ML20073R110
| ML20073R110 | |
| Person / Time | |
|---|---|
| Site: | University of Texas at Austin |
| Issue date: | 05/31/1991 |
| From: | TEXAS, UNIV. OF, AUSTIN, TX |
| To: | |
| Shared Package | |
| ML20073R108 | List: |
| References | |
| NUDOCS 9106050244 | |
| Download: ML20073R110 (308) | |
Text
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3.3 Spare Allueittion 1.4 Hearton Bay aint opei at ion Cont ol 37 l 13 1S suppoit l'a c i l i t i c < 1.5.1 Ile a l t h I'h y '. i c . 1 abot at on y ! 11 1.5 , Roople llaiidl i ng 1. abo t'a t o t y 3 13 f$.i 1:1 f luent Cont t'o l !- I $ 3 15
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SAR 5/91 k , EtC111'f] DEC 4.1.? Pi ort;it !;e r a t. t v e Te rtpe r a t u re Coe i f ic ii nt a 26 4.1.2.1 Codes Used for Calculations 4-28 i-4.1.7.? Zrli Model 4-?8 4 1.7.3 Calculations 4 31 4.1.3 Steady State Peactor lower 4 32 4.1. 3.1 Ent ranc e Loss 4-34 4.1.3.7 Ixit loss 4 30 ' 4.1.3.3 Los s 'Ihrough Por t ion o f Cluinne 1 Adjacent to Lower Reflector 4-36 4.1. 3.4 Loss 1hrough Port ion of Channel Adj acent to Uppet it e f l e c t o r 4 3h 4.1.3.5 Loss Through Each increment of the Channel i Adjacent to the l'ueled Portion 4 37 of the E len:en t s 4.1 3.6 Accelerat ion h im 4-37 o-40 4.1. 3. 7 i ric t ion Te rm 4.1.3.8 Gravity Term 4-41 4 45 4.1.3.9 f1omenclature 4.? Nucleat f>esign and Evaluation 6-47 4.2.1 Reactiv:ty Lflects 44/ 4.2.? Lvaluation of Nuclear Design 4 51 4,3 The mal and }{ydraulic Design 4-5? 4.3.1 D(sign bases 4-52
- 4. 3. 7 The rmal and llydraul ic De u i r.n Eval uat i on 4-54 4.4 Mechanical Design and Evaluation 4 54 4.4.1 General Description 4 - 5 '.
4.4.? Hellector Assenbly 4 57 4 5/ 4.4.1 Grid Plates 4.4.4 Safety Plat e 4 $9 4 59 4.4.5 f ue l -Mode rat e r 1:lement s 4.4.5.1 Instrument Fuel Litments 4 61 4.4.5.2 Evaluation of fuel Element De <. lon 4 62 4 64 l 4.4.6 ficutrun Source and Holder 4.4./ Graphite Dummy Elements 4-64 4.4.8 Control Syst(m Design 4 64 4 67
- 4.4. 8.1 Cont rol Road Drive As se mblie*
4.4.8.? Regulat ing Rod and St epping Mot or Drive 4 67 4 4.8.3 Transient Road Drive Assembly 4-69 4 /2 4.4.h.4 Lvaluat ion of Cont rol 11od Sv< t em 4 5 Safety Settings in Relat ion to Saf ety limi t s 4 73 4 74
Reference:
S Peactor Conlont System. 5-1 5.1 Design luu.e s 51 5 1.1 Reactor Core llent R e niov a l 5-1 I S.I.? Peactor Pool Heat Removal 52 5.1. 3 th a t E>: c h a n ge Design Bane > 5? 5.1 4 Water Purification Bases 56 11 f.
l n SAR 5/91 i I-e EttLinD E!!Le l 5.2 System Design 56 I 5.2.1 Cooiant System 56 5.2.2 Purification System 5 10 5.2.3 Vater System Instrumentation 5 10 1 5.3 Vater System Design Evaluation 5 12 ) References 5 15 1
- 6. Instrumentation and Control System..... ..... ........ .... 6 1 6.1 Design Bases 61 6.1.1 NM 1000 Neutron channel 65 6.1.2 NP 1000 Power Safety Channel 66 6.1.3 Reactor Control Console 66 6.1.4 Reactor Operating Modes 68 6.1.5 Reactor Scram and Shutdown System 6 13 l_
! 6.1.6 Logic functions 6-14 l 6.1. I Mechanical llardware 6 16 l 6.2 Design Evaluation 6 17 References 6 18 i
- 7. Design Features and Auxiliary Systems. . , .71 i 7.1 Design Bases 71 l 7.2 Desl6n Features 71
- 7,2.1 Reactor Pool and Shield Structure 7-2 1
7.2.2 Reactor Bay Vent 11ation Design 7-2 ! 7.2.3 Fuel Materiala 7 10 7.2.4 Safety Feature Evaluation 7-10 ,. 7.3 Auxiliary Systems 7 11 l 7.3.1 1.ife Safety and Fire Protr et ion 7 11 " 7.3.2 Passive Fire Protection Elements 7-11 7.3.3 Active Fire Protection Elements 7-12 7.3.4 IIVAc System 7 13 7.3.5 Communications and Security 7-13 7.3.6 Compressed Air and Deionized Water 7-14 7.3.7 Utilities 7 14 7.3.8 liar.ard Liquid Waste 7 14 7.4 Confinement Design Evaluation 7 15 t 7.4.1 Release of Nitrogen 16 and Argon 41 7 15 l- 7.4.1.1 Nitrogen 16 Activity in React or Room 7 15 i 7.4.1.2 Release of Argon 41 from Reactor Pool Water 7 19 l 7.4.1.3 Activation of Air in the Experimental Facilities 7+25 l ,. 7.4.2. Evaluation of Argon 41 Release '7-27 References 7 29 l O l l j 111 ) i
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1 i' i SAR 5/91 J 1 S.rs.Lisal fnts i
- 8. Experiment and Irradiation Facilities..... . . .8 1
( l 8.1 Standard Experiment Facilities 81 l 8-1 l 8.1.1 Central Thimble 8.1.2 Rotary Specimen Rack 81 l 8-1 8.1.3 Pneumatic Specimen Tube 8,1.4 Beam Tube Facilities 8-4 l 84 8.1.4.1 Tangential Beam Ports 8.1.4.2 Radial Beam Ports 84 i 8.1.4.3 Beam Tube Plugs 6-4 l' 8.1.3 Evaluation of Materials in Experiment Facilities B7
-8.1.5.1 Double Encapsulation 87 j 88 8.1,5.2 Explosive Materials j
8.1.5.3 Fueled Experinnents 8 10 l 8.1.5.4 Airborne Experiment Releases 8 11
- 8 12
! 8.2 Special Experimental Facilities 8.2.1 Reactor Core Facilities 8 12 8.2.1.1 Three Element Feature 8-12 i 8 12 8.2.1.2 Six Element Feature l 8.2.2 Gamma Irradiction Facility 8-13 g 8.2.2.1 Itazard to the Pool Water System 8-13 8.2.2.2 llazard to 1.aboratory Personnel B 17 l 9 8.2.2.3 Point Source Shielding Calculations 8.3 Other Experiment Facilities 8.3.1 Suberitical Reactor and Moderators 8 17 8-20 8 20 8 20 l 8.3.2 14 MeV Neutron Generator References 8-22 i l l
.. ... 9 1
- 9. Radioactive Materials and Radiation Measurement l
9.1 Radioactive Materials Control 91 ; 9.1.1 Reactor Fuel 9-1 i 9.1.2 Reactor Components 91 ! 9.1.3 Experiment Facilities 94 l 9.1.4 Activated Samplea 94 9.1.5 Radioactive Waste 94 9.1,6 Other Materials 9-4 9.2 Radiation Monitoring 95 9.2.1 Minimum Proceduren 95 9.2.2 Monitoring Techniques 9-6 9.2.3 Management Surveillance 9-6 9.2.4 Frequency and Accuracy 96 9.3 Instrumentation 97 9.3.1 Fixed Area Monitors 9-7 9.3.2 Airborne Radioactivity Monitors 9-7 9.3.3 Survey and 1.aboratory Instrumentation 98 9.3.4 Liquid Effluent Sampling 9-8 9 9.3.5 Range and Spectral Response 9.3.6 Calibrations 9-9 99 iv
i SAR 5/91 i j O Sfflin11 l'ent 4 9.4 Records 9-9 l i 9.5 Evaluntton of Monitoring System 9 10 l 9.5.1 Particulate Air Monitor 9 10 [ 9.5.? Argon 41 Monitor 9 15 j 9.5.1 Area Radiation Monitors 9+18 j 9.5.4 Monit or Availabilit y Condit ionn 9 19 References 9 21 l b
- 10. Conduct of Operations.. .. . . .. .. . . . 10-1 I 10.1 Facili ty Adininist r at .lon 10-1 10.1.1 Organirat_ ion 10 1 10.1.1 1 Structure _
10 1 1 10.1.1.2 Executive Vice President and Provost 10 1 l 10.1,1.3 Vice President for businesn Affairs 10 1 l 10.1.If4 Director of Nucicar Engineering Teaching 1.aboratory 10 1 10,1.1.5 Nuclear Reactor Committee 10 3 [ 10,1,1.6 Radiation Safety Officer 10 3 10.1.1. 7 Itadiat ion Sa fety Commit t ee 10 3 l 10.1.1.8 lteactor Supervisor 10 3 l I 10.1.1.9 licul t h l'hysic ist 10 4 ! 10.1.1.10 Profeustonal and Classified Staff 10 4
- 10.1.2 Quali f icat ions 10 4 10.1.7.1 Job Descrint ions 10 4
! 10,1.7.2 racility Director 10 5 10.1.2,3 Henct or Supervisor 10 5 l 10.1.2.4 llenith Physicist 10 5
- 10.1.7.5 Professional and Clanulfied Staff 10-5
_10.1.3 lteactor Operations 10-5 l t 10.1.3.1 Stalfing 10 6 ; 10.1.3,2 Procedures 10 6 l j 10.1.3.3 Experiments 10-7 ! 10.1.4 Actions and Reports 10 7 l 10.1.4.1 Operating Reports 10 7 ! 10.1.4.2 Safety _l.imit Violation 10 8 . 10.-l.4.3 Release of Radioactivity '10 8 } 10.1.4.4 Other Reportabic Occurrences 10 9 l 10.1.4.5 Other i<eport s 10 9 ,
.10.1.5 Records 10 9
, 10.1. 5.1 1.i f et ime Records 10 10 10.1,5.2 Fivo Year Period 10 10 j 10._1_.5.3 Training Cycle 10 10 10.7 Operator Requallilentton -10 11
- 10. 7. I liit rodue t ion 10 11 10.2.2 Operator 1.icense Staturs 10 11 10.2.3 Requalification Program P.as.es 10 11
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SAR 5/91 O Section D Lt 10.2.4 Requalification Program 10 11 10.2.5 Schedule 10 12 10.2.6 List. of Subjects 10 12 10.2. 7 On the Job Training 10 12 10.2.7.1 List of Annual Training Taska. 10 12 10.2./.2 List of Training Tasks; system malfunctions 10 13 10.2.7,3 On the Job Training Checks 10 13 10.2.8 Evaluation 10 14 10.2.9 Records 10 14 10.3 Radiological Protection Program 10 15 10,3.1 Managenient and Policy 10 15 10.3.2 Responsibilitles 10 15 y i 10.3.3 Organizational Access 10 15
- 10. 3.4 Equipment and Sup;illes 10 16 10.3.5 Training and Safety 10 16 l 10 17 l 10.4 Fire Protection Program 10.4.1 Facility Fire Protection Elements 10 17
-10.4 ;2 Facilit y Fire Prot ect ion Cont rol 10 18 10.4.1 Fi re Saf et y Assurance 10 18 10'.S Security and Emere,ency Plans 10 18 10.6 Quality Assurance Program 10-19 i: 10.6.1 Introductton 10 19 10.6.1.1 Purpose 10-21 j.
10.6.1.2 Responsibility 10 21 l 10 21 ! 10,6.1.3 Organiration 10.6.1.4 Documentation 10 23 ( 10.6.2 Quality Assurance Controls 10 23 l 10.6_.2.-1 Design Controls 10 23 10.6.2.2 Procurement cont rols 10 23 10.6 ?.3 Document Control 10 25 10.6.2.4 Material Control 10 25 10.6.2.5 Process control 10 25 10.6.3 Inspection and Corrective Actions 10 25 10.6.3,1 luspection' Program 10 25 10 6.3.2 Test Program 10 26 10.6.3,3 Measuring and Test Equipment 10 26 10.6.3.4 Non Conforming Material and Parts 10-26 10.6.3.5 Corrective Action 10 26 10.6.3.6 Experimental Equipment 10 27 10.6,3.7 Replacement.s, Modificationn, or Changes. 10 27 10.6.4 Records.and Audits 10 27 10.6.4.1 Quality Assurance Records 10 27 10.6.4.2 Audits 10 27-O vi
SAR 5/91 L l i 9 Ett11.011 l'att 10.7 Stastup Program _10 29 10.7.1 Storage of fuel and Acquisition of temponents 10-30 10.7.2 Tests of Systems Bef ore Core Loading 10 30 10.7.3 Core 1.oad for Initial Criticality 10 30 i 10.7.4 Tests Subsequent to Core Criticality 10 30 10.7.5 Acceptance for Operation 10 31 l I References 10-32
- 11. Safety Analysis. ... . ........... . . .. . ......... 11 1 11,1 Reactivity Accidont 11 1 l 11 1
! 11.1.1 Summary ! 11.1.2 Analysis of 2.8% insertion at I LW 11 2 j 11 1.3 Analysis of 2.8% Insertion at 880 kW 11 8 11.2 Lonn of Reactor Coolant 11 11 l l 11.2.1 Summary 11-11 11.2.2 roc 1 Temperature and Clad Integrity 11 15 l-- ! 11.2.3 After Ileat Removal rollowing Coolant Loss 11 18 11.2.4 Radiat ion Levels 11 22 l 11 26 j 11.3 risnion Product Release ' 11.3.1 Fission Product Inventory 11-26 11,3.2 Fission Product Helease Fractions 11 26 j 11.3.3 Downwind Dose calculatton 11 28 i 11 3.4 Downwind Dowe. 11 30 ! Re f e re nc e s. 11 32 ll l t a } t l 1 i a 4 (O t j Vii f p 3- ,)
S Alt 5/ 91
- 1. i s t of ligutes 1,'e ty D gult S t. a t e of '14 xas Count les ?-2
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/-? Travia. County ?3
/-1 City of A u *. t i n 24
/4 b a l r o t ic ', Poseatcli Centet ?-5
/S Tsavls Count y 1911 Cete.us Tract bouiuln i t e s 2S
/~6 City of Austin Census Itact boutula r ie r. 2-9 Austin Clinvitology Data 7 il
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?-8 Aust lii Wiial ltose Data ?-12
?4 Texan lotnado Irequencien 2 16 1-10 Texat, 1101 i i c ane Paths. ?-17
/.11 1.ocal l'unnel Cloud Sitings 2-18
/.1/ halconet lault Zone 2 20 7-11 Texas, l.a t t hquake Data 7 ?!
2 14 1.ocal Water Aquifcis 2 27
? lha itosentch Cent er Area 1960 2 25 7.lSh Balcone+. Hesevich Center 1960 2 26
/.16 tial c otir % Pescut ch Cetit or 1990 2 27 11 141 T1. M i t e I'l aa f oi hab ones Ret.eatch Centes b2 3 ';
1-? i:levat ton I'lans l- 1 Building Section Plans 36 l4 ) inst 1.evel Floot Plan 38 t-5 Second 1.evel l'loor Plan 3-9 t6 Thlid 1.evrl l'looi l' l a n blo 1/ I'ourth 1.evel l'loor Plan 3-11 18 1(e a c t o t Bay Area 3 14 61 Ph9se Dianiam of the Zinconium ilydrogen S;o. tim 43 47 rquilthrium llydrogen Ptessure Versus Tempetatute for Z i r coni uin -ilyd ro g e n 4-5 4-3 Strength of Type 304 Stalnless S t c e l a v. a i unct ion of Temperature 4-6 6-4 St rengt h and Applied St iess as a l'unt ion o f E.luillht ium liydr ogen Di ssoc i at ion Ptessure 4-8 Tempetatute, 4'. 11adial Power Dist ilhut ion in t he ll Zril luel l'. l e n e n t 4-10 46 Axial Powet Dist ilbut inu in t he U Zrit l'oel I,l en e r t 4 11 4/ Subc ooled hol 11:y*, lle n t Transte for Water 4 12 48 Clad lemperature at M i dpoi nt of Well Bonded l'uct Element 4 - l 'l 49 Fuel Body Temperatutes at Midplane of Well honded fuel Llement Atlet a Pul s. c 4-14 4--10 Surface lle a t I' lux at Midplnoe at Wel1 - Bointed l'ue1 1:1ement Attei a Pulse 4-15 4 '1 Susface lle a t Ilux for Standard Non Gapped 4-17 (hgap-SOO) Fuel Element Aften a Puls.e 4 1? Surface lle n t Flux !on St aretar d Non-Capped 4 18 (hgnp-375) Fuel I:lement Aster a Pulse 4 13 Sul f are llent 1lov for e candard Non-Capped 4 19 (hgyp-25D) F';n l 1:l eme n t After a Polw viii
. I l SAR 5/91 1 i ypgg ! ELLut i 4-14 Surface llcat Flux at Midpoint Versus Time for :Itandard Non Capped Fuel Element After a Pulse 4 20 i l ! 4 15 Transport Cross Section for Ilydrogen in Zril l and Average Neutron Spectra in Fuel Element 4 27 4 16 A Comparison of Neutron Spectra Between Experiments and Several llydrogen Modelu 4 29 ) '. 44 11 Effect of Temperature Variation ! on Zirconium flydride Ne utron Spectra 4 30 ; 4e in Prompt NegatIvo Temperature Coef ficient . l Versus Average Fuel Temperature for TRICA 4+33 j' l 4 19 Cencral fuel Element Configuration for Single Coolant Channel in the TRICA 4+35 { 4 20 Exper imentally Determined vapor Volumes I for Subcooled Boiling in a Narrov Vertical Annulus 4 38 ' 4 21 Cross Plot of Figure 4-?O Used in Calculations 4 39 4-72 Plot for which DNd Ratto is 1.0 l of Maxirnum llent Flux Versus coolant Temperature 4 44 i i 4-23 Estimated Reactivity 1.oss Versus Power 4 49 ! l 4 24 Estimated Maximum B Ring and l ! 4 50 ! F Average Core Teroperature Versus Power 4-25 Reactor, Reflector, and Shielding 4 55 l 4 26 TRICA MARK II Reactor 4 56 i l l 9 4-77 4 28 Core Arrangement TRICA Stainless Steel Clad Fuel Elernent with End Fittings 4 58 4 60 4 63 l l 4-?9 Inst ruinented Fuel Element l j 4-30 Fuel-Iollowed Conttol Rod 4 66 t 4 31 Rack and Pinion Control Rod Drive 4 68 l l 4 70
- 4 37 Adjustable Transient Rod [
4 33 Transient Rod Operational Schematic 4 71 j 51 Coolant System 1.ayout 57 ; 5-2 Purification System 1.nyout 58 { 53 Water System instrumentation 5 11 i 61 Control System Block Diagram 6-3 i 6,7 Neutron Channel Operating Ranges 64 6-3 Layout of the Reactor Control Console 67 69 l; 6-4 Console Control Panels ' 6-S Video Display Data 6 10 66 Rod Control Panel 6 11 i 61 1.ogic Diagram for Control Syr. tom 6 15
-/ l Reactor Shield Structure 73 7? luodose Curves for Shield Structure 74
/-3 Reactor Bay Air Ventilation System 76 '
14 Reactor Bay Auxillary Exhaust System 7/
/-5 Schematic of Ventilarton Systems 7-9 f
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i I I i i SAll S/91 i . EltittT hrt J 81 Rotary Specimen Facility E2 ( 82 Pneumatte Transfer racility 83 8-3 Beam Tube Configutatton 8.$ i 84 Camma Irradiator 8-14 I H$ Special Txperiment Equipment 8 21 I I l 9-1 Radioactive Material Use and Storage Areas 9? < ! 9-? Radioact ive Mat erial Ef fluent Cont rol Syst ems, 93 l l 93 Activity Accumulation on Particulate Filte 9 13 ) i l 10.I Admini r,t rat l on 10 2 . I 10 2 Quality Assurance organizat ton 10 22 I 11 1 Calculated Pulse Shape. Energy, and Ten'perat ure 11 3 11-7 fuel Temperature Distribution before and Alter Pulse 11 12 } 11-3 ruel Temperature and Power Don 91ty i for Llement Cooling Times 11 13 -i j 11 4 U Zrli (1.6) Strength and Stresa Versus Temperature 11 14 j ll+S Cooling Times After Reactor Shutdown to j 1.imit Maximum Puct Temperature Versus Power Density 11 16 j i t ! I L ! h i I f l- i t i ; l b v
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9 1 I I t X
- t i I ' i SAR 5/91 ( i i 1.i s t of Tables ; L l- Inhle Du ( l-1 Principal Design Parameters 12 , I ! l ?l Travis County 1980 Populat ion Density Distribut ion 27 i l 22 1987 Meteorological Data for Austin. Texan 2 14 [ l ?.3 llintorical Meteorological Data for Austin, Texas 2 15 l l 2-4 Cround Water Activity 2 23 ! 25 Tank Sludge Sanples 2-24 { i 41 Physical Propet t les of Dolt a Phase U-Zrli 4-2 l 42 Ilydraulic Flow Parameters 4-32 l l 4-3 Typical TRICA Core Nuclear Parameters 4 47 i
- 44 Est imated Cont r ol Rod Net Worth 4 48
! 4-5 Estimated Fuel Element Reactivity Worth j Compared with Water an a Function of Position in Core 4 48 , 46 f.xpected Reactivity Effects '; Associated with Experimental Facilities 4 51 47 Comparison of Reactivity insertion Effects 4 52 48 1000 kW(t) TRICA Itent Transf er and liydraulic Parameters 4 53 49 Thorinocouple Specifications 4 61 ; l i e 4 10 Summary of Fuel Element Specifications 4-11 Summary of Cont rol Rod Design Parameters 4 12 TRICA Safety Settings, Reactor Coolant System Design Sumniary 4 02 4 65 4 73 59 l l 51 5-7 Ileat ' Exchanger; llent Transf er and livdraulle Parameters 5 13 l 7-1 Saturated Argon Concentration in Water 7 20 7? Volumes and Thermal Fluxes of Facilities ?,-26 , I 81 Material Strengths 88 ; 82 Container Diameter to Thickness Ratio 8-9 l 8-3 Calculated Isotope Release Values 8 10 ; 8-4 Micror.hield Data 8 16 l f 4-1 Significant Fission Products: contribution to Total Activity, Percent 9 12 9? Beta Emitting, Caseous Radionuclides of Interest 9 16 l i 10 1 Q l.ist Ior IMW UT TRIGA 10 20 10-? Responnibilities and Eey Personnel 10 22 10 3 Format for safety Related QA checks 10 24 4 Quality-Ashurance Program Audit Procedutes 10 28 i I O l xi _ _.a_,,_. _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _-
I SAR S/91 { O I.nhlt l'att i' t t 11 1 Reacttvity Transient Input Parameters 11 4 ; i 11 ? Reactivity Transient input Parameters 11 9 f i 11 3 Calculated Radiation Dose Rates for 1oss of Shield Vater 11 22 j 11 4 Noble Cas and Halogens in the Reactor 11 27 3 1 11 $ Assurved tireathtug Rates 11 30 [" j 11 6 Average Gamm a Ray Energy and Int ernal Dose 1:f f ectivity j- for Each Pission Product Isotope 11 30 l l 11-1 Doses from Fission Product Re l e a t.e s 11 31 ! 4 l 1 l -i i ; t f t l-i- i VO ' i
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SAR 5/91 Chapter 1 INTRODUCTION AND
SUMMARY
I reactor and The University of This report describes the TRICA Texas 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. A *rRICA type reactor was first operated between 1963 and 1988 on the main campus of The University of Texas at Austin. Subsequent operation experience included safe operation of the facility at stendy state thermal power levels of 10 kW and ?$0 kV, and pulse powers of 250 Mi operation of a TRICA reactor at the Italcones Resentch Center of The University of Texas is expected for steady state power levels of 1.1 HV(t) and pulse powers of 1700 MV(t). Sotne values used in this report represent the latest design patameters, or maximum values as a means of evaluating the safety of the system. For this reason, these values may differ from those quoted in other docuitte nt s or from those that will be toca sured in the operating reactor system. Saf ety analysis demonstrates sale operation at power leveln as high as 1.5 MW steady state and 8400 MW peak pulse power. Operation of the TRICA reactor is one function of the Nuclear Engineering Teaching Laboratory. Laboratory functions consist of faci 1itles and prograrts for the purpose of education, research, and technology development. 1.1 PRINClpAL DESIGN CRITERIA
=
The reactor will be operated in two inodes: pulse and steady-state Reactor power levels in the steady state mode will range up to and include 1.1 MW(t). pulse mode operation will take place by step reactivity insertions with the reactor initially at a power icvel less than 1 kW. The maximwn step reactivity insertion vill be 2.2% 61< /k ($3.14) which will produce a peak reactor power of approximately 1700 MW(t) with a p rornpt energy release of about 21 MW - s e c . A summary of principal design parameters for the reactor is given in Table 1-1. 1.? DESIGN lilGilLICitTS The reactor will be located in a reactor pool structure. Reactor cooling will be provided by natural circulation of pool water which is cooled and purified in external coolant circuits. Reactor experiment facilities will include a rotary specirnen rack, a pneumatic transler system, core irradiation tubes, and horizontal and vertical beam tubes. 1 I Manufactured by General Atomics, TRICA Reactor Division l l 1-1
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I l SAR 5/91 9-Table 1 1 PRINCIPAL DESIGN PARAMETERS , Reactor type TR1CA Mark 11 Steady state power (maximum) 1.lMW (1.5MW design) Pulse power (maximum) 2.2% 6k/k ($3.14) ! Fuel element design ; Fuel moderator material U ZrH I H/Zr ratio 1.6 (1.65 maximum) Uranium content 8.5 wt 2 Urantua enrichment 19.7% U 235 Shape Cylindrical ; Length of fuel 38 cm (15 in.) overall Diameter of fuel 3.63 cm (1.43 in) o.d. Cladding material 304 stainless steel Cladding thickness 0.051 cm (0.020 in.). Number of fuel elements Critical core ~64 i Operational core -90 Excess reactivity, maximum 4.9% 6k/k : Number of control rods 4 transient ( w/ air follower) 1 regulating (w/ fuel follower) 1 shim (w/ fuel follower) 2
-Total-reactivity worth of rods 8.7% ok/k Reactor cooling Natural convection-of pool water O
5 12
