-1 3.2 - REACTOR AND REACTOR SYSTEM ............................................... 3-5 3 . 2.1 . Reactor Pit . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3-5 i 3.2.2 Core.....;..................................................................................... 3-7 +
- 3. 2. 3 : Reflector .. . . . . . . . . . . ; . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-11 3.2.4 Center Channel Assembly and Reactor Tank Covers ......... ........ 3-11 3.2. 5 Neutron Source . . . . . . . . . . . .. . . . .. . . . .. .. . . . . . . . . . . . . . ....... . . . . .. . . . . . ... . . . . .. . . . .. . . . . . 3-11 c 3 2.6 I rradiation Facilities .. ....... .. ..... .. ... . ....... .. ..... ....... .. .. ..... . . .. .. . . . .. .. .... 3-13 3.2.6.1 Rot ary Specimen Rack... ........ . .. .. .. ... ... . .......... . ... . .. ... .. ..... .. . .. .. 3-13 3.2.6.2 Pneumatic Transfer Tube ...... ... . .. . ..... .... ...... . . .. ... ... .......... .. .... . 3-13 3.2.6. 3 - Central Thimble . .. ... . ... . ... . . . .. .. . . . . . . . . . . .. . . .... . ..... . . .. . . . .. . . . . .. .. . .. . .. . . 3-14
' 3.2.7 Control Rods and Guide Tubes .................................................... 3-14 3.2. 8 " Control-rod Drives ;.. . . .. . . . . . . . . . . . . . . .. . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . .... . . .. . . . . . .. . . 3-15
. 3.2.9 ? Reactoi Water Cooling and Purification System.........,................. 3-15 3.2.10 ' Ability of Reactor Facility Structure, Systems and Components to Function Properly and Safely for the Term of the License.......... 3-18 3.2.10.1 Reactor 1 ank . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . 3-18 ...
+ 3.2.10.2-Core...........................................................c........................ 3-18
- 3. 2.10. 3 Electronics and Mechanical . . . . . . .. .. . . . . . . . . . . .... . . . . . . . . . .. . . .. . . . . . . . . . .. 3 ;3.2.11 Confinement Design Evaluation ...... . .. .. .. .<.. ...... ........ ..... .... ...... .. . 3-20 13.2.11.1 n Activation of Soil Surrounding Reactor Pit............................ 3 13.2.11.2( Production of Radioactive Gases by the Reactor................. 3 3.2.12 - Limiting Design Basis .. ...... . ..... ..... . ..... ........ ....... . .. .. .<.... . ... .. .c. ..... 3-21 3.2,13 ; Dynamlc Behavior of Reactor................c. c.................................
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t p Chapter 4 l 5 lNSTRUMENTATION AND CONTROL SYSTEMS............................. ... .... 41 t 4.1 S Y STE M S S U M MAR Y. . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-1 i 4 .2 C O NTROL CON SO LE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . 41
- 4. 3 C ONTROL SY STE M . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . 4-4 4.3.1 N uclear Control System . . . ... . . . .. .. ... . . ... . .. ... ... . . ... .... . ..... . . .. ..... ..... .. . . 4-4 -
4.3.1.1 Nuclear Instrumentation ...... ......... ... ... .... ........... .... ............ 4-4 1
4.3.1.2 Reactor Power Safety Channels . .................................. ....... 4-7 4.3.13 Internal Diagnosbcs.... ..... . . ... .............. ....... ......... ..... ...... . 4-7 4.3.1.4 Additional I nformation.... ... ...................... . ......................... ...... 49 4.3.2 Process Instrumentation ........... .......... ........................ .,............ 4-9 '
4.3.3 Electrical Power System . ... ................. ... ... . ....... .... .. ....... ........ 4-10 Chapter 5 COREPHYSICS............................................................................., 5-1 5.1 C R I TI C A L M A S S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 51
- 5. 2 VOI D COE FFI C I E NT . . . .. . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 51 5.3 MODERATING PROPERTIES OF ZlRCONIUM HWmDE............... ^
5-1
- 5.4 TEM PE RATU R E COE FFICI E NTS ........... .... ..... ... .. ..'. ....... ... .... .......... 5-1
- 5. 5 R EACTlVITY PE RTU RBATION S .. .. ..... . ........................... ............ ..... 5-3 Chapter 6 C ON D U CT OF O P E RATI O N S . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 7 '
-.k 6.1 FACI LITY ADM I N I STRATI O N . . ..... . . ........ ..... . .. .. .. ......... . ...... . . ... ..... 6-1 6.1.1 Ove rall O rg a nlz a tion . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-1 6.1.1.1 Director Veterans Administration Medical Center...... . .. ....... 6-1 6.1.1. 2 C h ief of St atf . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-1 6.1.1.3 Associate Chief of Staff Research............ ............................. 6-1 6.1.1.4 Reactor Safeguards Committee ........ ................................... 6-1 6.1.1.5 Radiological Safety Officer .............. . . .. ... ............. ................. 63 6.1.1.6 Re actor S u pe rvisor . . . .. . . .. .. . . . . . . . . . . . .. . . . . . . . . . ... . . . .. . ... . . .. . . . . ... . . . . .. . 6-3 6.1.1.7 Professional and Classified .......... .. ......... ....... .................... 6-3 6.1.1.8 Facility Staff Qualifications ......... ............... ... . .............. ....... B-3
. 6.1.1.8.1 R e act o r S u pe rvisor. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-3 6.1.1.8.2 Senior Reactor Operator .... .. .. ...... ....... ..................... .. 6-3 6.1.1.8.3 Re actor Operator . ... . . . .. . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . .... . ... . . . ... . 6-4 6.2 ' R EACTOR OPE RATI ON S .. .... . . . . .. . . . . .. . . .... . . . . . . .. . .. . . . . . . . .. . .. . . . . .. . . . . . . . 6-4 i- 6.2.1 Pro ced u re s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-4 6.2.2 Routine Operation Procedures.. . ,............. . ... ................ . . ...... 6-4 6-4 Chapter _7 RADIOACTIVE MATERIALS AND RADIATION MEASUREMENT... ........... 7-1 7.1 RADI OACTIVE M ATE R I AL C O NTR OL... ... .... .. . . . .. ..... . .... . ..... . . .. .. . ... .. . 7-1 7.1.1 R e a ct o r F uel . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7-1 7.1.2 Re a ct or Com ponent s . . . . . . . . . . . . .. .. . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . 7-1
- 7.1.3 Isotope Storage Cell. ... ....... .. .. .. .. . . .. ... . . .. .. . . .. .. . . . . .. . . . . . . . .... . . . . . . .. 7 '.
~7.1.4 Experiment Facilities .... . ... . . . . .. .. .. .. .. . . . . . . .. .. . . ... . .. . . . . . . . . . . . . ... . . 7-1 7,1. 5 . A ctivated S a m ple s . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . 7-2 s 7.1. 6 R adioa ctive Wa ste . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . 7-2 il
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1 N 7.1.7 Other Radioactive M aterial ... ..................................... ......... ......... 72 72 V) 7.2 RADIATI ON M ONITORIN G .... . . .. . . . . . .. ....... .. . .. . . ... ... . . . . ... . .. . ... . . ... . . . . .
- 7. 2.1 Minim um Procedure s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . .. . . . . . . . .
7.2.2 M onitoring Tech n iq ue s ... .. .. . . . . . . . . . . .. ... . .. . .. . .. . . . . .. .. . .. . .. . .. . . ... . . . . .. . .. .
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7.2.3 M anage ment Surveilla n ce ...... ... ........ .......... . ... ............... .. ... . . . 7-5 i
- 7. 3 I N STR UM E NTATI O N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 75 1 7.3.1 Fixed Ar e a M onit ors . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7-5
- 7.3.2 Airborne Radioactivity Monitors............. ...................................... 75 .
7.3.3 Survey and Laboratory Instruments .... .............. ....... .. ............. 7-6
- 7. 3. 4 Liq uid E ffl ue nts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7-6 ,
- 7. 3. 5 C a libra tio n s . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . 7-6 i 7.3.6 Records..................................................................................... 7-6
. 7.4 EVALUATION OF MONITORING SYSTEMS ...................... ................ 7-6 7.4.1 P articulate Air M onit or. .. . .. . . .. . . . . . . . . . . . . . . .. . . . . . . .. . . . . .. . . .. .. . . . .. . . .. .. . .. . .. . . . .. 7-7 7.4. 2 Are a R adiation M onitors . .. . . . . .. . . . . .. . . . . .. . . . ... .. .. . . . . . . . .. . .. . .. . .. .. . . . . . . . . .. . 79 Chapter 8 ,
AC C I D E NT A N ALYS I S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . .. . . .8-1 8.1 HAZARDS ASSOCIATED WITH THE OPERATION OF THE REACTOR..................................................................................... 81 8.1.1 Radioactive Contamination of Shielding Water ................. .......... 81 8 1.2 Loss of Shielding Water .... ........ .. . . . ... .. .. . ... . .. ... . .... . .. . . ... .. . . . . . . . . . . . . . 83 8.1.3 Handling Irradiated Fuel Design Basis Accident......................... 8-5 8.2 HAZARDS NOT ASSOCIATED WITH THE OPERATION OF THE REACTOR.................................................................................... 8-6 8.2.1 Mechanical Damage to the Reactor ..... ......................... ............. 8-6 8.2.2 Failure of Ele ctric Power ....... ... ...... ..... ....... . .... . . ..... .. . .. ... .. .. .. . 8-6 4
( 8.2.3 Fire......................................................................................... B-6 Chapter 9 OCCUPATIONAL RADIATION EXPOSURES .................. . .... ...... .... . ... . 9-1 9.1 PERSONNEL MONITORING PROGRAM ............ ........... ...... ......... 9-1 9.1.1 Pe rs o n nel E x pos u re s . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9-1 '
9.2 E FFLU E NT M ONITORING . . . . ... . . . . .. . . . .. . . . . . . . . . . . . . . .. . ... . . .. . . . . . . . . . . . . . . . . . . . 9-1 9.2.1 Airbo rne E ffl ue nt s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9-1 9.2. 2 Liq uid E fflue nt s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9-1 '
9.3 ENVIRONM ENTAL M ONITO RING .. . . ... . . .. ... . . ... .. .. .. ... ... ... ... ............. 9-1 9.3.1 Potential Dose As sessments..... ....... . ....... . . . ........... .... ............. .. 9-2 ,
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i h- List of Figures- q V) t j E.taat Enon 2.1 Om aha Are a M a p . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-2
- 2.2 Omaha VA Medical Center Complex Map ........................ ................................ 2-3 :
2.3 Omaha Population ProJ'ections . ... . . ... .. ... . ... . ..... . .. . ........... . ... .. . .. ......... . ... ...... ....... 2 !
, - 2.4 Omaha Area County M ap . . . . . .. . . . . . .. . . .. . . . . ... . . . . . . . . . . .. . . . . . .. . . . .. . . . . . . .. . . . ... . . . . . . . . . . .. .. . . . 2-5 :
- 2. 5 ' Om aha C lim atological Data . ...... .... . . ......... .... ... . . . .. .. .. .. ...... .. ... .. ...... .... ... . .. ... . . ..... 6 !
I 2.6 Omaha Meteorological Data for 1991............... ................................................ 2 1 2.7 Omaha Area Topography Map ... .. . .. .. ... . .. .. . . .. .. ... . .... ... . . ...... .............. .. ...... ........ 29 l 2.8 Seismic Risk Map of The United States ....... .................................................... 2-14 ;
2.9 M odified Mercalli I nte nsit y Scale .... ... ...... .. . ..... . ... . . . . .. ... ... .. . .. . .. . ..... . . . . . ...... . . ........ 2-15
- 2.10 E arthquakes in Nebraska . . . . . ... . . . . . . . . . ... . . . . . . . . . . . . . . . . . . . . . .. .. .. . . .. . . . . . . . . . . .. . . . . . . . . . . . . . . . . . 2-17 ;~
2.1 1 - Seismotectonic Provinces .. .. .... . . . ... ....... ... .. ...... . ..... ....... .. ..... . ..... ... .... .. .. ... ........ 2-18 1
3.1 Omaha VA Medical Center Elevation Drawing................................................... 3 : 3.2 Reactor Laborttory - Basement .. .. . . ....... .. . .. ..... . .. . . ... . . ....... . ... . ........ .. ..... . .. .. . . .... .. 3-3 :
3.2 a R oof Exha ust Fa n Placement . . . . . .. . . . ... . . . .. .. .. . . . . . . .. . .. .. . .. .... . . ... .. .. .... . ... . .. .. .. ... .. . .. . 3-4 ;
- 3. 3 ' R ea ctor a nd Pit . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .-3-6 ...
-3.4 . C are and Reflector Assembly .. . . . .... . . ..... . . . . . . .. .. . .. . ... ... . . ... . . . ....... ...... ... . . .... .... ..... . 3 '
3.5 Fuel Moderator-Element Assembly .... ......... ................................... .......... .......... 3-9
- 3 . 6 S o u rce H olde r . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3-12 .... r
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3.7 Control Rod Drive Mechanism .. . . . . . . . . ... . .. . . . .. . . .. .. . . . . ..... .... .... . . .... . .. .. .. ..... .. . . ... ... . 3-16
- 3. 8 . Reactor Cooling S ystem . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . .. . 3-17 3.9 Two Dollar Reactivity Transient ...... ................... ............................................... 3-22 4.1 - Block Diagram of Instrumentation ;.............. ........................... .......................... 4-2 4.2 Functional Diagram (NM- 1000) .. ... ........ .. . ... . . .. . . . ..... . ..... ... .... .. ...... . ...... . . .. .. . . ... ... 4-5 )
s 4.3 Schematic Representation of Conditions Leading to a SCRAM on the TRIGA '4-8 Reactor......................................................................................................
6.1 Fa cility Organization . . . . . . . . . . . .. . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . .. . . . . 6-2 6.2 D a ily .C hecklist . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B5 b
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- 1.1 Principle Design Parameters... ........... .. .. ..... ....... . .. . . . .. .. .... . . .. .............. . .. .. ... ... . ... . 1-2 -
2.1 l Log of B oring . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . ... . . . . . . . . . . . . . . , 2-11 -i n 2. 2 Laboratory . Te sting . . . . . . . .. . . . . . .. . . . . . . .... . . . . . .. . .. .. . .. . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . ... . 11 ,
2.3 Five Largest Earthquake Events in the Region ................................................. 2-16 j 2.4 Nebraska Tornado Frequency by Month............ ............................................. 2-20
- 2. 5 ' Climatological Normals . . . . . . . . . .. . . . .. . . .. . . . ... . .. . . . . . . . . . . . . .. . . . . .. . . . .. .. . . . . .. . . . . . . . . . . . . . . . ... . . .
2 21-
. 2.6 Maximum Recorded and Me an Wind Speeds ................................................... 2 -
4.1 Minimum Reactor Safety Ch innels..... ...................... ........................................ .4-3 >
7-3 7.1 Acceptable Surface Contam.1stion Levels for Unconditional Release .............. ,
7.2 Significant Fission Products Contribution to Total Activity, Present.. ................- 7-8 1 Recent Exposure History of Fuactor Facility Personnel...... ........................ ... 9-3 >
l 4
1 1
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Tcble of Cont:nts
- Appendices ADDoDdix_A
/O i
-Q RELEASE OF ARGON FROM FACILITY DURING NORMAL OPERATION .......... .. A-1 A.1 ' Release of Argon-41 from Reactor Water............... ............................. ............ A-1 --i A.2 Offsite Dose Calculations .. ... . .... .... ... .. .. .... ... ........ . . . ..... ... . . .. .. . . . .. . ...... ... .. .. . ... .. . . A-6 A.3 Nitrogen-16 Activity in Reactor Room........ ....... ........... ................. ........ . . .... A-6 A.4 Activation of Air in the Experimental Facility ......... .................... ................. ..... A-10 ' ,
Appendix Q CALCULATED MAXIMUM FISSION PRODUCT RELEASE AFTER A FUEL '
E LF. M E N T FAI L U R E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . B-1 F
B.1 Gaseous Fission Products Released from Cladding Failure.... .. .... ................. B-1
= B.1.1 Results of contamination of shielding water............... ................................. B-2 B.1.2 Internal Exposures from Breathing Fission Product Cloud........................... B-3 >
B .2 Fuel H a ndli ng Accident . . .. . . . . . . . . . . . . . . . . . ... . . . . . . . . . . . . . . . . . . .. . . . . . . .. . . . . . . . .. . . . . . . . . . . .. . . . . .. . . . . . B-4 Apoendix C 1
LOSS OF COOLANT WITHOUT FUEL ELEMENT CLADDING FAILURE..... ........... C-1 Appendix 0 DETERMINATION OF SOIL ACTIVATION OUTSIDE OF REACTOR TANK............. D-1 V
D.1 Neutron flux atte nu ation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . D-1 D . 2 S oil Activ a t i o n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .D-4 u;
vi
. _ _ - _ _ _ _, , . _. _ - ~ ~ _
Ltt of Tcbles Appendices L
Appendix A i
- A-1 Summary of resuits from N 16 and Ar-41 ............................................... A-17 l
Aooendix B B . Gaseous fission products produced in fuel element (saturated activity at ,
20KW)..................................................................................................... B-8 j
- B-2 Gaseous fission product activity in the TRIGA element containing the i maximum activity following reactor operation at 0.26 mwd...................... B-9 :
B Activity fiom radiocesium and radiostrontium in the TRIGA element containing the maximum activity following reactor operation at 0.26 mwd B-10 i B-4 Air concentration and radionuclide intake from radiocesium and radiostrontium following hypothetical maximum credible fuel-handling ,
accident.................................................................................................... B-11 l Appendix C
.C-1' Calculated radiation dose rates for loss of shielding water..................... C-1 1
6pandix D A .
D-1 Removal coefficients and attenuation lengths ......................................... D-2 iV D 2 Reactor vessel materials and geometry................................................... D-2 D-3 Soll activation (at saturation) from OVAMC TRIGA Mark I nuclear reactor D-6
'D-4 10 CFR effluents for soil activated isotopes in hypothetical water table .. D-7 4
w 4
/
vil ;
4
- - ., _ . . . _ , , - . - - . - , _._ _.,a m +_ . _ . - - . - . _ - , -
, x-
73 CHAPTER 1
( )
Q)
INTRODUCTION AND
SUMMARY
This report describes the TRIGA reactor end the Omaha Departrnent of Veterans Affairs Medical Center (OVAMC) facility, and provides a safety evaluation which shows that the reactor or facility does not cause undue risk to the health and safety of the public. The OVAMC reactor has been operated safely at the facility between June 1959 and April 12,1991 at 18 kW and at 20 kW from April 12,1991 to date. There has been no change in the method of operation since the issuance of the last facility license in 1983. Gafety analysis demonstrates safe operation at power levels well above the requested licensed power. The reactor will be used for the conduct of research, development and educational activities.
1.1 PRINCIPAL DESIGN CRITERIA The reactor will be operated only in the steady state mode. Reactor power levels will range up to and include 20 kW. A summary of principal design parameters for the reactor is given in Table 1-1, 1.2 DESIGN HIGHLIGHTS The reactor will be located la a below ground reactor pool structure. Reactor cooling will be provided by natural circulation of pool water which is cooled and purified in an external
[ } coolant circuit. Reactor experiment facilities will include a rotary specimen rack, a pneumatic
(/ transfer system, and a core irradiation tube.
The inherent safety of this TRIGA reactor has been demonstrated by the extensive experience acquired from similar TRIGA systems throughout the world. Forty-eight TRIGA reactors are now in operation throughout the world. TRIGA reactors have more than 450 reactor years of operating experience, and more than 15,000 fuel element years of operation.
The safety arises from a large, prompt negative temperature coefficient that is characteristic of uranium zirconium hydride fuel-elements used in TRIGA systems. As the fuel temperature increases, this coefficient immediately compensates for reactivity insertions. The result is that reactor power excursions arc terminated quickly and safely.
The prompt shutdown mechanism has been demonstrated extensively in many thousands of transient tests performed in two prototype TRIGA reactors at the GA Technologies laboratory h San Diego, California, as well as other pulsing TRIGA reactors in operation. These tests included reactivity insertions as large as $ 2.00 with resulting peak reactor power of 250 kW, on TRIGA cores containing similar fuel elements as are used in this TRIGA reactor [1 & 2).
[*
/ )-
C/
i 1-1
. - - ~ . - . . - - . - - - . . . . . . - . . . . . . - . . . . _ . - . _ _ - - . - . . . - . - . - - . . . . . -
l l
J l
Table 1 l PRINCIPAt. DESIGN PARAMETERS -
. Geta .
Fuel elements:
F Fuel-moderator material.................. 8 wt % urenium, j 91 wt-% zirconium, ;
1 wt-% hydrogen. l t
- Uranium enrichment....................... . 520% ;
4 Fuel :::. Tent dimensions................. 1,48 dia. by - ;
28.4 in long - ;
- Cladding . . . . . .. . . . . . . .. . .. . . . . . . . . . . . . . . . . . . . . . . 0.030-in. . thick i
' aluminum or. I f 0.02-in. thick !
! Stainless j Active lattice dimension................... a14 in. dia.'by )
] 14 in. high .;
1 Reflector .
M aterial .. . . . . . . . . . . . . .. . . . .. . . . . . . . . . . . . . . . . . . . . . = Graphite . i e
m Cladding material...................... ... ... - Aluminum Radial Thickness.................... . . . . . .. .. . 12in. i t
n ..
Top and bottom thickness................ 4 in. ,
t.
Nuclear Characteristics t
Thermal neutron flux at 20 kW:
- Average in core............... . .. . . . ... .. . . . . a3.2x10" n/cm'-sec L At rotary specimen rock.................. a1.35x10" n/cm'-sec a9.23x10" n/cm'-sec ,
At central thimble.~..... ........ ... ... ... .. .
At pneumatic transfer tube.............. a3.71x10" n/cm'-sec ,
Initial excess reactivity............. ... ... . . ... . .. .. a0.7% Sk/k Reactivity loss per year (20-kw opera- !
tion , 8 hr/ day) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . a0.14% Sk/klyr-l i= ' Reactivity loss per year (averaged from [
, 1980 to 1994).
- = 0.02% Sk/k/yr. ]
r' Core loading . . .. .. . . . . . . . . .. . . .. .... . . . . . . . . . . . . . . . . . . . . a2.0 kg U-235 J
k 4 }
1 i i
V I
. . _ . _ . . ~ , _ , _ _ . . _ . . _ , _ _ . _ . . _ _ . - . _ _ _ . , . _ - _ . _ - _ . . . . - _ _ . ~ . _ . . . _ _ . _ - . . . . . _ . . , _ .
i t
~
Total roectivity worth of control rods............ a5.0% 6Wk i Void coefficient of reactivity in core............. a-0.0016Wk/1% void ;
Prompt neutron lifetime................ ... ... ... ... .. 8x104 sec !
l Thermal Characteristics }
, r a
Power...................................................... 20 kW J - Method of cooling.................... . . . . . . .. ... . . . . . . . Natural convection of water through core; pool ;
water circuleted through 5-ton Freon water chiller - !
with air-cooled condenser - l u .;
Control ;
! Boron carbide control rods......................... 3
. Drives....................................................... Rock and pinion ;
t
' Maximum reactivity insertion rate............... m0.03% 6WWeec i i
I t
I i ID.Strumentation O.
lV i-NM 1000 microprocessor based neutron monitoring system with: i e Count-rate channel l Log n channel
- Period channel Linear channel..................... .. . .. . ..... 1 Independent % power channel..........
1 j'~ >
i- Servo amplifier for rod controller....... 1 i i ,
Rod-position indicator...................... 2 Magnet scram amplifier.' ..................
. 1 i Water radiation monitor.................... 1
- Water conductivity monitor............... 1 ,
s
.i
- Water temperature mon 2 tor............... .1
- .
i v O }
11 3 -
F
- (
47
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Start-up Source l' i
2-cude americium-beryllium source............ . 1-I Structures !
. 1 Reactor pool... .... . . . .. .. .. .. . .. . .. . . . . . . . . . . . . . . . . . . . . . . 6ft Sin ID by 21 ft deep !
1 i 6
e l Shieldina ,
i Vertical. . . .. . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .Water,16 ft above core i i litadiatioriEacilMan {
4- l t
i Rotary specimen rack................... . .. . ... ... ... 40-position rack located in graphite refLt:: tor i :
Pneumatic tube........ ..... . . ........ . . . . . . . . . . . . . . . . . . 1
! Central thimble........ .. .... . . ... . .. . . . . . . . . . . . . . . . . . . .. 1 1 1
1 I I s
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c Because the reactor fuel is similar, the previously cited experience and tests apply to NW k this TRIGA system. As a result it has been possible to use accepted safety analysis techniques applied to other TRIGA facilities to update evaluations with regard to the characteristics of this e
facility (3).
1.3 CONCLUSION
S
' Past experience has shown that TRIGA systems can be designed constructed, and safely operated in the steady state mode of operation. This history of safety and the d
conservative design of the reactor have permitted TRIGA systems to be sited in urban areas a using buildings without pressure type containment.
Resuhs of this safety analysis indicate that the TRIGA Mark i reactor system will pote no health or safety problem to the public when operated in either normal or abnormal conditions; Abnormal or accident conditions considered in this analysis include:
- a. A step insertion of reactivity, i
- b. Complete and instantaneous loss of coolant water in the reactor pool,
. c. And fission product release from a fuel element ruptured in air, 4
For both the postulated insertion of excess reactivity and the loss of cooling water accident conditions, fuel and clad temperatures remain at levels below those required to generate stress conditions which would cause loss of clad integrity. However, the results of a clad failure are analyzed and it is shown that such a failure will not cause excess radiation exposure, The loss of pool water has been examined from the standpoint of direct radiation to
, operating personnel as well as in terms of maintaining fuel integrity, The effects of argon-41 and nitrogen-16 production during normal operation of the reactor have also been evaluated. Results of these analyses show that production of these radioactive gases will present no hazard to persons in the reactor room or to the general public.
i -
1 1-5
. . ~ _ , . - . _ _. .. ___ _ _ . . . _ _ _ . _ . _ . . -
l i
l
' Chapter 1 References ;
l 4
,I l
1, "TRIGA Transient Experiments" Genera! Atomic GA-0531, Sept.1958
- 2 ' " Hazards Report for Torray Pines TRIGA Reactor" GA-0722,1959 i
I 4
- 3. " Safeguards Analysis Report for TRIGA Reactors using Aluminum-Clad Fuel", General Atomic Division Repcrt GA 7860, March 1967 I
)
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ry CHAPTER 2 (v)
SITE DESCRIPTION 2.1 GENERAL LOCATION The reactor is located in the Department of Veterans Affairs (OVAMC) Medical Center in the City of Omaha, Douglas County, Nebraska (see Fig. 2.1). The reactor is housed in the basement of the southwest wing of the Hospital building (see Fig. 2.2).
The Hospitalis built on the high point of a knoll. To the north is a County hospital, to the south a commercial district, to the west a residentisi area, and to the east a golf course. The reactar site is 2 miles northwest of a large railroad yard and 8 miles northwest of Offutt Air Force Base, headquarters for the Air Combat Command of the U.S. Air Force.
21 POPULATION DENSITY
- The city of Omt.hr. had a 1990 population of 335,795 with 618,262 people in the metropolitan statistic 6 area (MSr Omaha MSA population projections can be seen in Fig.
2.3. Growth continues to grow to me southwest with the largest percentage increase expected in Sarpy County. Figure 2.4 t, hows the counties mentioned in the population projections.
2.3 METEOROLOGY v Omaha is situated on the west bank of the Missouri River; the river level at Omaha is normally about 965 ft above sea level. The rolling hills in and around Omaha rise to about 1300 ft above sea level. The climate is typical continental, with relatively warrn summers and cold, dry winters. It is situated midway between two distinctive climatic zones-the humid est and the dry west. Fltictuations between these two zones produce periods of weathor condition that are either zone or combinations of both. Omaha is also affected by most storms or imd that cross the country. This causes periodic and rapid changes in weather, espec511y deing the winter.
Morst of the precipitation falls du:ing sharp showers or thunderstorms, and these occur mostly during the growing season, April to September. Of the total precipitation, about 75%
falls during the 6 month period April to September, predominantly as evening or night showers and thunderstorms. Although winters are relatively cold, precipitation is light, with only 10% of the total annual precipitation falling during the winter.
Sunshine is fairly abundant, ranging from around 50% of the possible in the winter to 75% or the possible in the summer.
The mean riate of the last killing freeze in spring is April 14, and the mean date of the first killing freeze ;n autumn is October 20. The longest freeze-free period on record is 219 days in 1924, and the shortest period,152 days in 1885. The average length of the freeze-free perbd is 188 days. Cigures 2.5 & 2.6 are a summary of the climatological data for Omaha.
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in the Omoha Economy 1984 - 1991 i
o, aeor omare camer 9 Commute 1301 Homey $i. O mct s HE68102 (4023M44000 FAX l4CIl3 3464000
\
i The following report is a <=paastom of assal populaglea Gr ,wth arends ha the Omaha econosey as sofloaed by the l=i"** IMO - 1990 ledenaar dass anatar.ad eart morsk by she N* ,,,
research staff.
! The indk of the seport concerns chargee la employ- y s. IJ l ment in she Naha metropolitan maanir. area (MSA). b see .
l Major uciari coered inchale construction arul minie :
i rmance, inawaus and esal estase (FRE): gmenment: a,, 2",
ms l a,.dse m.: .ede: a.-es: =d ansportaam. a co'%=e and esilities (TCU). e
, Also briefly covered are popdanan, age distrits- d see-8 tian,lhe r . ; ., r rats, set taxable sales, airline l
passengers, lhe Orsaha Help-Wanted indes. she changs ,
' in she Caneamer prios indes for All Usben W g,,, 3,,, 3,,, 3,, 3,,, 3,,,
I), and buildes permits for the Ory of Omaha.
l The ananal averages for each category are in the APPendia en she last page c( this report.
(
Age Distribetion 1991 Fwaistlos Omaha's economy can be characteriaed by slow g ,,.e. c3un,si sheady growth ont the pass tew decades. Betwoca 1940 and 1990 tue Ony of Omaha's populsion grew froen ,, , , , , , ,,,
l **
! 223.844 to 335.795, am incwase c( 50.0 percers. Frora 1940 through 1990, populance of the Omaha MSA rose l
. 33.6 pertem, fresa 336,731 to 618.262. Since 1990, the m.m 2, i metropolitan area's populanon has grown 5.7 percess. .
The sacady growib of de Omaha sw*opolitan area =
l tan = m .
- population is espected to corainue, prehnunary popula-l ticn prof =*- poist to an asummed nine pestest . . . . ,, ,, .,
increase between 1990 and 2010 for the metro area. The ,,,,,,
larges: percemange increase is especsed in Sarpy County.
Non-Farm Wage and Salary Eaaploysment omahaMsA Omaha MSA Population Project'.eas j
sea.aw
. me e _1 er.ine I Ceesty Cseems 1986 2006 1998 3016 Joe,em D=slas 414.444 430.361 440.373 442,624 445.722 O5arpy 102.583 110J23 116.314 12i.292 126.217
- Vwinepan luo7 17.549 17.e35 it.304 : 8.441 Pasawansmie 82.628 84.537 85.966 85.846 15,752
_ nauha un 1 s.262 sd2.32st na eel utes s7ain 2-4 o--. - . . . . . , ._..y. ,,m_. n,. ,,e_.-e.... m.,,. ,.,e, . ,,m , .
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13SN 0190-3872 I LOCAL CLIMATOLOGICAL DATA ,f'" 's, C199 .
ixxvii. suvuiar wirs covPiniriva uri i OMAHA (EPPLEY AIRFIELD), \, /
. NEBRASKA Daily Data e ,,,
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PflBCIPTTATI0tt E IfEASUEED Bl DICHEE. aCA12 N M0ft-LINEAa STaft0ff PilR35URS E HEASUIED DI 'MCHES OF RfE3tCURY I
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i METEOROLOGICAL DATA FOR 1991 OMAHA (NORTH), NEBRASKA ,
\ L All190( 2 - 41 '22' h LON5!!UO!! % 'Al' W [ LEVA 110N: FI. GRNO 1309 BARD 1315 11Mt 20NE: CEN1RAL WBAN: 94918 -
1 JAN FEB MAR APR MAY JUNE JULY AUG SEP OCT NOV DEC YEAR tr netaAf A.e e,eunt.*r t Dent, Meeleve 25.6 45.0 53.9 64.0 75.6 83.7 85.4 84.2 77.2 63.1 !37.6 39.3 61.2 Deit, nani.w. 8.5 - 25.6 30.8 44.3 58.1 65.9 66.2 65.0 55.3 40.7 23.4 25.2 42.4 17.1 35.3 42.4 54.2 66.9 74.8 75.8 74.6 66.3 1.9 30 32.3 51.8
-Monthlyneathi, De=st. 86 24.1 29.8 41.2 55.9 63.6 (0.1 63.0 51.4 }6.7 e 23.5 9 24.5 40.5 tatreses >
Nighest 44 65 83 87 92 96 96 94 91 84 55 54 96 >
Dete to 20 26 6 28 26 5 25 11 1 20 7 JUL 5 Lo eet -8 .? 15 33 38 58 57 58 31 18 -1 4 -8 <
Date 7 15 2 10 6 7 29 19 19 30 7 4 JAN 7 D(GR([ Oats BA$t 6$ 'f:
heensas 1879 825 699 338 118 0 0 0 116 412 1027 1006 6020 1 Coolsag 0 0 3 21 182 303 341 307 160 13 0 0 1330 L I 0F POS$18Lt $UNSHitet 48 62 54 53 50 70 78 76 76 65 37 f. 62 AVG. $KY COVER lteathst
$warsee - Swa et 6.0 5s 6.5 7.3 7.4 6.5 4.6 4.7 5.3 5.5 7.2 4.9 5.9 Medalpht
- Midnight 5.6 4.9 5.0 6.7 7.4 6.2 4.5 4.3 4.4 4,7 6.4 4.5 5.5 R Of DAt$r see to $wn eet-
- ( l e se 8 8 7 5 5 7 13 13 9 9 5 15 104
- Peetly Cloudy 10 12 11 7 5 11 11 9 14 11 8 6 115
-Cloudy 13 8 13, 10 21 12 7 9 7 11 17 10 146 Precipatetloa
. 01 inc hes or mor e 6 3 11 15 14 11 11 7 6 6 9 7 106
-$ao .lce pellete. hell 1.0 inches or more 5 0 1 0 0 0 0 ( 0 1 3 0 10 Thwa deestores 0 'O 6 7 9 9 7 6' 3 3 2 0 52 J-4evy f og, visabe lity 1/4 elle er less 5 1 4 0 1 0 1 1 0 1 5 8 27 r feepe..tue f
-Me. ows 90 ead -bo.. 0 0 0 0 1 0 9 8 3 0 0 0 29 32 ead belo. 23 4 2 0 0 0 0 0 0 1 12 8 50 Miaf
';2;andbelo= 31 24 18 0 0 0 0 0 1 9 23 27 133 0 and belo- 9 0 0 0 0 0 0 0 0 0 1 0 to j
AVG. $1All0N PRESS. Imbi 972.9 969.9 961.4 964.4 965.8 966.8 967.8 969.9 970.5 967.8 969.9 972.2 968.2 i R(LAllVC l# M10ltV 111 '
, Howe ou 70 73 72 78 75 70 76 69 66 81 82 74 Howe 06 4 79 77 83 87 83 80 64 78 71 83 82 80
'l' I'**' e4 58 58 62 i,3 60 61 55 3 76 67 61 kowe 12 58 Howe 18 70 59 54 58 59 59 56 59 50 3 76 76 61 PA(CIPif A110N llacheelt Watec Eowevelent letal 0.92 0.46 4.24 5.74 3.42 6.52 3.49 3.87 2.06 3.02 1.67 1.87 37.28 i -Greatest 124 hest 0.44 0.44 1.33 1.73 1.14 2.74 2.75 2.03 1.02 1.39 1.43 1.13 2.75 Date 45 18 1 26 16-17 13-14 0- 9 16 11 31 PM- 1 11*12 JUL 8- 9
$ao=Iotalles
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- Dete 20 14 27 6 5 14 28 7 25 17 23 13 M A81 27 Poen GusI -
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24 GEOLOGY
(
The area lies within the Dissected Till Plains of the Central Lowland Physiographic Province of the United States. The topography is gently rolling, and the ground surface at the Hospital varies in elevation between 1200 and 1230 ft above MSL. These elevations represent some of the highest ground within the Omaha city limits, being approximately 275 ft above the level of the Missouri River. Figure 2.7 is a U S. Geological Survey map of the Omaha area surrounding the Hospital site.