l l SAR 5/91 i The inherent safety of this TRICA reactor has been deinonstrated by . the extensive experience acquired froin similar TRICA systems throughout l the world. Porty.cight TRICA reactors are now in operation throughout the world and of these 31 are pulsing. TRICA reactors have more than 450 reactor years of operating experience, over 30,000 pulsen, and more than 15,000 fuel clercent years of operation. The safety arises from a large, prompt negative temperature coefficient that is characteristic of uranttus r.irconium hydride fuel moderator elements used in TRICA systcss. Are the fuel temperatare increases, this coefficient iminedi a t ely coiripensates for reactivity insertions. The result is that reactor power ; excursions are terminated quickly and safely, t The prompt shutdown mechanism has been demonstrated extensively in many thousands of transient testa pe rfortned on two prototype TRICA reactors at the CA Technologies iabor0 tory in San Diego, California, as , well as other pulsin6 TRICA reactors in operation. These tests included : step reactivity insertitas ar large as 3.5% 6k/k with resulting peak reactor powers up to 8400 MV(t) on TRICA cores containing similar fuel elements as are used in this TRICA reactor. Because the reactor fuel is similar, the previously cited experience and tests apply to this TRICA system. As a result it has been possible to use accepted safety analysis techniques applied to other TRICA facilities- to update evaluations with regard to the i characteristics of this facility (1 6), 1.3 CONC 1.USIONS Past experience has shown that . TRICA systems can be designed, constructed, and safely operated in the steady state and pulsing modes of operatio;. This history of safety and the conservative design of the reactor have perraitted TRICA systerns to be sited in urban is areas using normally buildings without. pressure type containment such as associated with reactors of like power levels. Results of this safety analysis indicate that the TRICA Hark 11 reactor sys tern proposed for construction and operation will pose no health or cafety 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 with the reactor at low and high power levels,
- b. Complete and instantaneous loss of coolant wat er - in the reactor pool, c, And fission product releaso from a fuct element ruptured in air.
1-3
I l i 1 i SAR $/91 I i j-4 The insertion of excess reactivity may represent a norrnal reactor i operatina, condition, while the loss of pool water is expected to be an j abnormal condition. Conservative estimates of doses frois fission l product releases are made independent. of accident scenarios, j - In both these postulated conditions, fuct and clad temperatures i remain at icvels below those required to generate stress conditions
- which would cause loss of clad integrity, llowever, the results of a i clad failure are analyzed and it is shown that such a failure vill not cause excessive radiation exposures.
l I ! The loss of pool water has been exarnined f rorn the standpoint of , I uitect radiation to operating personnel as well as in terms of I l maintaining fuel integrity. l
- The effects of t.rgon 41 and nitrogen 16 production during normal l
- . operation of the reactor have also been evaluated. Results of these l l analyses show that production of these radioactive gases will present no ,
i hazard to persons in the reactor room or to the general public. ! j .. l I i O l l l l 1 1 O 14
l SAR S/91 Chapter 1 Retetences
- 1. "llarards Report for TRIGA Matk 11 Pulsing Reactor", General Atomic Division Report GA 1998. Februaty 1961.
- 2. *lla z a r d s Sunanary Report for a TRICA 1 Nuclear Reactor", University of Texas Bureau of Engineering Research, October 1961.
- 1. "Sateguards Analysis Repott for TRICA Reactors uning Aluminum Clad fuel", Genetal Atomic Division Report CA 7860, March 1967.
- 4. "Salcty Analysin Report for ?$0 Kilowatt Operation of a TRICA Mark 1 Nuclear Reac t or" . University of Texas, College of Engineering August 1967.
- 5. " Safety Analysis Report for the TRICA Mark 11 Reactor E 117-478, General Atomic Company, October, 1975.
- 6. *TRIGA Mark I Safety Analysis Report", University of Texas, January 1981.
1-S
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SAR $/91 i Chapter 2 ; i i SITE DESCRIPTION The site ior the TRICA reactor facilty will be located on the east .[ tract of the Balcones Research Center, a tract of land owned and operated by The University of Texas. The Research Center is located in northern Travis County and the City of Austin about 11.6 kilometers north-northwest of The University of Texas at Austin campus. Figures , 7-1 thru 2-4 display the facility locations in relation to surrounding areas located near the transition line between hill country and rolling plains, the site is situated about 7.4 kilometers from where the flood controlled Colorado river crosses the transition region and , Balcones fault zone. The Balcones Research Center east and west tracts span-part of the inactive fault zone. The cast tract is within the transition region to rolling plains. Site location of the TRICA reactor is in the northeast region of n the research center east tract. Adj acent to the north boundary of the research center and near to the eastern boundary. the site location is near the intersection of Braker Lane and Burnet Road. Figure 2 4 shows the site location within the Balcones Research Center. Reference guidance for site evaluation was ANS 15.7 (1). P 2.1 GENERAL lACATION AND AREA Major activities of The University of Texas at Austin, State of Texas government, and City of Austin business district are centered at respective distances of 11.6, 12.6, and 12.9 kilometers to the south. southwest. Distances to air traffic landing facilities in the area are 6.7 kilometers for private aircraft, 9,7 kilometers for commercial aircra*:t and 20.8 kilometers for military aircraf t. A total area of 1.87 square kilometers is contained within the Research Center cast and west tracts of-land. The east side of the Center is bounded by a . State highway, FM 1325, and the west side is bounded by a Federal highway, US 183. The two tracts are divided by a rail line, formerly the Missouri Pacific, with 0.93 square kilometers in the east tract and 0.94 square kilometers in the west tract of land, liighway intersections of US 183 with FM 1325 and with Loops i-and 360 are within two kilometers of the site. Both highway 1 cops are planned for extension into the area associated with the vest tract of land [2]. An area of about 9000 square meters in a rectangular shape of 120 meters by 75 meters will- comprise--the general site location. The 120 meter length is along the north research center boundary. Areas for parking, landscape and access roads are within the general site area. A butter zone exists between the site area and activities or structures to the east and west. To the west the buffer zone is about 55 meters by 75 meters with parking also about 60 meters by 75 meters. The east buffer l region is primarily open space that will provide the access - to other development projects north of the general site area. 21
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SAR 5/91 l l Most ar eas adjacent to the Research Center are developea for mixed conunercial and industrial activities including warehouses, manufacturing facilities, small business parks and a few undeveloped areas. Mixed commercial and indus t. rial areas south and east of the Research Center are bounded by highway US 183, highway IH 1325, and the Texas New Orleans Railroad to the e a r. t . Approximately 2.2 square kilorue te rs of land are enclosed by the area. Much of the remaining area to the west of the Research Center is bounded by highway US 183 and the Missouri pacific Railroad and is not developed. The area is planned for future road right-of-vays and includes the west tract area. Immediately north of the Balcones Research Center east tract is a 2.3 square kilometer complex operated by International Business Machines Corporation. Undeveloped areas around the Research Center are expected to develop as a mixed commercial, industrial, research or office park areas. Residential areas are located beyond adjoining areas around the Balcones Research Center with distances from the reactor facility site of 1.2 kilometers to 2.0 kilometers. Few residential structures cor either multitamily or single f amily units are located within a radius of 1.2 kilometers of the reactor site. 2.2 POPULATION AND EMPLOYMENT Austin is composed primarily of governmental, business, and professional persons with their families. The city has substantial light industry but practically no heavy industry. Many of the persons in the local labor force are related to activities of the City and its role as a State Capitol or the University and its educational and research programs. A substantial population growth rate was in process prior to the time of project proposal and development. By the time of project completion a significant reduction of the local aren growth rate was effective. These population trends were characteristic of the "Sunhelt" areas of the country. Population growth of Travis county between the 1970 and 1980 census was 427.. The substantial growth of Austin in particular, was the norm during the early 1980's with a significant reduction in the rate by the late 1980's. As one of the ten fastest growth areas, the Austin area growth was 35 percent between 1980-1986. By 1988 the annual rate of growth of the Austin metropolitan area was about one percent. Population data of the Travis county region is presented in Table 2-1 with supportive data in Figure 2-5 and Figure 2-6. Estimates of the 1990 census project the 1990 Austin population to be 465,600 The 1980 population census listed the Austin city population at 345,496 and Travis county population at 419,573 with the Austin Standard Metropolitan Statistical Area population at 536,450 (3). Three counties, Trevis, Hays and Williamson, compose the Austin Standard Metropolitan Statistical Area. Population densities in Travis county range from 6.4 persons per 1000 square meters encompassing the main university campus to less than 0.2 persons per 1000 square meters in growth areas north of the research center site. Population census tracts adj acent to the site exhibit densities of 0.2 to 0.3 persons per 1000 square meters. 26 l
SAR $/91 O I
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_____-__.__._.___._.m._._.___.___.____ . _ . _ _ . 1 SAR $/91 Approximately 800 persons were involved in activities on the east j tract of- the Balcones Research Center in 1980 with projected activities 1 at the site to add an estimated 900 to 1000 persons by the late 1980's. On the west tract the Microelectronics and Computer Technology Corporation setup operations wit.h approximate'ly 500 persons in 1985. Facilities north of the Research Center operated by International ' Business Machines Corporation employed an estimated 6500 persons in 1985 with approximately 500 additional persons by 1990. Research activities at the Balcones Research Center are diverse, consisting of many different research organizations of the university science and engineering colleges, As of 1980, several ressarch progress were already situated at the research unter site of which the Applied Research Laboratories, Center for Research in Water Resourc( ) and Civil Engineering Structures Laboratory are examples, Between 1980 and 1485, additional facilities and 3 major additional programs, the Center for Energy Studies, Center for Electromechanics and Bureau of Economic Ceology were added to the research center functions. Since 1985 program development, besides the Nuclear Engineering Teaching Laboratory (NETL), included a computer facility, the Center for High Performance Computing, and an engineering facility, for microelectronics and manufacturing research. Administrative functions of the Bureau of Economic Coology are situated adjacent to the NETL site. Expansion of other activities near the NETL site is possible in the future, However, as of 1988 there were no of ficial plans for the adjacent site areas. 2.3 CLIMATOLOGY Austin, capital of Texas, is located on the Colorado River where the stream crosses the Balcones Escarpment separating the Texas Hill Country - from the Blackland Prairies to the east. Elevations within the City vary from 120 meters to 275 meters above sea level. Native trees include cedar, oak, valnut, mesquite, and pecan, The climate [4] of Austin is humid subtropical with hot summers. Winters are mild, with below freezing temperatures occurring on an average of less than twenty five days each year, Rather strong northerly winds, accompanied by sharp drops in temperature, occasionally occur during the winter months in connection with cold fronts, but cold periods are usually of short duration, rarely lasting more than two days. Daytime temperatures in summer _ are hot, but summer nights are usually pleasant with average daily minima in the low seventies. l Precipitation is f airly evenly distributed throughout the year, i with heaviest amounts. occurring in late spring. A secondary rainfall l peak occurs in September. Precipitation from April through September l usually r< sults from thundershowers , with fairly large amounts falling l within short periods of time. While thunderstorms and heavy rains have occurred in all months of the year, most of the winter precipitation O. occurs =as-light rain. Snow is insignificant as a source of moisture, and usually . melts ss rapidly as it falls. The City may experience several seasons in succession with no measurable rain fall. 2-10 I i I
SAR 5/91 ( ( . CLIMATOLOGICAL DATA TEMPERATURE (oEcaEEs rAHRENNEst) 100
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_ m- _ - _._ -______ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ - - _ . _ _ _ . _ . _ . _ . _ _ _ 1 4 l 4 SAR 5/91 Prevailing winds are southerly throughout the year. Northerly , winds accompanying the colder air masses in winter soon shift to southerly as these air masses move out over the Gulf of Mexico. Climatology data is summarized in Figure 2-7. Typical Austin wind t- data are presented in Figure 2 8 [5]. The average length of the warm season (freeze-free period) is 270 days, Based on data from 1943 1961, the average date of the last occurrence of freezing or below has occurred as late as April 13 (1940), and as early as October 26 (1924). Heteorological data is tabuinted in Tahic 2 2 and Table 2 3 [4]. Destructive winds and damaging hailstorms are infrequent. On rare occasions, dissipating tropical storms effect the City with strong winds and heavy rains. The f requency of tornado type activity is illustrated in Figure 29 (6), Recent tropic storm paths and local uiting of tornadoes and funnel clouds are presented in Figure 2-10 and Figure 2 11 17,81 2.4 CE01.0GY The northwestern half of Travis county is part of the physiographic province 'i Texas known as the Edwards Plateau. In Travis County, this is a highty dissected plateau with wooden hills rising in some places more than 150 meters above the drainage pathways. In marked contrast, the southeastern half of the county is gently rolling prairie land which is part -- of the physiographic province known as the Gulf Coastal Plain. These provinces are separated by the scarp of the Balcones fault zone, which rises 30 to 90 meters above the Coastal Plain. The scarp, however, is not a vertical cliff; it is an indented line of sloping hills leading up from the lower plain to the plateau , summit. The rocks that outcrop in Travis County are primarily of sedimentary origin and of Mesozoic (Cretaceous) and Cenozoic age. They consist largely of -limestone, clay, and sand strata which dip southeastward toward the Gulf of Mexico at an angic slightly greater than the slope of the land surface. Therefore, in going from southeast to - northwest the outcrops of progressively older formations are encountered,- and the rocks lowest in the geologic column have the highest topographic exposure. At the reactor facility site on the eact tract, the geology is of the Austin Group defined as chalk, marly limestone, and limestone with light gray, soft to hard, thin to thick bed, and massive to slightly n nodular character. On the west tract, the geology changes to the Edwards Formation of limestone and dolomite with light gray to tan, hard 2 to soft, thin to thick bed, and fine to medium grain character. The separate formations are, respectively, the up and down side of a segment of the Mount Bonnell Fault that passes approximately along the boundary of the east and west Balcones Research Center tracts. Distance to the iau.t is about 500 meters from the reactor facility site. 2-13 _ _ ~ - _ - . . - - - - - . . . - - - -- . - _ - - - - . . - . --- --- , ,
SAR 5/91 Table 2 2 1982 METEOROLOGICAL DATA FOR AUSTIN TEXAS
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SAR 5/91 4 Table 2 3 .r a i HISTORICAL METEOR 01hCICAL DATA FOR AUSTIN TEXAS , Average Temperature Ho ___ stinga..Degre_e Days .;T. _,._r.
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/' TEXAS TORNADO FREQUENCIES Figut'e 7-9 2-16
SAR 5/91 l'
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/N TEXAS HURRICANE PATHS k
Figure 2-10 2-17
SAR S/91 O k 4(1) BURNET MtLAM L1.ANO I 6(0) 5(0) 5(0) N^# 13(6) gy 4(0) ( 4 4(4) 5(0) KENDALL. 6(1) CALDVELL FAWTTE l(0) 2(2) 2(3) r GUADALUPE 1(1) 0(1) Tornado and funnel cloud occurences in a 50 mile radius of Austin for the years 1975-1983; # Tornados (# Funnel Clouds). LOCAL FUNNEL CLOUD SITINGS O Fir,ure 2-11 2 18
SAR 5/91 The Balcones fault zone, which extends from Williamson County to Uvalde County, extends the full length of Travis County on a line passing through Manchaca, Austin, and McNeil. llere the orderly sequence of formations is replaced by an outcrop pattern controlled by the faults, most of which are normal faults with the down thrown side toward the coast. Most of the movement of the Balcones Fault zone occurred during the Miocene period. Since no movement has been detected during modern tin:es , this fault is no longer considered active [9]. location of the Balcones Fault zone and formations in the Austin area are depicted in Figure 2-12. 2.5 SEISMOLOGY Thirty three earthquakes of intensity IV or greater have had epicenters in Texas since 1873 [10,11]. The earthquakes were characterized using the Modified Mercalli Scale of 1931. The scale has a range of I thru XII, on which an intensity of I is not felt, an intensity of III is a vibration similar to that due to the passing of lightly loaded trucks, and intensity of VII is noticed by all as shaking trees, waves on ponds, and quivering suspended objects but causes negligible damage to buildings of good design and construction, and an intensity of XII results in practically all works of construction being severely damaged or destroyed. The strongest earthquake, a maximum intensity of V111, was in western Texas in 1931 and was felt over 1.165,000 square kilometers. Figure 2 13 shows the locations and intensities of all earthquakes in Texas since 1873. Of these, some are known to have been felt in Austin, but no damage has ever occurred to local buildings. 2.6 HYDROLOGY Almost the entire county is drained by the Colorado River and its tributaries. Lake Travis, which is formed by the Mansfield Dam on the Colorado River, is part of the power, flood-control, water conservation, and recreation project of the Lower Colorado River Authority. Other lakes are alsa operated by the Authority, such as Town Lake and Lake Austin, and are created by Longhorn and Tom Miller dams, respectively. Low level alluvial deposits of the river are commonly saturated with water at relatively shallow depths. Recharge is primarily from the river and local surface contaminations are easily transmitted to this shallow water table. Ground water from subsurface formation is found in basal Cretaceous sands referred to as the " Trinity" sands. Elevations of the Trinity aquifer range from depths commonly less than 300 meters east of the Balcones Fault Zone to greater than 450 meter s to the west of the zone. East of the Mount Bonnell Fault, dolomite and dolomite limestones provide a source of ground water at shallower depths. Access to the Edwards aquifer ranges from 30 meters to 300 meters with natural springs necurring in areas near the Colorado River. Minor aquifers associated 2-19 l
SAR 5/91 O [rT 7% ' fy d 47(( [) k ' ,
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~
dif ferent formations BALCONES FAULT ZONE Figure 2-12 2-20
SAR 5/91 Feb. 1974 Match 1940 hv unh 1917
-lO , _\---
June 1951 O_ _ _pg' - auty 192s June 1936 0 _l-j- July 1966 _-l-!-I-l-l, _ arru 1934
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;I sc.i. ce 19n
. umussrs zy _ .. --
y\-_ c intrustw v - /_ O n=" sin v2 g rum 4stry vn Texas Earthquakes, 1873-1983 TEXAS EARTHQUAKE DATA Figure 2-13 2-21
SAR 5/91 O EDWARDS AQUlfla ( ,'
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SAR 5/91 with the Clenn Rose Formation supplies small quantitles of water west of
-the Balcones Fault Zone. Water bearing areas in the formation are -at varying depths and I ?. terally discontinuous. On the Balcones Research Center east tract, wells drilled for environmental monitoring have produced ground watet at depths of less than 15 meters. Figure 2-14 shows the location of the ground water aquifers.
Water supply for the roscarch center and wastewater treatment is provided by the City of Austin. Although wells - into the aquifers provide substantial water the city supply is filtered river water. Other area municipalities and organizations utilire aquifer water. Control of private wells is the function of county and state llealth Departments. Gross beta radioactivity of city water has been mear.ured ar.d is reported in Table 2-4. Table 2-4 CROUND WATER ACTIVITY (gross beta) Travis County 6 x 10~9 pCi/ml Balcones Research Center 8 x 10 pC1/ml 2.7 HISTORICAL Relocation of the UT TRIGA reactor and related facilities to.the Balcones Research Center site is to help accommodate growth of programs both at the University main campus and at the Research Center site. The fac111ty location at the Research Center is to be in the north-east corner of the research center site. The original research center site area was operated as a magneatum manufacturing plant. by the Federal government in the 1940's. Subsequent arrangements and acquisition by the University would determine *
-acttvities of the - site .throughout 'the 1950's, 1960's and 1970's.
Ac t ivi ties . at the site were not fully developed prior to the 1980's. University functions or research activities were moved to the site when required accommodations were not available on the main campus. A few funct ions of the University at the sits had resulted in t.ho construction of major- f acilities suitable for long tet' use. Other activities at the site have utilized existing structures or.other buildings not suited for long term use.
- A major program (12) was established in the 1980's to develop the Balcones Research Center site activities. As part of the first phase of development, several major research programs associated with energy and engineering were moved to f acilities constructed . at the site. Features of the site, before the development activities by the University and af ter initial development in the 1980's, are illustrated in Figures 2 15 and 2-16.
2-23 _ . . , _ . - - _ . _ . . _ . , _ _ . _ _ _ _ _ . _ _ _ _ . _ . _ _ _ _ . _ . ~ . _ _ ~ . . . . _ _ _ . _ _ . . _ . .
SAR 5/91 A major program [12) was established in the 1980's to develop the Balcones Research Center site activities. As part of the first phase of development, several major research programs associated with energy and engineering were moved to facilities constructed at the site. Features of the site, before the development activities by the University and after initial development in the 1980's, are illustrated in Figures 2-15 and 2-16. Several activities at the Research Center prior to 1980 had been associated with radioactive materials. These activit ies ranged f rom the burial of low level radioactive waste materials such as tritium and carbon-14 in the northwest corner of the site, to water transport studies performed in 30 meter diameter surface tanks. Isotopen of cesium 137, cesium 134, and cobalt-60 were present in sludge samples of one of the tanks, and are reported in Table 2 5, Gross beta activity in the samples of the west tank measured 22 microcuries per milliliter (1979). Table 2-5 TANK SLUDGE SAMPLES Total (1979) 22 pCi/ml (gross B) Cs 137 1.3 x 10'3 pCi/ml Cs l34 2.5 x 10 pCi/ml Co 60 5,7 x 10-5 uCi/ml Radioactive waste and other materials at the Research Center site are part of the University broad license for radioactive materials which is managed by the University Safety Office and issued by the Texas Department of Health. 2-24
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SAR 5/91 Chapter 2 References
- 1. "Research Reactor Site Evaluation", American National Standard, ANSI ANS 15.7-1979 (N379).
- 2. " Basic Data", Cit.y of Austin Planning Department. June 1980.
- 3. "1980 Census of Population", Department of Commerce, Bureau of Census, Cit y of Austin Planning Department.
- 4. " Local Climatological Data; Annual Stur. mary wi th Comparat ive Data Oceanic and Atmospheric Administration, 1982", National Environmental Data and Information Service, National Cl in.a t i c Center Anheville, N.C.
- 5. "Climatography of Texas; Wind Rose-Austin, Texas", National Weather Service, Austin, Texas,
- 6. " Texas Annual Tornado Density", National Weather Service, Austin, Texas.
- 7. George W. Bomar, " Texas Weather", 1983.
Data", 1975-1983, National Oceanic and Atmospheric
- 8. " Storm Administration, Environmental Data and Information Service National Climatic Center, Asheville, N.C.
- 9. L.E. Garner and K.P Young, " Environmental Geology of the Austin Area: An Aid to Urban Planning", Report of Investigations No. 86.
3,
- 10. " Earthquake Inforraa tion Bulletin," May-June 1977 Vol. 9 No.
U.S. Department of the Interior Geological Survey,
- 11. Steven M. Caulson, " Investigations of Recent and llistorical Seismicity in East Texas", Masters Thesis May 1984, University of Texas.
Project Analysis", Volume I, The
- 12. "Balcones Research Center University of Texas, 1981.
O , 2 28
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_ - . _ - _ _ . . _ . _ _ _ _ _ _ _ _ . _ _ ____.___.__m______.. SAR S/91 Chapter 3 FACILITY DESIGN STRUCTURES, SYSTEMS AND COMPONENTS The facility in which the TRIGA reactor is to be located is a building with several special design features. Most of the building design specifications are for criteria that are independent of the reactor systems. Ilowever, several design features and specifications are to assure - safe facility operation and effective utilization of facility' equipment. Building areas include office space and laboratory space. Space for administrative functions, library or conference rooms, maintenance shops and ut.ility equipment will support all facility activa ties. The reactor, which is the primary piece 01 facility equipaent, will establish the fundamental requirements for two basic areas of the building. One area is the reactor bay that contains the primary reactor structures and experiment facilities. The other area will contain the operation control center with space for the control console, training ac t. t v i tie s , and records maintenance. Building orientation on the _ site plan is shown in Figure 3 1. Total gross floor space of the facility is about 1950 square meters (-21,001 sq .ft.). Several design requirements a protect against the release of radioactive _ materials from reactor operation are set by the compliance requirements _-of the appropriate standards or regulations. Other potential releases that might occur from specific experiments or special F materials 'o r conditions. are also subject to these standards and regulations. Facility design includes features for control of airborne release from reactor oprration and releases of liquid waste or solid waste from other facility activities. 3.I GENERAL CONDITIONS The design of a structure to contain the-TRIGA_ reactor facility depends at least partly on the protection requirements for the fuel elements. Among these requirements are the control of personnel access to special
'_ nuclear ; material and the control of personnel exposure to radioactive
-fuel materia 1. Design of the fuel elements as a physical containment-
. system is the basic _ derign assumption. This fuel design contrels the release of radioactive material during routine reactor operation and
. potential _ accident - conditions. Facility -design will- control the exposure to. the radiation - levels that the- fuel vill create during operation. Other-facility design requirements will control' the release of operation effluents such as radioactive gases fron. either normal operation or accident conditions. Rupture of a fuel element will be the design _ basi s accident. _ Design of the-reactor bay as an air confinement system will- protect the operation personnel and the general public against the operation hazards of the reactor. These hazards are release of air activation products during normal operation and the 3-1 i'
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SAR 5/91 4 re t e.me of fuel lission products during accident conditions. Release crit erion are based on Title 10 Chapter 20 of the U.S. Code of Federal
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Engineering Design, specifications, and construction for the building are set by the State of Texas Uniform General Conditions and The University of Texas at Austin Supplementing Conditions [2.3,4.S1 provisions of the Uniform Building Code [$) and other national codes for mechanical, electrical, and plumbing are applicabic to this project. Equipment requirements will apply Underwrlter's Laboratories standards or labols, when appropriate, to a piece, type, class, or group of equipment. Other specifications will conform to the standards of the American Society for Testing and Materials (ASTM). provisions of the Life Safety Code are appilcable. One code of importance, the National 1 r'i re protection Code, will 6etermine requirements that relate to lire safety for significant facility operation hazards. 4 Ihploration logs for the site consist of 4 borings of depths to 30 feet (~9 meters). Approximate ground surface elevation is ill feet (241 soe t e rs ) . At an elevation of about 787 feet (240 meters), each log records a consistent subsurface structure typical of the region. Installation of a pier.ometer in one bore hole allows a check of ground wa t e r .- Evidence and experience indicate a periodic presence of some ground water, alt hough seasonal variation of the level is probabic. Water surf ace, depth at the time of piezometer installation was 29.1 feet (8.87 meters) af ter 24 hours and 9.3 feet (2.83 meters) af ter 3 months. The building site is located on a rock subsurf ace of limgtone. Soil tests of the subsurface set the load capacity at 1690 kg/m (2.4 psi). Concrete piers and footings will provide building foundations. 4 Seismic design specifications will meet the requirements of the Uniform Butiding Code for zone 0. The zone designation is a reference that determines the appropriate earthquake activities in the continental United States. These specifications require no special provisions Normal building beyond those of standard building load requirements. loads from_ gravity and wind _ forces will exceed the seismic accelerations for hu11 dings in zone 0. A butiding of good construction should earthquake acceleration of about .75 g. Ground withstand- an accelerations that exceed this would be rare events in a region in which earthquakes are already infrequent. Wind load designs will meet requirements of the Uniform Building Code for- 70 mph (31.3 m/sec) winds. Normal wind and storm conditions will be within these design factors. Those specifications include factors for gusts, in excess _ of the wind load criteria. Iturricanes are not likely to be a direct threat because of the natural dissipation of energy on land, llowever, tornados are a concern with their extreme wind b velocities. Tornado type activity is roughly one event per year per Q 1000 square miles (2590 sq. kilometers) in the general g site area. This-activity represents a frequency of one per 2.5 x 10 yr for an area of a square with sides of 333 feet (31 meters) typical of the building site. 33 s
SAR 5/91 liesign for watei sonol( in the ptoject vicinity will provide for wat e t liom local talnfall r a t e r, t Im t ate frequtntly dispersal of spoiadic but sometinms totiential. Di ai na ge provisions for the building tool, site landscape, access roadways and subsuttace will control local runoit. l.o c a l flood cont rol will include gravity flow drainage and collection sumps with dual operation pumps. Root drainage and nite tunott will be by gravity f low. Separate sumps with pumps will control subsurface drainata at the building perimeter and beneath the reactor shield foundation. Centle slope chatacteristics in the insediate site vicinity provide an ample gradient of about 3 feet (1 meter) for surface wates runot!. Mean elevation at the local site is 791 feet (241 meters). Data f t om the National Flood Insurance Program indicates that no port ion of the research center site is within the 100 0: 500 year flood cono. Thus, the only flooding likely will be as a result of local runoff cotulit ionn 3.2 ARClllTLCTURAL AND STRUCTURAL ENGINEERlNC Architectural design of the building will develop two separate functlonal sectionn, one the reactor bay wing and the other an academic and laboratory wing. Structural design of trie building sections is of concrete columns and beams with steel reinforcement. The first level of the teactor bay wing is 7 feet (2.1 ineters) below the mean grade, while the academic wing entry level is 7 feet (2.1 meters) above the mean grade Figures 3.2 and 3-3 111ustrate the basic building design and arrangement. The reactor bay wing will consist of three basic parts. One is the reactor bay with a floor to roof level of 56.5 feet (17. 2 rne t e r s ) . The second is a f our level sec tion adj acent. to the reactor bay. Third is the radiation experiment room with 4.25 feet (1,3 meters) thick shield walls. The reactor bay wing construction c ons i s t r. of several types of concrete construction. The floor is a slab and beam design of reinforced concrete on compacted fill material. All building columns
., St walls aro concrete, :nst in place with steel anu level entoic meot. Exterior walls of the reactor bay are concrete and steel const act ton with tilt panels and attachment colunas The combination of panels and columnn set on top of the first icvel structure forms an int egral unit by placement of the panels, then placement of the columns Adjacent to the reactor hay, structural concrete and steel columns support slab and beam floors. Interior walls are primarily concrete blocks with a few plaster board tyoe walls. Both concrete and met al panels complete the exterior coistruction of the reactor bay wing. Roof structure is a steel joist sys t em with met al deck , concrete slab, and built-up composition roof that includes fire barrier and thermal insulation. A room of four walls and a roof of standard density concrete 4.25 feet thick forms a radiation shield room to c ottple t e the react or bay wing. The room is cast in place with key joints between concrete placementr,. Tilt panels and composition roof finish the structure. All doors are of hollow metal construction.
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i i SAR $/91 Two floor levels will comprise the academic and laboratory wing, The entry floor level (second level) is an administ rative and of f ice ; section, l.aboratories will be on the next level (third level). ; construction of this wing is reinforced concrete pier and columns wit h i poured beam and slab floors and roof. Exterior walls will consist of conetete tilt panel, metal siding and window units. In:crior walls are metal stud trames with gypsum board pancia. Doors are solid core wood. r Entry way area and door is glass and metal f rame. Stairvells at each i end of the building wing will provide access to each building level. )- Standards and specifications of snechanical, electrical, and other , systerns will be typical of university construction projects. In some " casen, facility requirements may also supplement the standard requirements. 3.3 SI' ACE AlthCAT10N Net assignable space for administration, laboratory. and supplemental areas is about 50% of the total floor space. Figures 34 through 3 7 show the floor plan arrangements. Space in the building will consist of two major experiment facilities, a reactor boy room and a radiation experiment room. The reactor bay area consists of 350 , square meters-(3600 sq. ft.-) of space for reactor structures, systems, components, and experiments. The radiation experiment room with about ' 85 square motora (900 sq. ft,.) is a free space cube roughly 9 incters (30 ft.) on a side with thick shield walls. Besides the major experirnent-areas, 8 general laboratory areas and 6 supplemental support areas will be available. The respective spatial areas are 635 and D5 square meters (6840 and 1430 sq. ft.). Offlee, conference, and administrative space will contributo another 230 square ine te rs (2510 sq. ft.) of assignable space in ten to twelve different rooms. These space allocations represent the result of planning estimates for a typical configuration of facility activities. t Functional arrangernent of facility activities is one result of the two section building design. One of these sections, the reactor bay, wing, contains the primary experirnent facilities, support facilities, such as shops and building mechanical and electrical equipment. The academic and laboratory wing is the other section, which includes , eJministrative, office, cominunic a t ion , conference, and laboratory facilities. 3.4 REACTOR BAY AND OPERATION CONTROL ; l Reactor-tank, shield and primary experiment facilities are located in a reac_ tor bay area that is about 18.3 ineters - on each side, A total of 4575 cubic- meters of volume is enclosed in the reactor bay above the 335 square meters of floor space. Operation control of reactor and of reactor expr eirnent' activities is provided by an area located adjacent to the reacte bay. Space in the operation control area is divided into cont rol - >m , conference roorn , office, and entry way. Total operation O) control ., r e a (7.3 by 18.3 m) is 134 square ineter of floor space and
-roughly 489 cubic ineters of air space. The stairwell in the academic wing provides access to the reactor bay and operation control areas 37
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Design of the reactor bay is specified by constraints on the access control for physical ; i lunctton of the architecture design, security, radiation protection for personnel safety, and applicable ? building code standards. All access poirts to the reactor bay are located inside the ! engineering building and enter from the operation cont rol area side. { j The remaining three sides of the reactor bay area are enclosed by ' exterior walla. Itoth ernergency exit s and equipment bay doors on the ftrut level open into the adjacent area within the building from which building exits are accessible. Adjacent areas on the operation control side of the reactor bay provide some laboratory support functions in i conjunction with other building support functions. On the first lwel adjacent to the reactor bay are staging areas for the equiptnent bay - doors, a mechanical equipment room, liquid waste storage area, radiation - experitnent control center, and other areas. l f Adj acent to the reactor bay on the second level is the building
- electilcal distribution, a mechanical equipment shop, electronics
- equipment shop, and darkroom laellity. The third level area adjacent to the reactor bay functions as the reactor operation control center. All i building air handling syst erns and hot water systems are on the fourth level adjacent to the reactor bay.
Access from the reactor operation control center level to the reactor bay in at the top of the reactor shield at the point of access to the reactor pool. A stairway next to the reactor shield structure provides access to the first, second, and third levels of the building adjacent to the reactor bay. Two rooms within the reactor bay will enclose reactor support , pool water treatment sys terns for purification and cooling systems, equipment are on the - first level. Auxiliary equipment for experin.cnt
-systems, such as pneumatic systems, will be in_the second level room.
Other features of the reactor bay include a five ton bridge crano and six fuel storage pits. The storage pits and reactor shield structure ; are linpo r tant systems to safely operate and store the reactor fuel materials. lloweve r , only the ventilation design for the reactor bay will.be an engineering safety feature. The operation cont rol level consists of the reactor control room, tout ine entry hn11way, operations office, conference room, and file storage. Windows between the operation control level and the reactor hay ..llow visual observation of operation activities. Routine access to the operation control area is by way of the academic wing stairve.y. l- Design of all access points and barriers between the reactor bay l' or operation control area and adjacent areas will depend on requirements for eccess, security, fire and ventilation. Design will place limits on the size of penetrations into the area and require scalants to control h v air leakage. 3 12 _ _ . . _ . _ _ . . _ . ~ . . . _ _ _ _ . _ _ . _ _ _ . . _ _ . - _ _ . _ . . _ _ . . _ _ _ . _ . , _ _ .
l SAR 5/91 l l j On the third level from the reactor floor the adjacent area to the reactor bay is supplemented by the control room area, conference area operation office, and routine entry point. The third level entry way is provided for access to the control area from the laboratory building and access to the reactor bay f rom the control area. Access at the third 2 . level is to the top of the reactor shield structure. A stair structure is attached to the reactor shield with a supplementary access point to the reactor bay on the second level. Design of access points and interior walla are specified for , security control, fire cont rol, and ventilation control. Penetrations, besides the doors, into the reactor bay and control areas are limited in s i r.e and are sealed to limit air leakage. Details of the reactor bay area are presented in Figure 3 8. A 5 ton bridge crane is installed in the reactor bay for movement of shield structures, heavy equipment, and ; fuel transport loadinB. ; 3.5 SUPPORT FACILITIES Support facilities for reactor operation include functions of the academic and laboratory wing of the building, two shop areas for mechanical and electronic work, and support laboratories. i. 3.5.1 litalth Physics Laboratory A ilealth physics Laboratory is situated adjacent to the reactor facility on the second level. Radiation counting systems for evaluation of radiation exposure or contamination are maintained in the llealth Physics Laboratory. Equipment such as a thermoluminescent reader and an , alpha-beta proport.ional counter are maintained in the laboratory. Other equipment and supplies operated or stored in the laboratory include portable radiation monitors, coveralls, gloves and relat ed items. 3.5.2 Sansole llandling Laboratorv
' A sample handling laboratory is situated adjacent to the reactor f acility on the third level for the processing of radioactive s at.ipl e s
' and materials. Access ports via air or gas transfer tubes are installed to move samples between the reactor area and the Samplo llandling Laboratory. Two separate sample transfer systems are provided, one for the pneumatic tube-Irradiation facility and one for loading the rotary specimen rack facility. A hood for handling radioactive materials and a sink for disposal " of radioactive liquids are installed in the Sample llandling Laboratory. A saloty shower is adjacent the laboratory doorway. 3.5.3 Effluent Control Effluent pathways for air, liquid, or solid releases of radioactive material provide control of material releases. Control pathways for air and liquid effluents are by way of two rooms, room 4.lM3 and room 1.108. 3 13
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1 l l SAR 5/91 l l f Control of air releases from reactor experiment areas is provided in room 4.1M3, which contains the air purge system isolation valve and filter bank. The tilter bank normally contains prefilters and one high l efilciency particulate filter. Space will be available for expansion ' that allows addition of a charcoal filter and additional high efficiency particulate filter. I control of liquid relenwes that contain radioactive snaterial is provided in rocin 1.108, which contains storage tanks for collectlon, processing, storage, or release of 11guld effluents. The reactor pool will not release liquid effluents as a part of normal operation. , Solid waste cont rol will typically consist of canisters, such as plastic bags and plastic or metal barrels. Location of these canisters for. . solid waste control will be adj acent to experiment areas that produce the waste. 3.6 DESIGN EVA1.UAT10N Building design, construction, and inspection will- provide a ! facility for safe operation of the TRICA reactor. The facility and it's features apply standard engineering practices to control quality of building systeus. The features are typical of other siin11ar installations and should provide a reasonable margin of safety for the operation of the TRICA reactor. No unusual site conditions exist to threaten the building, Seismic and wind loads determine building structural design. These structural conditions protect against building failure from common site conditions of geological or meteorological origin. Several special building features provide for control uf effluents and support- of --operation and experiment activities. One specific engineering design feature, the _ reactor hay ventilation sys t ern , will protect against uncontrollable releases of airborne radioactive materials. Other building design features, such as the reactor shield structure, fuel storage wells and liquid waste system, supplement the vent.ilation syntem to provide protective or control _ functions that prevent the uncontrollable release of radiation or radioactive material. 3-15
k 1 i SAR $/91 , Chapter 3 References I i .l 1. Code of Federal Regulations. Chapter 10 part 20, U.S. Cove rtute n t-Printing Office, 1982.