The surface soils in the Omaha r 's are primarily loess and glacial drift deposits. Two
. stages of glaciation, the Nebraskan and the Kansan, left thick deposits of till overlying bedrock.
It is beneved that much of the glacial till has been eroded in the vicinity of the Hospital and that not more than 100 ft remains. The till consists mainly of lean and gravelly clays with a few lenses of sand-gravel. The exact depth to bedrock directly below the Hospital site is not known but is estimated to vary between MSL elevation 1000 and 1050, on the basis of the nearest top bedrock information.
. The loess at the site is of Peorian and Loveland formations of the late Pleistocene -
i period. The soil classification of the Peorian indicates that the material consists predominantly of clayey silts and lean clay. The soil of the Loveland formation varies from clayey sitt to fat clay with minor amounts of sand and clayey sand in the basal part of the formation. At the Hospital site, the Peorian is from 30 to 45 ft thick and the Loveland is over 60 ft thick. This would mean that the total thickness of the overburden is approximately 200 ft. Bedrock in this area is limestone and shale of the Pennsylvania period. The surface of the bedrock is very irregular N because of an extensive period of erosion that followed the uplift of the area in early Pennsylvania time and continued to the Pleistocene period. This uplift brought the granite to within 600 ft of the surface in certain areas, forming a ridge known as the Nemaha Ridge or Arch. Also, extensive faulting occurred that developed a major fault, known as the Humboldt fault, which has a throw of over 900 ft. There is no evidence of activity along this fault in recorded time.
No piezometers were installed or observation wells drilled at the site, so there is no definite information as to the exact depth of the water table. However, on the basis of logs of the borings drilled in 1946, the zone of saturation appears to be below 65 ft, although there is some inriication of perched water levels in the soil strata as high as 15 ft.
The following report from the Omaha Testing Laboratories gives the results of a 30-ft test boring made at the center of the reactor location:
4- Field Work:
One auger boring was made on January 8 and 9,1959 at the following location:
Between Columns L and M,10' from L and Between Columns 8 and 9,10' from 9.
Soil samples were taken continuously for obtaining a Soil Log and in addition undisturbed samples were taken at 5' intervals with 2" O.D. Shelby Tubes for bearing index tests.
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' Laboratory Work:
. All soils were classified as to texture, origin of deposit, and consistency [ sic). These
' classifications are shown in Table 2-1 attached hereto. The undisturbed samples were tested for moisture content, dry unit weight, and unconfined compressive strength. The shear strength was calculated as one-half the unconfined compressive strengin. The test results are given in Table 2-2 attached hereto.
General CommentE 4 - The soil boring shows the top soil layers for a depth of 26' to be Peorian Loess, a wind-blown deposit of clayey silt having low plasticity. The next 3' consists of Loveland Loess, a windblown deposit of silty clay having medium plasticity. The bottom 1' of the boring was in a
. glacial ci:y having a higher plasticity.
The tesi data show the soils to be , strong near the surface then decreasing to medium
. strength st a depth just above the clay layer which is a very strong layer. No water table was encountered.
Summary:
Based on the test data, the soil should support itself in vertical walls of an 8' diameter
/ excavation 20' deep even with the surcharge of 3000 psf from the building foundations on the
'\. adjacent soil.
The allowable bearing value at the 21' deoth is calculated to be 4300 psf for a circular footing.
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2-10
,j Q a. TABLE 2-1 i:
_ , LOG OF BORING: ~
Degt3 Soil Tv3 - Ormin Consistency - -
1
- ._ 5 3/4" ' Concrete Floor Slab
- - S-3/4"-8" Sand Cushion i 4
- 8"-3' : Brown Clayey Silt .Peorian Loess Stiff (-)
- : 3'-13'. Light Yellow Clayey Silt ' Peorian Loess . Stiff (-)
13'-21' . Light Grey Clayey Silt '
Peorian Loess Medium Stiff (+)
' 21'-26' ' Tan Clayay Sitt Peorian Loess Medium Stiff 26' 29'. Reddish Silty Clay Loveland Loess _ Stiff (-)
29'-30'J Reddish Clay Glacial Stiff -
j
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1 TABLE 2-2 4
i LABORATORY TESTS -
Unconfined
- Sample Depth Mcisture . Dry Unit Compressive
( -
' (ft) Content (%) Waiaht (ocf) Strenath (osf) Strenath (osf)
U 3-3-1/2 20.1
. .8-8-1/2- .19.3 104.9 9910 495G 13-13-1/2 21.8 101.0 6780 3390 .
18-18-1/2 20.5 100.6 6780 3390 l- 23-23-1/2 23.6 98.1 3490 1745 is 28-1/2 22.6 99.6 3490 1745 29 1/2 -22.0 101.6 11900 '5950 l
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p) 2.4.1 Groundwater Hy& logy (J
Groundwater generally moves in horizontal and lateral directions. In determining the subsurface movement of water, the actual trails are assumed to travel smooth pathways known as streamlines. Thus, the water molecules are taken to travel directly through matter The laminar flow rate v of underground water is given by Darcy's equation [1&2).
v=Kx L
where:
v = Laminar flow rate K = Hydraulic conductivity H1 = Total head measured by a piezometer at site 1 H2 = Total head measured by a piezometer at site 2 L = Distance between the ends of the piezometers The hydraulic conductivity of soil can be affected by temperature, ionic composition of the water, and the presence of entrapped air. The density and viscosity of water changes with temperature. K values are normally exnressed at 20 C. The ionic composition of water can change the K value via ion exchange when exposed to porous material containing clay. In addition, the pores of these clays can be so small as to produce size exclusion for some of the
/ larger ions. Entrapped air in the soil generally causes the K value to be less. Air can become
'N trapped within the soil by a rise in the water table. It may also occur when colder outside water enters an aquifer. For the purposes of our calculations the hydraulic gradient of flow will be assumed to be unity. Therefore, v=K From the site of the reactor the ground water will flow to the south-west. Traveling downward by gravity through the relatively impermeable loess until it reaches the level of impermeable glacial till. As seen in the area map (Fig. 2-1), the Big Papillion Creek runs in an south-easterly oirection approximately two miles west and four miles south-west of the site.
With the water table troughing along this creek the underground water would migrate along the creek untilit returns to the Missouri river south of Offut Airforce Base. Once withia the Missouri river the water would be readily available to members of the public for ingestion.
Glacial Till has a hydraulic conductivity in the range of 1x10a2 to 2x104 meters r second. The time for radioactive isotopes to be carried from the site to the Big Papillion La m is as follows:
6435m + 2x104m/s = 102.03 years.
For comparison the slowest rate is
[] 6435m + 1x10a2 m/s = 204,052,511 years.
V 2-12
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i Thus, any radioactive isotopes produced via soil activation would decay out well before it was made available to the public for consumption.
2.5 EARTHQUAKES The site is located in Seismic Risk Zone 1 of the United States (Fig. 2.8) which is defined on the Modified Mercall:(MM) intensity scale as " Minor damage, distant earthquakes may cause damage to structures with fundamental periods greater than 1.0 seconds; corresponding to intensities V and VI of the MM scale" These earthquake intensities are defined in qualitative terms by the MM intensity scale (Fig 2.9). The site is subject to an earthquake risk estimate characteristic of half the area of the 48 contiguous states. There is a risk of slight damage, principally to poorly built or designed structures.
There have been no reports or physical evider.ce of earthquakes at the site. The five largest earthquakes which have effected the region, as determined from searches of local historical newspaper accounts and other published materials, are indicated in Table 2-3. A complete listing of earthquake occurrences is given in Fig. 2.10. Based upon analysis of newspaper accounts since 1967, the seismic events experie.1ced by the region have led to the following:
m -no significant building damage I 3 -no loss or threat of loss of life, and C -no livestock or crops affected.
This technically stable region is characterized by low level as well as low frequency of earthquakes. The lack of severity of the experienced events has led to their being treated as infrequent curiosities in the regional history.
A plot of epicenters of all earthquake occurrences is superimposed on the regional seismotectonic pattem in Fig. 2.11 [3], (The figure is from Fort Calhoun Station Environmental Report and consequently the site designation is that of Ft. Calhnun. Our site would be Omaha)
The long time history of no significant damage over so broad a region of essentially similar geologic conditions further supports the contention that seismic or fault induced hazards are minimal.
2.6 TORNADOES From 1953 through 1962 a total of 20 tornadoes occurred in a 1 degree square centered near the site, This gives an annual frequency of 2.0 [4]. The probability of a tomado hitting a point is P = zt/A. Where A is the area in square miles of a 1 degree square centered on the point, and z is the'mean path area of a tomado in square miles with t being the mean annual frequency of tomadoes in A. For our site A is approximately 3,578 square miles; z is 2,8208
- and t is 2.0.
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MODIFIED MERCALLIINTENSITY SCALE
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ABRIDGF9 MODIFIEI' MERCALLI 1.9 TENSITY SCALE I. Not felt except by a very few under especially favourable circtunstances.
II. Felt only be a few persons at rest, especially in upper floors of buildings. Delicately suspended objects may swing.
III. Felt quite noticeably indoors, especially on upper floors of buildings, but many people do not recognize it as an earthquake. Standing motor cars may rock slightly. Vibration like passing of truck. Duration estimated.
IV. During the day felt indoors by many, outdoors by few. At night some awakened. Dishes, windows, doors disturbed; walls make creaking sound. Sensation like heavy truck stricking building. Stnading motor cars rocked noticeably.
V. Felt by nearly everyone; many awakened. Some dishes, windows, etc., broken; a few instances of cracked plaster; unstable objects overturned. Disturbance of trees, poles and other tall objects sometimes noticed. Pendulum clocks may stop.
(qj VI. Felt by all; many frightened and run outdoors. Some heavy furniture moved; a few instances of
() fallen plaster or damaged chimneys. Damage slight.
VII. Everybody runs outdoors. Damage nigligible in buildings of good design and construction; slight to moderate in well-built ordinary structures; considerable in poorly built or badly designed structures; some chimneys broken. Noticed by persons driving motor cars.
Vill. Damage slight in specially designed structures; considerable in ordinary substantial buildings with partial collapse; great in poorly built structures. Panel walls thrown out of frame structures.
Fall of chimneys, factory stacks, columns, monuments, walls. Heavy furniture overtumed. Sand and mud ejected in small amounts. Changes in well water. Persons driving motor cars disturbed.
IX. Damage considerable in specially designed structures; well designed frame structures thrown out of plumb; great in substantial buildings, with partial collapse. Buildings shifted off foundations.
Ground cracked conspicuously. Underground pipes broken.
X. Some well-built wooden structures destroyed; most masonry and frame structures destroyed with foundations, ground badly cracked. Rails bent. Landslides considerable from river banks and steep slopes. Shifted sand and mud. Water splashed (stopped) over banks.
p i \
) FIG 2.9 Modified Mercalli Intensity Scale 2-15
. _ _ . . _ _ - _ _ . . . . . _ _ . . . ._ _ _ . _ _ . . _ _ _ _ _ . - - . _ _ . _ _ _ _ ~ . . _
Table 2 3 Five Largest Earthquake Events in the Region -
Epicentral--
i . Date : Nearest Town Coordinates Intensities' (MM) -
-24 April 1867 = Manhattan, Kansas 39.5N %.7W VII ,
15 November 1877. Garland, Nebraska 41.0N 97.0W VII 28 July 1902 - Battle Creek, Nebraska 42.0N 97.6W V 7 January 1906- Manhattan, Kansas 39.3N %.6W : VII s
1 March 1935 Tecumseh, Nebraska 40.35N %.15W VI
' Final evaluation of epicentral intensities from newrpaper accounts.
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' T. riartington Feb,4,1896' 11:45 a.m. ' !!!*, 30!Chadrott ' " ' [Marcti9,1963^ 3:25p.m;.. . ".31r
.5.,8. Hartington - " Sept.16,.1E98 9:59 a.m.-
IV 31/Syracus.' ' . Wane 6.1% 2.47 4.m. J H nt :
,9. Batt!e Creet July.28,1902 ~ 6 p.m. VI. 32. Merrtman . Marctr28,1964 10:08 a.m. #Vlf
' 10. West Point ' Dec.1.1904 9 a.m. ill ' 3 tiBassett - .' "" Oct:16,1972 5:47 a.m.. 4 11, Neo..S.D. Corcar May 10.1908 12:27 a.m. IV 35. Bartlett : May 13.1975 7:53 a.m. ' . W"VI 4 12. Plainview - ~ 7 an.26,1909' 8:15 p.m. IV c 38.SE Rea Willow Dec.1,1977. 1:04 p.m: , ..jil i 13.Columous Feo 26,1910 8 a.m. V 39. SE Red widow *
' Dec.1,1977 1:22p.m; 'lu
- 14. Kirkwcod Sept.16,1915
- 7 p.m. IV 42lSW Cherry , . May 7.1978 4 06 p.m. ' , .V i 15. Stacleton December 1916 na It! 46. SE Red Willow June 6.1979 4:16 p.m. 111 i 16. Texaman- Sept.10,1923 6:30 a.m. IV 48. SE Red Willow July 16.1979 12:03 a.m. 111 17 Gothenourg Sect,2 *924 11 a.m. IV 55. SE Red W1110w . Aug. 2.1979 4:10 a.m. 111
- 18. Ord Oct.14 1927 4
- 10 p.m. IV 59. SE Red Wlucw Aug. 31,1979 8 a.m. IV
- 19. Scottsefuff ~ Aug. 8,1933 na 'V 69. NE Cherry Saot. 7,1981 12:38 a.rn. IV
- 20. North Loup May 11,1934 10.40 a.m. IV 88. Wymore June 3,1982 2:20 p.m. 11 i 21. Chacton July 30,1934 7.20 a.m. VI 94. Dunbar June 26,1983 12:42 p.m. IV j 22. Wood Lake Nov. 8.1934 4.45 a.m. lit d
FIG 2.10 Wortd Heral.:
i Location of 43 Quakes felt by humans . . , latensity is measureo on Modified Mercalli scale, which ranges from I to XII (as opposed to the Richter Scale, which rtms from I to 8), A quake measuring i is felt only by a few people under especially
- f avorable condi
- ions, A quake measunng VI causes slight damage. A quake measurmg XII causes it,tal damage. Numbers between I and ICC not cn the map apply to quakes detected only by monitors, not by humans.
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A Thus, P = (2.821)(2.0)/3,578 = 0.00158. The return period, the reciprocal or p is 1/.00158 = 633 years. .
Therefore, the probability of a tornado hitting the site in any given year is 0.00159 with a return frequency of once every 633 years. Since 1962 there have been 10 tomadoes in Douglas County._
Tornadoes have occurred in the general area of the site and may be expected each'
- year. For example, the tornado that hit Omaha in May 6,1975, was considered to be the most
. . destructive one ever to hit a major American City. The one-quarter-mile wide and 8.7-mile-long path taken by the tornado resulted in three deaths and caused property damage estimated at
$150 million to $500 million. The annual average number of tornadoes for Nebraska during the 41-year period from 1950 to 1991 was 35. Table 2-4 [5] presents the tomado statistics for Nebraska from 1950 through 1991, The fact that the Omaha VA Reactor Facility is in the basement of the hospital surrounded by poured concrete walls with no windows and with 3-4" of concrete overhead
- makes tornado damage improbable.
2.7 ATMOSPHERIC STABILILTY
- Wind direction and speed data are presented i.: Tables 2-5 and 2-6.
O
, _ l 4
- C
-\
(
19
Table 2.4 NEBRASKA TORNADO FREQUENCY BY MONTH AND YEAR SINCE 1950 -
e MAY- JUN JUL AUG SEP OCT NOV DEC TOTAL.
(q F >
JAN) FEB- MAR'1 APR 2'
, Af ^1950; 0 -
_0 :0, .0 2 1 0 1 0- 0 0 __ 6 ,
-1951-'0;-
0; O 1 3- 2 2- - l-
- 0 0 0 0 9 ,
1952; 0 01 0 1 .3 2 l' 2 0 0 .0 0- 9-1953: 0. 0 0 2- 9 7' 2 0 0 -' O 0- 44
. !!954- 0 O' 0 1 2 8 3 2 0 0 0 0: 16-
~
-1955 0 0 0 1 4 9 10 3- 4 0 0 '0- 31 1956 0- 0- 0 2 10 8 2: 1. 4 I .0 34-1957: 0-. 0 0 9 '16 15- 11 2 0 0, 0- 0. 53 1958 0: 0 2 4 12 25 -10 1 0 0 0 54 1959 .0- 0. 0 0 26. .9 3 5 1. 0 0- 0 44' .
-1960. 0 0 0 -1 10 24 1: 7 0 0 0- 0 1961 0- '0 0 0 6 3' 2 0 0 1 0 0 12-1962 .0- 0 0 1. 28 9 1 2 .0 - 0 0 0 41 1963 - 0 0 0 5 3 7 1 1- 0 10 0 0 17
-1964: 0 0 0 8 13 13 5 2 0 0. 0 0 41
--1965.-0 0- 0 0 32 8 6 0 0 0 0 0 46
. 1966 0 0. 0 'O 5 4 1 0- 0 0 0 0 10
-1967 0 0 1 0 1 35 0 1 1 0 0 0 ~39 1968' O O O 2 2 9 7 1 0 0 0 0 21 1969 'O O O 0 3 8 7 2 0 0 0 0 20 1970 0 0 0 0- I 8 2- 2 1 0 0 0 14 1971 0 0 0 1 18 22 8 0 0 3 0 0 52
/ 1972 0 0 0 0 15 4 10 1 0 0 0 0 30 1 1973 0 .0 0 3 4 2 5 0 3 2 0 0 19 1974 0 0 0 14 7 I 2 6 0 2 0 0 32~
1975 0 0 .2 10 22 35 5 2 1 0 0 2 79 1976 -0 0 1 5 3 5 10 1 1 0 0 0 26 1977 0- 0 0 4 35 8 8 '8 5 0 0 0 68 1978 0 0 0 8 16 7 7 2 2 0 0 0 42 1979~ 0 0 0 3 0 8 6 2 0 1 0 0 20 1980 0 .. 0 0 0 10 16 1 3, 0 8 0 0 38 1981 0 0 0 3 0 5- 6 5 0 0 0 0 19' 1982' 0 0 0 1 15 13 1 2 0- 1 0 0 33 1983 0 0 0 0- 2 12 1 0 0 0 0 0 15
.1984 0. - 0 0 4 6 36 4 0 0 0 0 .0 50
-1935 0' 0 4 15- 14 4 2 6 7 0 0 0 52 -
1986 0 0 0 17- 6 10 12 3 6 0 0 0 54 1987. 0 0 3 0 6- 8 5 4 0 0 0 0 26-1988 Oc 0 0 1 '7 5 4 3 0 0 0 0 20 1989 0 0 0 8 4 .21' 4 4 0 0 0 0 41-1990 01 0' 15 1 23 39 6 4. 0 0 0 0 88
.1991 0 -0 _0 9 29 23_ 1 0 1 0 .0 0 63 Total 0 0' 26 143~ 421 504 212 103 36 22 1 2- 1471 Mean 0 0 0.6 - . 3.4 10.0 - 12.0 5.0 2.5 0.9 0.5 0.0 0.0 35.0 Most ~0 0 15 35 39 25 10 7 8 1 2 88 if Year i 1990 17 1986 1977 1990 1958 1958 1985 1980 1956 1975 1990 l
2-20 I
l
( t.
TA8tf 2.5 CLIMT010GICAL llDRMAlS ggfARISON OF NORill 0MAllA WS Willi FORT CAllruM STATION i
Uf8B SIRECIl0E j PEACEst) Wlap SPEED (NFII)
I*' '"* rart c.1heimi Wad Strustig serta ammha uWs merth h WWE rert Co hnen.
times-Isos) itees-1sent fleef Inest flees-19esl anE 38 2.6 8.2 4.9 i DE 3.3 2.4 S.9 44 l
, ENE 3.2 2.4 5.5 4.5 i
E 3.1 3.1 6.1 $.8
[$E 5.4 5.1 6.8 5.4 SE 7.1 0.0 1.8 6.8 55E 10.0 10.2 9.9 8.9 y 5 10.8 10.1 10.4 9.5 55W' 9.1 7.2 9.5 9.3 SW 4.3 3.7 8.5 7.5 W5W 2.4 3.0 7.2 5.5 W 3.9 4.5 1.1 4.4 WNW 5.3 7.5 10.1 4.6 and 8.5 10.7 12.9 5.6 .
IINW G.8 9.4 12.8 6.5 i
n 13.0 5.5 8.0 5.9 I
Missine ---
2.9 ---
f.9 Average --- '--- 8.1 6.3 NOTE: The wind speeds at the North Gaaha W S were recorded 20 feet aboWe ground level, and the wind speeds at Fort Calhoun Station were recorded at 10 meters, above ground level. Data obtained from the Local Climatological Data; see References 4 and 5.
I
I I
i TA8tE 2.4 l
l MAXIMUM RECORDED AND MEAN WIND SPELDS (MPfll
' ~
EPPLEY AIRFIELD NORTH OMAHA It!S Fastest Fastest Direction Mear Direction Mean t
grig[ Wind Speed (1949-1990) i h rees) IEAE (1936-1990) "I"d SP"d ( b ren ) IEAE
[1985-1990)
[1979-1990)
January 57 NW 1938 10.9 41 NW 1978 10.4 February 57 NW 1947 11.1 38 NW 1978 9.6 March 73 NW 1950 12.3 38 NW 1982 10.9 y April' 65 W 1937 12.7 46 NW 1982 10.6 b May 73 NW 1936 10.9 34 N 1983 8.9 June 72 N 1942 10.1 34 NW 1983 8.4 July 109 H 1936 8.9 46 NW 1980 7.5 August 66 N 1944 8.9 39 NW 1980 7.7 September 47 E 1948 9.5 35 W 1980 8.4 October 62 NW 1966 9.8 34 NW 1979 8.9 November 56 NW 1951 10.9 38 NW 1982 9.9 December 52 NW 1938 10.7 37 NW 1981 9.9 Year 109 N 1936 10.6 46 NW 1982 9.3 --
NOTE: The wind speeds at Eppley Airfield were recorded at 70 feet above ground level (agl) until 1974; 20 feet agl since that time. The wind speeds at the N^rth Omaha NWS were recorded at 20 feet agl. Data obtained from the Local Climatological Data; see References 4 and 5.
E
?
L f"
- Chapter 2 References -
1.' = Bouwer, ." Groundwater Hydrology", Mcgrew-Hill.- 1978.
- 2. : Todd, " Groundwater Hydrology", John Wiley and sons. 1980.
- 3. - Nuttie, O.W., " State-of-the Art f;r Assessing Earthquake Hazards in the U.S.", Misc.
' Paper 5-73-1, U.S. Army Engineer Waterways Experimental Station.1973.
- 4. Thom, H.C.S ' Tornado Probabilities", Monthly Weather Review, 91:730-736, 1963.
- 5. NationalWeather Service, Omaha, NE i
4 e
i M
J.
2-23 4
_. _ _m_.__________.____m_m_____._______ _.m._ _ _ _ _ . _ _.-____ _ . . . _ _ _ . _ _ . _ _ _ _ __.____ _ . _ _ _ _ . _ _ - _ . _
__ ._ _ _ , _ _ _ _ . _ . _ . . _ _. _ .._.. _ . _ ~ _ _ _ _ . ,
s
' CHAPTER 3
~
FACILITY'DESIGN:
STRUCTURES, SYSTEM AND COMPONENTS - }
y
~
2 13.11 REACTOR LABORATORY 1 4
3 o The TRIGA reactor _wi!! be located in the basement of the 11-story Hospital building (Fig. '
2 3.1)lwhich was erected in'1951. The Hospitalis of brick and reinforced concrete construction, including floors and ceiling, except that the walls between the rooms in the reactor area will be ^
of wood-stud plaster construction. . Entrance to the reactor laboratory will normally be through the door marked SW-2 (Fig. 3.2). The area to the left of the access door will serve as a health -
t physics control point where pocket dosimeters will be issued as required by appropriate .
f ' regulations and procedures. A log is also kept of all persons entering the area together with the ,
, ~ exposure they received while in the area.
i *
- _ Samples to be irradiated will be prepared in either room SW-2C or SW-2E. Isotopes will - '
g be stored in the isotope-storage area SW-2F. The pneumatic transfer system is located at #
position PT as indicated in Fig, 3.2. Gamma counting will be done in the area marked shield as
- l. shown in Fig 3.2.
b . The reactor room ventilation supply provides heated or cooled 100% outside air to the reactor laboratory at the rate of 1,520 CFM through six ceiling outlet ducts. The exhaust
' effluent of 2,970 CFM exits the reactor room into the outside air by means of an exhaust fan -
installed in the outside wall of the building. In addition, two laboratory fume hoods (Fig. 3.2) exhaust a total of 919 CFM by means of fans installed on the roof of the hospital. Thus, the l ~ reactor area is kept at a slight negative pressure. The reactor area exhaust fan is operated ,
continuously and has a starter switch mounted on the reactor console so that it can be manually _
started or stopped. The fan is equipped with an automatic damper on the exhaust side, so that
- when the fan is off the exhaust portal will be closed, in addition, when the fan is stopped a duct e pressure control closes an absolute damper in the air supply duct and simultaneourly causes :
F an alarm to be initiated on the hospital Honeywell Delta-2000 control system which is
- . continually manned. Thus a single switch on the reactor consolo can stop air from entering or leaving the reactor laboratory and if the exhaust fan stops, the hospital ventilation engineers are
- Immediately notified by the Honeywell computer. The two fume hoods as shown in Fig. 3-2 are
' operated continuously and are exhausted by means of independent exhaust motors on the roof
. of the hospital as shown in the attached Fig. 3.2a. The output of the pneumatic tube is piped from the blower in the reactor water treatment pit outside and adjacent to the basement reactar
- room to the duct of the fume hood shown in the room labeled Radioisotope Reactor Research
,. . Lab (Fig 3.2). This hood has a flow switch alarm that emits an audible signal if the hood blower j ,
stops.L L _ .
!The principle use of the Omaha VA Medical Center TRIGA reactor is for neutron l activation analysis of biological samples and since they must be counted in a well-Ge detector -
Eshortly after they are activated, the radioactiv_ity of the samples must be low. Typical irradiation h times are 3 minutes in the pneumatic transfer tube.-Samples are irradiated in 5 mL vials and the !
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Figure 3.2 Reactor laboratory basement 4
d 3-3 W
4
_ _ _ _ . . . _ _ . ._ _ . . _ ~ _ . . _ . . _ . . . _ . . _ . _. . _ _ ___ _ _ _____._ _ _.___. -
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- RESEARCH BLDG.
HOODS EXHAUST ~~*
i- - ELEVs 52'-0' .
4 RESURCHBl00.
NUCLEAR MEDICINE SUITE 2 EXHAUST
, ELEV 371'-O' l NUCLEAR MED1CINC 1 " REACTOR LAB-l SUITE 1 EXHAUST M30DS ELEV 371'-O'
' ELEV 371'-O' t f
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- MAINHOSPITAL I l I 3 I l
XENON EXHAUST .
ELEV 405'-O' ~
OPCCLINIC i a _
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a 1
DEPARTMENT OF VETERANS AFFAm8
[- Medealcenter j 4101 Woolworth Avenue
. Omaha NE 68105 ~
v i Fig. 3.2a 4
-f
__________ ____-__s a __._m, - _ _ _ _ - - . _ _ _ _ . , ,
. . _ . ..w_ w.s,w...,..,.m-- , . . , .- , , , . , - , _ . , _ , . . ,
,.g .,
t j- vial is opened in the fume hood to allow the "Ar to vent to the atmosphere. Assuming the irradiation vial was empty, a 3 min irradiation of 5 mL of air would emit 0.127 Ci of *Ar and since the sample occupies at least 2 mL of the vial, each sample would contain 0.05 Ci of *Ar.
All experiments irradiated in the reactor must be approved by the Reactor Supervisor or his delegate and the irradiation of individual samples must be approved by the reactor operator or SRO in control of the reactor. Approval of irradiation takes into consideration exposures in the fume hoods. The maximum potential dose vented to the public from the fume hood comes primarily from the pneumatic tube. This has been calculated in Appendix A. The fact that the pneumatic tube is within 6 feet of the operator and the rotary specimen rack is within 14 feet of the operator allows direct control of the experimental facilities. Since the blower for the hoed exhaust is on the roof resulting in the entire duct havir g a negative pressure, any leakage would be into the duct. Consequently, there is no potential exposure within the hospital. This is further monitored by two area monitors (one GM and one scintillation detector) and two continuous air monitors.
If the exhaust fan stops while the reactor is operating the hospital HVAC engineer manning the central system will notify the maintenance man on duty to repair the fan. During off hours the hospital maintenance crew will also respond. If the SRO on duty determines that there is a potential hazard; the reactor will be scrammed immediately.
The areas of potential air exchange are predominantly at the doonways (the doors are p normally closed). There are no gaskets, packing or other materials to prevent or inhibit air i exchange betwaen the reactor room and other spaces within the hospital to which the public (d has access. L eeches in the walls due to conduit, pipes, and other structures are sealed with concrete. With the ventfan off there is still a slight negative pressure in the reactor room caused by the two fume hoods 3.2 REACTOR AND REACTOR SYSTEM 3.2.1 Reactor Pit The reactor is located near the bottom of a cylindrical pit 6.1 m below ground level, as shown in Fig. 3.3. The pit contains a steel tank of 208 cm ID and 0.64 cm wall thickness; the tank rests on an 28 cm concrete slab. Approximately 25 cm of poured concrete will surround the outside of the tank. This steel-and-concrete structure was fabricated in 1.2 m sections at basement level above the pit, and was installed b/ simultaneously excavating the earth and lowering the tank sections into place without disturbing the adjacent soil. When all circumforential sections were installed, the bottom concrete slab was poured and the bottom of the steel tank welded in place. All tank welds were Zyglo tested to ensure leak tightness.
The inside of the steel tank is covered on the sides by a layer of gunite approximately 5 crn thick and or, the' bottom by poured concrete approximately 10 cm thick. The entire inner surface is cor.ted with two applications of a waterproof epoxy resin coating. '
Shielling above the reactor core is provided by 4.9 m of water. The reactor pit has been designed to ensure against leakage of the water: The gunite and its waterproof coating protect the steel tank against corrosion by water, and if a small defect in the coating should occur, the steel tank will provide a secondary containment vessel.
3-5
'k A
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i Figure 3.3 Reactor and pit
>, i V
3-6
! i l
i
,m k" Three emergency storage pits are located immediately adjacent to the reactor tank. The pits are vertical steel pipes 25 cm in diameter and 305 cm long, and are lined with an organic coating. The pits may be filled with water and used for the temporary storage of irradiated specimens or failed fuel elements prior to their ultimate disposal.
The storage pits have been kept dry and have never been used. If used, all fuel elements will be readily supported during storage in a safe geometry (ke less than 0.8 under all conditions of moderation). Irradiated fuel elements will be stored in an array which will permit sufficient natural convection cooling by water or air such that fuel element temperature will not exceed des:gn values. The storage i.its have a nominal inside diameter of 10 inches and the fuel elements he,ve a nominal outside diameter of 1.48 inches. Calculations indicate that if the dimensions fall on the high and low limits, respectively, it might be just possible to weage 37 fuel elements into a sagte layer in one of the pits. From the experimental data cbtained during the initialloading of our reactor (6/59), the most conservative reciprocal multiplication factor (1/M) is 0.55 for 37 fuel elements. This gives a k, of 0.45 which is safely sub critical. From experimental tests conducted by General Atomic, it is known that the standard core spacing of elements in the TRIGA is the optimum configuration. Therefore the close-packed spacing in the fuel storage pit would represent a sornewhat less-than-optimum condition, and would give an even lower value of k,.
For practical reasons, if the reactor core had to be unloaded, the fuel elements would be divided between the three storage pits and consequently, the maximum number of elements that would ever be placed in a storage pit would be 20. From the loading curve, the 1/M value for 20 elements was 0.89 giving a k, = 0.11 which is safely subcritical. If the pits were ever used for storage both the water and surface of the pit would be would be monitored for radiation.
(3V} The spent fuel storage pits are designed with sufficient spacing to ensure that the array, when fully loaded, will be substantially subcritical. For comparison, actual measured multiplication in an array of five fully loeded (19 elements each) storage pits of similar design yields a k, of 0.45 (dry).
3.2.2 Core The core forms a right circular cylinder and consists of a lattice of cylindrical fuel-moderator elements and graphite dummy elements immersed in water. Figure 3.4 shows the reactor cure and reflector assembly.
The active part of each fuel element (see Fig. 3.5) is approximately 3.6 cm in diardeter by 0.36 m long and is a solid, homogeneous mixture of hydrided uranium-zirconium alloy containing 8 wt-% uranium enriched to less than 20% in U-235. The hydrogen-to-zirconium atomic ratio is approximately 1.0. A thin alurninum wafer at each end of the active fuel contains samarium oxide, a burnable poison. Each element is jacketed with a 0.076-cm thick aluminum can. Ten centimeter sections of graphite are inserted in the can above and below the fuel to serve as top and bottom reflectors for the core. Aluminum end fixtures are attached to both ends of the can. The over-all length of the fuel-element is approximato y 72 cm.