- 2. " Specifications for Nucicar Engineering Teaching Laboratory", The University of Texas at Aur. t i n , Project No. 102 568, Sept. 15, 1
l j 1986, j
- 3. " Construction Administration Manual for Nuclear Engineering '
Teaching 1.abonatory", The University of Texas at Austin, l'roj ec t Ho, 102 568, Decetaber 1986. ; i j 4. "NETL l'roject Non. 1, 2, f 3", The Universit y of Texas at Austin, . I l'roj ec t 102 568 Armendment s , December 1986. l S. "Uniforrn liullding Code", International Conference of 15uilding f officials, May 1, 1985. O l 4 i. t ' l
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q. SAR 5/91 l i Chapter 4 t TRICA RtMCTOR The reactor design bases are predicted on the maxitewn operationcl capability for the fuel clernent s and configurat ion described in this report. The TRICA reactor system has three snajor areas which are used j to define the reactor design bases. ..
- n. Fuel temperature,
- b. Prompt negative temperature coefftelent,
- c. Reactor power.
Of these three only one, fuel temperature, is a real liini t a t ion. A suruma ry is presented below of the conclusions obtained from the reacto:* design bases described in this section, i 4 4.1. DESIGN BASES The fuel temperature is a litni t in both steady state and pulse , m(dd operation, This limit sterns from the out gassing of hydrogen from tb Zril -(H/Zr ; x) fuel and the subsequent stress produced in the fuel e leinent clad material. The strength of the clad as a function of temperatura can set the upper limit on the - fuel t unperature. Fuel
- temperature limit s of 1150'C (wit.h clad 5 500'c) and 970'C ' (with clad 2 500'C) for U Zril (ll/Zr ; 1.65) have been set to preclude. the loss of clad integrity. Simnad [34] summarizes the properties of U Zril x fuel materials for TRICA reactors, includins the limiting design bases and !
parameters, t The basic parameter which provides the TRICA system with a large l safety factor in steady state operation and under transient conditions is the prompt negative terwerature coef ficient- which is rather constant with t emperature -(-0,01% 6k/k'C), as described later. This coef ficient is a function of the fuel composition and core geometry. Fuel and - clad temperature limit the operation - of the reactor. Ilowever, it is more convenient to set a power level limit which is based on temperature. The design bases analysis indicates that operation at ; up to 1900 kW (with an 85 element core 'and 120'F inlet water temperature)_with natural convective flow will not allow film boiling, and theref ore high fuel and clad temperatures which could cause loss of t clad' integrity could not occur. , l 4.1.1 React or Fuel Temperat urc The basic- safety limit for the TRIGA reactor system is the fuel temperature; this applies for both the steady-state and pulse mode of l operatlon. 4-1
SAR '>/91 j 1 i i 4 Tso limiting temperatures are of i nt e re s t. , depending on the type , of Tit!CA fuel used. The TRICA f uel which is considered low hydride , that with an II/Zr ratio of less than 1.5, has a lower temperature limit , than fuel with a higher il/Zr ratio. Figure 41 indicates that the l higher hydride compositions are single phase and are not subject to the large volume changes associated with the phase transformations at , approximately 530'C in the lower hydridos. Also, it has been noted [1] that the higher hydrides lack any significant thermal diffusion The of hydrogen. These two facts preclude concomitant volume changes. important properties of delta phase U Zril are given in Table 4 1. Table 4 1 PilYSICAL PROPERTIES OF del.TA PilASE U Zrli Thermal conductivity (93'C 650*C) 13 Btu /hr ft2*F Elastic modulus: 20'C 9.1 x 106 p,g 650'c 6.0 x 106 p,g Ultimate tensile strength (to 650'C) 24,000 psi _; Compressive strength (20'C) 60,000 psi Compressive yleid (20*C) 35,000 psi j lleat of formation (611 298'C) 37.72 kcal/g mole Among the chemical propert.ies of U Zril and ZrH. the reaction rate of the. hydride with water is of particular interest, Since the hydriding reaction is exothermic, water will react. more. readily with Zirconium is - frequently zirconium than with zirconium hydride systems. used in contact, with water in reactors, and the zirconium water reaction-is not a safety hazard. Experiments carried out at GA Technologies show
-that the zirconium hydride systems- have a relatively low chemical reactivity with respect to water and air. These tests have involved the
-quenching with_ vater of both powders _and-solid specimens of U ZrH after heating to as high as 850*C, and of solid U Zr alloy after heating to 'as high as~1200'C. Tests have also been made to determine the extent to; which fission -products are removed from the -surfaces of the fuel elements at room temperature. Results prove that, because of the high resistance to Lleaching, ' a large fraction of the fission producto i s.
retained in even completely unclad U ZrH fuel. , 42 _ _ _ . _ . _ _ , _ . . ~ _ _ _ _ _ _ _ . - . _ . . _ _ _ _ _ _ _ . , _ , . _ _ _ _ _ _ _ . . _ . _ . _ _ _ _ . _ . _ _ . _ _ _ - -
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SAR $/91 For the rest of the diecussion of fuel t empe raret c.1, we will concern ourselves with the higher hydride Ol/Zr 2 1.5) TRICA fuel clad with 304 stainless steel 0.020 in. (0.508 mm) thick, or a cladding . material equivalent in strength at the temperatures discussed. a At room temperature the hydride is like ceramic and shows little ductility. Itowever, at the elevated temperatures of interest for pulsing, the material is found to be more ductile. The effect of very t large thermal stress on hydride fuel bodies has been observed in hot cell observations to causa relatively widely spaced cracks which tend to be either radial or normal to the central axis and do not interfere with ' Since the segments tend to be orthogonal, their e radial heat flow. relative positions appear to be quite stable, The limiting offeet of fuel temperature then is the hydrogen gas over pressure. Figure 47 relates equilibrium hydrogen pressure over the fuel as a function of temperature for material with three different < il/Zr ratios. The_ hydrogen gas over pressure is not in itself' detrimental but il the stress produced by the gas pressure within the fuel can exceeds the ultimate strength of the clad material, a rupture of the fuel clad could occur. While the final conditions of fuel temperature and hydrogen .' prensure in which such an occurrence could come about are of interest, the mechanisms in obtaining temperatures and pressures of concern are different in the pulsing and steady state mode of operation, and each mechanism will be discussed indel.endently of the other. In this discussion it will be assumed that the fuct consists of tb Zril O(/Zr ; 1.65) with the uranium being 8.5 wt. 1 and further that the cladding can is 304 stainless steel. The clad thickness is 0.020 ; in. (0.508 mm) with an inside clad diameter of 1,43 in. (3.63 cm). - Theue fuel parameters have been chosen since they represent the nominal specifications- for TRIGA fuel elements. Figure 43 shows the characteristic of 304 stainless steel with regard to yield and ultimate strengths as a function of temperature. 3 In determining the stress applied to- the cladding from the i internal hydrogen gas pressure the equation S - P-r/t , (1) I applies where , I I l S - stress in pai, 1 l p - internal pressure in psi, r - radius of the cladding can, f
} t - wall thickness of the clad.
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SAR $/91 Then for the cladding we have approximately , S - 36.7 P , (2) or the stress applied to the clad is approximately 36.7 times the internal pressure. It is of interest to relate the strength of the clad material at its operating temperature to the stress applied to the clad f rora the internal gas pressure associated with the fuel temperature. Fi6 ure 4 4 gives information as to the ultimate clad strength as a function of temperature and also describes the stress applied to the clad as a result of hydrogen dissociation for fuel having a H/Zr ratio of 1.65 as a function of temperature. There are several reasons why the gas pressure should be less for the transient condit. ions than the equilibrium condition values would predict. For example, the gas diffusion rates are finite; surface cooling is believed to be caused by endothermic gas croission which tends to lower the diffusion constant at _ the_ - surface; reabsorption takes place concurrently on the cooler hydride surfaces away from the hot spot; there is evidence for a low permeability oxide film on the fuel surface; and some local heat transfer does take place during the pulso ; time to cause a less than adiabatic true surface temperature. 4.1.1.1 ruel and Clad Temperature. The following discussion relates the element clad temperature and the maximum fuel temperature during a short time after a pulse. The radial temperature distribution in the fuel element immediately following a pulse is very similar to the power distribution This initial steep thermal gradient at the fuci shown in Figure 4 5. surf ace results in some heat. transfer during the time of the pulse so that the.true peak temperature does not quite reach the adiabatic peak temperature. A large temperature gradient is also impressed upon the clad which can result in a high heat flux from the clad into the water. If the heat flux is sufficiently high, film boiling may occur and form an insulating jacket of steam around the fuel elements permitting the clad teroperature to tend to approach the fuel temperature. Evidence has been obtained experimentally which shows that film boiling has occurred occasionally for some fuel elements in the Advanced TRICA Prototype React or_ located at GA Technologies _[2]. The consequence of this film boiling was discoloration of the clad surface. Therinal transient calculations were made using the RAT computer c od e' . RAT is a 2 D transient heat transport code developed to account flow and temperature dependent material properties. l_ l- for fluid i calculations show that if film boiling occurs after a pulse it may take-l place either at the time of maximum heat flux from the clad, before the bulk teroperature of the coolant has changed appreciahly, or it may take j place at a much later time when the bulk teinperature of the coolant has
- approached the saturation temperature, resulting in a markedly ' reduced threshold for film boiling. Data obtaired by Johnson et al. [3] for 47
_u. _ _ _ ,. _ . ._ ... _ _ _ ~ _ .- - - _ . _ - . - .~... . _ . _ . _ _ .
SAR 5/91 i transient 0.9 to 2.0heating Mbtu/ft pfhr ribbons in 100'F for e-folding periods water, from showed burnout 5 to 90 fluxes of milliseconds. On the other hand, sufficient bulk heating of the coolant channeled between fuel elements can take place in several tenths of a second to i lower the departure from nucleate boiling (DNB) point to approximately 0.4 MBtu/f t%hr. It is shown, on the basis of the following analysis, that the second mode is the most likely; i.e., when filta boiling occurs it takes place under essentially steady state conditions at local water temperat.ures near saturation. A value -for the temperature that may be reached by the clad if film boiling occurs was obtained in the following manner. A transient thermal calculation was performed using the radial and axial power , distributions in Figure 45 7 and Figure 46, respectively, under the [ assumption that the therreal resistance at the fuel clad interface was ! nonexistent. A boiling heat transfer model, as shown in Figure 4 7, was ' used in order to obtain an upper limit for the clad temperature rise. The model used the dat.a of McAdams (4) for the subcooled boiling and the work of Sparrow and Cess [5] for the film boiling regime. A conservative estimate was obtained for the ininimum heat flux in film bo11 tog by using the correlations of Speigler et ab [6), Zuber [7], and . Rohsenow and Choi [8] to find the minimum temperature point a t. which ' film boiling could occur. This calculation gave an upper lirait of 760*C clad temperature for a peak initial fuel temperature of 1000*C, as shown in Figure 4 8. Fuel temperature distributions for this case are shown in Figure 4 9 and the heat flux into the water from the clad is shown in Figure 4 10. In this limiting case, DNB occurred only 13 milliseconds af ter the - pulse, conservatively calculated assuming a steady state DNB correlation. Subsequently, experimental transition and film boiling i data were found to have been reported by Ellion (9) for water conditions similar to those for the TRICA system. The Ellion data show the minimum heat- flux, used- in the limiting calculation described above, was conservative by a factor of 5. An appropriate correction was made which resulted in a more realistic estimate of 470'c as the maximum clad t en.pe ra ture expected if film boiling occurs, _ This result is in agreement with experimental evidence obtained for clad temperatures of - 400*C to 500*C for TRICA Mark F fuel elements which have been operated under film boiling conditions [10).
- - The preceding analysis assessing the maximum clad temperatures asnociated with film boiling assumed no thermal resistance at fuel clad interface. Meesurements of fuel ternperatures as a function of steady.
I state power level provide evidence that af ter operating at high fuel teraperatures,- a permanent gap is produced between the fuel body and the u clad-by' fuel expansion. -This gap exists at all temperatures below the maxiourn - operating - temperature. -(See. for- example,- Figure 1-16 in l
- Reference 10.) The gap thickness . varies with fuel temperature and clad temperature so_that cooling of the fuel or overheating of the clad tends t
' to - widen the gap and decrease the heat transfer rate. Additional l thermal . resistance due to oxide and other films on the fuel and clad I surfaces.ls expected. Experimental and theoretical studies of thermal contact resistance have been reported [11 13] which provide insight into the mechanisms involved. They do not, however, permit quantitative prediction of this application because the basic-data required for input-49 j_ E._...__.____._., _ _ . _ _ _ . . . . _ . _ _ _ . _ . _ _ , , _ _ . _ _ . _ _~
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'3 40 100 o,coi o.oi o.i 1.0
[ LAP $[0 TIME FROM IND OF PUL5E (SEC) SURFACE HEAT FLUX AT MIDPLANE OF WELL-BONDED FUEL El.EMENT AFTER A PULSE Figure 4-10 4-15
SAR $/91 are presently not fully known. Instead, several transient thermal I computations were made using the RAT cade. Each of these was made with an assumed value for the effective gap conductance, in order to determine the effective gap coefficient for which departure from nucleate boiling is incipient. These results were then compared with the incipient film boiling conditions of the 1000'C peak fuel temperature case. For convenience, the calculations were made using the same initial temperature distribution as was used for the preceding calculation. The calculations assumed a coolant flow velocity of 1 ft per second, which is within the range of flow velocities ecmputed for natural convect ion under various steady-state conditions for these reactors. The calculations did not use a complete boiliric curve heat transfer model, but instead, included a convection cooled redon (no boiling) and a subcooled nucleate boiling region without employing t.n upper DNB limit. The results were analyzed by inapection using the extended cteady-state correlation of Bernath [14] which has been reported by Spano [15] to give agreement with SPERT II burnout results within the experimental uncertainties in flow rate. The transient thermal calculations were gerformed using effective gap conductances of 500, 375, and 250 Btu /hr ft *F. The resulting wall - temperature distributions were inspected to determine the axial wall position and time after the pulse which gave the closest approach between the local computed surface heat flux and the DNB heat flux according to Bernath. The axial distribution of the computed and critical heat fluxes for each of the three cases at the time of closest approach is given in Figures 4-11 thru 4-13. If the minimum approach to DNB is corrected to TRICA Mark F conditions and cross-plotged, an estimate of the effective gap conductance of 450 Stu/hr-ft - 'F is obtained for incipient burnout so that the case using 500 is thought to be representative of standard TRICA fuel. The surface beat flux at the midplane of the element is shown in Figure 4-14 with gap conductance as a parameter. It may be observed that the maximum heat flux is approximately proportional to the heat transfer coefficient of the gap, and the time lag after the pulse for which the peak occurs is also increased by about the same factor. The closest approach to DNB in these calculations did not necessarily occur at these times and places, however, as indicated on the curves of Figures 4 11 thru 4 13. The initial DNB point occurged near the core outlet for a local heat flux of about 340 kBtu/hr f t - *F according to the more conservative Bernath correlations at a local water temperature approaching saturation. This analysis indicates that after operation of the reactor at steady-state power levels of 1 MW(t), or after pulsing to equivalent fuel temperatures, the heat flux through the clad is reduced and therefore reduces the likelihood of reaching a regime where there is a departure from nucleate boiling. From the foregoing analysis, a maximum temperature for the clad during a pulse which gives a peak adiabatic fuel temperature of 1000'C is conservatively estimated to be 470*C, 4-16
SAR 5/91 7 i i i i i
^
ELAPSED TIME FROM S END OF PULSE - 0.247 SEC _ i 6 - 1 3 E
~
ACTUAL HEAT FLUX ~ T 5 S x 5 d _ CRITICAL HEAT FLUX - a E i i i I 1 3 7 8 9 to 11 12 13 OlSTANCE FROM BOTTOM OF FUEL (IN.) SURFACE HEAT FLUX DISTRIBUTION FOR STANDARD NON-GAPPED (hg ap-500) FUEL ELEMENT AFTER A PULSE Figure 4-11 4-17
2 SAR 5/91 j 1-i;- l l l- ? i ! i O i i i i i i l- y l. CRITICAL HEAT FLUX l-9 ; M
'6 E
~
2 t 5 o' ACTUAL HEAT FLUX M 5 ~ i d , ti E-3 _ ELAP5E0 TIME FROM END OF PULSE 15 0.314 SEC
~
2 , , 8 9 10 11 12 13 15 7 DISTANCE FROM BOTTOM OF FUEL (IN ) l SURFACE HEAT FLUX DISTRIBUTION FOR STANDARD NON-GAPPED (hg ap-375)-FUEL ELEMENT AFTER A PULSE Figure 4-12 4-18
. ~ . . . - -____.-- -_....._-.._--_ _.-.-..._..,.--.- ,_-_ .
- . .. ._._. ~_=_.-.- - . .- ...... _-_- -.=~.-.. .- _-.. .--.. . -
SAR 5/91 O 8 i , . i , , i
~
7 CRITICAL HEAT FLUX
^
u [6 - a
-\ E
?
=s -
m
'o
~
ELAPSED TIME FROM END 0F PULSE IS 0,440 SEC _ 4 _ 5 t
- 3 E ACTUAL HEAT FLUX 2 -
i i i e i i i
-)
7 8 9 10 11 12 13 14 15 OlSTANCE FROM BOTTOM OF FUEL (IN,) SURFACE HEAT FLUX DISTRIBUTION FOR STANDARD l NON-GAPPED (hg ap-250) FUEL ELEMENT AFTER A PUISE Figure 4-13 4 19
. -- . . . - ~ . . . . _ . - - . . . _ . _ ~ - _ ..- . .. .- .. - .. - - . _ - - - , SAR 5/91 106 i i i i 1 i i i l EFFECTIVE HEAT TRANSFER -
- COEFFICIENT IN GAP, BTV/HR-FT2 'F 500 375 250 _
N 106- - t _ E - E 5 c' ti y O 0 E ce - N D
" 10 FLOW VELOCITY = 1 FT/SEC -
GAP THERMAL RESISTANCES ARE _ REPRESENTATIVE OF CONDITIONS AT _ END OF POLSE ( 1. E. TIME = ZERO)
? -
' ' ' ' I ' ' ' '
10 3 O.01 0.1 1,0 ELAPSED TIME FROM END OF PULSE (SEC) SURFACE IIEAT FLUX AT MIDPOINT VERSUS TIME FOR STANDARD NON-GAPPED FUEL ELEMENT AFTER A PULSE Figure 4-14 4-20
SAR 5/91 An can he seen from Figure 43, the ultimate st rengt h of the clad at a temperature of 470*C is $9,000 psi, if the stress produced by the hydrogen over pressure in the can is less than 59,000 psi, the fuel element will not undergo loss of containment. Referring to Figure 4-4, and considering U-Zrli iuel with a peak temperature of 1000*C, one finds the stress on the clad to be 12,600 psi. Further studies show that the hydrogen pressure which would result trom a transient for which the peak fuel temperature is ll50*C would not produce a stress in the clad in excess of its ultimate strength. TRICA fuel with a hydrogen to
.itconium ratio of a t. l e a s t. 1.65 has been pulsed to temperatures 01 about il50* C wit hout damage to the clad [16).
4 .1.1. 2 fl,' _, t e Diffusion Rat t To assess the effect of the finite dillusion rate and the rebydriding it the cooler surfaces, the following analysis in presented. As hydrogen is released from the hot fuel regions, it is taken up in the cooler regions and the equilibrium that is obtained in characteristic of some temperature lower than the maximum. To evaluate this reduced pressure, we will use diffusion theory to calculate the rate at which hydrogen is evolved and reabsorbed at the fuel surface. Ordinary diffusion theory provides an expression for describing the time dependent loss of gas from a cylinder: c ep 4 (7n Dt
- -- exp - , (3) et - cr , (2 r2 n-1 n 0 where .
et,cf,c, - the average, the initial, and the final gas concentration in the cylinder, respectively, In - the roots of the Equation Jo(x) -0, D - the diffusion coefficient for the gas in the cylinder, ro - the radius of the cylinder, t = time. Sett ing the term on the right-hand side of Equation 3 equal to x, one can rewrite Equation 3 as: c/cg - cr/ct + (1 - ct/ci) x , (4) and the derivative in time is given by d(c/c t) dx (1 - ct/et) - (5) dt dt 4-21
- _ . -- . _ _ . - --- . - - . - . - . . . . ~ . . - I
- i SAR 5/91 :
)
[
\'
This represents the fractional release rate of hydrogen from the cylinder, f(t). The deris>tive of the series in the right hand side of Equation 3 was approximated by dx de
- - - (7,339e.8.34e + 29.88e*249') -
, @)
dt dt 2 whern e_- Dt/ro The diffusion coefficient for hydrogen in zirconium hydride in which the H/Zr ratio is between 1.56 and 1.86 is given by D - 0.25 e-17800/R(T+273) , (7) where R - the gas constant and, T - the zirconium hydride temperature in 'C. Equation 3 describes the escape of gas from a cylinder through diffusion until some final concentration is achieved. Actually, in the closed system considered here, not only does the hydrogen diffuse into the fuel clad gap, but also it diffuses back into the fuel in the regions of lower fuel temperature. The gas also diffuses through the clad at a rate dependent on - the clad temperature, Although this tends to reduce the hydrogen pressure, it is not considered in this analysis. When the diffusion rates are equal, an equilibrium condition will exist. To account for this, Equation 5 was modified by substituting for the concentration ratios the ratio of the hydrogen pressure in the gap to the equilibrium hydrogen pressure, Ph/Pe . Thus, d(c'/c g ) dx f(t) - - (1 Ph (t)/Pe) - (8) dt dt where P h (t) - the hydrogen pressure, a function of time and Pe --the equilibrium hydrogen pressure over the zirconium hydride which is a function of the fuel temperature. The rate of change of the internal hydrogen pressure, in psi, inside the fuel element cladding is dPh 14.7 f(t) Nh 22.4 T+273
- . (9) de 23 V 273 6.02 x 10 g where Nh - the number of molecules of H2 in the fuel, T - the gas temperature ('C),
f(t) - the fractional loss rate from Equation 8, Vg - the free volume inside the fuel clad (liters). 4 22
SAR 5/91 As the atom density of hydrogen in Zril (ll/Zr ; 1.65) is about 0.1 moles and the fuel volume is 400 cubic cm., Nh is 19.9 moles (112 ) . The free volume is assumed to consist of a cylindrical volume, at the top of the element, 1/8 in. high with a diameter of 1.43 in, for a total of 3.3 cubic cm. Also, the temperature of the hydrogen in the gap was assumed to be the temperature of the clad, The effect of changing these two assumptions was tested by calculations in which the gap volume was decreased by 90% and the temperature of the hydrogen in the gap was set up equal to the maximum fuel temperature. Neither of these changes resulted in maximum pressures different from those based on the original assumptions although the initial rate of pressure increase was greater. For these conditions Ph - 7,29 x 103 (T + 273) f(t) dt . (10) The fuel temperature used in Equation 7 to evaluate the diffusion coefficient is expressed as T(z) - To ;t<0 , (11) T(z) - To + (Tm - To) cos [2.45(z 0.5)] ;t20 , where Tm - the peak fuel temperature (*C), Os To - the clad temperature (*C), z - the axial distance expressed as a fraction of the fuel length, t- the time after step increase in power. It was assumed that the fuel temperature was invariant with radius. The hydrogen pressure over- the zirconium hydride surface when equilibrium prevails is strongly temperature dependent as shown in Figure 4-2 and, for ZrH (H/Zr ; 1.65), can be expressed by 9 -1.974x106/(T+273) (12) Pe - 2.07 x 10 e . The coefficients have been derived from data developed by Johnson. The rate at which hydrogen is released or reabsorbed takes the form [Pe (z) - Ph (t)I g(t,z) - f(t,z) , (13) Pe (z) i where f(t,z) - the derivative given in Equation 8 with respect to time evaluated at the axial position z, Ph (t) - the hydrogen pressure in the gap at time t, ('~' 5 Pe(z) - the equilibrium hydrogen pressure at the ZrH temperature at position z. 4-23
-. . _ . - - - . . - . . . . - . - - - . ~ . _ - - - - . . . - - . . . - - . _ . . - . . _ -
SAR 5/91
\
'N The internal hydrogen pressure is then Ph (t) - 7.29 x 103 (To 4 273) g(t,z) de dt . (14) 0 0 This Equation was approximated by n m '
Pg(ti.1) i 1 - Ph (ti) - 7.29 x 103 (To + 273) x
, PE (Zj) 1-1 j-1 x f(tt,zj) 6z 6t , (15) where the internal summation is over the fuel element length increments and the external summation is over time.