An attemative TRIGA fuel element uses stainless-steel cladding and is the current standard element. Like the aluminum-clad elements, the stainicss-steel-clad fuel elements are homogenous mixture of U-ZrH, alloy containing approximately 8.5 weigh percent uranium enriched to less than 20% in U-235. The nominal weight of U-235 in each fuel element is 38 g.
]g ) The hydrogen-to-zirconium ratio is approximately 1.65 to 1.7. The active part of each fuel LJ 3-7
/ i iO .
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l Fig. 3.4 Core and Reflector Assembly -
O ~
3-8
- - _ . - - - - . . - . . ~ . . - . - - - . - . . - . . . - . - . . . . - - - - . . - . . . _ - . . . . . - . . . . . -
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l
- O ALUMINUM TOP END-FIXTURE
. SPACEA 3RAPHITE
.I -
SAM ARIUM
. . RNABLE POISON I C 'l d>
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Fig. 3.5 Fuel-moderator-element assembly l
l l i
3-9 l
! I
N _ .
I
~[d element is approximately 3.6 cm in diameter by 0.38 m long. Aluminum-samarium wafers are -
located at each end of the active fuel as a bumable poison. Each element is jacketed wah a 0.05 cm thick stainless-steel can. Graphite reflector plugs (s9 cm long) are located above and
, below the fuel and serve as neutron reflectors. Stainless-steel end-fixtures are attached to both ends of the can.
General Atomic has been using a mixed core of stainless steel and aluminum clad fuel since 1960 when they were first authorized to use a limited number of stainless stcel clad together with aluminum clad elements, as long as fuel temperature in the mixed Al and SS core did not exceed 550'C. This was authorized by Amendment 9 to License No. R-38 in Oct.,
1960. Change #1 to License No. R-38 dated Sept.1905 authorized General Atomic to use .
stainless steel, aluminum, Hasteloy X or Incoloy 800 up to a full core loading. In addition Amendment No. 31 to Section 4.0 of GA TRIGA Mark I (R-38) Technical Speufications (dated March 1994) authorize various cladding materials and thicknesses, including a mixture of Al and SS clads. Consequently, since a mixed core of Al & SS has been used in the Mark I
. concluded that the health and safety of the public will not be endangered by operating with
' mixed SS and Al fuel.
The elements are spaced so that about 33% of the core volume is occupied by water.
This fuel-to-water ratio in the core was selected because calculations show that it gives very
. nearly the minimum critical mass. At the present time, the reactor contains 57 active fuel elements. The fuelinventory consists of 56 Al clad elements and 1 stainless steel clad 4
i
[d element. The SS element was added to the core on Oct. 2,1995. Eighty five fuel-element positions are available in the lattice; the unused positions will be occupied by graphite dummy elements, i.e., elements in which the uranium-zirconium-hydride fuel is replaced by graphite.
The elements are supported and spaced by top and bottom grid plates of 6061 aluminum. The bottom grid plate is 1.9 cm thick, with holes drilled in it to receive the lower end-fixtures of the elements. These lower end fixtures are 0.64 cm diameter cylindrical
- projections on the bottoms of the fuel cans. A 1.6 cm shoulder is provided on the end-fixture, and the hole in the bottom grid plate is countersunk by a corresponding amount. The weight of the element rests on this shoulder, not on the bottom of the end-fixture, which is used only to position the element as it is being put into place.
The top grid plate is also 1.9 cm thick and has 3.8 cm dia. holes. The top grid plate does not support any of the weight of the elements. The holes serve only to determine the lateral position of the elements and to permit their withdrawal from the core.
The core is cooled by natural circulation of water, which flows through the core from
- bottom to top. Space for the passage of the cooling water through the bottom grid plate is provided by 36 special holes, and through the top grid plate by the gap between the triangular section of the fuel elements and the round grid hole.
v 3-10
(qt.)
3.2.3 Reflector The core is surrounded by a cylindrical reflector 30.5 cm thick,43 cm ID,107 c,m OD, and 56 cm high. This reflector is completely encased in a welded aluminum can, and it is anticipated that flooding of the graphite, in the event that the can should leak, will decrease reactivity. The top and bottom reflectors are the 10 cm graphite sections encased in the fuel
. element cans, so that the reflector in this region is approximately 67 % graphite and 33 %
water, by volume. The reflector assembly is supported at the bottom by an aluminum structure, as indicated in Fig. 3.4.
If water were to flood the reflector housing, then the safety margin would actually be increased by raising neutron absorption and thereby reducing k,. There is no direct way to verify leakage short of detecting a decrease in neutron flux. In a paper given by B. Dodd, A.G.
Johnson and T.V Anderson at the 11th TRIGA User's Conference at Bethesda, Maryland in 1988, they discuss evidence of possible flooding of the reflector at the Oregon State University TRIGA Reactor. They experienced a 20% drop in neutron flux and after evaluating measurements taken by a number of different individuals, concluded that the reflector was probably flooded. One of the reasons for keeping the reflector dry is to avoid the possibility of galvanic corrosion between the graphite and the alun inum; however, corrosion y;;;! only occur in the presence of an electrolyte. The water in the OVAMC reactor has always been kept at a very low conductivity level. Discussion with personnel from General Atomic indicate that impurities in the graphits may be Fe, Si, Ti, Zr, and Ca. However, according to GA these are all well bound and therefore will not change the conductivity of any water inside the reflector. In view of the above, galvanic corrosion appears very unlikely; however, if there were any i corrosion inside the aluminum housing around the reflector it would not become a problem
,d
\ unless sufficient corrosion occurred to significantly reduce the structural integrity of the reflector housing. Galvanic corrosion produces pits and holes rather than an overall thinning of the material. Therefore, breakthrough corrosion would be easily detected by white powdery spots on the surface long before it has progressed far enough to weaken the reflector housing.
3.2.4 Center Channel Assembly and Reactor Tank Covers The center channel assembly across the top of the reactor tank provides support for the drive-and-indicator assembly for the isotope production facility, control rod drives assemblies, and tank covers.
The top, or a portion of the top, of the reactor tank is closed by aluminum grating tank covers that are hinged and installed flush with the floor. A sheet of Lucite pir%c attached to the bottom of each grating section prevents foreign matter from entering the tank but still permits visual observation.
3.2.5 Neutron Source The neutron source consists of a mixture of americium-beryllium, double encapsulated to ensure leak-tightness. Its initial strength at manufacture (1-29-68) was 2 curies. This source has a nominal outside diameter of approximately 2.5 cm and a height of 1.9 cm. The neutron source holder (Fig 3.6)is the same general size and shape as a fuel element; thus, it can be placed in any vacant fuel or graphite element location. The upper and lower portions of the v
3-11
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f upper end of the neutron source holder supports the assembly on the uppe.r plate. ,
I
'3.2.6 : Irradiation Facilities
. . Special irradiation facilities are provided for the production of radioisotopes.;These: ;
= include a rotary specimen rack located in the well in the reflector can,' a pneumatic transfer-i tube, and a central thimble (Fig.' 3A). In adiition,' odd-shaped specimens may be irradiated in ~
! ;the water outside the reflectori 3.2.6.1- Rotary Specimen Rock ,
u -
3 The rotary specimen rack consists bas!cally of an aluminum ring which can be rotated ;
~
around the core. Forty aluminum cups', evenly spaced, are hung from the ring and serve as :
- irradiation specimen holders. The ring can be rotated manually from the top of the reactor pit,'.
1so that any one.of these cups can be aligned with the single isotope-removal tube which runs P
up to the top of the reactor pit. This tube is used for removing and replacing irradiation -
specimens. An indexing and keying' device is provided to ensure positive positioning of the cups. _
e
- The rotary specimen rack is completely enclosed in a welded ahtminum box. The
~
aluminum ring is located at approximately the level of the top grid plate, with the specimen cups c extending from the ring down to about 10 cm below the top of the active lattice. In the radial direction, the centers of the cups are about 10 cm from the inner edge of the reflector
[\
assembly. The box enclosing the rotary specimen rack has been designed to ensure that it will
- remain watertight. Flooding of; iis box will decrease the reactivity of the reactor. The decrease In reactivity is due to the fact that there is a greater absorption of neutrons due to the presence
' of water. Four of the aluminum sample cups, spaced 90 degrees apart, have perforations in the
. walls. One of the four perforated tubes has a 0.625 inch (15.9 mm) diameter hole in the bottom.
L The hole permits testing of the rotary-specimen-rack housing to determine the extent of any
. accumulation of condensation or leaking water. N condensation occurs, as a result of high humidity in the reactor area and low operating temperature', the four perforated tubes can each be loaded (when the reactor is shut down) with a suitable porous container filled with a water absorbing agent.
3.2.6.2 Pneumatic Transfer Tube
{ The pneumatic transfer tube is provided for the production of isotopes with short half
' lives. It consists, in essence, of two tubes leading down through the water tank to a position at the outer edge of the core, where the tubes are joined. The specimen is fed in and out through p one of the tubes and a blower connected to the other tube provides the pressure difference
, ; required to inject or eject the specimen. Specimens are inserted into and removed from the 2 pneumatic system in the reactor laboratory. All samples inserted into the pneumatic transfer
[ system (PTS) must be approved by the Reactor Supervisor or his designate. Investigator must
,. l have irradiations apQved daily by the SRO who signs a posted checklist. There is only one PTS and'it is within 5.5 ft of the reactor operator so control is direct. The same enteria is used
- . . for evaluating samples insested into the PTS as is used for samples inserted into the other
- experimental facilities, [ANS 15.1, (1990) and OVAMC Tech. Specs.). Samples are irradiated in s
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k[ an 11.5 'cc carrier. The normal sample vial used is 7 cc plastic. The PTS is used mainly for neutron activation analysis where the radionuclides produced are short lived and analyzed within one hour of irradiation. Samples are monitored upon removal and ALARA principles are
- followed.
3.2.6.3 CentralThimble A central thimble is provided to per' nit irradiations or experiments in the region of maximum flux and maximum clatistical weight. It consists of a vertical aluminum tube 3.4 cm !D leading from the top of the reactor pit through the exact center of the reactor core and terminating below the bottom of the core. The bottom of the tube is capped, but holes drilled in the wall of the tube ensure that the position in the active lattice will be filled with water during reactor operation.
The shield water can be removed from the portion of the central thimble above the upper grid plate using air pressure to force the water out of the tube through the holes in the tube wall. This provides a highly collimated beam of neutron and gamma radiation for experiments. The radiation dose on the next floor directly above the reactor is 1.5 mR per hour with the thimble, which has 5 cm of lead shielding above it, operating as a beam tube.
Lead bricks ( 2 in thick) are stecked around the central thimble before the shield water is removed and the radiation dose in the reactor room, when the thimble is used as a beam tube is less than 2 mR per hour. The central thimble has only been used once since 1959 for a (o) v beam to determine radiation dose levels during such use and it was demonstrated that with minimal additional shielding ( 2 to 4 in. of lead) above the thimble it can be used with no cignificant hazard to the hospital staff or to the public.
As mentioned in this section, the waterless central thimble has been used only once, July 1,1969 to determine radiation levels. At the present time we have no plans to utilize the facility to extract a beam since it is too collimated. However, if it is ever used we will assure that the radiation dose rates in occupational areas will comply with the hespital's ALARA program, 10CFR20.1101, and Reg Guides 8.10 and 8.18. Since the ficor above is occupied by an ear and eye clinic, we will shield the beam so no radiation above background will be received.
The central thimble filled with water is used to irradiate samples by placing them in a water tight aluminam tube that is completely filled with samples and polyethylene so no extraneous air is within the device. The irradiation device is then is then lowered into the water filled tube. The dose rate at the top of the tube is no different than that at the top of the reactor.
Unauthorized use or inadvertent operation of the central thimble is prevented by the fact that de Reactor Operator is only 10 feet away and in direct visual contact with the top of the centrtl thimble. The central thimble irradiation device can only be inserted or removed by the reactor operator or his designee.
3.2.7 Control Rods and Guide Tubes The three boron carbide control rods operate in perforated aluminum guide tubes. The guide tubes are attached to the bottom grid, and the upper grid provides lateral support. The
'7
) control rod has an extension tube which connects to the control-rod drive mechanism. The
\j 3-14
q ,
safety rod, which ',aring normal operation is completely out of the core, and the shim-safety rod [
are each worth approximately 5 2.25. The regulating rod is worth approximately $ 0.85.
t 3.2.8 Control-rod Drives The control-rod drive mechanisms, located on the bridge at the top of the reactor pool structure, consist of a motor and reduction gear that drive a rack and pinion, and a ,
potentiometer for position indication. The control-rod extension tube and dashpot are connected to the rack through an electromagnet and armature. In the event of power failure or a scram signal, all three of the control-rod magnets are de-energizd and the rods fall into the core. All control rods are scrammable. There is an interlock to prevent any two control rods, including the regulatirg rod from being withdrawn simultaneously. The interlock function of the source neutron count rate is above 2 counts /sec. The drive is nonsynchronous, single phase, and instantly reversible. Electrical dynamic and static braking on the motor are used for fast stops.
Limit switches mounted on the drive assembly indicate the up and down positions of the magnet, the down position of the rod, and magnet contact. The complets drive assembly is enclosed in an aluminum can. The control-rod drive mechanisms have a stroke of approximately 38 cm. Tne maximum rod withdrawal rate is 30.5 cm/ min, and the maximum rate of reactivity insertion is about 5 0.038 per second. Rod-position indicators are provided on the regulating rod and on the shim rod. Fig. 3-7 shows the control-rod drive mechanism.
Interlocks arc provided to assure minimum neutron countrate of 2 cps before control rods can be withdrawn and to prevent withdrawal of any two control rods, including the Reg rod, O, simultaneously.
3.2.9 Reactor Water Cooling and Purification System The reactor is cooled by natural convection of the pool water. A 5 ton freon vapor-compression chiller with an air cooled condenser is used as the heat sink. Water from the reactor tank goes to the water monitor, where the temperature, gamma activity, and conductivity of the water are measursd. It then goes to the suction end of a pump and from there to the chiller unit. From the chiller at goes through a filter and then through a mixed-bed type demineralizer. A bypass line is provioed from the outlet of the chiller to the inlet of a rotometer, where flow rate 's mecsure. The water is then returned to the tant,. The flow inlet pipe is 13 feet above the top of the core, in the event of a rupture in the cooling system the maximum amount of water lost would N to this level. However, the water level would most likely lose only a few inches before the skimmer began to suck air. This would effectively cause i the pumping system to lose its prime. Figura 3.8 shows a schematic of the cooling and water-treatment system.
The water system serves four functions; it
- 1. Maintains low conductivity of the water to minimize corrosion of all reactor I components, particularly the fuel elements.
- 2. Reduces radioactivity in the water by removing r.early all particulate and ,
seible impurifes.
- 3. Mainteins optical clarity of the water.
- 4. Provides a means of dissipating tne heat generated it' the reactor.
3-15
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3.2.10 Ability of Reactor Facility Structure, Systems and Components to Function Property and b Safely for the Term of the License in the 38 year history of the facility there has not been any observed changes in the strength or integrity of the fuel element components, the tank or the lining material due to neutron or gamma radiation damage. Four fuel elements have been examined each quarter by removing each separately and placing it in a device that allows us to examine it in detail 4
underwater with a 20 power telescope each element has failed to show any significant change.
Likewise we have not observed any change in the reactor tank. Consequently, there is no reason to believe that there will be a breach of integrity of the components during the requested
. license extension.
3.2.10.1 Reactor Tank 3.2.10.1.1 The reactor tank is 0.64 cm steel conforming to ASTM standard A 7-56T. t Welding at perimeters and joints of all pieces of plate being a minimum of 3 mm fillet. All welds were zyglo tested to insure leak tightness. The tank rests on a 28 cm concrete slab and approximately 25 cm of poured concrete surrounds the outside of the tank. All concrete had a minimum allowable compressive strength at 23 days of 3,000 psi. The inside of the steel tank is covered on the sides by a layer of gunite approximately 5 cm thick and on the bottom by poured concrete approximately 10 cm thick. The entire inner surface is coated with two applications of a waterproof epoxy resin coating. The reactor
[ tank has never overflowed. Visual observation of the tank with binoculars shows l 1
absolutely no evidence of deterioration of the tank and consequently, it is reasonable to expect that operation for an additional 20 years will have no adverse effect on the tank.
, 3.2.10.2 Core 3.2.10.2.1 Fuel elements are jacketed with 0.076 cm thick aluminum, and visual inspection of 4 elements each quarter have shown no indication of any deterioration or swelling.
3.2.10.2.2 Grid plates are 1.9 cm thick 6061 aluminum and visual observation of the top plate with binoculars shows no evidence of deterioration.
3.2.10.2.3 All Control rods are visually inspected annually, at intervals not to exceed 15 months. Rods are physically removed from the core and visually inspected for signs of pitting or deterioration. To date we have replaced the following control rods for the reason shown:
Safsty rod replaced S/18/64 - Pitting Shim and Regulating rods replaced 2/28/66 - Pitting Regulating Rod replaced 11/18/73 - Pit thru cladding d
3-18 ,
1
p 3.2.10.2.3.1 We were informed by General Atomic, after analyzing the 1966 rod, that the pits
(
( were probably due to iron particles becoming embedded in the surface of the rod during the manufacturing process. After being alerted of this process General Atomic revised their manufacturing and insnection procedure to minimize the possibility of iron being embedded in the aluminum. Since 1973 visual and contact inspection has revealed no evidence of pitting.
3.2.10.2.4 In summary, visual observation of all parts of the core reveals no indication that the facihty can not operate safely for the requested term of the renewallicense.
3.2.10.2.5 Effects of Fuel Aging - There is some evidence that the U-ZrH, fuel tends to fragment with use, probably as a result of the stresses caused by high !
temperature gradients and high rate of heating during pulsing [12). Some of the possible consequences of fragmentation are (1) a decrease in thermal conductivity across cracks, leading to higher central fuel temperatures during steady-state operation (temperature distribution during pulsing would not be affected significantly by changes in conductivity because a pulse is completed before significant heat redistribution by conduction occurs), and (2) more fission products would be released into the cracks in the fuel. ;
With regard to the first item above, hot cell examination of thermally stressed hydride fuel bodies have shown relatively widely spaced cracks that would cause ;
minimal interference with radial heat flow [3). However, after pulsing, TRIGA-O type reactors have exhibited an increase in both steady state fuel temperatures and power reactivity coefficients. At power levels of 500 kW, temperatures have
)
increased by approximately 20*C and power reactivity coefficients by approximately 20% [4). General Atomic has attribute these changes to an increased gap between the fuel material and cladding caused by rapid fuel expansion during pulse heating, which reduces the heat transfer coefficient.
Experience has shown that the observed changes occur mostly during the first several pulses and have essentially saturated after 100 pulses. Because these effects are small and have been observed in many TRIGA-type reactors operated at pulses up to $ 5.00 and power levels as high as 1.5 MW and because the OVAMC reactor is not operated in the pulse mode, they are not considered to pose any hazard during continued operation of the OVAMC reactor.
Two mechanisms for fission product release from TRIGA fuel meat have been proposed [3,5). The first mechanism is fission fragment recoil into gaps with!n the fuel cladding. This effect predominates up to about 400*C and is independent of ;
fuel temperature. OVAMC operating fuel temperatures have never exceeded ;
400*C; thus, this will be the main effect. General Atomic has postulated that in a closed system such as exists in a TRIGA fuel element, fragmentation of the fuel material within the cladding will not cause an increase in the fission product release fraction [5). The reason for this is that the total free volume available for fission products remains constant within the confines of the cladding. Under
, these conditions, the formation of a new gap or widening of an existing gap must k
3-19 l
_1
fx cause a corresponding narrowing of an existing gap at some c'.her location. Such (V) a narrowing allows more fission fragments to tiaverse the gap and become embedded in the fuel or cladding material on the other side. In a closed system in which the density of the fuel meat is constant, the average gap size and therefore the fission product re! ease rate remains constant, independent of the degree to which fuel materialis broken up.
Above approximately 400'C, the controlling mechanism for fission product release is diffusion, and the amount accumulated in the gap is dependent on fuel temperature and fuel surface-to-volume ratios. In the OVAMC fuel this mechanism is not significant because of the low fuel temperature and low utilization factor.
Therefore it is concluded that the likely process of aging of the U-ZrH, moderator under low-power, steady state, nonpulsing operation would not cause significant changes in the operating temperature of the fuel or affect the accumulation of gaseous fission products within the cladding. Therefore, there is reasonable assurance that fuel aging will not significantly increase the likelihood of fuel-cladding failure, or the quantity of gaseous fission products available for release in the event of loss of cladding integrity.
3.2.10.3 Electronics and Mechanical 3.2.10.3.1 Electronics - Routine maintenance is performed on all of our electronics systems (Vn(, by the VA Research Service electronics technician.
3.2.10.3.1.1 The original neutron monitoring system was replaced with a state-of-the art micro-processor system in 1991.
3.2.10.3.2 Mechanical- Routine maintenance is performed on ali mechanical systems by hospital electrical, air-conditioning, refrigeration and plumbing personnel.
3.2.10.3.3 In summay, all systems are operating in excellent condition and there is no indication that their lifetime would affect the safe opration of the facility.
3.2.10.4 Conclusion - On the basis of the above evaluation there is a reasonable basis to expect the total reactor facility to remain safely operable for the requested period of license renewal.
3.2.11 Confinement Design Evaluation 3.2.11.1 Activation of the Soil Surrounding the Reactor Pit The soil adjacent to the reactor pit can capture fast and thermal neutrons which escape from the pit. - The magnitude of the radioactivity induced has been approximated for a typical soil in order to determine whether leaching of activity in the soil might constitute a potential
[ T environmental hazard to the ground water. We have been unable to obtain the depth of the d
3-20
A l water table, but when the new Clinic Addition to the hospital was built (west of the reactor as l
\ indicated in Fig 2.2) pilings were sunk 150 ft with no trace of water. In addition a test bore at the center of the reactor location (SAR pg. 210) indicated no water table was encountered.
Calculations as shown in Appendix D show that at saturation the activity that would leach from a saturated soil through a 1 cm thick annulus at the level of the top of the core directly outside of the tank and with a height equal to the height of the core, together with that from the bottom of the tank would be 6.32 mCl. Due to the short half lives of most of the radioisotopes analyzed, the volume of ground water and the fact that the reactor is only operated for a maximum of 7.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> per day the activation of the soil should have no environmentalimpact 3.2.11.2 Production of Radioactive Gases by the Reactor Radioactive releases as well as the potential exposure to radiation workers and members of the public are calculated and summarized in Appendix A of this part.
3.2.12 Limiting Design Basis The basic limit for TRIGA fuelis dictated by the dissociation of hydrogen from the uranium zirconium hydride fuel and the subsequent stress produced in the fuel element clad.
The results of General Atomic's experimental and theoretical determinations [2,3), together with the conclusions stated in NUREG-1282 [5] and NUREG 0988 [6), show that fuel element integrity is not compromised for cladding temperatures at or less than 500 *C.
l r' 3.2.13 Dynamic Behavior of Reactor j k
This section will consider the behavior of the reactor as a result of the sudden insertion of a large amount of excess reactivity into the core. General Atomic has continued the testing ,
and evaluation of TRIGA by undertaking a high-power transient test program under controlled '
experimental conditions on the prototype reactor. A speciallicense was obtained from the AEC for this series of tests. Some of the salient features of the tests are summarized here [1]. The test was done using the Torrey Pines TRIGA Mark I reactor identical in construction to the Omaha VA TRIGA. The principle design parameters are shown in Fig.1-1 of our SAR. The only difference is that the Torrey Pines reactor had two safety rods worth $2.50 and $2.00, a pneumatically driven regulating rod worth $2.50 and a shim rod worth $4.50.
A 2-dollar sten reactivity insertion has been demonstrated, without deleterious effects either to the reactor or to operating personnelin the immediate vicinity of the reactor. This 2-dollar insertion yielded a reactor period of 10 ms and a peak power of approximately 250 Mw.
This excess reactivity was rapidly compensated by the large prompt negative temperature coefficient, which is an inherent characteristic of this reactor core. Within 30 seconds after initiation of the transient, the reactor power level had returned to an equilibrium of 200 kW. The total energy release in the prompt burst was approximately 10 Mw-sec. The maximum transient fuel temperature was about 360* C.
Curves of the transient power level and of the fuel temperature during this transient are shown in Fig. 3.9. No boiling was observed in the reactor tank and no disturbance of the shielding-water surface was noted during the 2-dollar transient. The integrated radiation dose
'b 3-21
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that an individual would have received had he stood immediately over the reactor tank during (V this power transient would have been 21 miiliroentgen, equivalent, man (mrem).
I During the quasi-equilibrium experiments on the prototype TRIGA, the reactor was operated at a power of 330 kW for a period of approximately one hour with no indication of j instability or bulk boiling in the reactor core. The data obtained in these experiments provide an experimental value of 80 t 5 msec for the prompt generation time for this reactor. The temper-ature coefficient measured !n the quasi-equilibrium experiments can be fitted to good approximation by a constant over the experimental temperature range. This temperature coefficient has been measured to be 0.016 dollar reactivity loss per degree centigrade rise in fuel temperature.
On the basis of the ovidence presented above, it is concluded that there is no hazard associated with a rapid insertion of as much as 2 dollars excess reactivity (1.6 Sk/k) in this reactor. From the above experiment the following reactivity limits can be justified:
a) Excess Reactivity The objective of limiting excess reactivity is to prevent the fuel element temperature safety limit from being reached by limiting the potential reactivity available to the reactor for any condition of operation. The maximum power excursion that could occur would be one resulting from inadvertent rapid insertion of the total available excess reactivity. Limiting the fuelloading of the OVAMC TRIGA reactor to $ 1.00 excess reactivity under clean-cold critical conditions will assure that the fuel temperature will not reach the rnaximum fuel temperature of 500*C where a phose change great enough to cause cladding failure occurs [2,3].
b) Shutdown Margin V)
(
Requiring a minimum shutdown margin of $ 0.51 with the highest worth control rod fully withdrawn, the highest worth non secured experiment in its most reactive state, and the reactor in the cold critical condition without xenon, assures that the reactor can be shut down from any operating condition.
c) Reactivity limits on experiments Limiting the worth of a single experiment to $ 1.00 assures that sudden removal of the experiment will not cause the fuel temperature to rise above the critical temperature level of 500*C.
Limiting the worth of all experiments in the reactor and in the associated experimental facilities at one time to $ 1.00 will also assure that removal of the total worth of all experiments will not exceed the fuel element temperature safety limit o' 400'C.
/%
3-23
Chapter 3 References ,
i
- 1. R.S. Stone and H.P. Sleeper, *TRIGA Transient Experiments", Interim Report General Atomic GA 531, September 19,1958,
- 2. Simnad, M.T., Foushee, F.C. and West, G.B., " Fuel Elements for Pulsed TRIGA Research Reactors", Nuclear Technology,28:31.1976.
- 3. Simnad, M.T., "The U Zrh, Alloy: Its Properties and Use in TRIGA Fuel', GA-4314, E- '
117-833, GA Technologies, Inc., San Diego, CA,1980.
- 4. "Thermionic Research TRIGA Description and Analysis", GA 5400, Rec. C., Nov.1, 1965. [ transmitted by letter dated Feb. 28,1966 (Docket No. 50-277)]. ,
- 5.
- Release of Rare Gas Fission Products from U ZrH, Fuel Materla!", F.S. Fushee, GA-8597, Mar.1968.
- 6. " Safety Evaluation Report on High-Uranium Content, Low-Enriched Uranium Zirconium Hydride Fuels for TRIGA Reactors", Docket No. 50-163GA Technologies, Inc., U.S.
g Nuclear Regulatory Commission, August 1987.
- 7. Safety Evaluation Report related to the renewal of the operating license for the research reactor at the Omaha Veterans Administration Medical Center, Docket No. 50-131. U.S.
Nuclear Regulatory Commission, July,1983. ,
y l
3-24
f 3
- ) CHAPTER 4 U
INSTRUMENTATION AND CONTROL SYSTEMS 4.1 SYSTEMS
SUMMARY
The control and instrumentation systems for the OVAMC TRIGA reactor ate similar to those used in other research reactors in the United States.The nuclear fission process is controlled by using three control rods. The control and instrument systems are interlocked to provide the means for operating the various components in a manner consistent with design objectives. A block diagram of the OVAMC instrumentation and scram system is shown in Figure 4.1 and the minimum required reactor safety channels, functions and set points are shown in Table 4.1.
4.2' CONTROL CONSOLE The reactor control console contains the control, indicating and recording instrumentation required for operation of the reactor. All of the rcactor's essential functions are controlled from the console. On the control panel are (1) rod control switches for raising and lowering the control rods; (2) rod-position indicators tc show the position of the shim and regulating rods to within 0.2%; e.g. the exact linear rod positions can be reproduced by 0.2%.
- There is no position indicator for the safety rod. It must be either all the way down or all the way
[' up or the interlock will not allow the shim or regulating rod to be moved. (3) enunciator lights to
( indicate the up or down position of each rod and rod-magnet contact; (4) linear and log-N power recorders; (5) period, power level, pool temperature, and start-up channel meters; (6) monitor alarm lights; and (7) edditional pilot lights to indicate power on, cooling system on, and startup source countrate. Other enunciator lights on the console indicate the source of a scram signal.
Automatic scram is initiated by (1) an excessive reactor power level as indicated by a signal from either a wide range fission counter or an uncompensated ion chamber, (2) a wide range fission counter or uncompensated ion chamber power supply failure (loss of high voltage), (3) an electrical power failure, or (4) a signal from the wctchdog timer. Manual scram can be initiated by the operator by means of the console scram button or the magnet current key switch. The magnet current key switch breaks the rod-magnet circuit so that the console may be operated without rod withdrawal if the switch is off. Following a scram or the dropping of a rod and after the rod reaches the full-in position, the drive mechanism automatically follows the rod down to reestablish contact.
For steady state operation, the control rods are withdrawn slowly by manual control until the desired power is reached. A servo loop may be used to hold the power constant at the desired level by movement of the regulating rod. The desired power level, expressed as a percent of the full scale power calibration is set on the % DEMAND dial of the Regulating Rod Servo panel and the mode switch is turned to AUTOMATIC position.
The purpose of the Reg rod drive servo is to regulate reactor power to a value set by the operator. The servo compares the reactor power with the power demand as set by the operator and adjusts the regulating rod position in accordance with the difference.
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WlDE RANGE NM 1000 3-i FISSION - PRE AMP AND - - PERICO - i m gespta
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- MICNOPROCESSOP. "APH 4
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% POWER POWER MAGNET m a*===
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m w POWER-DEMAND SERVO TO 4pg CONTROL- AMPLIFIER g i
3 UNCOMPENSATED CHAMBER * ~
POWER LEVEL h-I MONITOR i
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NNTER ER votrER-MApeATION MONSTOR h ,
i CONDUCTIVITY MTER-CONOUCTIVITY
, PROSE MONITOR j.
l RES881ANCE THERMOMETER 1RRTER-TEMPERMURE MONITOR' h
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Table 4.1 Minimum reactor safety channels Safety Channel Function Set Point Percent power
Log N (period) Prevents withdrawal of Minimum period of 3 seconds all control rods Startup (Neutron count rate) Prevents withdrawal of Less than 2 counts per second all control rods Console scram button Scram Manual Fission counter power supply Scram Loss of high voltage lon chamber power supply Scram Loss of high voltage Watchdog Timer Scram Key software tasks take longer than 1.5 seconds Magnet current key switch Scram Manual Simultaneous manual Prevents withdrawal withdrawal'of two rods' -
Withdrawal of shim or regulating Prevents withdrawal rod with safety rod not all the way out or seated
- Withdrawal of safety rod with Prevents withdrawal shim or regulating rod not seated
- Poollevel Alarm when water levelis les[
than 12 ft above top of core ,
Pool water temperature Meterindication 35'C (Administratively controlled)
- Uncompensated ion chamber antilog system
" Fission counter digital system
' May be defeated for control rod calibration k.
t E
4-3
During servo operatien, the reactor is period limited to either 30 sec or 60 sec as V' determined by the position of the servo manual switch on the front of the console. A change in power caused either by a demand change or by recovery from a transient can take place on a period no shorter than the value indicated by the switch. The rod drive servo receives the following three types of information:
- 1. Power demand, from the demand control
- 2. Reactor power information, from the linear recorder
- 3. Reector period information, from the NM 1000.
The reactor power information comes from a retransmitting slide wire on the linear recorder. This signal feeds one leg of a bridge while the power demand signal feeds the other.
Bridge output, representing the difference between reactor power and power demand then feeds the servo. The period information from the NM-1000 also feeds the circuit and limits the period that the reactor can be automatically put on to 30 or 60 seconds. The retransmitting slide wire on the linear recorder is designed to operate from 0 to 100% of the recorder scale. If the recorder is noisy while in auto mode it will cause the reg rod to oscillate, while if the recorder goes out the servo will di ve the rod up being limited by the preset period and the per cent power scram setting.
4.3 CONTROL SYSTEM The control system is composed by both nuclear and process control equipment and is designed for redundant operation in case of failure or malfunction of components essential to the safe operation of the reactor.
In\ '
V 4.3.1 Nuclear Control System The operation of the reactor is monitored by two seperate detector channels. A wide-range fission chamber and a boron-lined uncompensated ion chamber constitute the reactor core monitoring system. These detectors monitor the neutron-flux density of the core and provide trip signals to the safety circuits.
4.3.1.1 Nuclear instrumentation This instrumentation provides the operator with the necessary information for proper manipulation of the nuclear controls (Figure 4.1 & 4.2).
(1) The General Atomic NM-1000 Monitoring and Safety Channelis an industilal neutron monitoring system which !s used both in research reactors and in nuclear power plants.
It utilizes a fission chamber for the neutron detector, pulse processing electronics and a microcomputer to process instrument readings. Output from the microcomputer is routed to an alphanumeric display terminal with date entry and control capabilities. Log and Linear Power can be read on the display terminal and are also riisplayed on a chart recorder. Reactor period can be read on the display terminal and also on a hard wired bar graph. The linear power recorder is auto ranging and the range is indicated on a hard wired bar graph.
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(2) The NM-1000 uses a 1.3 counts /sec-nV encased fission chamber to provide 10 decades of power indication - from shut down (source) level to full power - hence it is also 4 -4
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( referred to as a wide-range power monitor. A count rate circuit is used to monitor power for six decades up from source level; the top four decades are monitored by a Cambelling circuit.
When neutron flux levels become high enough so that the detector cannot be operated in the count rate mode (power proportional to the pulses from the detector) without excessive pulse pile-up problems, the Cambelling technique is used. This technique consists of electronically deriving a signal which is proportional to the root mean square of the current fluctuations present in the fission chamber.