For the case in which the maximum fuel temperature is 1150*C, the equilibriurn hydrogen pressure in Zril ' (H/Zr ; 1,65) is 2000 psi. Calculations indicate, however, that the internal pressure increases to a peak at about 0.3 sec, at which time the pressure is about one-fifth of the equilibrium value or- abut 400 psi. After this time, the pressure s slowly decreases as- the hydrogen continues to be redistributed along the length of the element from the hot regions to the cooler regions. Calculations have also been made for step increases in power to peak fuel temperatures greater than 1150*C, Over a 200*C range, the time to the peak pressure and the fraction of the equilibrium pressure value achieved were approximately the same as for the 1150*C case. Thus, if the clad remains below about 500*C, the internal pressure that would produce _ the -yield stress in the clad _ (35,000 psi) in about 1000 psi and the corresponding equilibrium hydrogen pressure corresponds to a maximum fuel temperature of about 1250'C in ZrH (H/Zr ; 1.65).
-Similarly, an internal pressure of_1600 psi _would produce a stress equal to the ultimate clad strength (over.59,000 psi). This corresponds to an equilibrium hydrogen pressure of 5 x 1600 or 8000 psi and a fuel temperature of about 1300*C.
Measurements of hydrogen pressure in TRICA fuel elements during steady-state operation have not been made. However, measurements have been made during transient operations and compared with the results- of These measurements an analysis similar to that described here. indicated that in a pulse in which the maximum temperature in the fuel was greater than 1000*C . the maximum pressure was only about 6% of the l equilibrium value - evaluated at the peak - temperature. Calculations of l- the pressure resulting from such a pulse using the methods described ! above gave_ calculated-pressure values about three times greater than the l measured values. l An . instantaneous increase in fuel ternperature will produce the l most severe pressure conditions. When a peak fuel temperature of 1150*C is - reached by increasing the power over a finite period of time, the resulting pressure will be no greater than that for the step change in 4-24
___-....-___-__.__.-m___....__ _ __ . SAR 5/91'
. power analyzed above. .As the temperature rise times become long compared with the diffusion time of hydrogen, the prescure will become increasingly less than for the case of a step change in power. The reason for this is that the pressure in the clad element results from the hot fuel dehydriding f aster than the cooler fuel rehydrides (takes
-up the excess hydrogen to reach an equilibrium with the hydrogen over pressure in the can) . The slower the rise to peak temperature, the lower the pressure because of the additional time available for rehydriding 4.1.1.3 Summary The foregoing analysis gives a strong indication that the clad will not be ruptured if fuel temperatures are never greater than in the range of 1200'C to 1250'C, providing that the clad temperature is less than about 500*C. However, a conservative safety limit of 1150'C has been chosen for this condition. As a result, at this. safety limit temperature the pressure is about a factor of 4 lower than would be necessary for clad failure. This factor of 4 is more than adequate to account for uncertainties in clad strength and manufacturing tolerances.
Under any condition in which the clad temperature increases above 500*C, the temperature safety limit must be decreased as the clad material loses strength at elevated temperatures. To establish this limit, it is assumed that the fuel and the clad are at the same e temperature. There are no conceivable circumstances that could give rise to a situation in which the clad temperatures was -higher than the fuel temperature. In Figure 4-4 there is plotted the stress imposed on the clad by l the equilibrium hydrogen pressure as a function of the fuel temperature, again. assuming a clad radius of 0.73 in, and a thickness of 0.02 in. Also shown is the ultimate strength of 304 stainless steel at the same temperatures. The use of these data for establishing the safety limit is justified as a, the method used to measure ultimate strength requires the imposition of the stress over a longer time than would be L imposed for accident conditions,
- b. the stress is not applied biaxially in the ultimate strength measurements as it is in the fuel clad.
The point at which the two curves in Figure 4-4 intersect is the safety limit, that is, 970*C. At that temperature the equilibrium hydrogen pressure - would impose a stress on the clad equal to the ultimate strenSth of the clad. The same argument about the redistribution of the hydrogen within the fuel presented earlier is valid for this case also. In addition, at elevated temperatures the clad becomes quite permeable to hydrogen. Thus, not only will hydrogen redistribute itself within the fuel to reduce the pressure, but also some hydrogen will escape from the system entirely. 4-25 l
-~ . -_. -. . -- . . - . -
SAR 5/91 The use of the ultimate strength of the clad material in the establishment _ of the safety limit under these conditions is justified because of the transient nature of such accidents. Although the high clad temperatures imply sharply reduced heat transfer rates to the surroundings (and- consequently longer cooling times), only slight reductions in the fuel temperature are necessary to reduce the stress sharply. A 50*C decrease in temperature from 970'C to 920*C will reduce the stress by a factor of 2. As a safety limit, the peak adiabatic fuel temperature to be allowed during transient conditions is considered to be 1150'c for U-Zrill.65-4.1.2. Prompt Nerative Temperature Coefficient The - basic parameter .thich allows the TRIGA reactor system to operate safely - with large step insertions of reactivity is the prompt negative temperature coefficient associated with the TRICA fuel and core
' design. This temperature coefficient (a) also allows a greater freedom in steady-state operation as the effect of accidental reactivity changes occurring from the experimental devices in the core is greatly reduced.
CA Technologies, the designer of the reactor, has developed techniques to calculate the temperature coefficient accurately and t.herefore predict the transient behavior of the reactor. This temperature coefficient arises primarily from a change in the disadvantage factor resulting from the heating of the uranium zirconium hydride fuel moderator elements. The coefficient is prompt because the fuel is intimately mixed with a large portion of the moderator and thus fuel and. solid moderator temperatures rise simultaneously. A quantitative calculation of the temperature coefficient requires a knowledge of the energy dependent distribution of thermal neutron flux in the reactor. The basic physical processes which occur when the fuel moderatar elements are heated can be described as- follows: the rise in temperature of the - hydride increases the probability that a thermal neutron in the fuel element will gain energy from an excited state of an oscillating hydrogen atom in the lattice. As the neutrons gain energy from the Zril, their mean free path is increased appreciably. This is shown qualitatively in Figure 4 15. Since the average chord length in the fuel element is comparable with a mean free path, the probability of escape from the fuel element before capture la increased. In the water the neutrons are rapidly rethermalized so that the capture and escape probabilities are relatively insensitive to the energy. with which the neutron enters the water. The heating of the moderator' mixed with the fuel thus causes the spectrum to harden more in the fuel than in the water. As a result, there is a temperature dependent - disadvan.tage factor for the unit cell in the core which decreases the ratio of absorptions in the fuel to total cell absorptions as the fuel element temperature is increased. This brings about a shift in the core neutron k balance, giving a loss of reactivity. 4-26
SAR S/91 O 10 0 i O g 80 U a R 400*C t
$ 60 -
r2 Es w 40 -
"'b is nw
" " 20 -
1l t l l f f f t t 0.01 0.1 1.0 NEUTRON ENERGY (eV) TRANSPORT CROSS SECTION FOR llYDROGEN IN Zril AND AVEPACE NEUTRON SPECTRA IN FUEL ELEMENT Figure 4-15 l 4 27
.. -. - ~ ~ - - - - - - - - - - . ~ . - - - . - - - - ~ _ _ . .
SAR $/91
\ The temperature coefficient then, depends on spatial variations of the thermal neutron spectrum over distances of the order of a mean free
- path with large changes of mean free path occurring because of the energy change in a single collision. A quantitative description of these procensea requires a knowledge. of the - differential slow neutron energy transfor cross section in water and zirconium hydride, the encrgy dependence of the transport cross section of hydrogen as bound in water and zircontuu hydride, the energy dependence of the capture and fission cross sections of all relevant materials, and a multigroup transport theory reactor description which allows for the coupling of groups by speeding up as well as by slowing down.
4.1.2.1. Codes Used for Calculations. Calculational work on the t.emperature coef ficient made use of a group of codes developed by CA Technologies: GCC 3 [17), GAZE 2 [18), and GAMBLE 5 (19), as well as DTF-IV [20). . an S n multigroup transport coda written at Los Alamos. Neutron cross sections for energies above thermal (21 eV) were generated by the CCC+3 code. In this code, fine group cross sections _ (-100 groups), - stored on tape for all commonly used isotopes, are averaged over a space independent- flux derived by solution of the B1 equations for each discrete. reactor region composition. This code and its related u cross-section library predict the age of each of the common moderating materials to within a few percent of the experimentally determined values and use the resonance integral work of Adler, llinman, and Nordheim [21] to generate cross sections for resonance materials which are properly averaged over the region spectrum. Thermal cross sections were obtained in essentially the same manner using the CGC 3 code, llowever, scattering kernels were used to describe properly the interactions of the neutrons with the chemically bound moderator atoms. The bound hydrogen kernels used for hydrogen in the water were generated by the TilERMIDOR code [22] using thermalization work of Nelkin_[23). Early thermalization work by McReynolds et a1 [24) on zirconium hydride has been greatly extended at GA Technologies [25), and work by parks resulted in the SUMMIT [26) code, which was used to
. generate the kernels _ for hydrogen as bound in Zril. These scattering models have been used to predict adequately the water and hydride (temperature dependent) spectra as measured at the GA Technologies linear accelerator as shown in Figure 416 and Figute 4-17 [27) .
4.1.2;2. ZrH Model. _ Qualitatively, the scattering of slow neutrons by zirconium hydride can be described.by a model in which the hydrogen atom motion is treated as an isotropic harmonic oscillator with energy transfer quantized in multiples of ~0.14 eV. More precisely, the SUMMIT model uses a frequency spectrum with two branches, one for the' optical modes for energy transfer with the bound proton, and the other for the acoustical modes for energy transfer with the lattice as a whole. The optical modes are represented as a broad frequency band centered at 0.14 eV, and whose width _ is adjusted to fit the cross section data of Woods et'al. -[28). The low fraquency acoustical modes are assumed to have a Debye spectrum with a cutoff of 0.02 eV and a weight determined by an- [- effeetive mass of 360. i 4-28
SAR 5/91 6 ,,, , , , , , , , ,, , , , , 10 _
~
T = 316 *C "" . a:"';;.':.LT. _ . . . -. .J.".T~ ..n .
- . .m
'- -----'n---'-- ~
105 -
/ .=. -- ,iaa.,.~ =
/ /./ T = 23,2*C x e
n / .. . g .
/ ,
/ "T= ISO *C D 4 / .
'\ -
10 - f r'T = 30*C \ : e - -s
< ~ / / 's. -
tt F-
, / . '. '..
,/ s' .
tn / *. W s* .
< / ..
103 - . W - s -
~. '. .
2 g,
** ~'
lQ -
=.
, , ,,I i i iil i iiil e i iiI ' ' ' i 10l 0.0 01 0.01 0.1 1.0 10.0 10 0 NEUTRON ENERGY (EV)
A COMPARISON OF NEUTRON SPECTRA BETWEEN EXPERIMENTS AND SEVERAL HYDROGEN MODELS Figure 4 16 4-29
SAR S/91 b ( lO6 : , , , i n oi i i i i i o ni i e inmi i i i i s ut I ROOM TEMPERATURE , 5 r 10 :
~
~
150 C - 4 10 - u) : ' . '* . b : . : z . *$. - a _ s - 316'C ' ' .. 3 ~ . _ x ., *.,
<x -
s* .
% 3
- 10 -
b
- co g
E
- s. -
4 -
,s. , -
G m 468*C .* s \O 2 . 10 7 . s
~
n -
\., [
ZrH l75 00RON POISONED , 3.4 e ARNS/ HYDROGEN ATOM 10' -.
.s* ~
, ***** OATA , _
~ ~
EINSTEIN OSCILLATOR .
~
MODEL INCLUDING ACOUSTICAL. . TR ANSITIONS
- O imi...t - - *- u t i i i a i i i ii i - -
0.0 01 0.01 0.1 1.0 10,0 NEUTRON ENERGY (EV) EFFECT OF TEMPERATURE VARIATION ON ZlRCONIUM llYDRIDE NEUTRON SPECTRA Figure 4-17 4-30
SAR 5/91 O V This- structure then_ allows a neutron to slow down by the transition in energy units of -0.14 eV as long as its energy _ is above 0.14 eV, Below 0.14 eV the neutron can still lose energy by - the inefficient process of exciting acoustic Debye type modes in which the hydrogen atoms - move in phase -with the zirconium atoms, which in turn move in- phase with one another. These modes therefore , correspond to the motion of a group of atoms whose mass is much greater than that of hydrogen, and indeed even greater than the mass of zirconium. Because of the large effective mass, these modes are very inefficient for thermalizing neutrons, but for neutron energies below 0.14 eV they provide the only mechanism for neutron slowing down within the ZrH. (In a TRIGA core, the water also provides for neutron thermalization below 0,14 eV.) In addition, ir. the Zrli it is possible for a neutron to gain one or more ener6y units of -0.14 eV in one or several scatterings, from excited Einstein oscillators. Since the number of excited oscillators present in a Zril lattice increases with temperature, this process of neutron speeding up is strongly temperature dependent and plays an important role in the behavior of Zrli moderated reactors. 4.1.2.3. Calculattons. Calculations of the temperature coefficient were done in the following steps:
- a. Multigroup cross sections were generated by the CGC-3 code for a homogenized unit cell. Separate cross-section sets were generated for each fuel element temperature by use of the temperature dependent hydride kernels and Doppler g
broadening of the U 238 resonance integral to reflect the proper temperature, Water at room temperature was used for all prompt coefficent calculations. ! b. t, value for k. was computed for each fuel element temperature by transport cell calculations, using the P1 approximation. Comparisons have shown S4 and Sg results to be nearly identical. Group dependent disadvantage factors were calculated for each cell region (fuel, clad, and water) where the disadvantage factor is defined as the ratio:g trj eg" (region / cell).
- c. The thermal group disadvantage factors were used as input for a second CCC-3 calculation where cross sections for a homogenized core were generated which gave the same neutron
-balance as the thermal group portion of the discrete cell calculation.
- d. The cross sections for an equivalent homogenized core were used in a full reactor calculation to determine the contribution to the temperature coefficient due- to the increased leakage ' of thermal neutrons into the reflector with incraasing hydride temperature. This calculation still requires several thermal groups , but transport effects are no longer of maj or concern. Thus, reactivity calculations as a function of fuel element temperature have been done on the entire reactor with the use of diffusion theory codes.
4 31
. _. - .. . - . . -.- . _ - - - . _ _ - _ . . - - .- ~ - -
SAR 5/91 e \ Results from the above calculations indicate that more than 50% of the temperature coefficient for a standard TRIGA core comes from the temperature-dept.ndent disadvantage factor or " cell effect", and ~201 each from Doppler broadening of the U 238 resonances and temperature dependent leakage from the core, These effects produce a temperature coefficient of - -0.01%/'C, which is rather constant with temperatut e. The temperature coefficient is shown in Figure 4 18 for the high-hydride core of this TRIGA. 4.1.3. Steady-State Reactor Power The following evaluation has been made for a TRIGA system operating with cooling from natural convection flow around the fuel elements. This analysis investigates the limits to which such a system may be operated. The analysis was conducted by considering the hydraulic characteristics of the flow channel from which the heat rejection rate is maximum. The geometrical data from this channel are given in Table 4-2. All symbols in Equation 16 through 45 are defined in the list of nomenclature in Section 4.1.3.9. Table 4 2 IlYDRAULIC FLOW PARAMETERS f k Flow area (ft /2 element) 0.00580 Wetted perimeter (ft/ element) 0.3861 Hydraulic diameter (ft) 0.0601 Fuel element diameter (ft) 0.1229 Fuel surface area (ft2) 0.4826 The heat generation rate in the fuel element is distributed axially -in a cosine distribution chopped at the end such that the peak-to-average ratto is 1.25. The number of fuel elements in the core is assumed for 1 MW operation, but the departure from nucleate " boiling (DNB) ratio is conservatively evaluated on the basis of 85 elements. The driving force is supplied by the buoyance of the heated water in the core. Countering this force are the contraction and expansion losses at the entrance and exits to the channel, and the acceleration and potential energy losses and friction losses in the cooling channel itself. J 4-32
. _ . _ . ~ . _ _ _ _ _ _ . _ _ _ . _ _ . _ . . _ . _ _ . _ _ _ . . _ _ _ . . _ _ _ _ . _ . _ _ _ _ . . _ . _
SAR $/91 O
-14 h
N STAINLESS STEEL CLAD-
- 8,5 WT-% U-Zr H i,so CORE
-12 -
'o_.
x e _go _ H z W U E -8 u. W o U O- g _s _ o 4
.m-E -4 -
2-w V' H-
- o. -2 -
s-o e Q. L I I I I I L O 0 200 400 600 800 1000 /200 l-TEMPERATURE ('C) PROMPT NECATIVE TEMPERATURE COEFFICIENT O VERSUS AVERAGE FUEL TEMPERATURE FOR TRIGA l Figure-4-18 4 33
_ . _ _._. _ _.. -_ _ _ - - - . _ _ _ _ _ _ _ . _ - _ ~ _ _ . . _ - _ _ ~ . . - . _ _ _ SAR 5/91-
\ .
Figure .4 19 illustrates schematically the natural convection system established by the fuel elements bounding one flow channel in the core. The system shown is general and does not represent any specific configuration. Steady state flow is governed by the Equation 6pt + 6pe + 6pt + opu 4 Oj-It/Vo P , (16) jd1 where the lef t-hand member represents the pressure drops through the flow channel due to entrance, exit, friction, acceleration, and gravity losses and the right hand member represents the dt iving pressure due to the static head in the pool. The pressure drops throu6h the flow channel are dependent on the flow rate while the available static driving pressure is fixed for a known core height and pool temperature. The analysis, therefore, becomes an iterative one in which the left hand side of Equation 16 is evaluated on the basis of an assumed flow rate
.and compared with the known right hand side until equality is achieved.
The method has been programmed for digital computer solution. The methods of evaluating each of the 6p terms in Equation 16 for known power distribution and flow geometry and assumed flow rates are discussed below. 4.1.3.1, Entrance Loss, 6pt. ( . The entrance loss, 6pt, may be evaluated in the usual way as a 7 fraction-of the velocity head in the lower grid plate hole: (ki , k 12) (17) 6pg - (NW)2 , 2g Ai ! where N - the number of channels which receive their flow l from a single hole in the lower grid plate, k it - the loss factor for the entrance to the hole in the lower grid plate. For even slight rounding of the entrance, kit will be no greater than 0.30,
.k 12 - the loss' factor covering transfer of the flow from l the hole in the lower grid plate to the coolant l
channels. In most cases this can be N satisfactorily approximated as a sudden expar.aion k using i2 - 1.0. 4 34
SAR 5/91 b
-( CHANNEL SURFACE TO VOLUME RATIO, S/V CALCULATED FLOW AREA, Af FROM GIVEN HEATED PERIMETER, P OIMENSIONS EQU! VALENT DIAMETER, D e A
/
/ <\ FREE SURFACE
/ OF POOL y VIEW A-A p n\/a 2 . zt M 3"h E ^pe A A
{ 2 *zn+: -- - - " 6p p= +P amb __ t = n+1 j=n p T sat O n il P0OL AT CONSTANT
" = q"(z) TEMPERATURE, T g
GIVEN) j=1 k=1 z=zt
/ CROSS FLON, OR opg FLOW BETWEEN z=0 ADJACENT CHANNELS, a { Api: IS IGNORED
,_ q W475 Wb sp;'
/(
COOLANT INLET i HOLE OF AREA A j l l GENERAL FUEL ELEMENT CONFIGURATION FOR SINGLE COOLANT CilANNEL IN THE TRIGA Fip,ure 4-19 i 4 35 l
i SAR 5/91 4.1.3.2, Exit Loss, 6pe, The exit loss is expressed in terms of a coefficient Ke which is the - f raction of the velocity head in the flow channel which is not recovered: Ke v n+1 6pe - N * ( } 2gA[2 The term vn ,1 is the specific volume at the highest axial station dong the heated length of the core. It is evaluated from the temperature Tntl which is obtained from an overall heat balance: Tn+1 " 9t/VC + To ,- (19)
* *n+1 where qt - P q"(z)'dz .
Loss Throuch Portion of Channel Adjacent to Lower 4.1.3.3. Reactor 6pl. The flow ~is isothermal at the bulk pool temperature so that fm Vo 6zt y2 6zt 6 p . + , (20) 2g De At vo f, is evaluated from the Moody chart (assuming smooth surface) on the basis of a Reynolds number which is De vo Re W. '(2U Ag vo 4.1.3.4. Loss Throuch Portion of Channel Adiacent to Uoner Reactor 6pu-The flow is isothermal at .'n+1 where Tn+1 is determined by Equation 19 ' 1 fm Vn 0Zu 6z u 9
~
- + ~
6Pu 2g De ^r 2 vn f, is again. evaluated from the Moody chart, assuming smooth surface, on the basis of a Reynolds number which is De vn Re - V < (23) A( vu. 4-36
w il SAR 5/91 4.1.3.5. Loss Through Each Increment of the Channel Adjacent to the Fueled Portion of the Elementa. 6 pj. For the present, let us assume that the entire heated portion of the channel is in subcooled boiling. This implies that the wall temperatures calculated from subcooled boiling correlations are lower than those calculated for convection alone and that the liquid is below its saturation temperature at all locations. The pressure drop through an increment is given by v ,v f bj V ej 6z 6z 6pn - (n+1) - W + w + ~ (24) 2 2 v gat 2g At De mj (acceleration) (friction) (gravity) 4.1.3.6, Acceleration Term.
"a denotes the mean specific volume and is larger than the liquid specific volume, because of the vapor voidage:
vm - v/(1-a) . (25) o is the void fraction or the fraction of a channel cross section which is occupied by vapor, a may be calculated from the vapor volume (cubic in. vapor / square in. heating surface) and the flow channel geometry. Denoting the vapor volume as (, a - ( (S/V) (26) where S/V is the surface to volume ratio of the coolant channel. The parameter (, is dependent on the surface heat flux, the subcooling of liquid and the velocity of the liquid. It can be evaluated only by experiment. Data given by Jordan and Leppert [29] were used to estimate (; these data are plotted in Figures 4 20 and 4-21. Most of this represents a flow velocity of 4 ft/sec and appears to. be the only available data applicable under the thermal conditions encountered in TRICA type reactors. Extrapolations from these data are made for flow velocities different from 4 ft/sec. The extrapolations were based on a small amount of data given for flow velocities other than 4 ft/sec. The liquid temperature at a station. T ,k may be calculated from
- zk P q"(z)dZ zi Tk - + T o (27)
UC Therefore, one finds ( (Figure 4-21) from Tsat - Tk and qk", where Tsat and qk" are known. Since ok ~ (k (S/V) and vk is a function of T k Vmmay be evaluated from Equation 25. 4-37
.. . . ~ ,. .. . . . ~ - . - ~ . - . - . . . . . - = . . . . - . - . . - - . . - - - . - . ~ . . - . - . . _ . _ - - . . ~ . . -. . . , ~ _ .
SAR 5/91 O
- 1. . , , I , , ,
$UBC00 LING, (T *
- sat 48'F 10 2
=
; ~
78'F -
". 108'F
_5 . . u w 6 s, . - E t 10*3 - PRES $URE = 16.4 PSIA ~
~
Ft.0W vtLOCITY
- 4 FT/5EC 10' I t i t i I O 2 4 6 8 10 12 II.
2 HEAT FLUX, q" (BTU /liR FT X 10*5) EXPERIMENTALLY DETERMINED VAPOR VOLUMES FOR SUBC00 LED BOILING IN A NARROW VERTICAL ANNULUS t Figure 4-20 4-38
. _ _ _ _ ._..____....____._.____m._- ..,_._..m.~.________.-..._.__.._.__.___.__ _
.-. . . . . . . _ . ~ . . _ _ . .
SAR 5/91 O 10*I i , , , , , 10 2 y -
-/ . .
5 u w 5 C s> ( - qsf X 10* E D 5.00
> N 4.75 30 3 -. 4.50 -
~
4.25 ~ 4.00-
- 3.75 .
3.50 3.25 ~ 3.00 2.15 2.50 2.25 2.00 1.75 1.50 i i . t . 2 5, 4 1 e 1 0 20 40 60 60 100 120 140
$UBC00 LING. T ut .T CROSS PIDT OF Fibure 4 20 USED IN CALCULATIONS Figure 4-21 4 39 . _ _ _ . , _ _ _ . . _ . . , . ~ ... _ , _ ___,. . , . . _ . _ . _ _ . _ _ . , _ _ _ . . _ , . _ _ _ _ , . _ , . _ . _ _ _ _ . _ _ . , . .
SAR 5/91 b 4.1.3.7, Friction Term. vaj- denotes - a linear average of the mean specific volumes at the upper and lower boundaries of an increment. The approximate mean value is assumed to apply over the entire increment so that V mk 4 mk + 1 v mj
- (28) 2 A friction factor f bj is applied locally to calculate the friction pressure drop over the increment in subcooled boiling, Jordan and Leppert develop the correlation 8hb 8 q" 8St- - (29) fb - .
p CV p CV (Tw T) and provide experimental verification near atmospheric pressure in the range 0.0015 < St < 0.0050. This is simply an extension of Reynolds' analoFy to the case of subcooled boiling. The Equation of continuity is used to write Equation 29 as 8 q" At fb - . (30) WC (Tv - T) which may be evaluated if Ty is known. For subcooled boiling, the heat transfer is usually defined by an experimentally determined correlation of q" vs (Tw - Tsat) which has been obtained over a given range of flow velocity and pressure, McAdams (30] gives such a correlation for
. pressures between 2 and 6 atmospheres and flow velocities between 1 and 12 ft/sec. This correlation will be used to determine Ty for use in Equation 30.
Approximate mean values are assumed - to apply over the entire
. increment so that
~
I 8Af q"k 9"k+1 (31) 1/2 + bj - WC. ,Tw,k -Tk Tw,k+1 Tk+1 and ((q"k) + 4(9"k+1)- ' . Ty
, T ,tj 3 .
2 where d(q") is the correlation of McAdams previously cited. P (s 4-40
SAR.5/91 i t N 4.1.3.8. Gravity Term, The gravity term is evaluated from vj calculated from Equation 28. As implied in Section 4.1,3.5. , each increment must be checked to determine whether heat is being transferred by subcooled bofling or by convection. Tw is evaluated at the lower boundary of the incremenL on the basis of both the correlation from McAdams for subcooled boiling and a ' standard correlation for convection (Dittus Boelter). If the Tw calculated from convection correlations is less than that obtained for subcooled boiling, boiling is assumed not to be present in the
' increment. Equation 24 still applies, but since there is no boiling and hence no vapor void, vm becomes y and fb becomes fm-In the foregoing analysis an assumption was made that all of the vapor formed on the surface of the fuel element detaches and adds to the
. fluid buoyancy. This is not a conservative assumption. The position
-where vapor _ bubbles first leave the heated surface is obtained from two considerations; first, the balance of the forces exerted on the vapor bubble while it is in contact with the wall (buoyancy, surface tension, and friction), and, second, the temperature distribution in the single phase liquid away from the walls.
Determination of the buoyance forces resulting from the formation and subsequent detachment of vapor. bubbles is complicated by the g difficulty in predicting the point at which the. vapor detaches, and the fraction of that vapor which subsequently condenses. The problem was simplified by making use of an analysis performed by Levy [31] to determine the position at which the vapor detaches from the wall, assuming that at that point all of the vapor detaches and, finally, that there is no recombination of the vapor with subcooled fluid,
-According to Irry the position at which the vapor leaves the surface is.obtained from considering the balance of forces exerted on the vapor bubble while it is in contact with the wall, and the temperature distribution in the sin 61e Phase liquid away from the wall.
The forces acting on the bubble in the vertical direction consist of a buoyant force, F; B a frictional force, Fy, -exerted by the liquid on the bubble; and a surface tension force, F, S vertical component. The buoyant force, P, B is given by Cg rg3 (pt - py)g Fg - - , (32) Sc where rg is the bubble radius, Cg is a proportionality constant, pt and py are the liquid and vapor density, g is the acceleration due to. gravity and ge is a conversion ratio from lb force to lb-mass. The frictional force, Fy, is related to the liquid frictional pressure drop per unit length, (-dp/dz)p. The pressure differential acre s the bubble is proportional to the pressure differential times the bubb. radius and it acts across an area proportional to the square of the bubble radius. 4 41
_ ..~ ~ _ _ ~ Y l SAR $/91 Relating the pressure differential to the wall shear stress ty by [ j *(dp/dr.)y - 4 tw/D}{ , (33) I there results for ry: ' 3 (34) fp a CF ( 'w / DH ) rg , where Cp is a constant of proportionality and D}i is the hydraulic , diameter (four times the cross sectional area divided by the wetted perimeter). The surface tension force. Ts is 6 ven 1 by Fs - Cs rg o , (35) where C3 is a pioportions11ty const ant and a is the surface tension. Assuming upward flow the balance of these forces results in the following solutions for the bubble radius: > 1/2 C3o rg - - . (36) CB ( B/Sc ) (PL Pv) + Cp ( tw/Dn ) , l Assuming that the distance irom the wall to the tip of the bubble is . proportional to the bubble radius, a non dimens- nal distance > corresponding to this real distance can be gioch by 4 (a 6c Dl- pt)I/2 g (pt py) Dn 1/2 14 C' - (37) Ya - I'b - Ec Tw . where C and C' are appropriate constants. For those cases where the fluid forces are considerably creater than the buoyant forces this ; expression reduces to. Yg - C (o ge D}; pt)U2 1/pt . (38) != for the bubble to detach, the fluid temperature at the tip of the bubble must exceed the saturation tereperature by an amount such that the pressure differential acting across the interface at the tip of the bubbic balances the surface tension forces at the same position. By
- using the Claustus Clapeyron solution of this pressure differential one finds that the fluid temperature saturation temperature differenco can be assumed to be zero, The temperature at the tip of the bubble can also be specified f rom existing soluttens for the fluid temperature distribution. Thus, i if the flow is aswumed to be turbulent, and using the solution proposed by Martinelli, we have Tw T3 - # Pr YB ; OsY3s5 (39)
- 58 (Pr 4 in [1 $ (Ya/5 1))) ; 5 $ Yg s 30 s
- 58 (Pr + In [14 5 Pr) + 0,5 in (Y3/30]l ; Yg 2 30.
4 42 l-i
, a .--.. - ...-,, - - .-. ~. _ - . = _ ~,, _ _,,,,, _ . - - _ _ . _ . - - , - , . . , , . . . . - - . , , . ~ . _ , , . .
G SAR 5/91 The parameter # is a non dimensional term defined through the heat flux and liquid specific heat, that is, q/A
# - -- (40) 91, Cpt (fuge/pt) @
Levy obt ained values for the constanta C and C' by correlatten with availahic experimental data. Usin6 the accepted heat transfer relation from Dittus Boelter, one obtains hDg/kg,- 0.023 (WDg/pt,)0.8 (Pr)0.4 , (43) Calculating the friction factor from f - 0.0055(1 4 [20,000(e/Dg) 4 106/(WDg/pt,)]I/31 , (42) we are able to find the Wall shear stress from 2 tw - (f/8) (W /pt ge) , (43) The correlation with experiment yielded values for the constants of C - 0.015 , (44) C' - 0 . Finally, from the defir.ition of the heat transfer coefficient, one obtains-Tv - T - 4"/1. . (45) and _ setting _ the _ bubble tip temperature, _ Tg. equal to the saturation temperature. Tsat, we can express the relationship between the saturation temperature, tne wall temperature, and the fluid temperature at which the bubbic would detach from the wall by (Tw Tant)/(Tw T) - 0,023 (WDg/pt) 0 2 (Pr) 0.6(gjg)-0.5 i), (46) where 0 - Pr Ya ; OsY355
- 5 (Pr 4 in il 4 Pr (0.2 Y3 - 1))) ; 5 s Yg s 30
- 5 (Pr 4 In (1 + SPr) 4 0.5 in (Yg/30)) ; Yg 2 30.
( The solution of the force balance equation with void detachment was accomplished by iterating on the void detachment point to find where the right and left sides of Equation 46 were equal, The point at which the void was assumed to separaro from the surfo.:e was t.aken as the point at which equality obtained. ( 4-43 l
- - - - - - - - - - - . --- - - - ~ . . - - - . . - - . ~-
SAR 5/91
- The peak heat flux, that is, the heat flux at which there is a l departure froin nucleate boiling and the transition to filia boiling ber. ins, was deterinttied by two correlations. The fitet, given by McAdams l ~l/ l , linit ral en I hnt the penk heat ilun ia a Ionet ton of Ihe iluid velocity and the fluid only. The second correlation is due to be s tia t h
[33). It encornpasse s a wider range of variables over which the coraclation was made and it takes into account the effect of different i flow geometries. It generally gives a lower value for the peak heat flux and la the value used here for determinitig the rainiteurn DNB rat to, , that is, the minirnurn ratto of the local allowable heat flux to the actual heat flux. In general, the McAdates correlation gives a DNB ratio 50% to 801 higher than the Bernath correlation. i F16 ure 4 22 shows the results of this analysis, llere we have. Plotted the maximum channel heat flux for which the DNB ratio is 1, with ! bulk pool water temperature as a pararneter. It is assumed that all the - vapor above the detachement point separates from the heated surface. Froin the figure it can be seen that with the design cooling water ternperaturg at the core inlet *120'F) the maximurn heat flux is 325 kBTU/hr ft . For a 85 element core with an overall peak to average power density rat to of 2.0, this heat flux corresponds to a maxirnum reactor power of 1900 kW. .i
- 4. 5 -
L 4,0 '
..r T
4 h s.1 2 Ts a
- s. o
-4.5 to 90 100 110 llo 'llo I.S c..i.niini.i ..p..u. .t'ri plat FOR Wil!Cil DNB RATIO IS 1.0 0F MAXIMUM llEAT F1.UX VERSUS COO! ANT TEMPERATURE
-Figure 4 22 4 44 l
l
m.__ . _ . _ . _ - - d l SAR 5/91 4.1.3.9. Nomenclature A cross. sectional area, ft 2 channel free flow area, ft 2 Ar~ , C coolant specific heat, Blu/lb 'F d diameter, in. De channel equivalent diameter, it Du hydraulic diameter, ft ' friction factor with subcooled boiling, dimensionless l Ib fm friction factor without boiling, dimensionless F forces acting on vapor bubble 2 g constant, 4.18 x 100 ft/hr bb heattrangfercoefficientwithsubcooledboiling, Stu/hr ft 'r H distance from midplane of heated channel to free surface of pool, ft K pressure loss factor at channel inlet or exit, dimensionless n number of equal axial increments into which heated length of core is subdivided N Number of channels which receive their flow from a single opening in the lower grid plate p absolute pressure, Ib/ft 2 P heated perimeter of channel, it Pr Prandt1 number op pressure loss, Ib/ft 2 q heat load, Btu /br qt total heat load to channel, Btu /hr i q" heat flux, Bru/hr.ft 2 2 q"p peak or " burnout" heat flux, Bru/hr ft i rB bubble radius Re Reynolds number, dimensionloss
'S/V - channel surface to volume ratio, in.*I
[ T coolant temperature, 'F Tsat coolant saturation temperature, 'F v specific volume, it3 /lb 1 v flow veloetty, ft/hr W mass flow rate, Ib/hr 4 45 f.--.- . _ . _ _ _ ,
Salt 5/91 . Y non dinnensional radius j z axial coordinate in channel, ft ze total length of channel, it or length of a calculation increment in the channel, it p dynarnic viscosity, f t lb/hr a void fraction or fraction of a channel cross section which is occupied by vapor, dienensionless a surface tension, Ib/ft ( vapor volume, or volume of vapor produced per unit area of heated surface, cuble in./ square in. v kinematic viscosity, it 2 /hr t shear stress, Ib/ft 2 3 p densitj, Ib/ft
</De relative roughness '
4 Suhacripts e conditions at channel exit i conditions at channel entrance or inlet 1 conditions in portion of channel adjacent to lower reflector m conditions averaged over the liquid and vapor phases o bulk pool conditions u condittuns in portions of channel adjacent to upper i reflector j axial increment index k axial station index w conditions at cladding outer surface l y vapor L liquid i l' t 4 46
l SAR 5/91 4.2, NUCLFAR DESIGN AND EVALUATION The charactet1stics and operating parameters of this reactor have been calculated and ext rapolat ed using experience and data obt ained f r om exist ing TRICA teactors as bench Italks in evaluating the calculated data. There ate several TRIGA syntetts with essent ially the s.atte tote and reflector relationship as this TRICA so the values presented het e are felt to be accurate to within 51 but, of course, are influenced by speellic core configuration details as well as operational details. An operatlonal core of 86 fuel elements. 3 iuel followed control roda, and one air followed conttol tod is to be arranged in 5 rings with a cent ral, water t illed hole. Dittension of the active fueled core, a cylinder, is 15 inches in height with an avtrage radius of 8.6 inches. The cylinder radius in calculated as the average radius of a hexagonal fuel array with 91 unit cells (6 1111ed rings including the center hole) and a unit ec11 area of 2.55 square inchen. Table 4-3 summarizes the typical Mark 11 TRIGA re ctor parameters f or a core containing high hydride, stainless steel clad fuel elements Table 4 3 TYPICAL TRIGA CORE NUCLRAR PARAMETERS ruel elements SS-clad U Zrill.6 Cold clean critical loading -04 elements
-7.5 kg U 235 Operational loading ~90 elements
-3.4 kg U-235 f , Prortpt neutron lifetime 41 usec B, Effective delayed neutron fraction 0.0070 o, Prompt negative t ettpe r at u re coefficient
~1. 0 x 10',' 6k/h' C Tr Average fuel temperature (1.1 MW) 265'C TyAverage water temperature (1.1 MW) 65'C Water coolant volume to cell volume ratio -1/3 4.2.1. Ernetivity lffectr.
The reactivity associated with the control rod is of interest both in the shutdown margin and in calculations of possible abnormal conditions related to reactivity accidents. Table 4-4 gives approximate reacttvity values associated with a total control rod travel of 15 in. (38.1 cm), the full travel in the core. 4 47
SAR 5/91 Table 4 4 LSTIMATED CONTROL ROD NET WORTil Control Rod diameter bk/t in. (cm) 1 C ring transient 1.25 (3.18) 2.1 C ring regulating 1.35 (3.43) 2.6 D ring - shim 1 1.35 (3.43) 2.0 li ring - shim 2 1.35 (3.43) 2.0 The maximum reactivity insertion rate in that associated with the transient rod which can be fully temoved from the core in 0.1 see producing an average reactivity tiu.crtion rate of 21% ok/k-sec. The total react tvity worth of the control system is about 8.71. With a core excess reactivity of 4.91, the shutdown margin with all rods down is about 3.8% and with the most teactive rod stuck out is about 1%. The reactivity worth of the fuel elementu is dependent on their positton within the core Table 4-5 indicates the valuen that are expected in this installation. Table 4 5 ESTIMATED PUEL ELEMENT REACTIVITY WORTil COMPARED WITil WATER AS A FUNCTION OF POSITION IN CORE Worth (1 ok/k) Number of Core Position SS Clad U Zrill .6 ruel Positions B ring 1.07 6 C ring .85 12 D ring 0,54 18 E ring 0.36 24 F ring 0.25 30 C ring 0.19 36 Because of the prompt negative temperature coefficient a significant amount of reactivity is needed to overcome t enipe ra t u re and allow the reactor to operate at the higher power levels in steady-state operation. Figure 4 23 shows the relationship o r cenctor power level and associated reactivity loss to achieve a given power level. Figure 4-24 relates fuel temperature to a given steady-state reactor power level. 4 48
l SAR 5/91 ! l-I O ! l ! i l i.' i. u i i i b.0 i i i l ! [ 5 l 1.0 l e i l I O i l a f f I f ! 4 e l $s - [ ! 2.o - : g
~
l l 6 l : . l j ! io - i I
-t t
' ' i f ' '
0 0 200 400 600 800 1000 ] I PowtR (K.) f i i l i 1 t i t 9 ESTIMATED REACTIVITY 1hSS VERSUS POWER . t Figure 4 23 j. l
}
4-49 ; i
SAR 5/91 i e ; I I i 400-t T .
- 300- I E
9 < w 8 w
=
y l 200- - E i t , 100-l 0-
, ,go 6do sdo 860 10'00 POWER (W) l l
l I l-ESTIMATED MAX 1 HUM B RING AND AVERAGE CORE TE;4PERATURE VERSUS POWER Figure 4-24 4-50
SAR 5/91 The reactivity et fects associated with the insertion and removal of experiments in or around the core are difficult. to predict; however, Table 46 is supplied to provide a guide to the magnitude of the reactivity eficets associated with the introduction of experiments in the reactor core. Table 4 6 EXPECTED REAC'11VITY EFFECTS ASSOCI ATED WITil EXPERIMENTAh FACILITIES Vorth (% 6k/k) Central thimble, f ucl vs 110 2 40.90 : Central thimble, void vs H 2 O 0.15 Pneumatic transfer tube, (C ring) void vs 110 0.10 Rotary specimen rack,7 void vs H 2O 0.20 4.2.2. Eyltl uation of Nuclear Design The TRICA reactor system is well known for its conservative design. The stability of this reactor type has been proven both through calculations as well as through tests performed with the many TRICA , reactors in operation throughout the world. The stability of the TRICA type reactor stems from the prompt neSative temperat% e coefficient associated with.the U Zrli x fuel in conjunction with e suitable neutron thermalizing material. This TRICA will have the stab'.lity that has been demonstrated on other TRICA systems over tho years. A review of the reactivity worths associated with the reactor core indicates that no single item listed can produce a atep reactivity insertion greater than that offered by routine pulse operation. In the pulsed mode of operation the results of a step insertion of $3,00 are far below those attributed to test pulses on the advanced TRIGA prototype reactor in which 3.5% 6k/k was inserted in a step as is shown , in Table 4 7. A total reactivity limit for experiments is set at $3.00. This limit constrains the worst c a s.e transient accident to less than the design pulse insertion. The possibility of a reactivity accident which could produce reactor powers and fuel te.aperatures attributed to a $4,00 step insertion has been considered and evaluated in the accident analysis section of this report. It is concluded from this analysis that the i peak and average fuel temperatures resulting from this accident are well i below the temperatures indicated as safety limits described in the [ reactor design bases of this document. It is further concluded that the integrity of the fuel containment will not be jeopardized and no adverse effects to the reactor system or personnel will arise from the advent of such an accident. 4 51-t {
SAR 5/91 Table 4 7 COMPARISON OF REACTIVITY INSERTION EFFECTS Pulse Resulting from Max Pulse Tested Insert ton of Maximutn on SS Clad, lligh Excess Reactivity in llydride Fueled This TRICA TRICA types Reacttvlty insertion,
$ 3.00 5.00 Steady-State power before pulse, kW <1 <1 Peak power, MV -1400 ~8400 Total energy release, MW see -18 ~54 Pe riod , insec -3.1 -1.4 Maxirnurn fuel tenipe rature .