The amplifier / processor circuit employs designs which perform automatic on-line diagnostics and calibration verification. Detection of unacceptable circuit performance is automatically alarmed. The system is calibrated and appropriate scrams checked prior to operation during the prestart checks. Examples of " unacceptable" circuit performance are listed in Appendix 4A under Stack Errors. Thes* < rors can effect the operation of the NM-1000.
Failure of stack 1-9 to take longer than 1.5 sec. will cause the watchdog timer to be tripped.
Intemal diagnostics and self tests are performed continuously in the NM-1000, whether the reactor is secured or at power, to insure operation integrity. RAM, ",OM and battery backup RAM are continually monitored and tested. The accuracy of the channels is 13% of full scale; period and high power trip settings are repeatable within 1% of full-scale inout. The following are the Performance Specifications of the NM-1000:
Sensitivity Linearity Eunctior) (cos/nV) Ranoe { *i)
Log / linear 0.2 2x104 % to 120% 2 Percent power 0.2 1% to 110% 2 p
(
\ (3) A minimum source-neut'on count rate interlock from the NM-1000 prevents rod withdrawal unless the measure source level exceeds a predetermined value. l (4) Power level and scram channel no. 2 comprises a separate uncompensated ion chamber, power supply, and power-range adjustment control and meter to indicate power level from 0 to 110% of licensed power. Scram level on this channel may be adjusted from 20 to 110% of full power.
(5) The automatic regulating channel consists of a ser im,nlifier that controls the regulating rod and thus keeps the reactor power level constant the servoamplifier is activated by an error signal that is governed by the setting of the power-demand control in relation to the actual reactor power level. Because period information also is employed, the servo amplifier may be used to automatically bring the reactor up to power level, within the limits of the worth of the regulating rod, on a preset period of either 30 or 60 sec. Automatic changes in power level on these periods are possible. The servo amplifier will allow quick recovery to bring the power level back to within 1% of the original value, even when step changes in reactivity of up to several *.cnths of 1% of Sk/k is made.
(6) The two neutron-sensing chambers are hermetically sealed in aluminum or stainless steel cans and mounted on the outside of the reflector so that their positions are vertically ,
adjustable in oxler to change e asitivity. There is no apparent hdication of any rod shadowing m . or flux density chifts that effect the response of the two neutron detectors. The detectors are individually calibrated en a yearly basis.
4-6 l
l 1
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4.3.1.2 Reactor Power Safety Channels.
The TRIGA Mark I power safety system is designed to comply with IEEE Standard 3791977 [1] for single failure and common mode failures. A two-channel system is provided in a one-out-of two trip logic configuration.
One of the two power channels uses the output of an independent uncompensated ion chamber (Westinghouse 6937 or equivalent) and indicates percentage of power in the upper two decades of the power range. This chanr.elis part of the original TRIGA Mark i system and is housed in its own independent enclosure with separate power supply. When a preset power level is reached on the meter a relay is activated in the control chassis causing the scram loop to open.
The second power safety channelis provided by the digital wide range power monitor (NM1000). This channel has been designed to satisfy all requirements necessary to operate as a Clasc 1E system as a nuclear safety channel for the nuclear power industry [1-4). The NM1000 neutron monitor design utilizes high speed counting circuits, shielded signal and data communications cables, high speed digital (microprocessor) processing of the signal, and optically isolated output buffers for processing of power data from the fission chamber. As such, its use as a power safety channel meets the criteria of equipment diversity required for redundancy by ANSI /ANS 15.15 [5). To test its response to rapid power changes, the response time of the NM1000 to a sudden change in power (step changes in reactivity) has been measured and compared to the existing analog safety channels on the Mark i Torry Pines reactor by General Atomic [6]. These timing tests showed that the response time of the digital channel fully satisfies the requirements for a scram channel for the TRIGA. They also directly 3
C} compared, by measurement, the time required for detection of signals and low level, high level and period scrams of the NM 1000 with the TRIGA analog system. The times were found to be equivalent.Similar to the analog channel of the Per Cent Power channel, the NM1000 trip output is also hardwired into the scram loop; Thus any overpower condition in the NM1000 will als.o interrupt magnet current. The NM1000, therefore, also provides complete redundancy for operation as a license required safety channel with the analog per cent power channel.
The digital power monitor and safety channel (NM1000) uses the standard, well establishe'i technique of wide range power monitoring by the use of count rate and Cambelling techniques to monitor power from source range to full power [7]. However, the processing of the data from the amplifiers is performed digitally, using state-of-the-art, high-speed data processors. The response time of the digitally processed signal for performance of the required safety function has been shown through direct parallel testing by General Atomic to be equivalent, as regards TRIGA safety, to that from the older analog safety system.
A schematic representation of conditions leading to a scram on the TRIGA Merk I reactor is shown in Fig. 4 3.
4.3.1.3 Internal Diagnostics Intemal diagnostics and self-tests within the NM-1000 are performed to ensure NM-1000 operation integrity. RAM, ROM and battery backed-up RAM (BBRAM) are continually
,- m monitored and tested. The NM 1000 hardware is also equipped with a watchdog timer that will reset the NM-1000 software if it is not reset periodically. Its purpose is to prevent the NM-1000 (v]
47
4 l
NM 1000 ORIGINAL SYSTEM LINEAR 100% POWER POWER -
LOSS OF HV LOSS OF 100% POWER %
CHANNEL ,
. POWER.
MAGNET LOSS OF HV CHANNEL MANUAL
' : MAGNET CURRENT WATCH DOG TIME OUT MANUAL TIMER : ,
' CONSOLE SCRAM SUPPLY ~
FAILURE FACILITY w : _
AC POWER i
Fig.4.3 SCHEMATIC REPRESENTATION OF CONDITIONS LEADING TO A SCRAM
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ON THE TRIGA MARK I REACTOR
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. _ ._ _ _ _ . _ ~ _ _ _ _ . - _ _ _ _ . _ _ _ _ _ _ _ _ _ _ . . . _ _ - _ _
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Q software from failing and not performing its power monitoring function while giv;ng false results, it is reset every 16.7 msec while the NM 1000 communication is in sync with the counter / transmitter and the task level software is executing. Any failure will be indicated on the microterminal and a failure of the watchdog timer willinitiate a scram.
The watchdog system will cause a scram if key soft tasks take longer than 1.5 sec. The 1.5 seconds was selected because when the NM1000 is being used in a pulsing system the i computer is dedicated to gathering other data and the 1.5 sec. is a designed value. During development. ihen all the code had been completed this value was used to allow all subroutinas to update the programmed data and thus prevent spurious trips during pulsing.
4.3.1.4 AdditionalInformation Additional information requested by NRC with reference to our request for Facility License Amendment on October 15,1990, may be found in Appendix 4A. This information pertains to the NM-1000 system and covers the following topics:
(1) Verification and validation plan for GA Model NM 1000 Neutron Monitoring System.
(2) Loss of high witage to neutron detector scram function.
(3) Location and installation configuration for the new instrumentation and control ystem.
(4) Maintenance and surveillance program.
(5) Operator Training for the NM 1000.
(6) Hardwiring of NM 1000 trip output to TRIGA Control Unit.
(7) Description of Calibration Procedures .
4.3.2 Process instrumentation This instrumentation is used for (1) sensing and monnoring parameters associated with l the pool water and (2) radiation monitoring.
(1) The water-radioactivity monitor comprises a gamma-radiation detector and a count rate-meter circuit that gives both audible and visible alarms if the gamma activity in the pool water reaches a preset value. The water monitor is calibrated so that when it is removed from the box (Fig. 3.8) and exposed to a 100 mR/hr field, a full scale reading on the meter is equivalent to 0.1 pCi/cm* of 10-min-old fission products in the water system. The alarm is set for 80% of this value.
During the 38 years of operation the meter has not indicated over 2% of full scale. The water monitor reading is recorded on the daily check list before each startup.
(2) The water-conductivity monitor consists of a conductivity probe and Wheatstone bridge circuit. Daily measurements of the conductivity are made to ensure that neutron activation of pool water impurities will be small and that chemh.at corrosion of fuel cladding is limited. Experience has shown that the buildup or radioactivity is negligible if the conductivity does not exceed 5 micrombos per centimeter averaged over a month.
(3) The water-temperature monitor consists of a resistance-bulb thermometer that senses the bulk pool temperature. Temperature indication is provided on the
'j s control console while the thermistor is within the water monitor (see fig. 3.8). !
4-9
(
(j This system is required to be operational whenever the reactor is in operation.
The reactor is administratively shut down if the temperature exceeds 35'C which is the recommended upper limit for the ion exchanger.
(4). The water-level monitor consists of a float-swF.ch and associated circuitry. This provides both an audible and visual alarm if the water level is less than 3.6 m above the top of the core. An audible and visual alarm is also triggered at the Hospital switchboard which is manned 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> a day.
(5) The facility also has the following radiation monitors which are in operation whenever the reactor is operating:
- a. A calibrated, nonjamming gamma ray monitor with an audible alarm.
This monitor is position a short distance from the isotope removable tube and has an alarm set point of 2 mR per hour.
- b. A calibrated, continuous airborne radiation monitor (CAM) located in the reactor laboratory near the top of the reactor. The monitor can detect both gaseous and particulo.e radioactivities. The monitor also contains a charcoal filter to provide the capability of monitoring for radioiodine. The charcot! filter is not required for reactor operation. Procedures are availible for determination of the CAM's efficiency for radiolodine.
Monitoring for radioiodine is accomplished by analyzing a charcoal filter attached to the side of CAM and positioned so that CAM sir passes through it. The radioiodine monitor is always operable while the NMC CAM is in operation, however, in case of failure of the NMC CAM it is not required for reactor operation. Assay for radioiodine can either be done f
by removing the filter and counting on a Ge well detector and a multichannel analyzer or monitoring the filter in place in its holder with a radiation detector that has been calibrated so as to be able to detect an equivalent of lx10* pCi/cm'/24 hrs. .
4.3.3 Electrical Power System The electrical power requirements of the reactor facility are supplied by tnice circuii:
from the hospital electrical distribution system. The reactor facility has no emergency electrical power system except for two battery-powered lanterns that activate when the building power fails. In the event of loss of electrical power, the control rods are released to fall into the core by gravity, causing safe shutdown of the reactor. j l
1 4
4-10
Chapter 4 4
References _ f
- 1. " Application of the Single Failure Criterion to Nuclear Power Generating Class 1E Systems', institute of Electrical and Electronic Engineers Standard IEEE 37fs-1977.
- 2. ' Criteria for Protection Systems for Nuclear Power Generating Stations", institute of Electrical and Electronic Engineers Standard IEEE 279-1971.
- 3. "Dighal Neutron Monitor NM-1000', GA Technologies Inc. Document INS 25, GA Proprietary Information.1983.
- 4.
- Qualification Test Report: NM-1000 Digital Neutron Monitoring System', GA Document
' E-269-1239, GA Technologies Proprietary Information. May 1984. .
- 5.
- Criteria for the Reactor Safety Systems of Research Reactors *, American National' Standard Institute Guide ANSI /ANS 15.15-1978.
l
- 6. J, Razvi and W.K. Hyde "Results of Timing Tests on NM-1000 High Power Scrams",
General Atomics intemal Correspondence. May 1988.
- 7. 'Rediation Detection and Measurement', Chapter 14, John Wiley Sons.1979. .
o A
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_ } }'
4-11
i Additional Information Submitted for the Amendment j of Facility License No. A 57 ;
Omaha Department of Veterans Affairs hdical Center l
- !. Verification and validation plan for CA Model NN 1000 Neutron Monitoring System. ,
A. he complete reactor control console will not be replaced on1 the neutron monitoring systes for the linear, log and period, be power supply and pre amplifier (A) and computer modules (8) are ;
' j mounted Brown TMon 76the wall behind(C Microterminal theand console (see (D the Period Fig)1.).
and Power The Burr.
(E) largraphs are located on the) original console (Fig. 1). When i
changeover occurs the new Westronic Recorder (F) will replace thw eld recorder presently installed in the console.
Terminal Display items for th6 Microterminal are described in .
Attachment 1.
. S. Scras Verification . " Item" notation refers a#o eressina key on
! migIglerminal (See Attachment 1 for Reference)
- 1. Individually test the trip relays in the HM 1000 with a meter while they are isolated from the TRICA console to ,
assure that they are operating properly while they are put thru the tests outlined below. Relay connections shown in i
- Fig. 2. tecation of cards shown in Figures 3-6 4.
- a. Power level Trip. Relay Board A3, Pins 8&9. i (1) Hi level - push " Power Scras ' rest" bucten on
' console. ,
(a) Hi 14 vel Trip.
(b) Ites 41, Press F8 key (This puts you in .
Data Entry Mode).
(c) Enter 1.0E+02 and then press enter key.
This sets scram at 1004.
(d) -Verify that A2 light is on.
(e) Item 15 Verify that read out shows H for High.
- b. Period Trip. Relay Board A2, Pins 869.
(1) Item 43. Press F8 key ( h is puts you into Data Entry Mode.). >
(2) Enter 7 for 7 seconds and then press enter key.
' Once this value is entered it should not be changed.
(3) Ites 50, Press F8 key, Enter 5 (Campbelling High Test).-
(4) The above should cause a momentary hi period and activate' the period trip. gh positive (5) Verify that light A2 is on and Item 15 show a rsad out of R for rate. '
. c. High Voltage A3, Pins 869.
Trip. (loss of High Voltage) Relay Board '
(1) Push "High Voltage Test" button on console. >
' (2) Verify that A2 light is on.
(3) Ites 60. Verify that Burr Brown readout shows ,
10/ CXHIV.
(4) Item 61 09 until readout shows empty (this
- d. assures that there are no other errors). ->
Startup Channel (Low level Trip) Relay Board A3, Pins !
263.
(1) Ites 40, Press F8 key (this puts you into " Data Entry Moda").
(2) ' Enter 3.74E 07. This corresponds to 2 counts /sec (1.87E.7 - 1 sec.).
(3) Ites 10.
(4) Remove neutron source.
'\
P bl2 - APPENDIX 4A
(5) Verify that light A2 goes on when Burr Brown readout shows 3.7E.7 and Item 15 shows a read
/o\ out of L for low level trip.
t Q e. Watchdog Timer Relay Board A3, Pins 869.
(1) Disconnect plug labeled A3 CTX (card in position 4 in microp.ocessor card box).
(2) Creen light on 10 & Memory Card (card in position 9 in microprocessor card box) should go off and yellow light on.
(3) Verify that light Al is on and Item 60 shows 02 ;
CXFAIL.
- 2. When the operation of all scrans have been verified with the unit disconnected from the existing system, connect the relays in the NM 1000 microprocessor assembly to plug 38 in the TRICA Console as shown in Fig. 2. and raise control rod for each test.
- a. Repeat the procedures outlined it, ?aragraph I.B.1 above withut voltmeter, since control rod will drop signifying operation of relay,
- b. Do not proceed with installation unless all scrans are operating properly.
C. The NM 1000 Neutron Monitoring System has been installed in parallel with our existing system since October,1989, (scram relays not connected and reactor completely controlled with our existing licensed system) and the only major probica that we have had was a noisy fission chamber resulting in too high of a count rate when the neutron source was removed. This problem was resolved by replacing the high voltage poeer supply end assuring that the system was adequately grounded. Critical calibration values as described in paragraph IV below have been recorded each time the reactor is operated since Oct., 1989, and all values have varied less than 54 of the configured values.
D. Scram Response Time g 1. Since the scram relays in the NM 1000 are connected to the original TRICA ." ark I control console, the response time testing will utilize the same procedure as previously used.
- 2. Raise a control rod and measure the scram time with a stop watch. Compare the scram time with the scram time previously determined before the NM 1000 was connected.
E. Sensitivity of Detector (Calibration)
- 1. Before replacing the neutron monitoring system in the old I console the following procedure will be followed: (The i complete procedure described below was done on Aug. 13, 1 1990, and all values have agreed since then. However, the procedure will be repeated prior to switching over systems with the exception that the thermal calibration will be done after we have removed the linear ton chamber from its aluminum tube guide that is attached to the ion chamber mounting ring and inserted the new fission counter)
- a. Place the new Reuter Stokes fission chamber as close <
to the ori l possible, ginal linear compensated ion chamber as
- b. Align the fission chamber as outlined in Attachment 2.
- c. Thermally power calibrate the reactor as per existing SOP and axially move the linear, log, and per cent l power-ion chambers so that their output devices read the calibrated power. (linear and log on old r recorder)
(
k 46-13 APPENDIX 4A
l L d. Axially move the new fission counter so that the new
' linear recorder reads the calibrated per cent power,
- e. Verify that the new los recorder in also reading the i calibrated value. ~;
- f. Verify that the__ ton chamber readings compare with the
' fission chamber readings. ;
- 11. Loss. of high voltage to neutron detector scram function.
A. Original Systes For Cent Power chamber. i
- 1. The original per cent power lon chamber and scram circuit i vill be used as our second detector and consequently there ;
has been no change sads in the loss of high voltage scram l furction. t
- 3. ' New NM 1000 systoa.- i
- 1. 14ss of High Voltage to the new fission counter activates ;
l' Relay A3 in the NM 1000 assembly which is hard wired to the' '
- per cent power.scras in the TRICA console-(Fig'2). The !
[ system then scrans as in the original system.
III, 14cr*' and tastallation configuration for the new instrumentation and ' .
cc . system See Fi !
- -A. Specitication of g. 1.-the temperature and humidity _ conditions of the l systea.
- 1. The NM 1000 was tested to the following extreme _ conditions ,
and found te operate satisfactory.
I, a. Tegerature: 0 60'C. .
- b. -Relative humidity:-0 984. !
- 2. The Omaha V.A. Triga is installed in an air conditioned i hualdity controlled room.
- 8. ' Evaluation of enclosures, cabinets, and connections to building j
, structures for general ruggedness.under potential dynaale ;
i conditions. '
- 1. System was tested to meet-the requirements of IEEE 344 1975. '
- 2. Preamp 11fier and alcroprocessor assembly enclosures are . ,
sounted on basement cement wall (Fig.1, A is g) and hard wiring, to TRIGA console is done thru conduit (Fig.1, C), i C. Evaluation of potential contact chatter during dynamic conditions. l
- 1. All relays are normally eneraited and do not chatter under !
postulated seismic acceleration.
- 2. The system is designed to scras on potential contact chatter !
conditions, i D. Evaluation of cable and component shielding, configuration and/or :
isolation to sitigate the consequences of electro magnetic ,
interference (EMI).
- 1. Signal outputs are either OPTO or transformer isolated, .
inter connection is via twisted shielded pairs. !
l 2. EMI levels sufficient to cause a response would cause a-transient upscale response. If transient upscale response F exceeded 1004, scras would occur.
E. Evaluation of power supply buffers to mitigate power transient e f fec ts . L 1.- Power supplies are electronically regulated and input power ,
is buffered by a shielded passive line filter followed by a :
shielded steel NEMA active trackin 1 enclosure.g filter. . All are enclosed withir. a 1
- 2. After a power loss the surveillance program described in .
paragraph 4 of this document will be repeated. This is the ;
'~ - same prostaa used each day before starting up' the reactor.
t F. .
Evaluation of Instrument isolatlon devices. i
- 1. For the analog outputs isolation is provided by an " Analog
- Devices" isolation converter. Analog Devices rates the input to output isolation at V.,00 V._RMS. The device sects '
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d i
14 APPENDIX 4A o
, , , , , , -,..,+,,g-,www-,-- v --w- %, g r , . , = , r.,_,co-,e,,,.ww.-se-,v . s. m e -w w ww . v ow ,w.,m---,---+..--m-e - - - _ - _
l 1
the IEEE S;andard for Transient Voltage Protection (472 C 1974: . Surga Withstand capability) and offers reliable
( operation over .25' to +s5'c temperature range.
- 2. Trip outputs are provided by relay contacts. The isolation ratings at's not supplied by the manufacturer.
- 3. Isolation for communication is provided by optical isolation. Ceneral Atomic has tested the isolation of these optical isolators to 120V A.C.
l 4. Mechanical isolation is provided at the field termination points for all safety and non safety inputs and outputs.
IV. Maintenance and surveillance program.
A. At the start of each working day or after each major interruption of operation, the reactor electrical and machanical systems shall be checked out and certified to be in p. opes working order, in accordance with the check list shown in our existing S.O.P.
- 8. When the NM 1000 is installed the Chamber and Instrument
- i. Sensitivity section of the 5.0.P. will be replaced with the checks l 1 described-in Attachment 3.
I C. Maintenar.;e will *oe performed on any of the items in the Daily Checklist which can not be verified; so that the facility is In ;
compliance with the current Technical Specifications.
V. Operator Training for the new system (outline listed below). ,
A. Description and theory of fission counter.
- 1. Discrete neutron counting techniques.
- 2. Campbelling technioues.
B. General Description of W 1000
- 1. Physical Description.
Performance specifications.
2.
4
- 3. Amplifier Assembly.
4 Signal Process Assembly. .
f O C.
5.
1.
2.
Installation and Setup Calibration Functional Description General.
Source. Range Log Count Rate.
- 3. Wide Range Log Power.
- 4. Power Range
- 5. Multirange Linear Power.
D. Systee Description
- 1. Hardware.
- 2. Software.
E. NM 1000 Sof tware Description
- 1. Hardware /Sof tware Description.
- l 2. NN 1000 Systen Function.
- 3. Sof tware Organization.
- s. CPU Resst.
- b. Counter / Transmitter Message Character Received,
, c. Local Display Input Character Received.
- d. Local Display Out i 4. Database Organization, put Character Sent.
- s. Database Item Description. '
, b. Error Description,
- c. Daily Checklist. '
VI. Hardwiring of NM 1000 trip output to TRICA Control Unit
-A. The relay outputs cf the NM 1000 as described in paragraph I B above are connected to plug 38 or the Control unit of the TRICA-Console. Connections are shown in Fig. 2 and Fig. 7 2 attached.
(Please substitute the enclosed Fig. 7 2 for our original submission).
.B. - Connection is thru conduit C Fig.1.
1 9
h 4-15 APPENDIX 4A
. .- ._- .-_ ~ ___. .- - - - .-_- - - .-. - - - - .. - .- . - _ _ - .
l l
\ ,
l VII. Figure 7.2 (see revised Fig. attached) Replace Fig. 7 2 submitted October, 1989. ,
A. Connections'of V/F eut on J2A1 to counter transmitter J1A1
>( (drawing 0387 60820) to g3 to U14 to U19 to U12 of counter 1 1
(drawing 0387 60820). Drawin Maintenance Manual 31171000.gs are in CA Operations and
- 1. The connection between the V/F Converter and Counter i has been made on revised Fig 7 2.
S. We do not have remote display unit so it has been deleted on the
, diagram.
C. The calibration generator use the sunsation of the outputs cf a multi. frequency digital clock to produce a pseudo square wave in the Campbelling region. The calibration values are adjustable and stability is determined by power supply and passive component
. drift. In count rate mode discrete frequencies are counted and ,
stability is determined by the clock crystal. Function switching is performed by transistor switches controlled by the counter transmitter which is in turn sof tware controlled. If a j -calibration function is selected when at power a rod withdrawal prohibit function operates ex. spe in h15h Campbell calibration which causes a sc.an.
VIII. Proposed Technical Specification Changes Watchdog Timer.
A. See revision to Table 3 1 (page 7) and page 5 [ Attachment 4 & 5].
- 8. See revision to pa SER [ Attachment 6]ge 7 6 (Table 7.1) of proposed Amendment No.1, C. Please substitute the enclosed Attachments 4 5, and 6 for the corresponding pages submitted in October,19$0.
IX. Minimum count rate rod withdrawal interlock. .
A. The minimum count rate withdrawal interlock was set for 10 counts per second seconds in our October,1990, request due to the fact that the noise level of the detector was high. However, Jires the original request for change was submitted the noise has s,, / beea eliminated. Consequently, we request that 10 counts per second be deleted and replaced with the ori 2 counts per second (attachments 4 and 6). ginal licensed limit of k
.d
/ ! 49
(
4-16 APrENDIX 4A
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...______....e_...
- NM-1000 NEUTRON MONITOR QUICK REFERENCE CUIDE
[' : TERMINAL DISPLAY ITEMS
\ :
\ : CROUP 1 (F1 KEY) CROUP 2 (F2 KEY) CROUP 3 (F3 KEY)
- - COMPUTED VALUES SINGLE DETECTOR CAMPBELL DETECTOR
- los PERCENT POWER (S) 20s DET COUNTS (S) 30= CMB COUNTS (S)
- ils PERCENT POWER (F) 21 ALPHA 0FTSET (S) 31: NOISE OFFSET (S)
- 12s PERIOD (S) 622: (20) - (21) (S) #32= ((30)-(31))**2 (S)
- 13 PERIOD (F) 23u 33: CMB MULTIPLIER (S)
- - les MODE (I) 248 634: (33) * (32) (S)
- #15 RELAY STATUSES (I) 25 DET PPCONST (S) 35: CMB DET PPCONST(S)
- 16s 26: 36=
- 017s 200MS MESSAGE (a) 27 (S) 37=
- 18s EPOWER - MANTSA(S) 628: DET X0VR VAL (S) 638= CMB X0VR VAL (S)
- Its EPOWER - EXPONT(S) 29s DET X0VR SETF (S) 39: CMB X0VR SETP (S)
- CROUP 4 (F4 KEY) CROUP 5 (F5 KEY) CROUP 6 (F6 KEY) :
- TRIP SETPOINTS MODES OF OPERATION ERROR STACK :
- 40 LOW LVL TRIP (S) 50s OPERATION MODE (B) 60= ----
- Als HI LYL TRIP (S) 51: FLT TRIP MODE (B) 61= STK POSITION 1 (B) :
- 42s PLOAT LYL TRIP (S) *52: M-LINEAR MODE (B) 62: STK POSITION 2 (B) :
- 43= RATE TRIP (S) e53: LOCKED EXP (I) 63= STK POSITION 3 (B) :
- 44: 854= PERIOD DAC HI (B) 64: STK POSITION 4 (B) :
- 45: 55 65= STK POSITION 5 (B) :
./' : 46s 56: 66= STK POSITION 6 (B) :
( : 47: 57: MEMORY ADDRESS (I) 67: STK POSITION 7 (B) :
- 48= '58: MEMORY VALUE (B) 68: STK FOSITION 8 (B) :
- 49 59: VERSION NUMBER (A) 69: STK PcJSITION 9 (B) :
- Al LIGHT DEFINITIONS A2 LICL*T DEFINITIONS :
- ON = EQUIPMENT FALLURE (ERR IN STACK) ON = RATE OF CHANGE TRIP, HICH :
- OFF = NO EQUIPMENT PAILURES ,
LVL TRIP, OR LOSS OF HICH :
- VOLTAGE :
- OFF = NONE OF THE ABOVE STATUSES :
- STEPS TO READ OUT A TERMINAL DISPLAY ITEM: :
- 1. SELECT CROUP BY PRESSING TH! FOLLOWING: :
- F1 KEY = CROUP 1 (ITEM 1 WILL ALWAYS BE DISPLAYED) :
, : F2 KEY = GROUP 2 : :
- .... ETC .... :
- 2. C0 TO ANY ITEM IN GROUP BY PRESSING THE NUMBER KEYS (0-9) :
- OR ---> PRESS "." TO STEP FORWARD TO NEXT ITEM. :
- OR ---> PRESS " " TO STEP BACKWARDS TO LAST ITEM.
- TO CLEAR ALL ALARMS, PRESS F7 KEY, THEN ENTER CODE 90. :
- DATA ENTRY STEPS: 1. SELECT TERMINAL DISPLAY ITEM TO BE CHANGED. :
- (** LOCAL ONLY) 2. PRESS F8 KEY (THIS PUTS YOU INTO DATA ENTRY MODE).: '
~~s : (% REMOTE ONLY) 3. ENTER NEW VALUE FOR ITEM, THEN PRESS " ENTER" KEY. :
) : (#= DISPLAY ONLY) :
\._,/ : in single range. in campbelling range :
- Power Equation: POWER = (!!EM22)* ITEM 25 = (ITEM 34)* ITEM 35 :
+.............____.. .._..........__ ......_.............. ______..........+
(A) ASCII (B) = BYTE DATA (F) = F I X'9 (I) = INTEGER (S) SCIENTIFIC l 4-21 i
! : OPERATING MODES DESCRIPT10N (Control byte bits 7654 3 2 1 0) :
- 0: NORMAL Noract operation, No Self Test 11XXXX00 :
' : 1: CTR LOW Counter Low Test 10XXXX01 : I
/ ! 2 CTR MID. Counter Middle Test 10XXXX10 : l
( :
- 3
- CTR WI 4 : CMB LOW Counter High Test Campbelling Lsw Test 10XXXX11 01XXXX01
- 5: CMS HI campbelling Hisi Test 00XXXX01 : .
- 6 = SQUARE WAVE Squars Wave
- Automatically * : )
- 7 = PUL3E Reactor Puise **** Cycled ***** : )
(X means don't care)
- Operation Mode 6 causes Rate of Change trip to be inhibited for 10 :
- seconds. After 10 seconds, the NM-1000 automatically switches to : :
- Operation Mode 0. : i
- Operation Mode 7 causes the HM-1000 to internally sequence through :
- Operation Mode 4 for 10 seconds, Operation Mode 0 for 10 seconds, :
- followed by a switch to operation Mode 0. All trip statuses are :
- latched upon entry to operation Mode 7 and unistehed upon completion :
- of the timed operation Mode 0. :
- STACK ERRORS DESCRIPTION :
- 00 = EMPTY Stack position is empty (i.e. no error). : '
- 01
- BADER BAD Error (posting procedure detected a bad error code) : ,
- 02
- CXFAIL Ctr/Xmt Failure (No input to Micro from Ctr/Xat Assembly): '
- 03 : CXSYN1 Ctr/Xmt out-of-Sync (detected by the SCAN program) :
- 04 s CXSTN2 Ctr/Xmt Out-of-Syne (detected by the PROCES program) :
- 05 = CX8USY Ctr/Xmt Busy Error (Output Uart is not ready). :
[ : 06
- CXCBE Ctr/Xmt Rev Uart Error (No Concrol Byte Recieved). :.
( :
07
- CXCOMM 08 : CX-15V Ctr/Xmt Rev Uart Error (Parity, Framing, or overrun).
Ctr,Xat Assembly; Failure in -15V power supply.
- 09 CX+15V Ctr/Xat Assembly: Failure in +15V power supply. :
- 10
- CXXIV Ctr/Xmt Assembly; Pailure in High Voltage power supply. :
- 11
- CXHARD Ctr/Xat Hardware Error (statua bits 7, 6, or 5 is low) :
- 12 s MI-15V Microprocessor Assembly -15V failure. :
- 13 s MI+15Y Microprocessor Assembly +15V failure. :
- 14
- SDX0VR (Item 28 > Item 29) and (Item 38 < Item 39) :
- 15 : SOERR Single Detector offset Err (Jtem 20 - Item 21) < 0 :
- 16
- COERR Campb Detector offset Err (Item 30 - Item 31) < 0 :.
- 17
- 18 s BADROM BAD ROM error (hardware error, replace ROM board) :
- 19 CORUN Computing Task overrun (TASK 1 took more than 0.184 sec.):
- 20
- E0RUN Executive Overrun (all TASKS took longer than 1.5 sec.) :
- 21 ICRUN Interrupt Overrun (TASK 1 & 9 not run within 1.5 sec.) :
- 22 : T1 FAIL TASK 1 FAILURE (TOOK MORE THAN 200 MILLI-SECONDS) :
- 23 DBCHG Data Base Change Error (change in battery backed up RAM) :
- POWER LEVELS DESCRIPTION - ITEM 14 :
- 0
- CAMP Campbelling Operating Range (using only Camp, signal) :
- 1 : SINGLE Single Detector Range (using Single Detector) :
- POWER RANCE SWITCH OVER POINTS :
/N : :
('-- : 1) SINGLE -> CAMP ITEM 28 > ITEM 29 -
- CAMP -> SINGLE ITEM 38 < ITEM 39 :
+...__....... ....r...._........_........___.....-__......._..__...__......+
4-22
's NM-1000 NEUTRON MONITOR COMPUTED YALUE EQUATIONS f-( :
.( .
- 1) ITEM 10 & ITEM 11-- PERCENT POWER
. - (- :
- ITEM 14
- 1, SINGLE DETECTOR
- ITEM 10 m ITEM 22 (C0UNTS PER SECOND] *
- ITEM 25 [SDET PERCENT POWER CONSTANT]
- ITEM 14 8 0 CAMPRELLING DETECTOR
- ITEM 10 = ITEM 34 [ COUNTS PER SECOND] *
- ITEM 35 [CDET PERCENT POWER CONSTANT) l
- 2) ITEM 12 & ITEM 13 - RATE OF CHANGE
- ITEM 12 = LOG (ITEM 10 (CURRENT PERCENT POWER] -
- ITEM 10 [ PERCENT POWER 1 SECOND AC0]) *
- 60.0 / 26.05767
- 3) ITEM 28 - SINGLE DETECTOR CROSSOVER YALUE
- ITEM 28 = ITEM 20 :
- 4) ITEM 38 - CAMP 8ELLING CROSSOVER VALUE :
/N . .
I : :
\ ITEM 38 = ITEM 30 l
i :- :
i : :
l- : :
L : :
l : :
y : :
\ .
^~- .
l + ..................................-_.............................--- ....,
4-23
o.......................--.......................--......................., 1 l 3 : l l : MULTI. LINEAR MODES DESCRIPTION :
g .................. .............--.................................
- 0a AUTC Continulous tracking over all decades. :
Locked onto a specified decade.
['~'}
\ <
- 1 MANUAL :
\' # : Note: Use display item 52 to set the above. eulti-linear mode. :
- Also, display item 50 is used to set the operation modes. :
- RELAY DEFINITION WHEN ACTIVE :
..... .....__ ....................--.. ...._____................... : l FLOATING PERCENT POWER TRIP..... FOLLOWS CONDITION :
- 1
- 2 RATE-0F-CHANGE TRIP............. FOLLOWS' CONDITION. : l
- 3 LOW LEVEL PERCENT POWER TRIP.... FOLLOWS CONDITION. :
- 4 HIGH LEVEL PERCENT POWER TRIP :
- (OR) HIGH VOLTAGE F AILURE. . . . . . FOLLOWS EITHER CONDITION. :
- 5* ANY PROCESS FAILURE ............ PROCESS FAILURES ARE: :
- (TURN ON A2 LIGHT) 1. RATE-0F-CHANGE. :
- 2. LOW LEVEL PERCENT POWER. :
- 3. HIGH LEVEL PERCENT POWER. :
- 68 UNDEFINED....................... NEVER. :
- NOTE: CURRENTLY. THE TRIGA HARDWARE DOES NOT SUPPORT THESE RELAYS. :
- THE Al LIGHT WILL REMAIN ON UNTIL A CLEAR CODE 90 IS ENTERED. :
- THE A2 LIGHT WILL REMAIN ON UFTIL THE 77 KEY IS PRESSED. :
~, : :
I :
I :
r'~'
( -: .
+......................................................................... .
4-P4
l Alignment of NM 1000 Neutron Monitors I
(/ Theory: The NM.1000 neutron monitor is capable of measuring ten decades of ;
neutron flux with a single fission chamber. Alignment of channel requires a basic understanding of the software operation of the NM.
1000, which is detailed below.
The NM 1000 uses two techniques to calculate reactor power. For low power operation, the channel calculates reactor power utilizing counting techniques where discrete neutron counts from the fission chamber are directly proportional to reactor power. For high power operation, the channel calculates reactor power utilizing Campbelling techniques where the reactor power is proportional to the square of the ras value of the a.c. signal from the fission chamber. Combining these techniques, with sufficient overlap, allows the NM 1000 to cover a full ten decades To calculate the reactor power, the following two equations are used by the NM-1000.
Count Rate Region: (Equation 1)
Percent Power - [ Counts /Sec]*[ Count Rate Power Constant)
ITEM 10 - ITEM 20
- ITEM 25 Campbell Region: (Equation 2)
Campbell Linearizing Campbell Percent Power - [ Counts /Sec]8 * (Factor) * [ Power Constant]
ITEM 10 -
[ ITEM 30]8
- ITEM 33
- IuM 35 The following procedure details the complete calibration of the NM-
_g v) 1000 channel.
5.1, and 6.1.
For routine recalibration, follow steps 3.2, 4.2, Method: The first step in the NM- 1000 alignment is to properly position the fission chamber in the reactor core. The detector is a standard General Atomics supplied RSN-314 Reuter Stokes fission chamber.
1.1) The detector should be positioned to draw 1.0 mA from the high voltage power supply (800 volts nominal) at 100% power. Next, the PA-I5 preamp discriminator should ,be adj us ted. The number of shutdown counts (with the start up source in a cold core) will depend on the reactors license power, and will be a function of the crossover from count rate to Campbell.
' The cross over point from count rate to Campbell should be set about three decades down from the full power flux (about 0.1% power).
This gives three full decades of Campbell signal with adequate hysteresis for the crossover from Campbell to count rate.
2.1) To set the discriminator, change the count rate to Campbell crossover in the NM-1000 sof tware to 8X10' (ITEM 29). Bring the reactor to 0.14 power. Adjust the discriminator (R304 in the PA-I5) for 1.2X10' counts terminal (ITEM 20).per second This as will setting displayed give 12 oncounts the NM.1000 per second display at IX10** percent second at IX10' percent power (typical power shutdown (typical rodpower), and permit withdrawal 1.2 counts per Change the count rate percent power constant (ITEM 25) to 1. point).87X10*?
for a power indication of 0.1 percent power as read on the NM 1000 display terminal (ITEM 10).
J 4-25
i Y
. . c 3.1)' Change the count rate to Campbell cross over to 1.2X108 (ITM 29). :
- 3.2)' Increase reactor power to - full power and allow the . reactor to
.-(O 44
- d 1
.3.3) ' stabilize Adquet the (several Campbell minutes)'
amplifie r gain (R27 ir the Campbell amp) for - :
8XLO' counts per second as displayed on the % 1000 terminal (ITM 1 30). t
) 4.1) . Set the Campbell linearizing factor to 0.370'(ITEN 33), the Campbell 4
to count rate crossover to 1950 (ITM 39), and the campbell noise
, to 65.
d' 4.2) To constant set the(ITEN Campbe 31)ll percent power constant, take ten consecutive readings of the Campbell signal (ITEN 30) at full power and find-the average. Using equation (3), calculate the Campbell per:ent -
power constant and enter as ITm 35.
i ITEN 35 - "
T m 10 (Equation 3)
L [1Tm 30)3 * [1 TEM 33) ,
a .
i verify that the power indicated on the % 1000 is 1004 (ITM 10).
, t 5.1) To check the crossover alismsent from Campbell to countrate turn on.
the log chart recorder and scram the reactor from full power.
Examine the trace in the crostover region (about 0.1 percent power) and note any discontinuity. If a discontinuity is evident, observe i
whether the Campbell signal is too high or too low at the crossover.
6.1) To precisely- align the Campbell to countrate crossover if a discontinuity is evident, use the following procedure.
Campbell signal too high ~at crossover:
Increase the Campbell detector noise constant (ITEN 31) by about
.s five to ten percent of the current Campbell detector noise constant and repeat step 5.1 'above.- Note any discontinuity at the crossover, i and make appropriate adjustments -to the Campbell detector noise constant.- Repeat steps .5.1 and 6.1 as necessary.
l Campbell signal too, low at crossover:
Decrease the Campbell detector noise constant (ITEN 31) by about
! five to ten percent of the current Campbell detector noise constant '
l and repeat step 5.1 above. Note any discontinuity at the crossover, and make appropriate adjustments to the Campbell detector noise constant.-~ Repeat steps 5.1 and 6.1 as necessary.
a e
i s
h.'
4-26
-.-. - - .. -- - , , . - . - - - - . . i- . - . - . . - -
l l l ADDENDUM TO DAILY CHECKLIST r
\ CHAMBER AND INSTitU4ENT SENSITIVITY I. NN 1000 Cali> ration Constants A. Verify that all calibration constants entered into the NN 1000 agree with the values posted on the control console (see sample label below).
NM-1000 Calibration Constants 21 =
25 -
29 = ,
31 =
33 =
35 -
3C =
l 40 =
41 -
42 =
43 =
51 =
52 =
53 =
PA-15 DISC =
(
DATE:
BY:
)
II. Calibration modes 1,3,4 & 5 are sequentially tested for correct power level outputs. Item 50 of the NM 1000 is programmed to the a mode and the corresponding power level is read from Item 10. ppropriate The power level is then compared with the configured test levels and is deemed OK if it falls between 954 and 1054 of the configured value.
A. The configured values are stored in the following configuration channels, (Attachment 1, Page 2)
- 1. Iten 1 Counter Low Test 2.24E 5 t Power
- 2. Item 3 Counter High Test 5.74E-2 t Power
, 3. Item 4 Campb. Low Test 1.08E+1 % Power 4 Item 5 Campb. High Test 1.09E+2 4 Power
> B. The procedure is as follows: (Itea refers to pressing key on Burr Brown Microterminal).
- 1. Item F5, Ites F8, Item 1. Enter, (Reading-CTR LOW)
- 2. Item F1, Read t Power and linear recorder and record in log
- 3. Item F5, Iten F8, Item 3. Enter, (Reading-CTR HI) -
4 1 Item F1, Read t Power and linear recorder and record in log '
- 5. Iten FS, Item F8, Item 4, Enter, (Reading-CMB IDW)
- 6. Item F1, Read t Power and linear recorder and record in log.
7.- Iten F5, Iten F8, Item 5, Enter, (Reading-CMB HI) i
- 8. Iten F1, Read 4 Power and linear recorder and record in log i
- 9. At the end of the calibration test reset the NM 1000 to the
[n
(
.s normal mode.- Iten F5, Item F8, Iten O. Enter (Reading -
Normal) l 4-27 l
i I
1 . .
- .III. High Power Level Trip A. -Raise control rod.- '
B.. Push " Power Scram Test" button on console.
C.- Verify that control rod scrams. i D.- _ Verify that A2 on Burr Brown is.on.
E. Item F1, Item 5. Verify that road out shows H for High.
IV. Period Trip
-A.- -Raise control rod (Do not use same rod used for III.-
above). "
B. Item FS, Item F8 Item 5. Enter (Reading =CMB HI).
-C.- The above should cause a momentary high positive period and activate the period trip.-
D. Verify that control rod scrams.
E. Verify that A2 light is on.
F. Verify that Item F1.. Item 5, sh,ws a read out of R for rate.
V. Ioss of High' Voltage Trip
, A. Raise control rod.
B. Push'"High Voltage Test" button on-console.
~
C. Verify that control red scrams.
.D. Verify that A2 light is on.
E. Verify that Item F15-shows a read out of 32/V.
- VI. Startup Channel.(Low Level Trip)
F. Remove Neutron Source.
G. Allow enough time for the NM 1000 power to drop below the source trip limit.
H. Verify that light A2 goes on when Burr Brown readout >
shows 3.7E-7.
I. Verify that Item F1, Item 5 shows a readout of "L" for low level trip.
J. Try to raise control rod'. .
VI . - Watchdog Timer (Do first run day of each month)
A. Raise control rod.
B, Disconnect plus labeled A3 CTX (card in position 4 in microprocessor card box).
C. . Green light on IO & Memory Card (card in position 9 in microprocessor. card box) should go off and yellow on.
D. . Verify that light Al is on.
i E. Verify that Item F6(60) shows CXFAIL.
F. Go thrum Item F6, 1-9 until read out shows empty.
G. Verify that control rod scrams.
1 I
os 4-28
-- _ = . - . ,- ;_. _--
X L
l 4
. CHAPTER 5 ~
1 _ .
CORE PHYSICS lThe physics of the reactor has been studied in considerable detail on a critical-assembly;
"' _ mock-up as well as on the operating prototype. In the succeeding sections, the more important i features of the core' physics'are discussed.~
/ 5.1 - . CRITICAL MASS
- The prototype TRIGA reactor attained criticality with 54 fuel elements, or about 1.g kg of
-U 235.' The OVAMC reactor attained criticality on June 26.1959 with 54 fuel elements.
E 5.2 VOID COEFFICIENT y
- , . ' The void coefficient should be very similar to the values obtained on the l critical assembly, which were measured to be -0.15% Sk per 1% water void at 23' C in the
! - central region of the core and +0.04% Sk per 1% water void at 23' C at the core-reflector r
interface, whero there is a region of graphite-loaded dummy elements. The core average vakio I . was approximately -0.14% Sk per 1% water void, which agrees with calculations.
I LCalculations indicate that the total reactivity value of the water in the core is about 10%.
t 5.3 - MODERATING PROPERTIES OF ZlRCONIUM HYDRIDE Experiments performed by General Atomic personnel at the Brookhaven National Laboratory have shown that zirconium hydride has very unusual moderating properties for slow neutrons [1] . The results of these experiments can be explained by assuming that the hydrogen-atom lattice vibrations can be described by an Einstein model with a characteristic energy hv = 0.130 ev. This description is consistent with the theory that the hydrogen atom occupies a lattice site at the center of a regular tetrahedron of zirconium atoms. The basic consequences of this model, which have been experimentally verified, are that:
- 1. Neutrons with energios of less than hv cannot lose energy in collisions with zirconium hydride.
- 2. _ A slow neutron can gain an energy hv in a collision with zirconium hydride with a.
probability proportional to exp(-hv/kT), which increases very rapidly with temperature.
Since hv? kT, it has been found that zirconium hydride is not effective in thermalizing neutrons but that it can speed up neutrons already thermalized by water by transferring to them a quantum of energy hv.
5.4. TEMPERATURE COEFFICIENTS A particularly large effort has gone into designing the reactor in such a way that an .
increase in the t6mperature of the fuel elements will result in a relatively large decrease in
' reactivity /This large prompt negative fuel-tem'perature coefficient results from the following
{
( ' effects:
5-1
l l l
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.'"/ 1. Cell and inhomogeneities
- 2. Doppler
- 2. Core leakage Cell an Inhomooeneities. Although a TRIGA reactor is frequently referred to as a homogenous reactor, a large part of the prompt negative temperature coefficient arise.s because of the inhomogeneity associated with the 3.8 cm diameter fuel-moderator elements and the interstitial cooling water. This cooling water provides a part of the neutron moderator (metal: water ratio of a 2.1) and is instrumental in producing the large negative temperature coefficient by the process noted below and first identified by F. Dyson. When the fuel temperature increases, the zirconium hydride temperature essentially follows it instantaneously, thus increasing exponentially (Boltzman equation) the number of bound hydrogen atoms in excited levels. This increases the probability of speeding up the neutrons within the fuel element when they collide with the bound hydrogen and gain energy (hv from the lattice vibrations. This results in the hardening of the neutron spectrum, a decrease in the f:ssion probability, and an increase in the fraction of neutrons lost from the fuel element because of leakage from the element. When the neutrons leave the hot fuel element, they rethermalized in the cooling water and undergo increased parasitic capture in the interstitial core water and the cladding material, especially when steel clads are used. CELL EFFECT is an important contributor to the prompt negative coefficient; note that it is almost entirely dependent on the heterogeneous appearance of the core to thermal neutrons. If the core had no water and were constructed entirely of U-Zrh.., the prompt negative coefficient would of course still exist but l,-) would be much smaller because the harder neutron spectrum would not permit as much V parasitic capture in hydrogen and cladding material.
Doppler Effects. The uranium in the LEU is approximately 20% U-235 and 80% U-238.
The capture resonances in U-238 are Doppler-broadened by an increase in the fuel temperature causing a decrease in the resonance escape probability, p.
Core Leakaae . The core leakage contribution derives basically from the same mechanism which produces the cell effect. The core can be envisaged as a large super-cell
. with reflector acting as a moderator. When the core heats up, leakage is increased and relatively more captures occur outside the fuel.
Experiments at General Atomic have shown that the cell effect is the dominant contributor to the temperature coefficient. The fuel temperature coefficient of the TRIGA reactor has been experimentally demonstrated to be -0.01% Sk/k per *C rise in average fuel temperature. The temperature coefficient associated with heating the water and the fuel in the TRIGA reactor core is extremely small. The total reactivity contribution do to this latter coefficient over the range of 10*to 60*C is less then 0.08% Sk!k. The operational chcracteristics of the reactor are therefore primarily determined by the extremely large prompt negative temperature coefficient. The experiments performed to determine this temperature coefficient demonstrate that it is a prompt coefficient and it is nearly constant over the power range from 0 - 1.4 Mw.
!p i
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( ~ Reactivity effects associated with water temperature will have essentially no effect on either the normal operating characteristics or the transient behavior of the reactor for the ,
following reasons:
1.- :Under normal operating conditions,~ if the ' reactor were operated at a power level .
of 18 kW even with the water-cooling system tumed off, the average temperature l of the a 4000 gal of water in the core tank would be increaeed at a rate of less than 0.78'C/hr. this would certainly cause a negligible reactivity ' perturbation.
- 2. .The temperature of the core water does not change matcrially during a reactor -
trar.sient. The transient behavior of the reactor is determined primarily by changes in fuel temperature.
The transient behavior of the reactor has been studied in detail in the test program of ;
j- ' the prototype TRIGA [2] (see also Section 3 2.13 Dynamic Behavior of Reactor).
- 5.5 REACTIVITY PERTURBATIONS Perturbations of the reactivity resulting from physical changes in the core and reflector can be of importance to the safety of some reactors. It is possible for water to be introduct>J accidentally into the reflector region of TRIGA by flooding the reflector graphite, the specimen -
- rack, and the pneumatic tran:,fer tube. It is expected that water in any of these regions will cause a decrease in reactivity. This decrease in reactivity is due to the fact that there is a greater absorption of neutrons because of the presence of water.
The effect on reactivity of interchanging fuel elements and graphite dummy elements must also be considered. Experiments on the General Atomic subcritical assembly have shown
. that the cylindrically symmetricalloading of fuel elements surrounded by graphite dummy
! elements is the maximum reactivity configuration. Any rearrangement will result in a decrease
- in reactivity..
1 a
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+
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5-3 ,
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i . - _ . - _ _ -.- _ ,.._. _. _ -_ _ ___ ,_
i
-- , m Chapter 5 i
References
.1. - A.W. McReynolds, M. Nelkin, M.N. Rosenbluth, and W. Whittemore, " Neutron _
Thermalization by Chemically Bound Hydrogen and Carbon *, Proceedings of the ,
i Second United Nations International Conference on Peaceful Uses of Atomic Energy, i Geneva.- Paper UN/1540.1958.
.- - September 1958.'
l 4
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y + 1 h a 1.
M CHAPTER 8' 1
l CONDUCT OF OPERATIONS' j
-6.1- FACILITY ADMINISTRATION j
3 6.1.1 : ' Overall Organization L ,
Figure 6-1 illustrates the organizational structure that is applied to the management and
-~
operation of the reactor facility. These responsibilities include safeguarding the public and staff- l 1- from undue radiation exposure and adherence to license or other operation constraints.' The Reactor Supervisor is delegated responsibility for overall facility operation. {
c .
Facility operation staff is an organization of a Reactor Manager / Reactor Supervisor and .
cat least one equivalent person (Reactor Operator). This staff of two provides for basic _
e operation requirements.L A staff of one may occur during transitional periods (e.g. to replace a
- vacant position). At present both the Reactor Supervisor and the Reactor Operator are holders - '
- of Senior Operator Licenses. Students and researchers supplement the organization. Titles for ompone s of he o a!!o folo 6.1.1.1 Director, Veterans Administration Medical Center Has overall responsibility for all functions of the Medical Center and has delegated his authority to the Reactor Supervisor to assure the integrity and security of the special nuclear
. material and the safe operation of the reactor facility. .
6,13.2 Chief of Staff i~
- Is responsible for the professional staff of the hospital and is Chairman of the
. . Radioisotope and Radiation Safety Committee. 1 6'1.1.3 Associate Chief of Staff Resexrch Is the Director of the Research Service to which the Nuclear Reactor Facility is attached.
All administrative functions such as personnel matters, payroll,' purchasing, secretarial, etc. are
- . supplied by the Research Service._- ,
. 6.1.1 A Reactor Safeguards Committee
~
The Reactor Safeguards Committee has broad responsibilities to provide independent -
. review of facility activities for safe operation.
iThe committee shall meet at least onco per calandar quarter and shall be composed of at least tour members who represent a broad spectrum of_ expertise appropriate to reactor -
L -' technology.-
ss L
i 6-1'
- 4. :- j l
. . _ _ _ - _ - _ _ _ __ . -. . _ _ -.__ .~ . _ _. . . - ._ _ . ._. .
O O O '
t DIRECTOR OMAHA V.A.
MEDICAL CENTER i I
CHIEF OF i STAFF
/
l
- ASSOCIATE CHIEF 0F STAFF 7,
RESEARCal REACTOR SAFEGUARDS COMITTEE REVIEW #91T' ,
FUNCTION FUNCTION
~
RADIOLOGICAL REACTOR '
l
- SAFETY OFFICER'
- ~ ~~ ~
SUPERVISOR *
---~~- 1~~~~j I
REACTOR OPERAT10 tis
- Responsible for Facility Operation ,
Figure 6.1 Facility organization h
l I
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\ -
N 6.1.1.5 Radiological Safety Officer The Radiation Safety Officer acts as the de egated authority of the Radiation Safety Committee in the daily implementation of policies and practices regarding the safe use of
- radioisotopes and sources of radiation. He or his delegate will be knowledgeable of the facility radiological hazards. Responsibilities willinclude calibration of radiation detection instruments, measurement of radiation levels, contiol of radioactive contamination, maintenance of radiation records, and assistance with other facility monitoring activities.
6.1.1.6 Reactor Supervisor Reactor operation at the OVAMC is directed by a reactor supervisor. Responsibilities of the reactor supervisor include control of license documentations, reactor operation, equipment maintenance, experiment operation, instruction of persons with access to laboratory areas, and development of research activities.
Activities of reactor operators with USNRC licenses will be subject to the direction of a
. person with a USNRC senior operator permit. The reactor supervisor shall be qualifed as a senior operator. This person is to be knowledgeable of regulatory requirements, license conditions, and standard operating practices. The OVAMC Operations Manual will be maintained by the reactor supervisor.
6.1.1.7 Professional and Classified Staff
[
(/ Professional and classified staff, such as research scientists, reactor operators, technicians and secretaries, will supplement the organization as necessary to support facility programs. Personnel associated with the research reactor facility [1] shall have a combination .
of academic training, expeuwe, skills, and health commensurate with the responsibility to provide reasonable assurarw
- Nt decisions and actions during normal and abnormal conditions will be such that thu P.;iay and reactor are operated in a safe manner.
6.1.1.8 Facility Staff Qualifications 6.1.1.8.1 Reactor Supervisor At the time of appointment to the position the Reactor Supervisor shall have a minimum of 5 years of nuclear experience. He shall have a baccalaureate or higher degree in an engineering or other scientific field. The degree will fulfill 4 years of experience on a one-for-one time basis. Equivalent education or experience may be substituted for a degree.
6.1.1.8.2 Senior Reactor Operator At the time of appointment to the position a Senior Reactor Operator shall have minimum of a high school diploma or equivalent and should have 4 years of nuclear i
4 6-3
+
I
,/# %
( ) experience. A maximum of 2 years of experience may be fulfilled by related acadernic or technical training on a one-for-one time basis.
6.1.1.8.3 Reactor Operator At the time of appointment to the active position, operators shall have a high school diploma or equivalent.
l I
6.2 REACTOR OPERATIONS The TRIGA reactor for the Veterans Administration Medical Cente, ,n Omaha, Nebraska is designed to operate continuously at a power level of 20 kW , however, normal operation is usually 7-8 hours per day. The maximum available excess reactivity, for any temperature conditions which could conceivably be encountered, will be limited to $1.00. Experiments conducted with General Atomic's prototype TRIGA show that the reactor power level wil' be limited to safe values even if all of this available excess reactivity thould be suddenly introduced into the reactor (see " Reactivity Accidents").
Operation of the reactor and activities associated with the reactor control system, instrument systems radiation monitoring system, and engineered safety features will be the function of staff personnel with appropriate license certifications [1]. Operation willinclude the implementation of required procedures, execution of appropriate experiments, actions related to safety, and the preparaticn of required reports and records.
g) v 6.2.1 Procedures Written procedures shall govem many of the activities associated with reactor operation.
Preparation of the procedures and minor modifications of the procedures will be by licensed operator. Substantive changes or major modifications to procedures, and prepared procedures will be submitted to the Reactor Safeguards Committee for review and approval. Temporary deviations from the procedures may be made by the reactor supervisor or designated aenior operator provided changes of substance are reported for review and approvcl.
Activities subject to written procedures will include routir.e start-up, shutdown and operation of the reactor; control rod calibration; emergency and abnormal conditions; fuel loading, unloading, storage and inspection; control rod romoval cr replacement; criteria for evaluating experiments; reactor power calibration; radiation n'onitor calibration procedures; review and approval of changes to procedure; personnel radiation protection; administrative controls of operation and maintenance; and filling of reactor tank with make-up water.
6.2.2 - Routine Operation Procedures The Daily Checklist (Fig 6.2) will be completed prior to each daily start-up, at the completion of each days operation and before start-up after any maintenance on the reactor. This checklist allows the checking of each electronic and mechanical component to assure that it is l
l functioning properly. The check list also verifies the settings of the micro-processor constants.
'A i u
DAILY CEECKLIST ~l y.. x ,
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- :fCHECKERS NUMB"R/ i
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/ 7 ERATORi '
DATE- !;
llN),R.0; .
. .9 - -
l FRELIMINARY PROCEDURES {
r C;A.W Eon'and Operable: "
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7
. ccckground - (mR/Hri : ; VAMP--
iAlcrm1at?2'mR/Hr
}
LCoolinglsyst em. .on :- '
Tcnkiwater. level-
- Waters temperature - 'C ~
-Conductivity--uimhos. Inlet' Outlet
- Vicual check-of reactor'
- Alhrods-down -;
-Doors' closed 4 O $csion. product? monitor
[.sj low-reading-Water GM' alarm set- :
iPower on'all chassis
>:lRocorder-power and paper.
drive on;.. pens down . -,
DOWN ROD POSITION. Shim Reg. PREVIOUS COWN ROD POSITION Shim Rec.
1 VERIFICATION OF CONSTANTS: 4 21.=-0. -
40--= 3.74E-07 +
'25 = 1.87F-07 41 '= 1.00S+02 29?= 1.20E+06- 42 ='1.00E+03 .
_ -31}=^-6.50E402: 43---3.00E+00 33 = 3.70E 51-= 0, off
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DAILY CHECKLIST - CONTINUED I
CALIBRATION.
Average Burr- Average input Signal Setting -5% Brown Linear Log +5%
Counter Low 2.23E-05 2.50E-05 lF5F81E Counter High 5.56E-02 6.14E-02 lFSF83E Cambelling Low 1.03E+01 1.14E+01 lF5F84E Cambelling High 0.98E+02 1.08E+02 lFSF85E ,
FSF80E Reset
% Power (UIC) Zero Chamber and instrument sensitivity:
SOURCE IN SOURCE OUT today previous today previous linear x 0.2mW x 0.2mW x 0.002md x 0.002mW log x E-6% x E-6% x E-8% x E-8%
,m A Control Rod is raised from down limit and scrammed by the methods i 'isted below:
Q ,/
Safety = 1 Shim = 2 Reg. =3 High Power Level Low level A2 light on A2 light on 15 = H . 15 = L control rod Watch Dog A3CTX Period green light off A2 light on yellow light on 15 = R 60 = cxfail (F7 90) 60 - 69 = empty High Voltage Manual Scram A2 light on _
Uncompensated Ion Chamber 15 = V- High voltage control rod % Power mj !
Control Rod Interlocks Fig 6.2 6-6
SHUTDOWN CHECKLIS2 NUMBER DATE All rods down
. ,m l 7ter temperature - *C ,
/
, Vission product mc.nitor low reading
- Visual ~ int.pection of reactor Area monitor off ROcorder power and paper urive off; pens up Conductivity--u mhos Inlet Outlet. __
Number of samples irradiated today _
Number of samples irradiated this year Total hours of operation __
To.:a1 power generated up to date _
KW*hr Power generated today KW'hr Total power generation KW*hr Signed
-[
AREA SURVEY - VICTOREEN 450P
'U BEFORE REACTOR AFTER REACTOR OPERATION OPERATION
- GRID uR/Hr uR/Hr SAMPLE HOLE uR/Hr uR/Hr INDEX uR/Hr uR/Hr 3 Ft. HIGH @ im. FROM CENTER CHANNEL uR/Hr uR\Hr ,
J P.T. RETRIEVER AREA uR/Hr uR/Hr l l
hours after operation l
TIME OF FLIGHT OF ROD DROP ROD' UP INDEX RISETIME DROP TIME DROP TIME AVERAGE I SEC. #1 SEC. #2 SEC. SEC.
SAFETY l
SHIM l
REG.
Fig 6.2.
6-7.
1' l
(V Chapter 6 ,
- References
- 1. " Selection and Training of Personnel for Research Reactors", ANS!/ANS - 15.4 - 1988.
i 6
i 5
6 6
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6-7
y CHAPTER 7 y )
RADIOACTIVE MATERIALS AND RADIATION MEASUREMENT Radioactive materials and radiation control within the Reactor Laboratory will be subject to industry standards (1,2), license conditions, and 10 CFR 20.1001 - 20.2401 and appendices.
7.1 RADIOACTIVE MATERIAL CONTROL Physical control of radioactive materials shall be provided as an essential part of the radiological safety program. Control shall include identification of items or storage in identified locations. Controls such as shielding, isolation, containment and ventilation will be provided, as necessary, to control radiation exposure to the inventory of radioactive materials. Since most of the liquid waste from our neutron activation procedures involve short half-lived isotopes the waste is stored until it reaches background. If release into sanitary sewerage is inc';cated it will be done so as to comply with 10 CFR 20.2003 and the values recorded in the Omaha VA Medical Center sewerage disposal records. If essay of the waste is required it will be done with calibrated radiation survey meters or a well Ge detector coupled to a murtichannel analyzer.
7.1.1 Reactor Fuel irradiated reactor fuel shall be maintained in the reactor core, reactor pool storage (n) v racks, or the three emergency storage pits located immediately adjacent to the reactor tank as described in paragraph 3.2.1.
7.1.2 Reactor Con.ponents Each reactor component removed from the reactor pool shall be measured for activation levels and removable contamination. All components remaining in the pool shall be assumed to be radioactive. Components removed from the pool will be cleaned or covered as necessary to control radioactive contamination. Components that contain radioactive material will be labeled and stored in the isotope storage cell as shown in Fig 3.2 (SW 2F) 7.1.3 Isotope Storage Cell The isotope storage cell, SW 2F (Fig. 3.2), is underground and adjacent to the reactor laboratory. The storage cell contains 10 holes in the floor 20 feet deep and 64 holes in the wall 10 inches deep. Both types of holes have lead plugs. No part of the hospitalis directly above or adjacent to this storage.
7.1.4 Experiment Facilities Experiment facilities shall consist of the rotary specimen rack, vertical tubes, pneumatic transfer systems, central thimble and in-pool irradiation facilities. Removal of experiment
,, facilities from the pool or the beam originating from the reactor shall be subject to the same i/ controls as those for reactor components.
7-1
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'" 7.1.5 Activated Samples Materials that are inserted into the reactor experiment facilities or reactor beam shall be controlled as radioactive materials until disposed as radioactive waste, transferred to an authorized user, or decayed to releasable levels for non-radioactive materials. Any separation of radioisotopes that may t,a involved in our neutron activation procedures will be transferred to and come under the jurisdiction of the Omaha VA Medical Center Type A Broad Scope License.
7.1.6 Radioactive Waste Canisters shall be available and labeled for radioactive waste at locations wnere contamination from sample processing or other activities with contamination occur. Locaticns shall be designated for storage of solid wastes. Liquid and solid waste will be stored in the isotope storage cell in the reactor room until release criteria are determined such as decay, dilution, or processing. Specific sinks in the facility that are designated for radioactive materials shall be identified.
Gaseous wastes from experiments are vented in the two laboratory fume hoods by means of exhaust fans located on the roof of the hospital All proceduros invciving radioactive waste will follow the same criteria as described in
, the OVAMC Type A Broad Material License #26-00138-10.
i 7.1.7 Other Radioactive Material Radioactive reactor components, contaminated tools and fixtures and other radioactive materials shall be included in the radiation safety program as required on the basis of radiation or contamination levels. These materials shall be maintained in a rettricted area or be under i the control of authorized individuals. They may be released by authorized individuals for unrestricted use upon decontamination using the criterion of Table 7-1 or based on background equivalent induced gamma activity measuremsnts. Altematively, they may be disposed of as radioactive waste. Background equivalent gamma activity means an unshielded gamma measurement one meter from any surface net exceeding 5 mrem /h above ambient background. ;
Ambient background is define as follows: " Radiation from cosmic sources; naturally occurring l radioactive material including radon (except as a decay product of source or special nuclear material) and global fallout as it exists in the environment from the testing of nuclear explosive devices. Background radiation does not include radiation from sources controlled or regulated by the overseeing regulatory authority". l 7.2 Radiation Monitoring Radiation monitoring shall consist of fixed, portable, or sampling type systems. I I
Monitoring systems will be applied to measurement of radiation areas cnd high radiation areas around the reactor facility, significant contamination within and adjacent to the facility, and i l
radioactive materials and their concentrations in effluents. Monitoring shall be considered for
(,,h routine operations, abnormal conditions, and emergency situations. ,
) i v
l 7-2 l
L l
O TABLE 7.I'
( Acceptable Surface Contamination Levels i
for Uriconditional Release Removable'd Nuclide' Average fixed'd Maximum fi.ved'"
1 2 2 U nat,2"U, and 5,000 dpm /100 cm 15,000 dpm /100 1,000 dpm /100 cm
[1700 Bq m4 ]
i associated decay products - [8000 Bq m4 ) cm2 [25000 Bq m4 )-
2 2 2 Tranuranics 22'Ra,22*Ra 100 dpm/100 cm 300 dpm/100 cm 20 dpm/100 cm 4 4 2g,22 Th,23ip,,22:Ac, [170 By m4) [500 Bq m ] [30 Bq m )
- l, "'I 2
Th-nat, 2"Th, "Sr,222Ra, 1,000 dpm/100 cm 2
3,000 dpm/100 cm 2 200 dpm/100 cm 22'Ra,2nU, "'I, "'1, "'I [1700 Bq m4 ]- [5000 Bq m4) [300 Bq m4 ]
2 l - Deta-gamma emitters 5,000 dpm -/100 15,000 dpm - /100 1,000 dpm - /100 cm (nuclides with decay modes em:[8000 Bq m4 ] cm 2 other than alpha emission or spontaneous fission) except [25000 Bq m4 ] [1700 Bq m4 ]
Sr-90 and others noted above.
. ;t
- Where surface contamination by both alpha and beta-gamma-emitting nuclides exists, the limits v established for alpha and beta-gamma-emitting nuclides should apply independently.
6
.As used in this table, dpm (disintegrations per minute) means the rate of emission by radioactive material as determined by correcting the counts per minute observed by an appropriate detectcr for background, efficiency and geometric factors 79 ociated with the instrumentation.
2
' Measurements of average contaminant should not be averaged over more than 1 m . For objects ofless surface area, the average should be derived for each such object.
2
- The maximum contamination level applies to an area of not more than 100 cm .
?The amount of removable radioactive material per 100 cm2 of surface area should be determined by wiping that a:ra with dry filter or soft absorbent paper, applying moderate pressure, and assessing the amount of radioactive material on the wipe with an appropriate instrument of known efficiency. When removable contamination of objects ofless surface area is determined, the pertinent levels should be reduced proportionally and the ent:re surface should be wiped.
'The average and maximum radiation levels associated with surface contamination resulting from beta-gamma emitters should not exceed 0.2 mrad /h at I cm and 1.0 mrad /h at I cm, respectively, measured through not m' ore tl an 7 mg/cm2 of total absorber. l r
7-3
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%.)
7.2.1 Minimum Procedures Zone identification, access control, and protective equipment shall be designated. Zone identification for radioactive materials and radiation areas are das!gnated as specified by 10 CFR par' 20. Access control for zones shall be to control radiation exposures and physical security of the reactor facility and it's material as specified by 10 CFR parts 19 and 73, (Notices, Instructions, and Reports to Workers; Inspections and Physical Protection of Plants and Materials). Protective equipment for routine abnormal and emergency conditions shall include at least tape, plastic bags, absorbent paper, gluves shoe covers, covera!!s.
Contamination areas or areas that are routinely subject to contamination shall be marked clearly and control points established to monitor for contamination of personnel or -
equipment that leave the designated area. Measurements shall prov;de action levels for removable activities of 500 disintegrations per minute. Periodic monitoring of areas in which contamination is probable shall be of an adequate frequency to reveal significart changes in contamination levels. Decontamination of personnel, equipment, and surfaces shall be appropriate to requirements for control of radiation exposure and control of radioactive material containment. Release of radioactive components, etc. for unrectricted use may only be done by the Reactor Manager or a delegated Senior Redor Operator as referenced in the facility ALARA Program which is approved by the Reactor Safeguards Committee.
Airborne radioactive monitoring shall consist of continuous sampling of air particulate s
4 activity in the reactor area. Warning levels and action levels will be determined relative to V allowable DAC (Derived Air Concentration). Measurements should be sensitive to one DAC change in one hour Monitoring will occur during reactor operation or activitiea involving fuel, core, or expenmental facilities, and will provide measurements for routine, abnormal, and emergency conditions. Additional airborne monitoring equioment will be provided for special experiment needs Personnel dosimetry shall be required for access to reactor areas and some other facility activities. Monitoring devices will typically be film badges with pocket dosimeters for supplemental measurements. Other personnel monitoring, such as bioassays wir be applied as determined by the activity and conditions or radiation exposure situations. Pert,onnel shall use supplemental dosimetry during activities that deviate substantially from routine operations with supplemental dosimetry also provided for persons visiting areas with potential radiation exposure.