*C -540 -1050 Pulu width, resce ~11 -5.5 4.3 THERMAL,AND HYDRAUi.1C DESIGN This TRIGA reactor will be operated with natural convective cooling by reactor pool water. This method of heat dissipation is more than adequate for the power level of the reactor; i.e. 1100 kW(t). The therrnal and hydraulic design of the reactor is well within the safety limits required to assure fuel integrity.
4.3.1. Desirn Basu The thermal and hydraulic design for this TRICA is based on assuring that fuel integrity is maintained during steady-state and pulsed mode operation as well as during those abnormal conditions which might be postulated for reactor operation. During steady-state operation fuel integrity is maintained by limiting reactor powers to values which assure that the fuel cladding can transfet heat from the fuel to the reactor coolant without reaching fuel clad temperatures that could result in clad rupture. If these temperature conditions were exceeded, the maximum local heat flux in the core would be greater than the heat flux at which there is a departure frorn the nucleate boiling regime and consequently filen blanketing of the fuel. This heat flux safety limit is a function of the inlet coolant temperature. l l 4-52
SAR 5/91 Figure 4 22 summarizes the results of the thermal and hydraulic analysis for steady state operation of the TRICA. In the figure critical heat flux for departure f rorn nucleate boiling is plotted as a function of the coolant inlet t e nipe ra t ure . The maxirnum power density in a TRICA core is found by multiplying the average power density by a radial peak to average power generation ratio of 1.6 and an axial value of 1.25. The correlation used to determine the heat flux at which there is a departure frois nucleate boiling is frora Bernath [33). This correlation encompasses a wider range of experimental data than the usual correlations, c.6 . the correlation due to McAdams, and, generally gives a lower value for the DNB ratio than the other correlations. Table 4 8 1000 kW(t) TRICA HEAT TRANSFER AND HYDRAULIC PARAMETERS Number of fuel elements 90 Diameter 1.475 in. Length (hented) 15.0 in. Flow area 0.522 ft 2 Wetted perimeter 34.75 ft Hydraulle diarneter 0.0601 ft 2 Heat transfer surface 43.44 ft Inlet coolant temperature -120'F (48.9'C) Exit coolant temperature (average) 174*F (78.9'C) Coolant mass flow 63,700 lb/hr Average flow velocity 0.55 ft/sec Average fuel temperature 500'F (260'C) Maximum wall temperature 280* F (138'C) Maximum fuel temperature 842*F (450*C) Average heat flux 78,600 Btu /hr ft 2 Maximum heat flux 157,100 Btu /hr ft 2 Minimum DNB ratio 2.0 4 53
SAR 5/91 Tabic 4-8 lists the pertinent heat t r ant. f e r and hydraulic parameters for the TRIGA operating at a nominal power level of 1000 kW. These data were taken f roin the results of calculations described in Section 4.1. During pulsing operation the limiting thermal hydraulic condition is the fuel temperature and the corresponding 112 excess pressure beyond which clad rupture may occur. As indicated in Sec tion 4.1, coolant temperature is not a limiting condition in pulsing since heating conditions are essentially adiabatic and significant transfer of heat energy to the coolant does not occur until after peak fuel clad temperatures occur. The safety limit on fuel temperature occurring in the pulse mode of operation is 1150*C. This temperature will give an internal equilibrium hydro gen pressure (U Zrti fuel ll/Zr; 1.6) less than that which would produce a stress equivalcut to the ultimate strength of the clad at a temperature of 680'C. This clad temperature is higher than would actually occur and therefore conservative even in the case of a pulse producing a peak adiabatic fuel temperature of 1150*C. 4.3.2. Ibermal nnd Hydraulic Desinn Evaluation The validity and safety of the TRIGA thermal hydraulic design is established in Section 4.1. In that section it is shown that design-basis conditions evaluated for TRIGA reactors using stainless steel clad U Zrit (H/Zr; 1.6) fuct elements provide a generous safety margin for this TRIGA. These general evaluations are supported by extensive experience in operation of TRIGA cores at equivalent fuel temperatures and reactor power levels. No adverse results are reported from other similar TRIGA reactors with comparable fuel temperatures and power levels. 4.4. MECHANICAL DESIGN AND EVALUATION 4.4.1. General DescrintioD The TRICA Mark 11 reactor core assembly is located near the bottom of an elongated cylindrical aluminum tank surrounded by a reinforced concrete structure. A typical installation is shown in Fi ure t 4 25. The standard reactor tank is a welded alurninum vessel with 1/4 in, thick (0.64 cm.) walls, a diameter of approximately 6.5 feet (2 meters), and a depth of at least 25 feet (7.6 m.). The tank is all welded for water tightness. The integrity of the veld joints is verified by radiographic testing, dye penetrant checking, and helium leak testing. The outside of the tank is coated for corrosion protection. An aluminum angle used for mounting the ion chambers, fuel storage racks, underwater lights, and other equipment, is located around the top of the tank. Demineralized water in the tank is provided such that at l least 21 feet (6.4 m. ) of shielding water is above the core. The core is shielded radially by a minimum of 7.97 feet (2.43 m) of concrete with ; a density of 2.88 gm/ce , 1.5 feet (-45 cm. ) of water, and 10.2 inches l (25.9 cm.) of graphite reflector (see Figure 4-26). 4-54
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SAR 5/91 4.4.2. Iteflector Assenbly The reficetor is a ring-shaped block of graphite that surrounds the core radially. The graphite is 10.2 inches (2$ 91 cm.) thick radially, with an inside diameter of 215/8 inches (54.93 cin.) and a height of 21 13/16 inches (54.40 cm. ) Thr. graphite is protected from water penetration by a leak tight welded altuninurn can. A "Well" in the top of the graphite reflector is provided for the rotary specirnen rack. This well is also alurninurn lined, the lining i being an integral part of the alurninum reficctor can. The rotary specimen rack is a scif-contained unit and does not penetrate the sealed reficctor at any point. The graphite, and outer surf ace of the alurninura can are pierced by ; an alurninurn tube which forms the inner section of beam ports that penetrate the reflector. The reflector penetrat.ing bearn tubes are connected by alurninurn couplings to the correspondina bearn tube section fabricated as part of the reactor tank structure. The reflector assembly rests on an alurninum platforan at the bottom of the tank, and provides support for the two grid plates and the safety plate. Three lugs are provided for lifting the assembly, , 4,4.3. prid Plates The top grid plate is an aluminum plate 5/8 inches (1.59 cm.) thick (3/8 inches, 0.95 can., thick in the central region) that provides accurate lateral positionin6 for the - core components. The plate is supported by a ring velded to the top inside surface of the reflector container and is anodized to resist wear and corrosion. One _ hundred twenty one (121) holes,_1.505 inches (3.823 cm.) diameter, are drilled into the top grid plate in a modified hexagonal pattern (the vertexes of the. hex are omitted) around a central hole. - The holes are to locate the fuel moderator and graphite dummy elements, the control rods and guide tubes, and the pneumatic transfer tube. (See Figure 4 27.) An equivalent diarneter center hole accommodates the central thimble. Small holes at various positions in the top grid plate permit insertion of wires or foils-into the core to obtain flux data. A hexagonal section can be removed from the center of the upper grid plate for the insertion of specimens up to 4.4 inches (11.18 cm.) in_ diameter into the region- of highest flux; this requires prior relocation of the six fuel elements from the B ring to the outer portion of the core and removal of the__ central thimble. This removable seetion will not be used initially; a license amendment will be obtained prior to its use. Two generally triangular-shaped sections are cut out of the upper grid plate. Each cut out encompasses two E and one D ring holes. Fuel elements placed in'these locations, require lateral support by a special fi x t.ure . When the fuel elements and support are removed, there is room ' for inserting specimens up to 2.4 inches (6.1 cm.) in diameter. 4 57
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SAR 5/91 The bottom grid plate is an aluminum plate 3/4 inch (1.91 cin . ) thick which supports the entire weight of the core and provides accurate spacing between the fuel moderator elements. Six pads around a hexagonal ring which is welded to the reflector container support the bottom Brid plate. Holes in the bottom grid plate are aligned with fuel element holes in the top grid plate. They are countersunk to align the adaptor end of the fuel moderator elements and the adaptor end of other in core experitsent systerns, such as the pneumatic transfer tube. The differential area between the fitting at the top of the fuel eternents and the round holes in the top grid plate permits passage of cooling water through the plate. 4.4.4 Safety Plate The safety plate is provided to preclude the possibility of control rods falling out of the core. It is a 1/2 inch (1.27 cm.) thick plate of aluminum set on the core support structure below the reflector. The plate is placed about 16 inches (40.6 cm. ) below the bottom grid plate. A central hole of 1.505 inches (3.823 cm.) in diameter in the lower grid serves as a clearance hole for the central thimble. Eight additional 1.505 inch (3.823 cm.) diameter holes are ali ned 6 with upper grid plate holes to provide passage of fuel follower control rods. Those holes in _ the bottom grid plate not occupied by control rod followers are plugged with removable fuel element adaptors that rest on-the safety plate. These fuel element adaptors are solid aluminum cylinders 1.5 inches (3.81 cm. ) in diameter by 17 inches (43,18 cm.) long. At the lower end is a fitting that is accommodated by a hole in the safety plate. The upper end of the-cylinder is flush with the upper surface of the bottom grid plate when the adaptor is in place. This end of the adaptor has a hole similar to that in the bottom grid plate for accepting the fuel element lower end fitting. With the adaptor in l- place, _a position formerly occupied by a control rod with a fuel follower will now accept a standard fuel element. The adaptor can be l removed with a special handling tool. 4.4.5. E9el Moderator Elements The active part of each fuel moderator element, shown in Figure 4 28, is approxitnately 1.43 in. (3.63 cm.) in diameter and 15 in. long (38.1 cm.). The fuel -is- a solid, homogeneous mixture of uranium-zirconium hydride alloy containing about 8.5% - by weight of uranium enriched to 20% U 235. The hydrogen to zirconium atom ratio is about-1.6. -To facilitate hydriding, a small hole is drilled through the center of the active . fuel section and ' a zirconium rod is inserted in this hole after hydriding is complete. The hydriding hole and zirconium rod are not shown in Figure 4 28. Several types of end fittings exist, therefore, those shown are typfcal, l 4 59
SAR $/91 i STAINLESS STEEL TOP END Fi11ING II m GRAPHITE 3.45 IN. s
- c STAINLCSS STEEL TUBE CLADDING THICKNESS 0.02 IN.
URANIUM ZlRCONIUM HYDRIDE-WITH AXIAL ZlRCONIUM ROD 15 IN. 28.8 IN. O TOTAL 1.4 3 I N .-
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- SAR $/91 Each element. is clad with a 0.020 in, thick ( . 508 men. ) stainless
- s. teel can, and all closures are made by heltarc welding. Two sections ;
of graphite are inserted in the can, one above and one below the fuel, ; to serve as top and bottors reflectors for the core. Stainless steel end fixtures are attached to both ends of the can, making the overall length of tSe fuel moderator element 28.8 in. (73.2 cm.). The lower end fixture supports the fuel moderator element on the bottom grid plate. The upper end fixture consists of a knob for at t achrnent of the fuel handling tool and a triangular spacer, which permits cooling water to flow through the upper grid plate. The total weight of a fully-loaded fuel element is about 3.18 kg. (7 lb.). 4 I. 5.1 Instrument Fuel Elements An ins t rurne nt.ed fuel moderator element will have three t herreocouples embedded in the fuel. As shown in Figure 4 29, the sensing tips of the fuel eternent thermocouples are located about 0.3 in. 4 (0. 76 cm. ) from the vertical centerline. Thermocouple specifications are listed in Table 4 9. The thermocouple lead wires pass through a seal in the upper end fixture. A lead tube provides a watertight conduit carrying the lead wires above the water surface in the reactor pool. The rmocouple specifications are listed in Table 4 9. In other respects the instrumented fuel moderator element is identical to the standard element. Table 4 9 THERMOCOUPi.E SPECIFICATIONS t Type Ch rornel -alurnci , Wire diameter 0.005 in. Resistance 24.0B ohms /douhic foot at 68'F Junction Grounded Sheath material Stainless steel Sheath diarneter 0.040 in. Insulation Mg0 . 1.ead out wire Material Chroinel alumel Size 20 AVG Color code Chromel yellow (positive) Alumel red (negative)- Resistance 0.59 ohms / double foot at 75'r i 9 Y F 4 4 61
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$AR $/91 4.4,5,2. Ly.ithtallmuf fuel rienient De stigJJ.
General Atomic has acquired extensive experience in the fabrication and operation of high hydride, stainless steel clad fuel elements. As shown in other sections of this report, the stresses associated with the fuel and cladding temperatures in all modes of operation, normal and abnormal, are within the saf ety 11:nita described in the Reactor Design Bases. Dimensional stability of the overall fuel element has been exec 11cnt in the TRICA reactors in operation. Analysis of the heat removal irom elements that touch owing to transverse bending shows that the contact will not result in hot spots that darnage the fuel. Tests have been conducted on TRICA fuel eletnents to determine the strength _ of the fuel element clad when subjected to internal pressure. At room temperature the clads ruptured at about 2050 ps16 This corresponds to a hoop stress at rupture of about 72,000 pst which compares favorably with the minimum expected value for 304 stainicas 1 steel. It is concluded that the chemical stability of U Zril l .6 fuel. moderator material does not impose a safety limit on reactor operation (see Section 4.1.1) . Table 410 gives a summary of the fuel element specificacions,
-Table 4 10
SUMMARY
OF ltEL E11HENT SPECIFICATIONS
!!2minal value fuel Hoderator Material ll/Zr rattu 146 Uranium content 8. 5 -wt 1-Enrichment (U 235) 19.7 #0.2 Diameter 1.43 in.
Length 15 in. Graphite End Reflectors Unper Lower porosity 20% 20% ! Diameter 1.43 in. 1,43 in. Length 3.44 in. 3.47 in. Cladding Haterial Type 304 SS
-Vall thickness 0.020 in, L Length 22.10 in, End Fixt.ures and Spacer Type 304 SS overall Element Outside diameter 1.47 in. (3,73 cm)
Length 28.37 in. (72.06 cm) Weight 7 lb. (3.18 kg) I 4 62 l
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SAR 5/91 Most. of the fuel for the initial core loading will consist of elements with burnups of a fraction of a MV day to r,cveral MV days. It is anticipated that as, initial core loading of about 94 fuel elements, including instrumented elements, and fuel followed control rods, will produce a col 1, clean excess reactivity of ~4.9% ok/k. The operational I core configuration will contain two instrumented fuel elements with at least one located in the inner most reactor ring, j 4.4.6. Neutron Source and Holder [ A 2 or 3 curie americturn berylliurn neutron source vill be used for ' startup. The neutron source holder is rnade of aluminum, is cylindrical in shape, and has a cavity to hold the source. The source holder can be installed in a vacant fuel or graphite element location. A shoulder at ' the upper end of the holder supports the assembly on the upper grid plate, the rod itself, which contains the source, extending down into the core region. The neutron source is contained in a cavity in the lower portion of the rod assembly at the horizontal centerline of the , core. This cylindrical cavity is 0.981 in. (2.492 cin. ) in diameter and approxirnately 3 in. (7.62 cm.) deep. The upper and lower portions are screwed together. A sof t. alurninua ring provides sealing agains; water leakage into the cavity. Since the upper end fixture of the source holder is similar to that of the fuel element, the source holder can be installed or removed with the fuel handling tool. In addition, the upper end fixture has a small hole through which one end of a stainless steel wire may be inserted to facilitate handling operation from the top of the tank. 4.4.7. Ornehite Duremy Element s Graphite duminy elements may be used to fill grid positions not filled by the fuel moderator elements or other core compounds. They are of the same general disnensions and construction as the fuel moderator elements, but are filled entirely with graphite and are clad with a1urni nura. 4.4.8. Centrol System Desien The reactor uses four control rods: a, Shim rod 1 .
- b. Shim rod 2
- c. A transient rod d._ A-regulatory rod The regulating and shim rods are sealed 304 stainless steel tubes approximately 43 in. (109 crn) long by 1.35 in. (3.43-cm) in diameter in which the upperrnost 6.5 in. (16.5 cm) section is an air void and the next 15 in..(38.1 cm) is the neutron absorber (boron carbide in solid fo rrn) . Immediately below the neutron absorber is -a fuel follower section consisting of 15 in. (38.1 cm) of U ZrH l .6 fuel. The bottom section of the rod is 6.5 in. (16.5 cm) air void, i-4-64
SAR 5/91 The regulating and shim rods pass through and are guided by 1.5 in. (3.81 cm) diameter holes in the top and bottorn grid plates. A ' typical control rod with fuel follower is shown in the withdrawn and inserted positions in Figure 4 30. The safety transient rod is a sealed, 36.75 in. (93.35 cin) long by 1.25 in. (3.18 cm) dianteter tube containing solid boron carbide as a neutron absorber. Below the absorber is an air filled follower section. The absorber section is 15 in. (38.1 cm) long and the follower is 20.88 in. (53.02 cm) long. The transient rod passes through the core in a perforated aluminum Suide tube. The tube receives its support from the safety plate and its lateral positioning frotn both grid plates. It extends approxirmately 10 in. (25.4 cro) above the top grid plate. Vater passage through the tube is provided by a larSe number of holes distributed evenly over its length. A locking device is built into the lower end of the assembly. The control rods are connected to their individual drive units by screwing the upper end of the rod into a control rod drive assetobly connecting rod. Vertical travel of' each rod is approximately 15 in. (38.1 cm). - Reactivity worths and core positions for each rod are summarized in the section on nuclear design. A summary of other control tod design parameters is given in Table 4 11. Table 4 11
SUMMARY
OF CONTROL ROD DESICN PARAMETERS Shim and Transient Regulating Cladding Material Al Type 304 SS OD 1.25 in. (3.18 cm) 1.35 in. (3.43 cm) Length 36.75 in. (93.35 cm) 43.13 in. (109.5 cm) Vall thickness 0.028 in. (0.071 cm) 0.020 in. (0.051 cm) Absorber Material Boron Carbide (solid form) OD 1.19 in. (3.02 cm) 1.31 in. (3. 32 cin) Length 15 in. (38.1 cm) 14.25 in. (36.20 cm) Follower Material Air U Zrill .6 l OD 1.25 in. (3.18 crn) 1.31 'n. (3.34 cin) j' Length 20.88 in. ($3.02 crn) 15 in. (38.1 cm) 445
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1 SAR 5/91 1 1 4.4.8.1. Control Rod Drive Assettblica. The control rod drive assemblies for the shim rods are mountnd on a bridge assembly over the pool and consist of a motor and reduction gear driving a rack and pinion as indicated in pigure 4 31. A helipot connected to the pinion generates the position indication. Each control rod drive has a tube that extends to a dashpot below the surface of the water. The control rod assembly is connected to the rack through an electromagnet and armature. In the event of a power failure or scram signals, the control rod snagnets are de energized and the rods fall into , the core. The tirne required for a rod to drop into the core from the full out position is about I second. The rod drive motor is non synchronous, single phase, and instantly reversible, and will insert or withdraw the control rod at a rate of approxirnately 18 in./ min. (0.75 cia /sec) for the shim 1 and shim
? rods. The regulating rod design is sitellar to the shita rods except ,
for the type of. rod drive. A key locked switch on the control console power supply prevents , unauthorized operation of all control rod drives. Electrical dynamic and static braking on the actors are used for fast stops, l.itni t switches mounted on the drive assembly actuate circuits i ehich indicate the following:
- a. The "up" and "down" positions of the magnet ,
- b. The "down" position of the rod c._The magnet in contact with the rod 4.4.8.2. Regulating Rod and Stennine Motor Drive The icod drive mechanism for the regulating rod will be an electric stepping motor-actuated linear drive equipped with a taagnetic coupler and a positive feedback potentiometer.
A stepping motor drives a pinion gear and a 10 turn potentiometer via a chain and pulley gear mechanism. The potentiometer is used to provide rod position information. The pinion gear engages a rack attached to the magnet draw tube. An electromagnet, attached to the lower end of the draw tube, engages an iron armature. The armature is screwed and pinned into the upper end of a connecting rod that
- terminates at its lower end in the control rod. When the steppin5 inotor is energized (via the rod control UP/DOWN switch on _ the operator's console), the pinion gear shaf t rotates, thus raising the magnet draw tube, the armature and the connecting rod will raise with the draw tube so that the control rod is withdrawn frors the reactor core. In the event of a reactor scram, the magnet is de-energized and the armature will be released. The connecting rod, the piston, and the control rod will then drop, thus reinserting the control rod, i
l 4 67
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RACK AND PINION CONTROL ROD DRIVE Figure 4 31 4 68
SAR $/91 l Stepping motors operate on phase-switched direct current power. The motor shait advances 200 steps per revolution (1.8 degrees per step). Since current is inaintained on the inotor windings when the motor lu not being stepped, a high holding totque is maintained. The torque vs speed characteristic of a stepping motor is greatly dependent on the drive circuit used t o st ep the anotor. To optimize the torque characteristic vs motor frame size, a Translator Module was selected to drive the stepping motor. This combination of stepping motor e.nd translator modsle produces the optirnurn torque at the operating speeds of the control rod drives. 4.4.8.3. Innglent Rod Ddye Ansembly. The saf ety t ransient control rod on pulsing TRICA man 11 reactors is operated with a pneumatic rod drive (see Figures 4 32 and 4 33). Operation of the transient rod drive is controlled from the reactor console.