7.2.2 - Monitoring Techniques Implementation of radiation monitoring to maintain the goal of"As Low As Reasonably Achievable" should consist of: (a) preoperation planning, (b) operations techniques, and (c) post operation analysis.
\O) 7-4 s - _ . - - .
l l
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\ j 7.2.3 Management Surveillance ,
1 A review by the Reactor Safeguards Committee and the Medical Center Radioisctope and Radiation Safety Committee of radiation exposures related to operations that cause significant radiation exposuras compared to routine operations will be performed. The review will be 9pplied to determine whether facility modifications or procedures should be implement to maintain radiation exposures "As Low as Reasonably Achievable".
Befare a new experiment can be performed in the reactor a " Request for Neutron Activation" form must be completed by the requester. This form lists the phase of the sample, the requested irradiation time, the concentration of all elements which are or may be present in the sample, whether the sample contains compounds which are highly reactive with water, potentially expiosive, or fiss5nable material. The form also asks if the sample contains uranium, thorium, plutonium, horon, lithium or any other known high neutron absorbing element and how the sample will be encapsulated. This completed form together with an experimental protocol N submitted to the Reactor Supervisor. The Recctor Supervisor then completes a separate
" Check List for Reactor Experiment Approval" form making an independent analysis of the safety of the procedure and certifying that the experiment is in compliance with the Omaha VA Hospital Technical Specifications. The experiment is then submitted to the Reactor Safeguards Committee for Approval. After approval the experiment is rehearsed with the investigator.
During the first procedure radiation levels will be measured, observations of the sample before and after irradiation wili ce noted, if applicable, and an analysis made of the results. Each time an experiment or irradiation is to be performed it must be approved on the Daily Experiment g]
Checklist by the SRO.
v;
(
7.3 INSTRUMENTATION Instrumentation for the evaluation of radiation exposures from routine, abnormal and emergency situations shall consist of fixed area monitors, portable survey monitors, and appropriate sampling methode. The minimum instrumentation available during reactor operation shall consist of fixed area gamma dose rate monitors, continuous air particulate monitor, portable thin window GM tube survey meter and pocket dosimeters with charger. Othe' detecting eciuipment that is available includes an alpha-beta proportional counter, a multichannel gamma pulse height analyzer with Ge detector, a liquid scintillation detector, Ic'u range beta-gamma dose rate meters, a low energy icaization chamber type meter, and GM tube or equivalent friskers.
7.3.1 Fixed Area Monitors Fixed area monitors have audible and visual alarms. One monitor is permanently mounted approximately 1.4 meters from the isotope removal port and the other is near the pneumatic tube.
7.3.2 Airbome Radioactivity Monitors A continuous air particulate fixed filter monitor with audible and visual alarm shall be
/o\ functiona! in the reactor vicinity during reactor operation. A thin window GM detector will also d
7-5
l
,a monitor the activit; and provide alert and alarm conditions with an audible enunciator. Count rate of the instrument includes the range of 50 to 50,000 counts / minute The continuous air monitor also is provided with a charcoal filter for monitoring the presence of lodine radioactivity.
7.3.3 Survey and Laboratory instruments Portable survey monitors for alpha, beta, or gamma radiation shall be maintained for area surveys of laboratory and experiment areas. Survey instruments will ccnsist of the following instruments or equivalents: (1) a pancake style GM or low energy scintillation detector and (2) ionization chamber for radiation fields of 0 to 50 R/hr.
Supplemental measurements can be made with an alpha beta proportional counter, or a gamma ray pulse height analyzer. A liquid scintillation counter is also available in the adjacent research building.
7.3.4 Liquid Effluents The reactor generates no radioactive offluents. Radioactive liquid waste generated in the research program are govemed by the requirements of the OVAMC NRC Type A Broad Materials License # 26-00138-10. Reactor coolant water may be monitored for radioactivity as a supplementalindicator of water activity.
p/
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' 7.3.5 Calibrations Calibration methods, accuracy, frequency and fundamental checks are established for radiation monitors following the procedures specified in Regulatory Guide 10.8 " Guide for the Preparation of Applications for Medical Use Programs" and reactor operation specifications.
7.3.6 Records Records are specified for maintenance of radiological data that relate to reactor operation. These records shallinclude:
- a. Personnel dosimetry including bioassays or other special measurements made.
- b. Radiological control surveys required by facility specifications.
- c. Gaseous and liquid radioactive effluents released to the environment.
- d. Radiation Surveys.
- e. Instrument calibration rea-ds.
- f. Radioactive material rempt and transfer records.
- g. Solid radioactive waste disposal records.
- i. Data on radiologicalincidents.
7.4 EVALUATION OF MONITORING SYSTEMS 7-6
) The radiation monitors provide information to operating personnel c%nding or existing hazards from radiation so that there will be sufficient time to take ty messary steps to control the exposure of personnel and the release of radioactivity or to svacuate the facility.
Two types of radiation monitors are used: a continuous air particulate monitor with an attached charcoal filter for determ!ning radiation levels due to particulate radioisotopes suspended in the reactor room air and for monitoring the presence of iodine radioactivity, and area monitors for determining the gamma field at several locations in the facility. While CAMS are designed to be air particulate monitors, both will detect *Ar and other non-particulate airborne radioactive materials. The efficiency of the NMC CAM has been calculated to be 4x10-8 %. The efficiency J the Eberline ANS 3A Cam has been calculated to be 1.23x104 % and consequently a count rate 8
- build up of 180 cpm /hr is equivalent to an air concentration of 2.28x104 pCl/cm .
Since the DAC limit for *Ar is 3x104pCl/cm3, both CAMS can detect the DAC limit of 1
, *Ar Realisticat'/, we have not seen a rise of 180 cpm /hr in either CAM, which is consistent with our calculated argot emission as shown in Appendix A. Since it is not feasible to monitor such effluents from a 20 kW reactor in real time at the point of release, calculated releases will be 4
substituted. A count rate build up of 1CJ cpm /hr is equivalent to an air concentration of 14x10 pCi/cm'.
Each type of radiation monitor has a specific radiological purpose. The particulate air
- monitor !s used to detect radioisotopes released due to fuel element failure (a design basis accident), while the area radiation monitors are used to minimize personnel radiation exposures. The radiation monitors described below are typicalinstruments possessed at the
,( time this application was written. Replacements may have slightly different characteristics but 4 i will be at least equivalent.
7.4.1 Darticulate Air Monitor
- Set points for the particulate continuous air monitor warn of the presence of particulate fission product nuclides. Since gaseous and volatile elements such as krypton, xenon, bromine urid Mna hae particulato decay products, the presence of some or their radioisotopes should also be detected. The attached charcoal filter can also be monitored manually to detect 3
gaseous elements absorbed on the charcoal. An alarm set point at 6x10'S pCi/cm detects '
particulate activity concentrations below the occupational DAC values of Appendix B of 10 CFR 20.100120.2401 for the relevant isotopes in the ranges 84105 and 129149. These ranges of isotopes represent the one perr ent yield for fission products of uranium 235. Significant fission products as a percent of total reiease are shown in Table 7-2.
The air monitor in use is a Nuclear Measurements Corp. Model AM 2d gross beta-gamma air monitor configured for continuous sampling of airborne beta-emitters on a fixe,d filter. It uses an end-on coplanar GM type detector with a window diameter of 17/8" and a thickness of 5.6 mg/cm'.
Both detectors are calibrated with a "Tc standard of 0.005 pCi that has a 0.29 MeV beta. Since the monitors use a thin endwindow GM tube, a calibrated source of an E, greater than 0.29 MeV would have a greater off.::iency and consequently result in a larger epm /cm'-hr.
Using the efficiency determined from the "Tc standard the calibration of the CAMS are as follows:
7-7
. - . _ - _ - = .. - . - . - -.. - . _. . - - . _ ~ . . . .- . ._
TABLE 7.2 Significant Fission Prodnets Contribution to Total Activity, Percent ,
, t Element Isotope i Day 10 Days 30 Days 90 Days 1 Year B energy (McV)
Strontium Sr" 2.8 6.7 10.5 2.7 1.49 Sr" 1.8 .546 YttriumY" 2.29 Y" 3.4 7.6 12.5 3.9 1.55 Strontium Sr" 6.7 1.1,1.38,2.68 :
YttriumY" 4.2 3.62 Y" 7.6 2.29 Zirconium . Zr" 3.7 8.2 14.7 7.3 .366,.398,.888 Niobium Nb" 4.1 18.2 15 .159 Zirconium Zr" 9.0 1.93 Niobium - Nb" 9.6 1.28 Molybdenum Mo" 4.6 6.8 1.23
!- Rhodium Rh " 2.55 5.5 7.0
- Ruthenium Ru'" 2.65 5.7 7.2- .225 Rhodium Rh'" 1.35 .556,.25,.26
+
Rh'" 3.54 Ru'" 2.4 .0394 l}uthenium lodine 1 " '. 6.8 3.7 .606,.25,.81 !
I"' 2.7 5.3 .80 I"' 7.3 1.27
' Tellurium Te"' 2.6 5.1 .23 Xenon Xc 1.23 11.4 1.6 .346 lodine 1"' 4.7 1.0,.5,1.5 Xenon Xc"' 12.5 - .91 ;
' - Barium Ba"'" ,
Cesium Cs"' l.5 .512,1.173 ,
Barium " a'".
1.25 10.6 10.8 1.6 1.01,.47
, Lanthanum La'" 12.0 12.5 2.4 1.36,1.25,1.68 Cerium Cc" 6.3 11.2 8.5 .436 581 Lanthanum - La"' l.4 2.43 Praseodymium Pr"' 10.0 l1.2 1.9 .932 Cerium - Cc"' 61 1.09,.l.39 4
. Ce'" 2,0 6.0 26.5 .316,182 Praseodymium Pr"* 3.00 Neodymium Nd"' 4.8 .4.1 .804,.364 Promethium Pm'" i.45 1.072 4
- f 7-8 >
, e - -- , ,n n...-,. e -. , , , . - , , , . , - , . .,s., s.,.., - , , , - . , ,,,,,,,,.p.~,,.n,,,,. ~ - , , ,
i l
i l xj 8 NMC Recorder rise of 180 cpm /hr = lx10'* pCi/cm Ebertine Recorder rise of 185 cpm /hr = lx10-* pCl/cm' i
Since all of the radioisotopes that we may encounter at the facility have an inhalation DAC of less than lx10'* as listed in Table 1 Appendix B to 10 CFR 20.1001-20.2401, a rise of 180 or 185 cpm /hr would indicate that we had a possibility of exceeding the DAC limit. The above calibration values are for the present calibration and may vary from calibration to calibration.
Radionuclides would be identified by analyzing the filter with the calibrated Ge or Gell detectors. The Omaha VA Medical Center Reactor SOP lists the maximum weight of isotopes which could off-gas, sublime, or volatilize and produce an activity such that if 100% of the gaseous activity or radioactive aerosols escaped to the reactor room or the atmosphere, the airborne concentration averaged over a year would not exceed the limit of Appendix B of 10 CFR 20.100120.2401, Experiment approvalis dependent on not exceeding these weight limits.
The facility uses the Canberra ASAP Peak Search and Isotope Identification software together with the SAMPO 90 Interactive Gamma Spectrum Analgis package.
7.4.2 Area Radiation Monitors Several area radiation monitors which observe the gamma field are part of the permanent installation. Some locations are experiment areas in which shield configurations p determine the levels of radiation during reacto operation. Alarm set points for all a*ea radiation (vj monitors will be at either 2 mR/or or 5 mR/hr. The first number is obtained by dividing the maximum desired dose each week by the number of working hours each week. The second number is obtained from the definition of a radiation area in 10CFR20.
7.5. Radiation Hazards from Experiments or Fuel Because there can be Intense radiation fields from radioactive isotopes produced by this reactor, it is considered necessary that reactor aperations be supervised by individuals who are trained in the detection and evaluation of radiological hazards. It should be noted that these hazards do not differ from those encountered with any reactor operating at comparable power levels.
Calculations based on ""'Co indicate that the reactor is capable of producing an equilibrium concentration of radioisotopes of approximately '60 curies in the rotary specimen rack if the reactor is operated at 20 kW [3). The fuel-moderator elements are another source of radioactivity, and our calculations as shown in Appendix B, Table B-2, indicate that if the reactor is operated at 20 kW, the equilibrium activity associated with one of these elements is appnximately 12.2 curies at the time of shutdown.
The maximum production of 160 curies of radioisotopes is distributed in the 40 sample positions each containing two sample containers. The maximum amount of activity which can be withdrawn at one time is therefora approximately 2 curies. This constitutes an intense source
,m . of radiation; however, it is a source intensity that is routinely handled by competent, technically
( ) trained personnel. Becruse the isotope-Production capabilities of this reactor are so large,
'J 7-9
i shielded isotope-handling equipment will be used to reduce radiation dosage to reasonable
- levels in accordance with 10 CFR 20,1201 and ALARA. The material content of Experiments i will follow the guideline outlined in Section 2 of Regulatory Guide 2.2 and administrative !
vntrols will follow the guidelines specified in Section 3 of the same Regulatory Guide together [
with those in Regulatory Guide 2.4. Detailed specifications of Experiments are listed in the !
Technical Specifications.
The radioactivity hazards associated wkh fuel elements are of the same nature as those associated with isotope production. The calculated dose from a single fuel element after prolonged operation at 20 kW and at a distance of 6 ft is 200 r/hr at the time of shutdown [4].
It should be realized that the hazard associated with direct exposure of personnel to highly radioactive fuelis not unique with the TRIGA reactor. This hazard will be encountered to -
the same extent with any reactor operating at similar power levels for similar periods of time.
Because of the significant radiatio /el associated with the reactor fuel moderator !
elements, it will normally be necessary to k p the elements under water for shielding. If an element is to be removed from the reactor sr.ield tank, a conventional fuel-element transfer cask will be used to reduce the radiation level to within tolerable limits. ,
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( Chapter 7 References -
- 1. .' Radiological Protection at Research Reactor Facilities", ANSI /ANS 15.11.1993.
- 2. " Development of Technical Specifications for Experimenttiin Research Reach s",
Regulatory Guide 2.2, U.S. Atomic Energy Commission. Nov 1993. 1i
- 3. *Safegurds Analysis Report for TRIGA Reactors Using Aluminum-Clad Fuel *, GA 7860,- -l General Atomic. March 16,1967. ,
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- 4.
- Hazard Report for Torrey Pines TRIGA Reactor", GA 722 General Atomic. May 27, .
1959.
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CHAPTER 8 ACCIDENT ANALYSIS 8.1. HAZARDS ASSOCIATED WITH THE OPERATION OF THE REACTOR Certain potential hazards associated with the operation of the r tactor system have been studied and have been found to cause no senous environmental hazarl, nor any hazard to operating personnel. However, in the handling of all radioactive rnateria', because of its very nature, it is necessary to observe accepted safety precautions. In the n( rmal use of TRIGA it will be necessary to handle irradiated samples. and standard health-phyaics procedures will be followed. On the rare occasions when it is necessary to remove highly ra dioactive fuel elements from the water shield, special equipment and safety procedures will be utilized.
Specifically, the following problems were investigated: radioactive contamination of the shielding water after a fuel element leak, and the production of radioactiv+ gasses in the reactor area and the environment after a fuel element leak. The possibility of loss of shielding water is ateo discussed.
A.w 't Apotential accidents considered to be credible, the one with the greatest potential erw cu che environment and the unrestricted area outside of the OVAMC reactor facility is the loss of the cladding integrity of an irradiated fuel rod in air in the reactor laboratory.
, ('vp) This has been designated as th. design-basis accident (DBA) and, for the ourposes of classification, is referred to as the " fuel-handling accident".
A DBA is defined as a postulated accident with potential consequentat greater than those from any event that can be mechanistically postulated. We have evalu ated other possible accident sequences that originate in the intact reactor core and none pose a algnificant risk of cladding failure. However, it is possible that an operator, when removing a fu31 element from the core or relocating one previously removed following irradiation, could experience an accident that would break the integrity of the fuel cladding. If this cladding weta ruptureo then noble gases and halogen fission products could escape into the pool. It is assumed that the accident occurs but no attempt is made to describe or evaluate deterministically the mer.hanical details of the accident or the probabi'ity of its occurrence. Only the consequences are considered.
8.1.1. Radioactive Contamination of Shielding Water The hazards associated with a failure of the fuel-element cladding and connequent fission-product contamination of the water have been studied experimentally [1] and theoretically [ Appendix 8.1.1). The results show that in the improbable event of a cladding failure, the water activity level would be extremely low because of the very high water-corrosion resistance of zirconium hydride. The water may be decontaminated by using the deionizer, or it may be safely disposed of in the public sewer system. Manufacturing inspection and qual;ty control, however, ensure that the possibility of a cladding failure is extremely small.
,.m 8-1
L Direct measurements have been made at General Atomic of the rate of escape of [
fission products from a TRIGA fuel element in the event of a cladding failure [1]. In these ,
experiments, unciad samples of uranium zirconium hydride were exposed to neutrons la order to produce fission products in them, and the irradiated material was then exposed to t domineralized water, which simulated conditions in the reactor, in these experiments, the rates !
of solution of the typical fisalon-product isotopes Sr-89, Ba 140, and 1-131 in the water were l measured. It was found that 1-131 dissolved the most rapidly of the three and that it appeared in the cooling water at a rate corresponding to the solution of 100 mg of uranium-zirconium i t
hydride per day per square centimeter of fuel element exposed to the water. This quantity is used in the following calculation of the amount of l 131 that would appear in the tank water in the event of a cladding failure.
After the reactor has been running at an average power of 20 kW for sufficient time to !
produce saturation of the 1 131, the core will contain approximately 400 curies of I 131. The [
solution rate of 100 mg indicates that if a cladding failure exposes 10 cm of !
uranlum zirconium-hydride fuel surface, one part in 10' of the contained 1 131 will escape into i the reactor cooling water in 24 hr. The total amount of l 131 escaping into the cooling water 4
would then be approximately 4 pCl, yielding a concentration of approximately 2.6 x 10 pCi/ml in the 15,000 liters of cooling water. The measurements made on the escape of Sr-89 and Ba-140 from the fuel-element material show that these isotopes will dissolve more slowly in the cool og water than will 1 131, by a factor of 10 to 100. No fission products would be hberated more rapidly than 1131.
The total water activity due to lodines, xenons and kryptons as a result of a cladding ;
failure h:o been calculated by Generel Atomic for a reactor operating at 250 kW [1].
Converting these calculation to 20 kW operation, the results show that in an improbable event >
of a cladding failure, the water activity may reach a maximum level of 1.4 pc/cm'. Independent calculations in Appendix B show that the following would result from a cladding failure of a fuel l element within the reactor tank: !
- 1) Activity in reactor water 2.1x10d mci /cm3 (eq.1, Sec. B.1.1) ,
- 2) Activity in air c 7.38x104mci /cm 3 (eq. 2, Sec. B.1.1)
- 3) Dose rate to individual at the center of a hemisphere with a hemispherical volume of 7.075x10' cm'.
8.5x10i mR/hr (eq. 3, Sec. B.1.1) 4)- Internal exposure from breathing fission cloud for 2 minutes :
2.65x104 rads (eg 5, Sec. B.1.2) it is concluded that such a cladding failure could be detected by monitoring the shield water and that the resulting contaminated water could be readily demineralized 'or disposed of without hazard to operating personnel or to the neighboring community.
a 82
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Th; activity of contaminants other than fission products in the shield water is kept at low l levels by equipping the reactor with a demineralizing system. The demineralizer is in a pit 6' 10" below ground adjacent to the hospital basement (see Fig. 3 2, area adjacent to the exhaust l fan). Consequently, if the fission products were collected in the dernineralizer they would be shielded from the public and ALARA principles would be applied to limit exposure. In order to determine whether the demineralizer can adequately control activity in the system due to a minor fuel-element defect, a calculation was made by General Atomic [18) in which it was assumed that 10 cm' of fuel surface were exposed to water. Experimental work at General '
Atomic [18, pg 117] showed that the corrosion rate of the zirconium-hydride mixture in water is about 2.3 mg/cm'-yr. For the purpose of the calculation it was assumed that the exposed fuel
. surface corroded uniformly at the above rate and that only those fission products contained in the corroded material were released to the water. Fission products were assumed to be uniformly distributed throughout the fuel-bearing region. The reactor was assumed to have been operating for infinite time at a power level of 10 kW. The demineralizer and filter were 3 assumed to be 100% effective for all nonvolatile fission products, i.e., they removed all non-volatile fission products from 12 gal of water per minute.
In calculating the activity due to nonvolatile fission products, this activity was taken to be t the total fission-product activity. To account for the fact that many fission products are very short lived and would decay soon after entering the water, the total fission product activity was
< assumed to be that which would exist 5 hrs after shutdown, as given by the Wigner-Way equation [3); 5 hr is the effective " half life" of the demineralizer (i.e., the time it would take the demineralizer to reduce the nonvolatile concentration by a f actor of two if no impurities were being added to the coolant).
The result of this calculation indicate that the demineralizer can limit the specific activity of nonvolatile fission products in the coolant resulting from 10 cm' of exposed fuel surface to an equilibrium level of 1.6x107 mci /cm' of water. This activity levelis below most of the effluent concentration levels listed in Table 2 of Append:x B to 10CFR20.100120.2401 8 1.2 Loss of Shielding Water Because there are many floors in the Hospital building immediately above the reactor
- that are normally occupied, the possibility of loss of shielding water has been considered. This loss of water can occur by only twn means
- (1) the tank may be pumped dry, or (2) a tank failure may allow the water to drain into the soil. !
. The tank outlet water line extends only 3 ft below the normal water level. Therefore, even if the water system is operated carelessly if, for example, it is operated when the pump discharge line has been disconnected for repairs the tank cannot be accidentally pumped dry.
This can only be done by deliberate action on the part of the operating crew. In the unlikely--
event that it is necessary to drain the tank for repairs, the fuel will first be removed in shielded casks. Since the recirculating pump does not have sufficient suction head to drain the tank, another more powerful pump must be installed with its suction line inlet below the core.
J 8-3
Tank failure cou:d possibly be caused by severe earthquake or major settling of tha building foundation. The tank has been designed by Holmes and Narver, a firm experienced in the design cf sarthquake-proof structures, and there is no record of a structure of this type sustaining damage through earthquake. No earthquake damage has been reported in the Omaha area. The tank has been designed to withstand the existing bearing loads from the building foundation. As described in Section 3.2.1., " Reactor Pit", the reactor tank has been carefully installed so as not to disturb the soit under this foundation. The building has been in exir.tence for 41 years with no evidence of foundation failure. There are five barriers which will p, event water leakage from the tank. Two of these barriers are waterproof the epoxy resin coating and the welded steel tank. The other three barriers would present a very high resistance to water leakage. The gunite, the reinforced concrete, and the adjacent boilitself.
The core drilling made at the reactor location shows the soil to be clayey sitt and glacia clay, both of which are essentially impervious (3).
Even though the possibility of a loss of shielding water is believed to be exceedingly remote, a calculation has been performed to evaluate the radiological hazard associated with this type of accident. If the reactor had been operating for 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br /> at 20 kW before losing all of the shielding water instantaneously, the integrated dose that an individual would receive in the first-floor radioisotope laboratory immediately above the reactor would be about 23.7 R in the first hour as calculated in Appendix C. The radiation from the unshielded core would be highly collimated, so that if an individual did not expose himself directly to the core, he could work in the immediate vicinity of the tank for several hours. He would fill the tank with water from a fire hose and view the interior of the tank with ; mirror while making the necessary emergency repairs.
(
To ensure that this accident wot!d not go unnoticed, a float switch is :nstalled in the reactor tank to actuate an audible alarm located at the switchboard. The operator will then notify the Reactor Supervisor or designated citernates in the Hospital building, who will take immediate remedial action.
Because the water is required for adequate neutron moderation its removal would terminate any significant neutron chain reaction. However, the residual radioactivity would continue to deposit heat energy within the fuel. Calculations have been done by General Atomic [4] to determine the maximum fuel temperature rise re';ulting from a loss of coolant after operation for an infinite time at 250 kW. Results indicate that if the water loss in the core occurs immediately after the reactor has been shut down, the maximum temperature of the fuel, and consequently the aluminum cladding, is less than 150'C. This temperature is such that the pressure exerted by the trapped air and fission product gases is less than 30 psi. This pressure produces a stress of about 660 psi, whereas, the yield stress for the aluminum cladding is greater than 5000 psi at 150'C. Consequently, it is concluded that, subsequent to loss of cooling water after infinite operation at 20 kW, the release of hydrogen from the fuel and the expansion of air and fission gasses in the space between fuel and graphite end pieces will not result in the rupture of the fuel element cladding. We have analyzed the calculations performed by General Atomic and concur with their conclusion.
- A loss of cooling accident was also analyzed for the Reed College TRIGA reactor, a
()
V typical Mark I model fueled with aluminum clad elements using ZrH' fuel [5). The postulated 8-4
T C loss-of cooling accident showed that the maximum fuel temperature would be less than 150'C after the infinite operation at 250 kW was terminated by the instantaneous loss of water. At this temperature the equilibrium pressure from fission gases, entrapped air and dissociated hydrogen was reported to produce a stress of only 660 psi which is well below the yield stress of greater than 5000 psi for aluminum cladding at 150*C.
Consequently, it is concluded that afterheat in this reactor following a water-loss accident would be such that the system temperatures would be far below that required to melt tne aluminum fuel-element cladding. Therefore, no dispersal of fission products would take place.
It is a!ao concluded that the possibility of loss of shielding water is extremely remote, that the consequences would be unlikely to cause severe injury to personnel or damage to the reactor, and, therefore, that this type of accident does not present a significant hazard to the public.
8.1.3 Handling Irradiated Fuel- Design Basis Accident f
A fuel handling accident involving the breaching of the cladding of a TRIGA fuel element while it was out of the reactor tank was thoroughly analyzed in NUREG 2387 [6]. We have analyzed their calculations and assumptions and adapted their calculations to the power and 2
operating history of the Omaha VAMC TRIGA Reactor (Ave. neutron flux = 4.8x10'" n/cm /sec; operation at 20 kW for 0.26 mwd). Our analysis can be found in Appendix B, section B.2. In this scenario, which has potential consequences greater than those from any event that can be
\ mechanistically postulated, it was assumed that during the unloading of a fuel element it is hit by a shipping cask while the element is out of the core. The blow is assumed to have created a large cladding rupture and severe physical damage to the fuel. Table B2 of Appendix B shows the calculated noble gas and radiciodine in the TRIGA element containing the greatest activity, following operation at 0.26 mwd, at shut-down and 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after shut-down. As can be seen from the table, at 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after shut down the major activities are f;om *Xe. "'I and *l.
Applying the release fraction of gaseous activity of 1.5x10-5 as used in Foushee and Peters [7], to the values in Table B2 the activity within the reactor room would then be:
a) At shut-down = 1.84x104 Ci b) 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after shut-down = 3.73x10'5 Ci Wholebody and intemal doses have been assessed. Potential offsite doses have been analyzed for the Most Exposed Member of the Public (MMP) and the Nearest Permanent Residence (NPR) and are summerized in Sec. B.2.
In addition to the gaseous radionuclides, other fission products will also be present, with the radiostrontiums and radiocesiums being the mesi significant from a hazards standpoint.
Table B3 of Appendix B shows the quantity of significant radiostrontiums and radiocesiums containing the greatest activity at zero and 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after shutdown following operation at 0.26 mwd.
O 8-5
I O l Table B4 in Appendix 0 gives air concentrations and Derived Air Concentrations (DAC) of the radiostrontiums and radiocesiums following the hypothetical maximum credible fuel-handling accident. DAC values are listed in Table 1 of Appendix B to 10 CFR 20.100120.2401.
These 10 CFR 20 concentrations are equivalent to radionuclide concentrations which, if inhaled or ingested continuously over a course of a year, would produce a total effective dose equivalent to 0.05 rem. As can be seen, the air concentration values from the hypothetical accident are all less than the 10 CFR 20 values.
8.2. HAZARDS NOT ASSOCIATED WITH THE OPERATION OF THE REACTOR 8.2.1 Mechanical Damage to the Reactor ;
it is conceivable that a heavy weight, su::h as a lead transfer cask, could be dropped on the reactor core from above and could smash the core in such a way as to change the fuel-to-water ratio. The designed fuel-to water ratio in tne core was selected because this ratio was calculated to give very nearly the minimum critical mass. Consequently, smashing the core is likely to decrease the reactivity, and at worst cannot increase it appreciably. Mechanical damage to the reactor core could cause a fuel-clad failure within the reactor tank and consequently a release of fission products into the water. This type of accident has been analyzed in Section 8.1.1 and Appendix B of the SAR.
8.2.2 Failure of Electric Power
( The reactor control system is fail-safe in the event of power failure; i.e., loss of power will de-energize the magnets and release the control rods.
8.2.3 Fire The Hospital building is constructed almost entirely of fireproof materiais. The load-bearing walls, the ceilings, and the floors are of reinforced concrete. Carbon dioxide fire extinguishers are located in all halls and laboratories. The reactor room also is equipped with a sprinkler system that is dry until it is charged by heat sensors. The system has heat sensors in the ceiling which activate at 135'C to fill the system and activate an audible alarm. Thr.
sprinkler heads do not activate until the temperature reaches 165'C. The sprinkler system is independent of the hospital system and can be tumed off in the room designated SW1 A on Fig 3.2, if activated, the system will release about 50 gal / min. Aid from the Omaha Fire Department is available in less than ten minutes. The sprinkler was installed because of the requirements of the Veterans Administration that the entire hospital be have automatic sprinklers. The VA would not allow a dry sprinkler to be installed due to the possibility of Toxic vapors. If the sprinkler system was activated it could possibly flood the reactor tank causing the water to overflow, in that case all of the water, the area, and the outer clothing of allindividuals invoived in the fire would have to be monitored for radioactivity. The Fire Department is briefed yearly on this possibility and is instructed not to leave the area until they are monitored. Through their HAZMAT training they are familiar with radioactivity. The radioactivity of the water, if measurable, would depend on when and for how long the reactor was operated. Six ml
[ j samples of reactor water taken at shut-down and analyzed for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> in a 70 cc well type Ge V
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r'N l' detector have not shown any significant gamma peaks. Because of the proximity of the fire (V) extinguisher, most fires would be extinguished before the sprinklers are activated.
Consequently, the possibility of fire will not contribute significantly to any radiological hazard.
B.3 Conclusions in accordance with the discussion and analysis above we conclude that even in the case Of the design basis accident, the radiation doses to both occupational personnel and to the public in unrestricted areas would be far below the guidelines and limits of 10 CFR 20.1001-20.2401. This fact was also stated in the previous Safcty Evaluation Report related to the renewal of the operating license for the research reactor at the Omaha Veterans Administration '
Medical Center, NUREG-0988, dated July 1983.
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Chapter 8 References I i
, 1
- 1. " Technical Foundations of TRIGA", GA 471 (1956), pg 114-115.
- 2. Same as 1 above. pg 113-114 l
. 3. Way, K. and Wigner, E.P.,
- Rate of Decay of Fission Products", Phy. Rev. ,' 73:1318, . :
1948. !
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- j. 4. Batch J.M. and Foushee, F.C.,'" Safeguard Analy.,is Report for TRIGA Reactors using
' Aluminum Clad Fuel", GA 7860, General Atomic,1967. '
l
. 5;-
- Safety Analysis Report',1967, Reed College, Portland, Oregon,1967.
- 6. Hawley, S.S. and Kathren, R.L.,
Fueled Reactors, NUREG/PR-2387, PNL-4028, April 1982. !
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- 7. Foushee, F.C. and Peters, R.H.," Summary of TRIGA Fuel Fission Product Release '
Experiments", Gulf Energy and Environmental Systems, Gulf EES-A10801,1971.
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77 CHAPTER 9
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OCCUPATIONAL RADIATION EXPOSURES 9.1 PERSONNEL MONITORING PROGRAM The OVAMC reactor facility personnel exposures are measured by the use of film badges asslaned to individuals who might be exposed to radiation. TLD neutron dosimeters capable of detecting neutrons with an energy of 0.06 ev-4.4 MeV are issued to the two reactor operators. In addition, self reading pocket ion chambers are used. Instrument dose rate and time measurements are used to ensure that administrative occupational exposure limits are not exceeded. These limits are in conformance with the limits specified in 10 CFR 20.1001-20.2401 and Amendments.
9.1.1 Personnel Exposures The OVAMC reactor facility personnel annual exposure history for the last few years is given in Table 91.
9.2 EFFLUENT MONITORING 9.2.1 Airborne Effluents n
As discussed in Chapter 3, radioactive airborne effluents from the reactor facility consist (V) principally of activated gases. The airbome radioactivity is monitored to provide prompt indication of any abnormal concentrations being discharged to the environment. This is accomplished by withdrawing a representative stream from a point near the top of the reactor through a continuous air monitor. This monitor also is provided with a charcoal filter for monitoring the presence of iodine radioactivity. The output of the monitor is indicated on a meter having adjustable alarm set points, and a continuous record also is provided. The characteristics of the monitor are described in Section 7.4.1. As shown in Appendix B and Section 3.2.11.2.1.3, the total"Ar vented to the reactor room and thc9 ;o the environs represents an exposure of less than 1 mR/ year and a yearly concentration of less than 1 % of the DAC for occupational exposure or effluent concentration as presented in Tables 1 and 2 of Appendix B to 10 CFR 20.1001-2401.
9.2.2 Liquid Effluents The reactor generates no radioactive liquid effluents. Radioactive liquid waste generated in the research program is discussed in paragraph 7.1,6.
9.3 ENVIRONMENTAL MONITORING Radioactive gas is the only potentially adioactive material released to the environment as a result of the routine operation of the OVAMC reactor. The routine gaseous effluent
,_ measurements consist of those recorded by the continuous air monitor, and the monthly
,' exposure date obtained from film badges located within the reactor room, at the exhaust port 9-1
output, and at the water treatment pit output. The latter represents the airborne exhaust to the environment because the reactor room air is discharged through the water treatment pit. The net integrated exposure at the water treatment pit output for 1992 was 21 mrad. The positiun of the pit output monitor was under the 1/32" corrugated steel roof of the pit,8'7" from the mixed bed resin tank. Considering that the natural background levels in the Omaha area are about 80 ,
mrem per year causing the film badge monitor to detect some background radiation together l with some radiation from the resin tank, the exposure levelis consistent with less than 1 mR/yr i calculation mentioned in Section 9.2.1. l l
i - A continuous air sampler was operated by the State of_ Nebraska Health Department's Division of Radiological Health on the roof of the Omaha Douglas County Hospital (300 m from ,
the OVAMC) for a number of years, primarily as a weapons testing fallout monitor. At no time .
- was any activity detected ths.t could be attributed to the OVAMC reactor operation. This !
monitoring program was discontinued because there was no further need for the program. {
9.3.1 Potential Dose Assessments Recently the facility has placed a Landauer " Low Level Environmental" monitor outside
, of the hospital, and for the third quarier of 1994 obtained a net reading of 23.8 mrem. This would be consistent with the 80 mrem / year background level stated above.
Based on the calculations in this SAR it is felt the release rates from this facility fall well 1 within the requirements of 10 CFR 20.1001-2401 and comply with ALARA concepts.
L ,
i i
i L
3 O
9-2
~ . . . _
- - - . - . . - - - . - - . - . - - - _ . - - - . - - - - . - . . ~ . - . - . - - .
i !
I l,
Table 9-1 ,
4 i
Recent Exposure History of Reactor Facility Personnel ,
t i i
I 1
Whole body exposure mrem /vepr j l
Individual 1998 1989 1990 1991 1992 1993 1 20 30 10 40 515 515 I
t J 2 515 515 30 515 515 t
- i
.c d
+
4 l
a 4
1 J'
t i
i i
i I
t t
i
, s 9-3
f APPENDlXA Release Of Argon Fram Facility During Normal Operation A.1 Release of Aroon-41 from ReactorWater The argon-41 activity in the reactor pool water results from irradiation of the air dissolved in the water.
, The following calculations were performed to evaluate the rate of argon-41 escaping from the reactor pool water into the reactor room. The calculations show that the argon-41 decays while in the water, and most of the radiation is safely absorbed in the water. The changes in argon-41 concentrationin the reactor,in the pool water extemalto the reactor, and in the air of the reactor room are given by.
V ,, = V @ Nf a * - Nf(vi + Vo @a + L Vs) + Nf vi, i (1)
Y dN,, = -L Ny2 V + v (Nf - NfJ-(/2 ,Nf 2 V -/u, Nf V3), (2)
/-
( V>dN = , f1.,, Ny Vz - /u, Ny V3)- Ny(X"Y> + q), (3)
Where:
subscript 1 = Reactor region (water region in core),
subscript 2 = Reactor tank water region external to the reactor, subscript 3 = Reactor room region, superscript 40 = Argon-40, superscript 41 = Argon 41, superscriptA = Argon-40 plus argon-41, V = Volume of region, cm',
N = Atomicdensity, atoms /crif, A = Decay constant, sec4 ,
o = Absorption cross section, cm',
q = Volume flow rate from reactor room exhaust (cm'/sec),
v, = Volume fic : rate through region No.1 (cm'/sec),
$ = Average thermal neutron flux in Region No.1 (n/cm2 -sec),
fg = Fraction of argon-41 atoms in region i that escape to region j per unit time, seed.
A (v)- A-1
To estimate the volume flow rate of the water in the reactor, the following equation is used:
f^
\
l Q (4) !
vi = C,5Tp. l.
Where: ,
v, = Volume flow rate of the water through the core, ;
Q = Reactor power 20 X 10' watts, _
t C, = Specific heat of water a 4.19 watt-sec/g 'C, .
- BT- = Temperature rise across the core a 11*C, r r a Exit water density = 0.99862 g/cm'. +
Thus:
20x10' i cm' = 435x10' cm' '
v, = (4.19)(l l)(0.99862) sec sec
. Equation (1) can be reduced to Vi dt
= Vo@Nf a"-(Nf-Nf)vs
[
( by considering the following numbers:
v, = 4.35 x 10' cm*/sec, V, = 1.45 x 10' cm',
$ = 4.8 x 10" n/cm'-sec, ,
i i
. c = 0.060 x 10' cm',
.A = 1.06 x 10d sec', ,
I to showthat: ,
4 va + Vo @a + k Vi 5 vs .
During equilibrium conditions the three equations reduce to: ;
Vi@Nf a" = (Nf-Nf)va (6) \
i Nf AV + f2 3 3 V : "(Ni - N)Vi + /s2NV 2 3 3 .(6)
A-2 i
,o- Combining equations (5) and (6) gives I
Nf 'LY> + q + /s-2V>l = f2 s N?V2 (7)
, VoWa* 4 />-> &s ,
ggy
-gy , LY + /2->V2 #
L Y + /2 s V2 which inserted into equation (7) for N/' yields
~
si 'k V> + 4 + f 3 s Ys f3->Vs = Vo@N?a*
Ne - (8)
. /2 s V2 1 V2 + /2 s Va. L V2 + f2 s V2 The values of constants in equation (9) are ,
V2 m 0.189 x 10' cm' ( 6.5 ft. dia. x 20.08 ft. high) .
V, = 7.075 x 10' cm* (1951 ft: x 12.83 ft)
- q = 1.12 x 10' cm'/sec o* = 0.47 x 10 cm8
'\
which leaves the following to be evaluated:
Nf, fs s,13 2, N?
The argon activity in the reactor pool water results from, argon dissolved in water. In the calculation to determine the amount of argon dissolved in the pool water, the assumption was made that argon follows Henry' slaw. If the water temperaturels taken to be 70*F, the correspondingwater vapor pressureis 26 mm Hg. The partial pressure of air is then 760 - 26
734 mm Hg. The argon content of air is 0.94% by volume; hence, the partial pressure of argon is 734 x (9.4 x 10-8) = 7 mm Hg. -
The saturatedconcentrationof argonin water, accordingto Henry'slawis:
P (10)
X=A-6 L
t A-3 1
I
.~.i-,
(~] where:
( ) X = Mole fraction of argonin water,
'# P = Partial pressure of argon above water, K = Henry's constant,2.84 x 10' at 70*F Thus, X = 2.46 x 10 mole A* per mole of (H2 ) + A*, or X = 1.367 x 104 mole A* per 1 cm' H,0.
This yields 1.367 x 104 x 6.02 x 10 = 8.22 x 10'8 argon atoms /ctrf H,0.
An estimate of the parameterf, 3 (that fraction of argon atoms in the poolwatar that escepe each second) can be obtained by examining the mobilities o. ions in dilute solution. Most ions have velocities of the order of 3 to 8 x 104 cm/sec under a potentialgradient of 1 volt per em
[A1). Since the argon atom will not have the advantage of being affected by a potentialgradient, its velocitywould be less than 3 x 10d cm/sec. Therefore, only the argon atoms within 3 x 104 cm of the pool surface will be in a region in which the argon atoms can leave tt:e water within any given second. Actually, even this source volume is still too large. Nevertheless,it gives an upper limit for the fraction of the total argon atoms that can leave the water per second.
3x10" ,
fr , = 3x10" " ~ *
- q. 6.12x10'
') where: H = water height (cm).
%./
During equilibrium conditions and assuming no difference in the rates of escape fraction for argon-40 and argon 41, the number of argon atoms that escape from the water into the air equals the number of argon atoms that enter the water from the air, i.e.,
fr , Nll's = /s , Nll's. (11) where:
N^3 = 2.1 x 10" argon atoms /cm' of air a N.3",
N,^ = 6.77 x 10'5 argon atoms /cnf of water a Ni ",
solving for f 3_ , gives f,_,, = f,.,, Nll'#, = 7.85x10'" sec (12)
\ l V
l l
Since A > fus > fu,, equation (g) reduces to
%J g;i_ , Vi* N 0 /9s (13)
(kVs + q) L
i where:
V, = 1.45 x 10' cm' ,
4 = 4.8 x.10" n/cm'sec -l '
N," = 8.22 x 10 argon atoms /cnf o"- = 0.47 x 10~ cm' fus = 4.9 x 10r'stom/sec A = 1.06 x 10d seed
- q. = 1.12 x 10' cm%oc V3 = 7.075x10' cm' Solving for N yields 0.10 atoms /cm' This dorrespondsto a concentrationof argon-41 ;
activity of A N l.06x10" x0.10 A = = = 0.286x10 c / cm' (14)
C 3.7x10,
- /~
where:
k A* ' = Argon-41 concentiation.pc/cm'
.C = Conversionfactorfromdisintegration/secto c.
t This is below the effluent concentration for dose to the public of 1 x 104 pc/cm' and the 4 '
occupationalinhalation DAC value of 1 x 10 c/ cm' as listed in Appendix B of 10CFR20.1001-4 20.2401 and complies with the provisions of 10CFR20.1302.
1
- The argon-41 activny discharge rate from the reactor ro'om is obtained by multiplying the ,
activity concentration by q, the value of air discharged per unit time, that is A, = 0.286 x 10* c/cm' x 1.12 x 10' cm'/sec
= 0.320 x 10~' pc/sec (15)
The whole body dose rate for the most exposed worker in the reactor room can be ,
estimated by assuming that the room is a hemisphere with the equivalent volume of the room and the individualis positioned at the center of the hemisphere. Therefore: .
(
A-5
_ _ _ . . , . - . - ,_ . _ . . . ., . . . - - -..a
l g
~
D= (18) 2 p, g where:
S,, = 0.28gx10'pCi/cm 8x 3.7x10' dis /s-pCl
= 1.07x105 dis /s-enf-R = Radius of hemisphere = 696 cm - :
, = linearabsorptioncoef.cm d
= 8x105 forair g - = dose conversionfactordis/s-enf mR/hr
= 5.5x10' for Ar-41 1.07x10 (1 - c"D'*)
d
~
2 2(8x10")(5.5x10 )
= 6.5gx10' mR/hr Regarding the above, it has been determined that the whole body dose for the Most Exposed .
Worker (MEW) would be 0.013 mR per 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br />. The above values would actually be less since the hatches to the reactor tank are kept closed and have 1/4 inch plastic sheets under the grates which allow only 2% of the tank to be uncovered A.2 Offsite Dose Calculations Calculations of dose to the Most Exposed Member of the Public (MMP) and the Nearest Permanent Resident (NPR) were calculated using assumptions and methods for finite clouds prescribed in NUREG 0845. Whereas the use of semi-infinite plume modeling techniques have been used in the past for determining dose; it does not take into account the exposure from shine due to higher concentration of radionuclidesin the plume passhg above ground-level. As a result, it is possible that this technique may underestimatethe actual dose.
When using the noinograms provided in the NUREG for Ar the wholebody doses 4
received in a year by the MMP and NPR are 1.89 x 10 mR/hr and 1.68 x 10-5 mR/hr respectively. ,
4
' The skin dose was also calculated to be 2.14 x 10 mR/hr for the MMP and 2.0 x 108 mR/hr for the NPR.
A.3 Nitrocen-16 Activity in Reactor Room The cross section threshold for the oxygen-16 (n.p) nitrogen-16 reaction is 9.4 MeV; however, the minimum energy of the incident neutrons must be about 10.2 MeV because of center of mass corrections. This high threshold limits the oroduction of nitrogen-16 since only about 0.1% of all fission neutrons have an energy in excess of 10 MeV. Moreover, a single
- (f}).
f m t -
l
,n, hydrogen scattering event will reduce the energy of these high-energy neutrons to below the
.I threshold.
The measured effective cross section for oxygen-16 (n,p) nitrogen-16 reactions and a fission neutron spectrum is 1.85 x 10r" cm' [A2). This value agrees well with the value obtained from integrating the effective cross section over the entire fission spectrum.
The concentration of nitrogen-16 atoms per cm' of water as it leaves the reactor core is given by l - c'# (18)
N " = *' A Where:
N" = Nitrogen-16 atoms per cm2 of water,
- e, = Virgin fission neutron flux a 4.8 x 10" n/cm'-sec at 20 kW N = Oxygen atoms per cm' of water = 3.3 x 102aa toms /cnf 2
a = Absorption cross section of oxygen = 2 x 10'" cm d
). = Nitrogen-16 decay constant = 9.7 x 10-2 see t = Averagetime of exposurein reactor The average exposure time in the reactoris given by t=V1 (19)
V/
l O~
where % is the core water column exposed to flux %. and vi si the volume flow rate through the core (see section A1). Thus, l.45x10* cm' I= = 333sec 435x102 cm' / sec , ,
' 8 2 Solving for N" from equation (18), one obtains 3.14 x 10 nitrogen-16 atoms per cm of water leaving the core. With a flow of 435 cm'/sec, the rate of nitrogen-16 leaving the core is therefore 1.36 x 10' atoms /sec.
In the TRIGA Mark I, the measured transport time for the water to travel the 16 feet from the reactor core to the surface of the tank is 42 seconds when the reactoris operating at 100 kW
[A3]. To a first approximationthe velocity of the rising water is proportionalto the density difference between the pool water and the heated water from the core, that is, V = f(p o - P,,, )
l l
v A7 l
ex Thus the velocity of the rising water column for the OVAMC TRIGA Mark i Reactor can be
! ) estimatedfrom v
Y'(P o - P,a )
y ,
(P'o-P',a )
Where, V, = the velocity of the rising water column (OVAMC) po = initialdensity(OVAMC)
= 0.99987 g/cm' p.,, = exitdensity(OVAMC)
= 0.99862 g/cm*
v' = the velocity of the rising water column density (reference reactor [A3])
= 16 feet /42 sec p'o = initialdensity(referencereactor[A3])
= 0.99987 g/cm*
p' . = exit density (reference reactor [A3])
= 0.99862 g/cm' The transport time for nitrogen-16 through the 16 ft. of water above the OVAMC TRIGA reactor core is then 226 seconds.
This assumes that the nitrogen-16 rises straight up toward the pool surface. In practice, j'] however, the nitrogen-161s slowed down by the interruption of the vertical convective currents from the discharge of water horizontally from the coolant flow return pipe in 226 seconds, the
.V nitrogen 16 decays to 2.87 x 10"' of its initial value. Thus, the number of nitrogen-16 atoms that reach the water near the pool surface is about 0.39 atoms /second.
Only a small portion of the nitrogen-16 atoms present near the pool surface are transferred into the air of the reactor room. When a nitrogen 16 atom is formed,it appears as a recoil atom with various degrees of ionization. For high-puritywater (~ 2 mho), practically all of the nitrogen-16 combines with oxygen and hydrogen atoms of the water. Most of it combines in an anion form, which has a tendency to remain in the water [A4). It is assumed that at least one-half of allions formed are anions. Because of its 7.1 second half-life,the nitrogen-16 will not live long enough to attain a uniform concentrationin the tank water. Assume that the nitrogen-16 atoms will t;e dispersed in the 1 ft, of water at the top of the pool directly above the core, in the area directly above the core, the dominant contribution to the dose rate is the direct radiation from the core. The interest from a hazard point of view is then the number of nitrogen-16 atoms escaping into the air and diffusing away from the area above the core.
The maximum fraction of nitrogen-16 atoms that can escape from the water to the air per second can be estimated similarlyto the case of argon (see Appendix A.1). Thus, 1 3x10"cm / sec = 0.5x10-' sec ,
f,x,, s 2 (20) 30cm
[ }
l' \
x ') A-8 L
l L-
l.
i f where it is assumed that one-half of the ions formed, namely the anions, remain in the water. .
Thus the number of nitrogen-16 atoms entering the air is given by 4
f, 3(0.039) = - -
0.5x10-'(039)
/##' N"r = (21) f,,, + A" 0.5x10-8 + 9.76x10-8
= 2.0 x 105 atoms /sec For the purposes of analysis,it is postulated that the water-bearing N-16 rises from the L
core to the surface and then spreads across a disk source with a radius of 100 cm and area A, = -
3.14 x 10' cm'. "
For a constant velocity of v,; the cycle time for distributing the N-16 over the tank surface would be:
t, = L =
- 46s v, 2.16cm / s
-The average concentration at this time is
,,,m e '"
y , ',[Noe '" dtl , N4 ,.u,) , 9.0lx1 0 " , l = 2.00x10" #### (22)
[s a t, At, 9.7x10'2x46 cm'
- The thickness of the layer of N 16 bearing water is:
h = "2 = = 3.16x10-'cm (23)
A, 3.14x10' The dose rate at the tank surface arising from the N-16 near surface is ,
D = 13 1 - E2( h)- (24) 2K- -
4 d
Where p is the attenuation coefficient for 6 MeV photons ir water with a value of (0.0277 cm ), K is the flux-to-dose-rate conversiort 5
p totonlsC"' '
- 1.6x10 Yhr 1
A9
_ _ _ _ _ _ . _ . _ . _-. ~- - . _. _ _ _ ~ _. _ _ . . _ _ _ _
P O and E2 is the second exponential function O)
E ( h)n [# prT, dT 2
For x 1 the interval of integrationis taken to be from 1 to 10/x as a close apprnximation. It can also be seen that since -
e"' = 1- ht + (yhi)' (ni,i)' +-
then e* s 1- ht. Thus, the dose equation becomes H
AN " l- hT dT (25)
D = 2pK <l ~ T* , ,
This yields a dose rate of 7.00 x 104 mR/hr. This negligible exposure precludes the necessity of
.any further dose calculations to individuals at greater distances. The production of N-16 is of no significancein terms of the dose received and poses no hazard.
A.4 Activationof Airin the ExperimentalFacility In the. Omaha VA Medical Center TRIGA reactor facility, the rotary specimen rack and v pneumatictransfertube contain air. Of the radioisotopesproducedin these air cavities, argon-41 (with a half-life of 110 min.) is the most significantwith respect to airbome radioactivity hazards.
Nitrogen-16 (7.11 see half-life) and oxygen-19 (26.9 see half-life) are considerably less significant.
The saturated activity of argon-41 in an experimentalcavity is calculated from i d
AS (26)
A = N"L" = C(k" + q / V) pCl/cm' where:
q = Volume flow rate, reactor room exhaust (cm'/sec)
V = Volume of region (cm') ;
$. o Ave.thermalneutronflux d
4" = N"o," = Macroscopic absorption cross section of argon, cm S = $L n/cm'-sec J A-10
p Volumes and thermal neutron fluxes of facilities are as follows:
Ave. ThermalFlux at 20 kW pac;gy Effective air volume (cm')
(n/cm'-sec)x10" Rotary Specimen Rock 8.7 x 10' 1.36 5
PneumaticTube 1.6 x 10 3.74
- 1. PneumaticTube(PT) .
This facility exhausts into a fume hood which ews to the roof of the hospital.
In air-filled cavities from which air is removed at the rate of q cm' per second, the i accumulation of argon-41 as a function of operating time is given by dN = , - Nd ' A' - N (27) dr i
! where:
N = Atomicdensityof argon-41, atoms /cm'of air r
G = Average thermal neutron flux in cavity, n/cm .,
I," = N*a." = Macroscopic absorption cross section of "Ar, cm., = 0.986 x 10' cm '
d A = Decay constant of"Ar = 1.06 x 104 see V_ = Volume of air cavity, cm'
' At equilibrium conditionsdN/dt = 0, in which case e I" N = ,, , ' g, atomsI cm' The correspondingactivityof argon-41is A $ I""
A = N A = C(L + q / v) pCl / cm 4 where C = 3.7x10' disintegrationsper sec per microcurie.
5
. The effective air volume of the PT that exhausts its activity through the stack is 1,6 x 10 r
cm'. The neutron flux in the PT has been measured to be 3.74 x 10" n/cm -sec and the exhaust rate for the PT is 9 x 105 cm8/sec. Under these conditionsthe activity leaving the PT per second is given by _
^#
n A-11 b
-_ _. , ., _ . _ _ ___ y ,_
1
{
\v/ - A,,q = L@E.*V(q / V) pCl/ sec C(L*' + q / V)
(28) 1 Where:
q = Volume flow rate,(cm'/sec)
V = Volume of region (cm')
@ = Ave.thermalneutron flux I," = 0.986 x 104 cm - ,
S = @I. n/cm*-sec ,
2 For pneumatictransfertube q = Volume flow rate = 3100 ft/ min x 6.157 x 108ft'
= 9.01 x 10' cm'/sec zV- = 1.6 x 10'cm' '
-@ = 3.74 x 10" n/cm'-sec 4 8 8 (1.06x10-')(3.74x10")(0.986x10 )(1.6x10')(9.01x10 )(1.6x10 );
. -Aq= 7 9.0lx10'.
3.7x10' 1.06x10" +
( ,1.6x10, ,;
Aq = 0.169pCl/sec (29)
The activity leaving the pneumatic tube is therefore 0.169 pCi/s.
The ventilation exhaust rate of the hood is:
150 LFM x 3.348 ft' = 502.2 CFM
= 2.37 x 10' cm*/see providing dilution of the argon-41.
4 .;
The concentrationin the exhaust air due to the pneumatictube is then about 7.13 x 10 pCi/cm'.
_. _ However, during the last 3 years the fellowing total number of samples were irradiated in
- all of the experimentalfacilities. .
s_
A-12 ,
, , w ,w .-,,-~r -,,--,,,,+a , , . + , ' , -- - - - - , - - - - - - - ~ - . - - - - - - - -
4 j i
~'
Yag . Numbnr of samples 1994 (6mo)c _ 2,294 (2x1147). ;
1993 2,813 = l 1992 1,995 Assurning all of the samples were irradiated in the pneumatic tube and that the blower for the pneumatictube was used for 30 sec for each irradiation (sample transfer time = 2 sec), the total
- time the blower would be on would be: 30 sec x 2813 = 8.44x10' sec. .
f . P Therefore, the quantity of argon-41 released from the pneumatic tube would be 0.169
- pCi/sec x 8.44x10' sec.-
= 14.26x1& pCi d'Ar.
c ? A check, w2h an volometer, of the input air ducts to the pneumatictube showed that the only time
. air is circulated thru the PT tube is when the blower is on for the 30 seconos per sample.E The -
4 blower is turned off while the sample is being irradiated.
Since the fume hood is operated continuously,the annual average concentration of argon-41 released to an unrestricted area is as follows:
" 8 Fume hood exhaust = 2.37x11 crn /sec Number of sedyr = 31.536x1& sec/yr
)[ '
Therefore:
2.37x11 cri.8/sec x 31.536x11 sec/yr = 7.47x10'8 cm' air exhausted.
- The average annual concentration o' d'Ar vented from the pneumatic tube to the hospital roof is 3- then:
p 14.26mCl 4
= 1.91x10-" mci / cm' = 1.9x10 pCi / cm' (30)
- 7.47x10" cm' .
- - Calculations of dose to the most exposed member of the public (MEMP) and the marest -
permanent resident (NPR) wera again determined as in section A.2 and are summerized in Table 1.
- ' - 2. Rotary Specimen Rack .
The rotary specimen rack contains 40 evenly spaced aluminum containers each with'a
- maximum.intemal space.of 31.75 mm dia, by 27.4 cm in length giving a volume of 216.93 cm' ,
each. The neutron flux is 1.36x10" n/cm -sec.
The exhaust rate from the rotary specime'n rack removal tube is no greaterthan 10 linear i . feet per minute (LFM) as determined by volometer readings.The exhaust occurs from natural ;
x: heat convection out of the long elender removal tube with a inner diameter of 33.9 mm. Again, we
- . use the saturated activity under equilibrium conditions -
4 j,
. A-13 4 g y I ,
'i
. - - - m. _; .___ _ . _ __ __
i!-
'jet 9-A
- I9 O pCi/ sec. (31)
C(L*' + q / D Wriere:
8 q = volume flow rate = 8.5 cm /s _
8 V = Volume of region = 216.93 cm x 40
= 8.68x1d .
$. = Ave, thermal neutron flux -
= 1.36x10" n/M-s I," = 0.986x10' cm4 A = 1.06x104 .
f '
8 8.5 (1.06x10-*)(1.36x10")(0.986x10-')(8.6x10 )(8.86x10's Af = , ,
gg e 3.7x10' 1.06x10-' +
( 8.86x10'ss I
= 0.30 pCl/sec.
The ventilation exhaust rate of the reacto: room is 1.12x105 cm*/sec providingdilution of
.- the argon-41. The concentration of argon-41 exhausted to the outside and in the room is then
. (. / about 2.66x10 pCi/cm 8-s if the rotary specimen rack was continually venting and left open.
However, when samples are not being pJt into or retrieved from the rotary specimen rack the sample tube is capped. 'In practice, the Rotary Specimen Rack is seldom used for activation.
The higher flux and the shorter sample retrievaltime make the pneumatic tube our principle irradiation facility.Therefore, the assumption of 100 samples per year as well as a sample removal time of 2 min per sample (sample transit time for removal or retrieval = 7 seconds), the s
total time the argon woud be venting (inserting & removing sample) would be 1.2 x 10'sec/ year (assuming 5 days /weekor 260 days! year the removal).
Therefore,the pCl/yr vented would be:
0.30 pCi/sec x 1.2x10* sec/yr = 3.60 x10 Ci/yr Further assume that only 25% of the rotary specimen rack is assumed to exhaust, since experimentswill replace some of the air, and the exit of the argon will be hindered by the slender
. bend removaltube.
- Therefore,the pCi/yr vented would be:
3.60x10 x 0.25 = 0.00x102 pCi/yr_ l N
, 7%
' ' v 'l -
A-14 1 1
w
?
77 Since the reactor room is continually, the amount of air exhausted per year is d 8
,1,12x11 cm /s'x 3600 x 24 x 365_= 3.532x10'8 cm'/yr Thus, the average concentration of argon-41 vented to the outside or inside from the rotary specimen rackis:- ;
9.00x10' pCi/yr + 3.532x10'8.cm*/yr 1
= 2.55x10" Cl/cm' Which is less than the occupational DAC concentration and the environmentalair effluent concentrationlisted ir. App:,ndix B to 10CFR20.1001-20.2401. Typically,the lazy susan samples -
are run separately so that only one can is actually vented. Thus, under normal circumstances p : would be 2.5% of the above calculations.
a a.- The whole body dose rate for a worker in the reactor room can be estimated by -
L assuming that the room is a hemisphere with the equivalent volume of the room and the individual is positioned at the center of the hemisphere. Using this assumption,the dose rate is determined by:
D= (32) 2 p,g 1 ' where:
S, = 2.55x10"pCi/cm8x 3.7x10' dis /secyCi 3
= 9.43x10* dis /s.cnf
-R = Radiusof hemisphere- 696 cm (volume of hemisphere = 7.075x17 cm8 p, = linear absorption coef (cm ')=8x105 for air g = dose conversion f% tor dis /s-cm' , mR/hr
= 5.5x102 forargon ,
4--
a therefore:
4 D = 9.43x10^ (1 -4e*V"*h*)
- 2x10-'(5.5x10
' = 5.80x104 mR/hr
= 1.10x108mR for 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> operation Calculations of dose to the MMP and the NPR were also calculated as in section 4.2 and F . summerizedin Table 1.~
p
.d A-15 .
, - , . _ - , , , , , s
a; n.
Y Experimentaldeterminations of Argon levels have been p$rformed with little success.The -
theoretical calculations suggest that the levels are to low for instmmental determination.-
...-1 4
(
b 1
L
- A-16 -
,=--_____x-_-__ _ _ _
Q x ,
Table #1 Summary of results '
from N &d 'Ar Wholebody Wholebody Skin DOSE Yhiobody Skin DOSE mci /yr Release Point DOSE ' DOSE mR/yr mR/yr DOSE mR/yri (outside) NPR m R/yr . . MMP MMP mR/yr NPR MEMF MEW 2.1 Ventfan 1.3x10' 1.9x105 2.1x105 1.7x105 2.0x10-5 d'Ar water 0 1.28x10' 1.43x10* .1.28x10 d
1.43x104-d'Ar PT 14.3 Hood exhaust l
Ventfan 1.2x16 6.3x10* 9.0x10* 4.5x10' 8.1x10* -
! Ar LS 0.9 d 1.3x10' 4
1.53x10 - 1.73x10' 1.5x10 1.7x10' -
Total *'Ar 17.3
'N water - NA 1.4x105 A-17
O
.(j AppendixA References A1 . Daniels, F., " Outlines of Physical Chemistrf, Wiley and Sons, New York, p. 414.1949.
A2 Henderson,W.J., and Tunnicliffe, P.R.,"The production of N-16 and N-17 in the cocling
- water of the NRX Reactor", pp. 145-150. NSE 1958.
A3 " Hazard Report for Torrey Pines TRIGA Reactor, General Atomic GA-722, 1959.
A4 - Mittli, R.L. and Theys M. H.,"N-16 Concentrationsin EBWR", Nucleonics, p. 81. March 1961 h
~
A-18
1 APPENDIX B Calculated Maximum Fission-Product Release After A Fuel Element Failure is,)
(
Calculations and a related experiment have been made to determine theoretically the maximum concentration of fission products that might be present in the reactor-room air and the environs following a fuel-element cladding failure.
The calculations are based on the fact that as the reactor operates, fission products will build up in the uranium - zirconium hydride fuel mixture until an equilibrium concentration is reached for each nuclide, dependent on (1) the total energy release in the reactor, (2) the decay process for each nuclide, and (3) the yield of the species from fission. Of the various fission products produced in the fuel material, only certain nuclides will migrate into the gap between the fuel material and the fuel element cladding. These nuclices are the iodines, the xenons, the kryptons, and the decay products of all of these elements.
B.1 Gaseous Fission Products Released from Claddina Failure In the event of a rupture of a fuel element cladding the fission products released to the water in the reactor tank will be limited to all or a portion of those fission products that have collected in the gap between the fuel material and the aluminum cladding. A portion of these fission products will then go into the air above the water, this fraction depending on the solubility of the species in water.
The quantity of gaseous fission products produced in the fuel element was determined by Blomeke and Todd [B1]. The amounts of krypton, xenon and iodine produced in a typica' element after infinite operation at 20 kW are given in Table B-1. These data are based on a loading of 36.8 grams of 235U per element, where Nf3 is the initial number of 23sU nuclei v (9.4x1022), N, is the number of nuclei of the isotopes in the fuel element, and the average flux is 4.8x10" thermal neutrons /cm'-sec. l In order to determine the actual percentage of fission-product gases that escape from l the fuel material and collect in the air gap between the cladding and the fuel material, a series l
of experimente "Sre done in the TRIGA reactor at General Atomic between 1960 and 1971
[B2]. These experiments involved measuring the Xe-133 that findu its way into the gap during the exposure of the fuel element to a known amount of reactor flux. From this, the percentage of Xe-133 and consequently the percentage of all other fission-product gases that collected in I I
the gap were determined. From this work it was concluded that for low temperature operation at temperatures below 350*C a release fraction of 1.5x10~5 for a standard TRIGA fuel element is l a conservative value to use for safety considerations [B2]. l Thus, applying this release '"ction to the values listed in Table B-1, the total gaseous activity in the gap is calculated to be 9.25 mci.
Therefore, the nsaximum amount of fission products that could be released in the event of a cladding failure is: ,
lodines 3.84 mci Xenons 1.97 mci j Kryptons 3.44 mci 1
Total fission products 9.25 mci I 1
pm I
) i a
B-1
B.1.1 Results of Contamination of Shieldina Water 2
O The v+1me of water in the TRIGA reactor tank is approximately 1.9x10 7 cm8
. For purposes of this calculation, it is assumed that the total 9.25 mci of gaseous f""on products in
- the gap escapes into the water. However, when these fission products (the ic. es, kryptons, and xenons) are released from the fuel element into the water, the lodines will be dissolved in. .
the water. Since the krypton and xenons are quite insoluble in water, it has been assumed, for the purposes of this analysis that a major portion of these isotopes will escape into the reactor-
- room atmosphere.' Based on the assumption that xenon and krypton follow Henry's law, it is estimated that 95% of the xenon and 98% of the krypton will escape into the air and that all of
.the iodines will remain in the water Therefore, the activity in the water will be:
,pCl 3.84 mci + 0.05 x 1.97 mci + 0.02 x 3.44mCl
- A = x 10'pCs.
cm' l.9x 10' cm,
= 2.1x10" mci /cm8 (1) and the activity in the air will be
,'" , 0.95A xenon (pCi) + 0.98A krypton (pCl)
- volume ofair (cm')
0.95 x 1.97 mci + 0.98 x 3.44 mci
= x 10' Cl 7.1x 10* cm' h = 7.38x104 mci /cm 8 (2)
I Since krypton and xenon are inert gases, the hazard presented by their being in the air is from the dose a person in the room would receive from their decay. To estimate this dose, it
. is assumed that each decay of a krypton or xenon nucleus results in the emission of a 0.5-Mev gamma photon. Then the dose rate at the center of a hemispherical volume of 7.075x108 cm*
(equivalent radius = 696 cm) is:
y,S,0-c%
2EC Where:
D = dose rate, Mr./hr S, = photons emitted per sec per em8 of air S = attenuation coefficient of air.
- R = outer radius of hemisphere C = flux to dose conversion factor d.
B-2
Therefore:
,o
(', ) D=
738xl 0-'2Ci / cm'x3.7x10 photons / sec- Ci(1 - e~' "'*)
2x8x10cm~'x1.085xl O' photons I cm' - secI rad I hr
- = 1.57x10 (1-e**7) = 8.5x10 2 MrJhr (3) where the attenuation coefficient for air is taken as 8x10 cmd and the flux-to-dose conversion factor is 1.085x10 8.
.. B.1.2 Intemal Exposures from Breathina Fission Product Cloud When considering the effect of inhalation of insoluble volatiles escaping from a ruptured fuel element, the critical organ is the lung. The beta-emitting nuclides become more important than the gamma emitters. This situation is different from external exposures where betas are ini'1nificant unless they are deposited on a person's clothing. Table B-1 shows the activity of the Deta-emitting nuclides produced in a central fuel e;ement. The fraction of the activity released is again taken as 1.5x10-8 There is sufficient time for a person to evacuate the room before breathing a significant amount of beta-emitting nuclides because the fission products do not come out of the water all at once and because the air in the room is not stagnant. This conclusion assumes that the evacuation alarm sounds with the release of fission products and that personnel leave the reactor room within minutes of the release. Calculations have been made to doteimine the internal exposure from breathing the fission products after they have been Mspersed uniformly
,/^g in the reactor room. This situation approximates the condition of a person leaving the reactor V room after the fission products have been dispersed. In the case in which the person spends a short time in the reactor room (compared with the shortest half-life of the beta-emitters, i.e.,
about 2 minutes), his lungs will be filled with the concentration existing in the room, which for the reactor room is:
A (0.95 x 1.95 mci + 0.98 x 3.39 mci) x 10' pCi E 7.1x10'cm 3
= 7.28x10 mci /cm Since there are approximately 3 liters of air in the lungs, the beta activity in the lungs would be:
2.18x10 2 mci on leaving the room. (4)
The beta exposure to the lungs is calculated from the following formula:
Dose = ACR [s f,E,(1-e~4) rads (p)
~)
\
B-3 i
n where:
A = activity in the lungs on leaving the reactor room (V) C = conversion factor (3.7x104 b/sec-mci)(1.6x10-8 ero/MeV)
R = Retention factor m- = mass of lungs (1000 grams) f, = fraction of total activity Ei = energy of beta from nuclide I, MeV I, = decay constant t = time of exposure A fraction of the inhaled activity will be retained in the lungs. In as much as gases have a tendency to callect on dust particles, a 1/8 retention factor is customarily used for the inhaled activity. The decay constant is given by the sum of the radioactive decay constant and the biological release constant (6.7x10-8 per sec).
For a beta activity in the lungs of 2.18x102 mci the maximum integrated exposure to the lungs from inhalation is:
2.65x10-5 rads. (5) 8.2 Fuel Handlino Accident A fuel handling accident involving a TRIGA reactor was analyzed in NUREG 2387 [83).
In this analysis a TRIGA fuel element was removed from a TRIGA reactor that was operated at
.Q 1 MW for 1 year prior to shutdown, or 365 mwd It was assumed that during fuel unloading a
(") fuel element is hit by a shipping cask dropped from an overhead crane. The fuel element was assumed to have a large cladding rupture and severe physical damage to the fuel. Calculations in the NUREG assume a flux of 1x10" n/cm2 -s. Modifying the calculations for the average OVAMC TRIGA reactor neutron flux of 4.8x10"n/cm 2-s and operation at 20 kW for 1 year or 7.3 mwd. This is a highly conservative assumption when compared with the actual yearly average power generated of 0.492 mwd. It is also assumed, to be consistent with worse case conditions, that the element was near the center of the core and thus received the highest bumup, in the TRIGA reactor, the maximum burnup is typically twice the average. Dividing by the 57 elements in our core and multiplying by two, the power history of our single element will be 0.26 mwd. The calculated noble gas and radioiodine activities released from this element are shown in Table B-2.
As can be seen from the table, at 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after shutdown, the major activities are from "2Xe, "'I and "31. Xenon-133 is a nob!e gas that decays by the emission of a beta particle with a maximum energy of 0.346 MeV. Since this nuclide is chemically inert, it mainly poses ar.
extemal exposure hazard from immersion in a gaseous cloud containing the " Xe. On the other hand, the iodine isotopes pose an intemal hazard. Radiciodine, if inhaled or absorbed directly through the skin, is absorbed into the blood stream and localizes in the thyroid, which receives the greatest dose equiva'ent. All of the radioactivity would not be released since the fuel matrix acts to strongly retain the fission products. As previously mentioned in Appendix A, Foushee and Peters (B2] have collected and summarized empirical data regarding release fractions.
Even with unciad, heated, irradiated fuel the fraction of gaseous activity released did not
(~N ; exceed 7x10'8 and was generally 2 to 3 times below this value. The gap activity fraction was approximately 1.5x10~5, and negligible amounts of nonvolatiles were released to the
)
B-4
f- atmosphere. Release of nongaseous fission products is typically one or more orders of cs magnitude lowe* [7], and thus exposure from the source is secondary to exposure from f ') gaseous radionuclides.
.%.j' As summarized in Table B2, the total activity released from a single element at shut-down fol!owing operation of 20 kW at 0.26 mwd is 0.18 mci and after a shut-down of 48 hrs is 0.04 mCl.
The concentration in the Reactor Room would then be:
0.18xl0 Ci 5
Cl I a) At shut - down: = 2.54x10
7.1x10* cm' cm' O.04 Cix10' = 5.63x10~* pCi b) 48 hrs after shut -down: 7.1x 10*cm, cm, Applying the same reasoning as was used in Section B.1.2 where an individual leaves the room after tlra accident, the beta activity in the lungs would be: ;
7.62x104mci on leaving the room.
The beta exposure to the lungs is calculated to be:
9.26x10~7 rads.
(D j Following the criteria specified in Regulatory Guides 1.109 and 1.145; calculations were made to determine the potential dose to the public outside the facility. The total body dose for ground release of noble gas is given by D.r(r,0) = Sd x,(r,0)DFB, (S) where:
DFB, = the total body dose factor for a semi-infinite cloud of the radionuclide i which including the attenuation of 5 g/cm2 of tissue in mR/pCi-yr;
( o,r (r,0) = is the annual total body dose due to immersion in a semi-infinite cloud at the distance r in sector 0, in mR/yr; and x,(r,0) = the annual average ground-level concentration of radionuclide i at the distance r in sector 0, in pCi/m x,(r,0) is determined by x,(r,0) = 3.17 x 10' Q,[x/Q) (r,0) (7)
[x/Q) (r,0) is the gaseous dispersion factor (corrected for radioactive decay) in the sector at angle 0 at the distance r from the release point, in s/m* 3.17 x 10*is the number of pCi/Ci divided by the number of seconds in a year.
B-5 l
, , , , .. - _ - . - - . - . . ~ . - . . -.. . . . .
I i (For atmosphenc stability conditions when the'windspeed at the_10 meter level is lessf 4 l
t than 6 meters per second horizontal plume roeander may be considered. x/Q values may be ~
- l,(qjf 7 determined from the set of equations specified in Regulatory Guide 1.145;! For this
'N A '
design basis accident the windspeed is presumed to be 2 m/s with a Pasquill stability class of F J to generate a conservative x/Q. :With respect to the to our specific parameters, the atmospheric.-
O - ~
- diffusion can be described by.
c . y (8)' +
G = Un E , a ,
l where:
. 4 x; = the short term average centerline value of the ground level concentration in -
J l Cl/m8 '
Q' = the amount'of material released Ci/s h
LU) = the average windspeed m/s~
f of = is the lateral plume spread which is a function of the atmospheric stability and .
- distance -
4 L= is vertical plume spread with meander and building wake effects which function ,
of atmospheric stability, windspeed and distance. For less than 800 rnsters . ;
' distance 4 = Mo, where M is a correction factor based on stability class.-
' ~
The annual organ dose from inhalation of radionuclides in air is given by .
O, 'Dj(r,0) = Rd x,(r,0)DFA, (9)
Q,); '
~where:
Dj(r,0) = the annual dose to organ j of an individualin the age group a at location (r,0) due to inhalation, in mR/yr; :
DFA, = the inhalation dose factor for.radionuclide I,= organ j, and ar, .
group a, in mR/pCi; .
~
R. = the annual air intake for individuals in the age group a, in m3 /yr all other factors are' defined above. The dose specified for eq.9 can be used for radioiodines and other gases other than the nobel gases. In this case the organ with the highest dose will be the thyroid.' Whole body deses are also given.J Postulated dose assessments are
. summarized below.
R .
SUMMARY
-MMP 1 TOTAL WHOLE BODY DOSE = 1.33E-04 mrem /yr MMP% TOTAL THYROID DOSE S = 2.67E-02 mrem /yr.
ENPR 1.-TOTAL WHOLE BODY DOSE = 4.67E-05 mrem /yr NPR 5 TOTALTHYROID DOSE . = 1.14E-02 mRemlyr -
- \
w.
I + '
4 e -- - v- - -
-_--L_ -------.__:--
In addition to gaseous radionuclides, other fission products will be present, with
~s radiostrontium and radiocesiums being most significant from a hazards standpoint. As previously mentioned the release fraction of nuclides nther than the noble gases are lower by at
[( /j least one or two orders of magnitude. With this in mind, a release fraction of 1 x104 is used.
Table B3 gives the quantity of significant radiostrontium and radiocesiums at zero and 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after shutdown in the fuel element containing the greatest activity with a power history of 0.26 mwd as well as the activity released.
Table 4 shows the calculated air concentration of radiostrontiums and radiocesiums assuming the initial activity divided by the room volume. No reduction is calculated with respect to air exhaust. These values are compared with the Derived Air Concentrations specified in Appendix B of 10 CFR 20.
Therefore, even under the worst of circumstances, the potential exposure to personnel outside the facility from our credible fuel-handling accident would be small and of little or no health significance.
l%.)
,V B-7
TABLE B-1 GASEOUS FISSION PRODUCTS PRODUCED IN FUEL ELEMENT
--[-) (Saturated Activity at 20KVW ,
- G/ .
ISOTOPE N #n N. DECAY- ACTIVITY, - TYPE OF CONSTANT. CURIES RADIA.
-1 SEC TION
- Kr 1.1 x 104 1.0 x 10 1.0 x 10d 3.4 y
- Kr ' 7.8 x 17' 7.2 x 10 4.4 x 175 10.2 0y
'SKr 3.0 x 104 2.8 x 102 2.2 x 104 1.9 py
Kr 3.2 x 10# 2.9 x 10 1.5 x 10d 14.3 py
"*Kr 1.2 x 10* 1.1 x 10" 7.0 x 10' 25 py
- Kr 2.0 x 104 1.8 x 10'S 3.6 x 10-3 21.6 py
- Kr 5.8 x 10* 5.3 x 10" 2.1 x 102 36.5 0y "Kr 2.7 x 10* 2.5 x 10 2.3 x 104 18.7 py
- 1 1.6 x 105 1.5 x 10" 1.5 x 104 7.2 0y
"'I 7.0 x 104 6.4 x 10 1.0 x 104 21.0 py
- Xe 1.0 x 104 7.2 x 10'8 6.7 x 10# 0.2 y f "2l 1.3 x 104 1.2 x 10" 8.0 x 104 31.2 Py
('
- l 1.6 x 104 1.5 x 10 9.3 x 10* 44.7 py
"'*Xe 1.0 x 104 9.2 x 10 3.5 x 105 1.1 y "3Xe 1.0 x 10d 9.2 x 10'8 1.5 x 104 44.9 py
- l 8.0 x 104 7.4 x 10 2.2 x 10d 52.8 py
"'I 4.7 x 108 4.3 x 10" 2.9 x 10 4 40.8 py "5*Xe 5.7 x 10* 5.2 x 10" 7.4 x 104 12.7 py "5 Py Xe 4.8 x 10* - 4.4 x 10" 2.1 x 104 30.2
- l 8.8 x 10-* 8.1 x 10" 8.1 x 104 21.4 py "7Xe 4.6 x 10' 4.2 x 10'5 3.0 x 104 42.3 py
" 81 6.7 x 10* 6.2 x 10'8 - 1.2 x 104 24.1 py
- *Xe 1.9 x 10 # 1.8 x 10 6.8 x 10d 39.4 py
- l 1.6 x 10* 1.5 x 10" 2.6 x 10d 12.5 py
- Xe 6.5 x 104 6.0 x 10" - 1.7 x 102 33.0 py
" Xe 2.0 x 104 1.8 x 10" 4.3 x 104 25.8 py
- (s - -
TOTAL 616.9Ci B-8
._ _ _ __. . _ . . _. _ _ . _ - . _ _ . - - = . _ .
TABLE B 2 ,
O- Gaseous Fission Product Activity in the TRIGA Element Containing the Greatest Activity Following Operation at 0.26 mwd NUCLIDE . T 1/2 ACTIVITY AT ACTIVITY AT 48h ACTIVITY ACTIVITY RELEASED SHUTDOWN' POSTSHUTDOWN,CI . RELEASED AT 48h 4 Cl AT SHUTDOWN,Cl POSTSHUTDOWN,Cl_
- Kr 1.9h 8.71E-02 0.00E+00 1.31 E-6 0.00E+0
- Kr 4.4h 2.06E-01 1.42E-04 3.09E-6 2.14E-9
Kr 10.8y 3.56E-03 3.56E-03 5.34E-8 5.34E-8
'7Kr 1.3h 3.92E-01 0.00E+00 5.88E-6 0.00E+0 "Kr 2.8h 5.62E-01 3.56E-05 8.43E-6 5.34E-10 "Kr 3.2m 6.93E-01 0.00E+00 1.04E-5 0.00E+0
' Kr 32.0s 7.84E-01 1.98E-02 1.18E-5 2.97E-7
'*Xe 2.3d 2.77E-02 1.40E+00 4.16E-7 2.11 E-5
- Xe 5.3d 1.62E+00 0.00E+00 2.43E-5 0.00E+0
'3*Xe - 0.3h ~ 4.27E-01 0.00E+00 6.41 E-6 0.00E+0
- Xe 9.1h 7.32E-01 3.96E-02 1.1 E-5 5.94E 7
l 8.1d 7.68E-01 7.20E-01 1.15E-5 1.08E-5
- l 2.3h 1.18E+00 5.14E-07 1.77E-5 7.72E-12
- l 20.3h 1.38E+00 2.89E 2.07E-5 4.33E-6
- l 0.9h 1.81 E+00 0.00E+00 2.72E-5 0.00E+0
- l 6.7h 1.58E+00 1.11 E-02 2.36E-5 1.66E-7 TOT 1.22E+01 2.49E+00 1.84E-4 3.73E-5 AL x
B-9
(( ' TABLE B-3 Wll
, Activity from Radiocesium and Radiostrontiumin the TRIGA Element Containing the Maximum Activity Following Reacbr Operation at 0.26 mwd NUCLIDE. T%- ACTMTYAT ACTIVITY AT 48 h ACTIVITY RELEASED SHUTDOWN,Cl POSTSHUTDOWN,Cl AT SHUTDOWN,Cl "Sr 52.7d 7.12E 1 0.00E+0 7.12E-7 :
- "Sr - 27,7y 2.37E-2 6.96E-1 2.37E-8
- OSr . 9.7h 9.22E-1. 2.37E-2 9.22E 7
'2Sr 2.7h 1.04E+0 1.58E-3 1.04E-6
- Sr - 8.3m - 1.18E+0 - 1.42 E-6 1.17E-6
"'"C s 2.05y 1.19E-3 0.00E+0 1.19E-9 ,
"'Cs 2.9h 1.98E 3 0.00E+0 1.98E-9
"'Cs ' 13.7d 1.98E-2 1.98E-3 1.98E-8 "7Cs - 30.0y . 3.52E 1 1.58E-2 3.52E-7 "8Cs 32.2m 1.47E+0 3.52E-1 1.47E-6 ii i
u B-10 es.zr e +a e
TABLE B 4 -
4
- 4O .
' Air Coi entrationsand Radionuclidesintake from Radiostrontiumsand Radiocewiums Following HypotheticalMaximum Credible Fuel-Handling '
Accident NUCLIDE- AIR - DERIVED AIR CONCENTRATION CONCENTRATION t (Cl/m') (DAC) -
"Sr 4.9 x 10" 4E-7 "Sr- 1.6 x 10-'2 8E-9
Sr . 6.5 x 10-" 2E-6 e2Sr 7.4 x 10-" 4E _;
- "'C s 9 x 10~" 6E-5
- Cs 1.5 x 1048 4E-8
- Cs 1.3 x 10-'2 3E-7
"'Cs - 2.5 x 10-"- SE ~^
- /
\
e
\
l{
u B 11 l
I 1 _ . ,- . - -. . - . . _ , .
Appendix B References '
t B1: ' Blomeke, J.O. and Todd, M.F., " Uranium-235 Fission-Product Production as a Function of -
- Thermal Neutron Flux, Irradiation Time, and Decay Time", Pt. I, Vol.1, Oak Ridger.
National Laboratory Report ORNL-2127. 1957.' ,
82 Foushee, F.C. and Peters, R.H., " Summary of TRIGA Fuel Fission Product Release Experiments", Gulf Energy & EnvironmentalSystems. Gulf-EES-A10801.1971.
B3. Hawley, S.C. and Kathren, R.L., " Credible Accident Analysis for Triga and TRIGA-Fueled Reactors", NUREG/CR-2387.PNL-4028. April 1982.
.B4 Hawley, S.S., R.L. Kathren and H.A. Robkin." Analysis of Credible Accidents for Argonaut Reactors". NUREG/CR-2079, PNL-3691, U.S. Nuclear Regulatory, Washington, D.C.
1981.
J A
i i
i n
- - B ' .I
- . _ . . m- _ _. - __ --
l
i n APPENDIX C
'l ..
~V LOSS OF COOLANT WITHOUT FUEL ELEMENT CLADDING FAILURE Even though the possibility of the loss of shielding water is believed to be exceedingly remote, a calculation has been performed to evaluate the radiological hazard associated with this type of accident (see Table C-1). Assuniing that the reactor has been operating for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> and 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br /> at 20 KW prior to losing all of the shielding water, the radiation dose rates at two different locations are listed in the table.
Time is measured from the conclusion of operation at 20 KW. Dose rates assume no water in the tank. The first location (direct radiation) is 6.4 meters above the unshielded reactor core, near the top of the reactor tank. The second is at the top of the reactor; this location is shielded from direct radiation but is subject to scattered radiation from a thick concrete ceiling 4.6 m above the top of the reactor shield. The third location is the first floor of the hospital which could be occupied by members of the public. An individual situated above the core at this location is an additional 4.12 meters from location two. In addition, there is also 10 cm of concrete.
Table C-1 Calculated Radiation Dose Rates For Loss of Shield Water Decay Reactor Reactor Reactor Reactor First floor First floor direct
) Time room direct room direct radiation room scattered room scattered direct radiation radiation v/ radiation radiation radiation R/hr @ 10 R/hr @ 1000 R/hr @ 10 R/hr @ 1000 hour0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br /> hour R/hr @ 10 R/hr @ 1000 hour hour hour hour 1 minute 52 64 0.048 0.060 5.06 6.23 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 12 23.7 0.011 0.022 1.17 2.31 1 day 1.1 8.9 0.001 8.3E-3 1.07E-1 8.67E-1 1 week 0.13 3.7 1 E-4 3.4E-3 1.27E-2 3.6E-1 1 month 0.026 1.4 3E 5 1.3E-3 2.53E-3 1.36E-1 For persons outside the building, the radiation from the unshielded core would be collimated upward by the shield structure and therefore, would not give rise to a public hazard.
The method of calculation is as follows:
The core, shut down and drained of water, was treated as a bare cylindrical source of 1-Mev photons of uniform strength, its dimensions were taken to be equal to those of the a.:ive core lattice. The' source strength as a function of time was determined from Way and Wigner's
[C1) data on fission product decay (Eq.1). No accounting was made of sources other than fission product decay gammas (i.e. activation gammas from the steel cladding and the aluminum grid plates) or of attenuation through the fuel element end pieces and upper grid plate The first of these assumptions is optimistic, the second conservative; the net effect is conservative. The conservative assumption of a uniformly distributed source of 1-Mev photons was balanced by not assuming any buildup in the core.
f.-
t i
C-1
i m - An approximation of the fission product energy release term is taken as:
!l l 42 ;
Mf G(t) _= 1.26 t (1)
.where:
G(t) = energy release in MeV/sec-fission t = time after fission in seconds-By integration the total source term is
<+r S(t,7) = 3.1x10 P. F (t)dt.
= 1.95x10"P,{1 - [1 + T/t}*2}t*2 where;, .. -
S(t,T) - = energy release in Mev/sec-watt,
- P. - = reactor power, watts TE ' = period of time at power.
The vc!umetric source of 1 MeV photons is S(ti,T)
S' = nr,2X, The direct dose rate at a point outside and on the axis'of a cylindrical source is given by:
Q' . -
Ds = S, c'" 2nrdrdx kaa 4n R' where:
S, = source strength in photons,1 MeV/cm*-sec, 2
K = flux-to-dose conversion factor,5.77x10 5photons /cm -sec per rad /hr.
2nrdrdx = dV = Cylindrical volume element r, = core radius,26 cm .
- x. = core height,38 cm
= core attenuation coefficient,' O.207 cm4 pe R = distance from volume element to receiver, em z = slant penetration in core = xR/(a+x), cm a = distance from top to core to receiver,640 cm
, ! For distances far from the core (i.e., for a > r, and x.) the above expression reduces to S.rc -
p, a 4 p, a2K,(, _ ,.,,,,
The scattered dose rate was calculated from [C2]
D, = 6.03x10 2
) Z 1,CQ" p A K(E)x' il A
C-2=
)
-, where:
= Density of scattering material, concrete,2.3 g/cm2
[V T r.
Z/A l.C
= Ratio of average atomic number to atomic mass of the scatterer, (= 0.5),
= Incident current times section of beams, photons /sec K. = Photon current to dose rate conversion,2.75x108 photons /cm2-sec per rad /hr, E = Energy of scattered photon, MeV x = Distance from scattering point to detector,400 cm, Q. _1 So
= ,
popqc os0o /cos0,)60 po,pi = Aitenuation coefficient in scatter for incident and scattered photons, cmd, 0.146,0.292, Oo,0, = Incident and scattered angle (measured from the normal to the scatterer),0, 25 degrees, do,dO = Differential Klein-Nishina scattering cross section, cm2/ electron-steradian it was assumed that all of the source photons that exit the top of the reactor pool were incident normally to the concrete ceiling (i.e., Qo = 0) at a point directly over the core, thus loc = sow where:
2 p So = S,n -
( .
g, ,
.yl(ro -xi)+rifr$+xi). n 1
- l. . (ri + xi)(ri + y$) _ 2 2n and to = Distance from core to top of the pool, ~ 6.4 m xo = Half wid'h of the pool. -
yo = half length of the pool, S ,r., ,., have a: ready been defined The energy of the scattered photons is given by E, Eo 1 + Eo(1 - caso)/ 0.51 where E is the incident photon energy (1 Mev) and e is the scattering angle = p -(Oo + 0,)
The differential scattering cross section is given by Sa r, ' 'E E. E '
=
- - (-smo )' + (-)s BD 2 .Eo Eo Ea .
O were r is the classical electron radius = 2.818 x 1043 cm.
C-3
F .
t
- APPENDIX D -
Determination of Soil Activation Outside of Reactor Tank .
D.1 ' NEUTRON FLUX ATTENUATION a
' Purpose L .
Calculate the penetrating ~ neutron flux for the Veterans Administration (VA) Medical
. Center. ncelear reactor. The results of the flux calculation will be used to estimate soil and .
ground water activation. l Assumptions 1, The non;iydrogenous materials are less than 5 relaxation lengths thick. This assumption -
is reasonable since these shiolds have a relaxation length ofp,t a 3.4.
= 2. ' The hydrogenous materials are least than 50 centimeters (cm) between the source and the shield. Since there are 45.7 cm of water and 5 cm of Gunite concrete between the
~ core and the side of the steel vessel, this assumption should be considered valid.
8
- 3. ' The fission neutron spectrum flux on the outside of the core reflector is 3x10" n/cm - ,
f . sec, [D1). .
4.. - The fission neutron spectrum flux on the bottom of the core is 1.2x10" n/cm'-sec. [D1).
- 5. - The reactor vessel shell is assumed to be made of carbon steel.
3
,n 6. - . The thickness of Gunite was neglected in these calculations. This is a conservative
-( assumption, because flux calculations will be slightly over estimated by not including the 5 cm thickness of Gunite.
Inputs 4
- 1. Technical information found in reference D6.
- 2. - Attenuation coefficients found in reference D2.
.~
. 3. Equations and methods found in reference D4.
- 4. Calculations where performed for a one use, using MathCad Version 5.0 plus by
'Mathsoft Corporation.
g . Method The attenuation coefficients are based on an idealized fission source embedded in an infinite homogeneous hydrogenous medium ato which a nonhydrogenous component in the form of a
- homogeneous slab of thickness t is niserted. According to with reference 3 experimental results
- have shown the attenuation is related to the flux with the slab in position (i.e., at distance x of
, hydrogenous medium plus a thickness t of the nonhydrogeneous component) to the flux at a -
4: - distance x from the source, witnout the slab, by
& ** &, e*"
l VGJ
' [4)'where pn is the removal coefficient and has'a constant characteristic cf the nonhydrogenous
' ,Q compone'nt for a given fission neutron energy spectrum. Two important requirements must be D-1
- i
[T j jmet for the removal constant to be valid. First there must be at least 50 cm of hydrogenous -
t rnatorial between the source and the nonhydrogenous component. Second, the thickness t
-_must be such that p,t is less than 5.- ~
- To calculate the removal coefficients for various compounds, the following empirical formulas (in units of em'/g) have been developed to permit interpolation between measured -
- Values.
N p--
.206X O YZ"
[4} where A and Z are the atomic mass and atomic number, respectively, for the elements of
' concern.
< Table 1 Removal Coefficients and Attenuation Lengths ,
Removal Coefficient Attenuation Length Material - - p,(cm4) 1/p,(cm)
-Water 0.103 - 6.7 ;
Paraffin - 0.106 9.4 fron .
0.158 6.34
- Concrete (6% H,0) 0.089 11.3
_ Graphite (density 1.54) 0.079 12.7 _
g --Taken from Protection Against Neutron Radiation, NCRP Report No. 38
=t
The steel shell of the reactor vessel was assumed to be carbon steel. Therefore, using reference 2, a compound specific removal coefficient was calculated (see section D.2 ).
According to reference 2, carbon steel is made of 1% Manganese,0.9% Carbon, and 98.1%
Iron. These~ weight percents were used to calculate the removal coefficient for carbon steel.-
Results -
The results of the calculation suggest the fission spectrum flux is reduced by over 99% at the vessel soll interface. Table 2 shows the relative geometry and thickness of materials used to calculate the reduced flux Table 2 Reactor Vessel Materials and Geometry
- Side ' Bottom Flux f.
Material - (cm) (cm) - n/cma-sec L Water : 45.7 - 61.0 Side (3x10")- '
. Gunite -- 5.0 : 5.0 Bottom (2.1x10")
Steel.- 0.64 0.64.
Concrete 25.0 38.0 The calculated neutron flux at the surface of each side of the reactor vessel is estimated at
} - 2.6x107 n/cm'-sec, and 6.9x108 n/cm 2-sec at the bottom.
D-2
. -..__ - c_ . _ _ _ ___ o
i i
! i I[
i The fotowng calculations are based on an idearuted Gesen source embedded h en merdle
'( 'i-- ;za hy*ogenous medium into wtuch a nonhydrogenous component in me form of a homogeneous slab of thickness xis'rmerted.
SElI1E11 0a ses .
p g
- l- p2 *III P $ ** I'70 k I.
.m .m .
11 y g . 0.103 em M 2 0 089cm + 2.s 1.210 ,g 2.,,,4 x4: 3&am x 5 ;* 8I'*"
! x g := 45.7 ce x2.: 25 em x3:0.64 em .
%' - AttaP= '--, Coef",c'est for Carter; Steel 1
Cason steelis made of 1% Mn, .9% C, and 98.1 Fe reference D3 y Z g =0.0125 + 0.00096 + 0.98l-26 l Z g = 25.76I .
i A g. 00135 +.0.000912 0.98156 A g = $$.497
' 1
.1 1 8 ** reference D4 igerp :=0 4 206 A g .Z g**,
5 i= p 3 :spperpp 3 p 3 = 0.161 cm
N- b:-n of React'of Flur at Outside Surface of Reactor Vessel from the Side e, ;* p g x g + p 2'"2 '" k 3'" 3 i
j e, = 7.033 -
(,s4gs 2
+ , - 2.641 10' c d .
^ ' ' "- - of Rex;e- Flur at Or ' '- St.'eee of M--
-V ' "n, the " ^^ --
efb;* M l'*5 + H 2'"4 + p 3'*3 eb"'188
_k
(,:s4ge**6 l ( g "NI M ' .ses*I Da3 -
D.2 SOIL ACTIVATION Tr. determine the radioactivity of the soil induced by reactor neutrons, the volume was determined of an annulus 1 cm thick at the level of the top of the reflector directly outside of the tcnk and with a height equal tc the height of the core. This was labeled side. In addition the volume was determined of a 1 cm thick slab directly below the reactor core a'id having a di: meter of thtt of the reflector. This was labeled bottom The composition of the commoner chemical elements in the earth's crust was taken from the Heidbook of Chemistry and Physics [D3). From c. book entitled " Soil Survey for Doug!cs and Scrpy Counties, Nebraska", U.S. Dept. of Agriculture Soil Conservation Service, dated December 1975 the soil was determined to be Monona silt loam with a permeability of 0.6-2.0 in/hr. Where permeability is defined as an estimate of the rate at which saturated soil transmits water in a vertical direction under a unit head of pressure. Groundwater hydrology is presented with more detail in Sec. 2 A.1.
We have been unable to find the depth to the water table, but since the original 30 ft test boring at the center of the reactor location indicated no water table was encountered (SAR page 2-10),
the total number of nanocuries that reached the 30 foot depth after the soil was saturated with water was calculated. For the purpose of radioactive decay it was assumed that the time started when the leached water reached the depth of the level of the bottom of the core for the side calculation and 1 cm below the bottom of the tank for the bottom calculation. It is not readily apparent that the activated products within the soil about the reactor tank would have any method of transport to the water table. The reactor tank within a poured concrete l basement floor. The closest area of exposed soilis a minimum of 28 feet radially form the top of the tank. It is assumed that all the soil activity was removed by the water. Using the maximum permeability value of 2.0 in!hr, tne time for the combined radioactivity to reach the 30 foot level was determined. The above volumes were then multiplied by the soil density, the nCi/g soil (at saturation of the radioisotopes selected) and the appropriate decay factor to give the total nanocuries in the volumes specified above. Results of the calculations are shown in Table 3. The results show that for this worst case scenario only 6.32 mci would reach the 30 foot level.
The above analysis is based on assuming that water reaches the area around the reactor core and flushes all of the induced radioactivity into the water table. Table 4 shows the quantity cf water necessary to dilute the radioactivity of the individualisotopes to the ef'luent concentrations listed in Table 2 column 2 of Appendix B to 10CFR 20.1001-2401. How much of the element is mobilized and actual!y reaches the water depends upon the interaction between the phases of the soil i.e. liquid, solid and gas. For example, the general roles goveming the mobilization and fixation of Fe are that oxidizing and stkaline conditions promote the precipitation of Fe, whereas acid and reducing conditions promote the solution of Fe compounds. The retsased Fe readily precipitates as oxides and hyd oxides, but it substitutes for Mg and Al in the minerals and often complexes with organic ligands. The solubility of Fe in soils is extremely low in comparison with the total iron content [D7]. Consequently, even in the worst ase scenario, after chemical reactions, radioactive decay and dilution with the water table the amount of radioactivity that would be available t0 the public would be well below 10CFR20 environmentallimits.
I o.4
i p) a M
l
- Appendix D References D1 J.E. Larson, " Calculated Fluxes and Gross Sections for TRIGA Reactors", General Atomic GA-4361, Suppl. B.1966;.
D2' NCRP No. 38. " Protection Against Neutron Radiation". National Council on Radiation Protection and Measurements Washington DC.1971.
.D3. CRC Handbook of Chemistry and Physics 61" Edition.1981. q D4 Arthur B. Chitton, J. Kenneth Shuitis, Richard E. Faw. " Principles of Radiation Shielding". Prentice-Hall.1984.
.D5 James E. Tuner. " Atom, Radiation, and Radiation Protection". Pergamon Press.1986.
D6 - Section 3.2. " Reactor and Reactor System". Construction Permit for VA Medical Center.
D7 -. A.Kabata-Pendias and H. Pendias, " Trace Elements in Soils and Plants *, CRC Press, Boca Raton, Fl.1984.
v.0 I
b
- D 5.
- ,-~
,I L $ / .k /
f
(./ V
\s/
TABLE 3 SO!L ACTrVATON [AT SATURATON) FROM OVAuC TRGA MARK I NUCLEAR REACTOR
! CORRECTED FOR DEC AY
{ ATOMS / RADOACTIVE SIDE rc/ BOTIOM rCV HALFUFE SIDE BOTTOM A TOMO TARGET % NATURAL CROSSECTON g SOIL ISOTOPE g Soil g SOfL HOURS TOTALrCI YOTAL to WEIGHT % ATOM % VOLUME *A WEfGHT ISOTOPE ABUNDANCE 8ARNS ELEPENT 0.00E+00 0 00E+00 0204 0.00016 3 58E+ 19 O-19 0.004 0.001 7.47E4J OXYGEN 46 60 65 55 9*.92 15 9994 O-18 l
0.10700 1.73E+20 SW1 12.9TT 3 444 2.62E+00 3 26E41 1.59EC2 StuCON 27.72 21.22 Om0 29 9738 SL10 3 100 l 0 00E+00 0 00E+00 A147 !C0 000 0.23300 1.81 E+ 21 Al-28 296.995 78.818 3.75E42 ALUMINLW 8.13 6.47 0.77 26.9815 1.45 E+ 18 Fe-59 1.227 0.326 1.07E+03 5.72E+ 05 1.1 t E + 03 IRON 5.00 1.92 0 68 57.9332 Fe-58 0.200 1.20000 5.33E+ 19 Mg47 1.349 0.350 1.58E41 0.00E + 00 0.00E+ 00 MAGNES 1UM 2.09 1.84 0.56 25.9826 Mg.26 11.010 0 03600 1.90E+16 Ca47 0.009 0.002 1.09E+ 02 3.04 E + 03 6.33E+00 CALCIUM 3 63 1.94 1.48 45 9537 Ca-46 0.004 0.70000 Ca49 0.659 0.175 1.45E41 0.00E + 00 0.00E +00 3.63 1.94 1.48 47.S525 Cu8 0.187 1.10000 _852E+17 CALCIUM 5.36E+06 1.82E+ 04 100.000 0.40000 7 41E+20 NA,24 208.296 55.278 1.50E+01 SO"MJM 2.83 2.64 1.60 22.9898 NA-23 1.46000 2.56E+ 19 K-42 26.282 6 975 1.24 E + 01 3 65E+05 1.39E + 03 biASSILM , 2.59 1.42 2.14 40.9618 K41 6.730 l 6.30E+06 2.07E+ 04 TOTAL ACTIVtTY SOlt rCI =
h O BOTTOM OF TANK SiOE OF TANK 2.60E+07 '
hEUTFON FLUX (rVem2'sec) = 6.90E+ 06 NEUTRON FLUX (rVerrQ'sec) =
1.282+06 . VOLUME OF SOtt (cm3) = 9.33E + 03 VOLUME OF Sort (ctrG) =
4.,"+
P 05 MASS OF SOfL (g) = 3.53E+ 03 MASS OF SOlt (g) =
" 43 TIME FOR WATER TO TRAVEL 260.2cm (hr) = 51.22 i !ME FOR WATER 10 TRAVEL 365 Ocm (hr) =
+
.p).
( TABLE 4
'%J -
10CFR20 EFFLUENT LIMITS FOR Soll ACTIVATED ISOTOPES IN HYPOTHETICAL WATER TABLE l CORRECTED FOR DECAY 10CFR20 WATER VOLUME
- RADIOACTIVE SIDE BOTTOM WATER LIMITS SIDE BOTTOM ISOTOPE TOTAL nCl TOTAL nCi uCi/ml mi ml 0-19 0.00E + 00 0.00E+ 00 '
si.31 3.26E-01 1.59E-02 1.00E-04 3.26E+00 1.59E-01 Al-28 0.00E + 00 0.00E + 00 Fe-50 5.72E + 05 1.11 E+ 03 1.00E-05 5.72E + 07 1.11 E + 05 Mg-27 0.00E + 00 0.00E + 00 Ca-47 3.04E + 03 6.33E + 00 1.00E-05 3.04E + 05 6.33E + 02 Ca-49 0.00E + 00 0.00E + 00 NA-24 5.36E + 06 1.82E + 04 5.00E-05 1.07E + 08 3.64E + 05 K-42 3.65E + 05 1.39E + 03 6.00E-05 6.08E + 06 2.32E + 04 A
- volume of water necessary to dilute isotope to effluent concentration listed in Table 2 Column 2 of Appendix B to 10 CFR 20.1001-2401 s
'w D-7 i