'l h e transient rod is a scrammable rod operated in both pulse and stead state modes of reactor operation. During non pulse operation, the transient rod will function as an alternate safety rod with air continuously supplied to the rod. Rod position is thus contro11.ed by operation of an electric motor that positions the air drive cylinder.
The position of the transient conttol rod and its associated reactivity worth will generally dictate removal of the rod as the undid step of a startup for steady state operr.tlon. The transient rod drive is mounted on a steel frame that bolts to the bridge. Any value from zero to a maxinwn of 15 in. (38.1 cm.) of rod rnay be withdrawn f rom the core; administrative control is exercised to restrict its travel so as not to exceed the maximum licensed step insertion of reactivity ($3.14 or 2.21 ok/k). The transient. rod drive is a single-acting pneumatic cylinder with its piston attached to the transient rod through a connecting rod assembly. The piston rod passes through an air seal at the lower end of the cylinder. Compressed air is supplied to the lower end of the cylinder from an accumulator tank when a three-way solenoid valve located in the piping between the accumulator and cylinder is energized. The compressed air drives the piston upward in the cylinder and causes the rapid withdrawal of the transient rod from the core. As the piston rises, the air trappud above it is pushed out through vents at the upper end of the cylinder. A t. the end of its travel, the piston strikes the anvil of an oil-filled hydraulic shock absorber, which has a spring return, and which decelerates the piston at a controlled rate over its inst 2 in. (5 cm.) of travel. When the solenoid is de energized, the valve cuts off the compressed air supply and exhausts the pressure in the cylinder, thus allowing the piston to drop by gravity to its original position and restore the transient rod to its fully inserted position in the reactor core 4-69
- - . . . - . . . - - . . . . _ . . . - . . ~ ~ . - . . . . . . - . . . . . . _ . - . _ _ ~ - . .. =.... _ .~ . . . . . - . , ~ . . . ~ . - - SAR 5/91
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SAR 5/91 Ert0CK ABSOROLR
. SHOCK ABSORBER h / anvil VENT HOLES =~
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=
BOTTOM LIMIT CONTROL A00 PISTON ROD 5 TRANSIENT ROD OPERATIONAL SCHEMATIC Figure 4-33 I 4-71
- , .~ . - - - - . ~ . . - . ~ . ~ ~ ~ ~ - - . - . - _ . - - - - - . - . . . - -
SAR 5/91 The extent of transient rod withdrawal from the core- during a pulse is determined by raising or lowering the cylinder, thereby controlling the distance the piston travels, The ' cylinder bas external threads runnin6 most of its length, which engage a series of ball- bearings contained in a ball nut mounted in the drive. housing, As the ball-nut is rotated by a worm gear, the cylinder moves up or down depending on the direction of worm gear
- rotation.
A ten-turn potentiometer driven by the worm shaft provides a signal indicating the position of the cylinder and the distance the transient rod will be ejected from the core. Motor - operation for pneumatic cylinder positionin6 is controlled by a switch on the reactor control console. The magnet power key switch on t he - control console power supply prevents unauthorized firing of the transient rod drive. Attached to and extending downward f rom - the transient rod drive housing is the rod guide support, which scrves several purm s. '% air-inlet connection near the bottom of the cyli der prsjct's thrcugh a slot in the rod guide and prevents the cylinder from rotating . ached to the lower end of the - piston rod is a flanged connector that is attached to the connecting rod assembly that moves the trandent od. , The flanged connector stops the downward movement of ..t &cs. : cod when the connectar strikes the damp pad at the bottom of too rod side
- support, A_microswitch is mounted on the outuide of the ga' % tube with its actuating lever extending inward through a slot. Vnwn tne transient rod is fully inserted in the reactor core, the flange c . m.ect.-- engages ~
the actuating lever of the microswitch and indicates on 'or (natrument console that the rod is in the core. In the case of the transien cod a scram signal de-energizes the solenold valve which supplies the air required to hold the rod in a withdrawn position and the rod-drops into the core from the full out position in about I second. 4,4,8.4 Evaluation of Control Rod System. The reactivity worth and speed of travel for the control rods are adequate to allow complete control of the reactor system during operaticn from a shutdown condition to full power. The scram times for the rods are quite adequate since - the TRIGA system does not rely on speed of control as being paramount to the safety of the reactor. The inherent shutdown mechanism of the TRICA prevents unsafe excursions and the control system is used only for the planned shutdown of the reactor .
- and to control the power level in steady state operation, p
4 72
SAR S/91 4.5 SAFETY SETTINGS IN RELATION TO SAFETY LIMITS As has been indicated, fuel temperatures are the main safety considerations in the operation of the TRIGA system. The temperature of the fuel may be controlled by setting limits on other operating parameters. The operating parameters of interest for TRICA are:
- a. Maximum licensed steady-state power level
- b. Fuel temperature measured by thermocouple
- c. Maximum reactivity worth of transient rod
- d. Core inlet coolant water temperature The safety settings s listed in Table 4-12 are such that in all operation, normal and abnormal, the safety limits indicated in the reactor design bases will not be exceeded.
Table 4 12 TRICA SAFETY SETTINGS Parameter Limited Safety Setting Function Maximum steady-state 1100 kW (t) Resctor scram power level Maximum measureme: ' 500*C Reactor scram of fuel temperature
-Administrative limitations are imposed for the excess reactivity, transient conditions and coolant water temperature as follows:
- a. Maximum core excess reactivity of 4.9% 6k/k with a shutdown margin of at least 0.2% 6k/k, assuming one rod withdrawn.
b, Maximum transient control rod worth of 2.8% 6k/k with a limit of 2.2% 6k/k, for any transient insertion,
- c. Core inlet water temperature of 4!$.9"C.
These safety settings are conservative in the sense that if they are adhered to, the consequence of normal or abnormal operation would be fuel and clad temperatures well below the safety limits indicated in the reactor design bases. Because of the conservatism in these safety settings, it is reasonable that at some later date less restrictive saf ety system settings could be assigned in conjunction with upgrading of the reactor ta operate at higher steady-state power levels or in the pulsing mode, while still using the same fuel and core configuration. 4-73
SAR 5/91 -4 C Chapter 4 References
- 1. Merten. U., et al., " Thermal Migration of Ilydrogen in Uranium-Zirconium Alloys," General Dynamics, General Atomic' Division Report GA 3618, 1962.
- 2. Coffer, C. O., et al s, Research in Improved TRICA Reactor Performance, Final Report," General Dynamics, General Atomic Division Report CA 5786, 1964.
- 3. Johnson , 11. A . , e t al., " Temperature Variation, Heat Transfer, and Void Volume Development in the Transient Atmosphere Boiling of Water," SAN 1001, University of California, Berkeley, 1961.
- 4. McAdams, W. H., Heat Transmission, 3rd ed. McCraw Hill Book Co. ,
New York, 1954.
- 5. Sparrow, E. M. and R. D. Cess, "The Effect of Subcooled Liquid on Film Boiling," He.it Transfer, f!i, 149 156 (1962).
- 6. Speigler, P., et al., " Onset of Stable Film Boiling and the Foam Limit," Int. J. Heat and Mass Tran h , f., 987-989 (1963).
- 7. Zuber, W., " Hydrodynamic Aspects of Boiling Heat Transfer," AEC Report AECV 4439, TIS, ORNL, 1959.
\ \ Rosnehow, W., and H. Chot, Heat. Mass and Engentum Transfer, 8.
Prentice Hall, 1961, pp. 231-232.
- 9. Ellion, M. E., "A Study of the Me chani e"n 20 Ik i'. ing heat Transfer," Jet Propulsion Laboratory Memo. Na 20 88, 1952.
- 10. Coffer, C, 0., et al., "
Characteristics c' L, ;' ;:n 's ity
-Insortions in a High Performance TRIGA U ZrH Cc.e," General Dynamics, General Atomic Division Report GA-6216, 1965.
- 11. Fenech, H., and W. Rohsenow, " Thermal Conductance of Metallic Surfaces in Contact," USAEC NYO-2130, 1959,
- 12. Graff, W. J., " Thermal Conductance Across Metal Joints," Machine Design, Sept. 15, 1960, pp. 166 172.
- 13. Fe ne ch , H . , and J. J. Henry, "An Analysis of a Thermal Contact Resistance," Trans. Am. Nucl. Soc. ), 476 (1962).
- 14. Bernath, L., "A Theory of Lecal Boiling Burnout and Its Application to Existing Data," Heat Transfer -
Chemical Engineering Progress Symposium Series, Storrs, Connecticut, v. 56, No. 20, 1960,
- 15. Spano, A. H., " Quarterly Technical Report SPERT Project, April,
-.. May, June, 1964," ISO 17030.
4 74 t-- -"+r v - - * -r-.='ev m e-i-r w*-M e -m- -"
SAR 5/91 I
- 16. Dee, J, B., et al., " Annular Core Pulse Reactor," General Dynamics, General Atomic Division Report GACD 6977 (Supplement 2).
1966.
- 17. Adler, J., et al., " Users and Programmers Manual for the GCC-3 Multigroup Cross Section Code," General Dynamics , General Atomic Division Report GA 7157, 1967.
- 18. Lenihan, S. R., " GAZE 2: A One-Dimensional, Multigroup, Neutron Diffusion Theory Code for the IBM 7090," General Dynamics, General Atomic Division Report GA 3152, 1962.
- 19. Dorsey, J. P., and R. Forehlich, " GAMBLE A Program for the Solution of the Multigroup Neutron Diffusion Equations in Two Dimensions, with Arbitrary Group Scattering, for the UNIVAC-1108 Computer," Gulf General Atomic Report GA-818P, 1967,
- 20. 1.athrop, D. K., "DTF-IV, A FORTRAN IV Program for Solving the Multigroup Transport Equation with Anisotropic Scatterings," USAEC Report LA-3373, Los Alamos Scientific Laboratory, New Mexico, 1965.
- 21. Adler, F. T., G. W. Hinman, and L. V, Nordheim, "The Quantitative Evaluation to Resonance lategrals," in Proc . 2nd Intern. Conf.
Peaceful Uses At. Energy (A/ CONF. 15/P/1983), Geneva, International Atomic Energy Agency,1958,
- 22. Brown, H. D., Jr., Culf General Atomic. Inc., "THERMIDOR - A FORTRAN 11 Code for Calculating the Nelkin Scattering Kernel for Bound Hydrogen (A Modification of Robespierre)," unpublished deta.
- 23. Nelkin, M. S., " Scattering of Slow Neutrons by Water," Phys. Rev.
11.9, 741-746 (1960).
- 24. McReynolds, A. W., et al., " Neutron Thermalization by Chemically-Bound Hydrogen and Carbon," in Proc. 2nd Intern. Conf. Peaceful Uses at Energy (A/ Conf. 15/F/1540), Geneva, International Atomic Energy Agency, 1958.
- 25. Whittemore, W. L., " Neutron Interactions in Zirconium Hydride,"
USAEC Report GA-4490 (Rev.), General Dynamics, General Atomic Division, 1964.
- 26. Bell, J., " SUMMIT: An IBM-7090 Program for the Computation of Crystalline Scattering Kernels ," USAEC Report, General Dynamics, General Atomic Division Report GA 2492, 1962.
- 27. Beyster, J. R., et al., " Neutron Thermalization in Zirconium Hydride," USAEC Report, General Dynamics, General Atomic Division Report GA-4581, 1963.
4-75
SAR S/91 t .
- 28. k'oods , A. D. B, et al , " Energy Distribution of Neutrons Scattered from Craphite, Light and Heavy Water, Ice, Zirconium liydride , Li thium Hydride, Sodium Ilydri de , and Artuuonium Chloride ,
by the Beryllium Detector Method,' In Proc. Symp. Inelast ic Scattering of Neutrons in Solids and Liquids, Vienna, Austria, Oct. 11-14, International Atomic Energy Agency, 1960.
- 29. . Jordan, D. P, . and G. Leppert, " Pressure Drop and Vapor Volume with Subcooled Nucleate Boiling," Int. J . tirat Mass Trang. h, 751-161 (196?).
30 McAdams, ny , e t L , pp . 390-392.
- 31. Levy, S., " Forced Convection Subcooled Boiling-Prediction of Vapor Volumetric Fraction," Int. J. llent Mass Trans 1Q, 961 965 (1967),
- 32. McAdams, op. cit., p. 392.
- 33. Bernath, op. cli., pp. 95-116.
- 34. Simnad, M. T., "The U-Zrilx Alloy: Its Properties and Use in TRIGA 5'uel ," CA Proj ec t No. 4314, E-ll7 833, February 1980, pp. 4-1,7.
I 4-76
SAR 5/91 I Chapter 5 REACTOR COOLANT SYSTEM The TRIGA is designed for operation with cooling provided by natural convective flow of demineralized water in the reactor pool. The suitability of this type of cooling at the power levels for this TP.ICA has been demonstrated by numerous TRIGA installations throughout the world. The primary functions of the coolant system are,
- a. to dissipate heat generated in the reactor,
- b. to provide vertical shielding of radiation from the reactor and allow access to the reactor core.
Heat dissipation is satisfied by natural convective flow of pool water through the reactor core and forced circulation of the pool water through an external heat exchanger. The pool coolant volume is composed of approximately 41.0 cubic meters in a two by three meter oval pool with a vertical depth of 8.1 meters. A vertical shield is provided by about 6.8 meters of water above the reactor core. Other functions provided by the coolant system are:
- a. minirnize corrosion of all reactor components, particularly the fuel elements,
- b. maintain a minimal level of radioactivity in the reactor pool water and
- c. maintain optical clarity of pool water.
These three functions are accomplished by a purification system that is included as a part of the coolant system, 5.1 DESIGN BASES The design basis for the reactor coolant system is predicated on its primary function, reactor cooling. Other coolant system functions establish the design bases for the purification circuit. 5.1.1. Reactor Core Heat Removal To assure adequate reactor cooling, the e f fec tiveness of natural convective cooling has been evaluated with respect to the peak heat flux which may be achieved in the reactor. This evaluation then establishes the maximum heat flux beyond which fuel element cladding integrity cannot be assured. 5-1 l
_ . . . _ ._. .. _ _ _ _ _ ~ _._._. _ ,_ _ . _ ._ . _ _ _ _ _ - - _ _ _ _ _ _ SAR 5/91 Based on these evaluations, it is concluded that for steady state operation the coolant inlet temperature and maximum heat flux at which fuel clad integrity '.s no longer assured is determined by the curve relating heat flux and coolant temperature for the hottest coolant channel, The maxiwum design temperature of the coolant system, coolant inlet temperature, is 120*F (48.9'C). The maximum allowable peak heat flux at this temperature is 325 kBtu/hr ft2 (103 watts /cm2) corresponding to a power level of 1900 kW for an 85 element core. Since the maximum-licensed power Icvel is 1100 kW, the resulting maximum heat flux will be 188 kBtu/hr-f t2 (59.4 watts /cm2) which is well below the value at which clad integrity may be questioned. 5.1.2. Reactor Pool Heat Removal Supplemental cooling of the reactor pool is required for continuous operation at the rated power level. A heat rate of 20,7'C/ hour is expected with . the reactor operated at 1000 kW. Heat removal from the pool is provided by heat exchange with a chilled water supply. The chilled water supply is operated by the University for cooling of Research Center buildings and equipment. Chilling capacity is provided by multiple 1200 ton (4229 kW) units. At reactor rated
. power the heat removal capacity required is represented by about 25% of the chilling system capacity of one unit. A tube and shell heat exchanger is installed for heat removal from the reactor pool to the available chilled water sysrem.
5,1,3. Heat Exgj) anger Design Bases Heat excl. anger capacity is designed for a stable operating temperature of the reactor pool at or below the coolant design temperature, 120*F (48.9'C) for convective reactor core cooling. The stable temperature 1: maintained by a heat exchanger capacity equivalent to the reactor core tharmal output capacity. Other heat losses such as evaporation, or heat gains from the pump,' are considered negligible. Heat transfer is defined by q - U A 6T m (1) where .U - overall heat transfer coefficient (watt /m2 ..C) A - surface area for heat transfer (m2)
-6Tm - true mean temperature difference (*C)
For a tube and shell heat exchanger the overall heat transfer coefficient is_- composed of three terms, the convective heat- transfer f rom L the fluid in the tubes to the tube walls, the conductive heat transfer thru the tube wall, and the convective heat transfer f rom the outside tube wall to the fluid in the shell of the heat exchanger. Based on the outside tube area for heat transfer, the overall heat transfer coefficient is defined as [1], f 5-2
- _ . -~_ - - - - - - - - . . .-. - - . . - - . _ - - .. _.- .-_ .
SAR 5/91 Ao Ao i n (r o/ri) 1 - l Uc - + + , (2) Ahti 2*kl ho where 2 Ao - total outside tube area (m ) Ai - total inside tube area (m 2)
- ri - tube inside radius (m) l ,
ro - tube outside radius (m) hi - convective heat transfer coefficient between fluid in tubes and tube wall (W/m2.*C) ho - convective heat transfer coefficient betveen fluid in shell and tube wall (W/m2.*C) k - conductive heat transfer coefficient .n the tube wall (V/m2 *C) 1 - tctal tube length in heat exchanger :m) A correction is applied for fouling of heat exchanger caused by
' - buildup of various deposits. - The overall heat transfer coefficient for a fouled heat exchanger is determined by 1
Ur - (3) Rt + 1/Uc where Rg is the-fouling factor, (non-dimensional). The convective heat transfer coefficient is defined as
. Nu k h- (4) d where Nu - Nusselt Number k - thermal conductivity of the fluid evaluated at the appropriate average temperature (W/m *C) d - tube diameter or applicable hydraulic diameter (m)
The complicated nature of turbulent flow heat transfer is described by a Nusselt number determined by experimental correlation 5 with the Reynolds and Prandtl Numbers. Dittus and Boelter [2] recommend the following relation for fully developed turbulent flow in tubes: 5-3 q w v 4 ~ m o v e m --n.- s-,n+q . e--w +,,,-t w me er = ~
2 SAR 5/91 O Nut - 0.023 Re 0,8 Pr" , (5)
-where parameters are measured inside the tubes Re - Reynolds Number based on tube diameter, Pr - Prandtl Number at average fluid temperature, n - 0.4 for heating, n - 0.3 for cooling.
The relation for the shell side of a baffled cross flow heat exchanger is suggested by Colburn (3) as follows: Nu, - 0.33 Re 0.6 Pro.33 , (6) where parameters are measured outside the tubes and Re - Reynolds Number based on tube outside diameter and velocity at minimum shell cross sectional area, Pr - Prandtl Number at average fluid temperature. The product terms, A 6Tm, are defined consistent with the definition of U and heat exchanger design. The total cross sectional area of the tubes is represented by the heat transfer area, A, as specified by the heat transfer coefficient, U, The true mean temperature difference, 6Tm, is related to the heat exchanger type by a correction factor, F, and a log mean temperature difference, 1RTD [4]. The correlation relates a simple single pass heat exchanger with more complex multiple pass baffled units. A relation is defined by 6Tm - F
- 1RTD , (7) where F - correction factor [5,6],
s IRTD - (Ta - Tb )/In(Ta/Tb ) , (8) For a counter flow heat exchanger Ta - (T hot fluid in - T cold fluid out) . (9) Tb - (T hot fluid out -T cold fluid in) . (10) Actual heat exchanger capacity is calculated using an energy balance on either the shell or tube fluid. The heat transfer is defined as: q - C (T in - T out) . (11) 5 -_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . .
]
.. . . - . . _ . . -- .. . - - - - . . .- ~ ..
SAR 5/91 ) U where C-mc p .. m - mass flow rate, cp . fluid specific heat, Tin - Temp of fluid entering heat exchanger, To ut - Te.mp of fluid exiting heat exchanger, In the current case Tout of either fluid is not known. Only Tin (100* F pool water, 48* F coolant water) and the mass flow rate of both fluids are known. To determine Tout the effectiveness /NTU method (4,7] is used. The dimensionless parameter called the heat exchanger effectiveness e is defined as Actual Heat Transfer
<- , (12)
Maximwn Possible Heat Transfer where the maximum possible heat transfer is 9 max - Cmin (Thot in
- Teold in). (13)
Substituting (11) for each flu!d and (13) into (12) results in Chot (That in - T hot out? c- , (14) Cmin (T h at in - Teold in) for the hot fluid and C cold (Teold out Tc old in) c- ,. (15) Cmin (Thot in - Teold in) for the cold fluid, The heat exchange effectiveness determined by.[7] for a shell and tube heat exchanger with one shell pan and any multiple of tube passes is-1 + exp ( NB) -
-1
( -z 1+r+B( ) (16) 1 - exp ( NB) - where r- Cmin/ Cmax (17) N- UA / Cmin (10) 5-5
SAR 5/91 U - overall heat transfer defined in (2) A - surface area for heat transfer B-(1+r)b 2 (19) Once the effectiveness is calculated, (14) and (15) are used to determine Thot out and Teold out. These may then be used in (11) to determine the capacity of the bett exchanger. 5.1.4 Vater Purification Bases The functions of corrosion control, radioactiv'.ty control, and optical clarity of the coolant water are provided by filtration and ion exchange. Control of the water purity is performed by analysis of the water conductivity. Measurements of water conductivity as low as 2.0 micromho per centimeter ( or resistance of 1 megohm per centimeter ) are maintained by filtration and ion exchange. The condactivity is reduced further by control of materials exposed to the reactor coolant, minimizing dust settling to the pool surface, and occasional cleaning of pool surfacca. Experience has shown that conductivities of 5.0 pmho/cm are sufficient to maintain acceptable limits on corrosion plus good water optical quality and removal of activation products in the water, a 5.2 SYSTEM DESIGN
~
Principle components of the coolant system are the aluminum reactor pool tank, the external cooling loop consisting of heat exchanger and pump, and the purification loop consisting of filter, resin bed and pump. Most of the total coolant volume is represented by the approximately 41.0 cubic meters of water in the reactor pool. Typical flow diagrams for the systems are shown in Figures 5-1 and 5-2. 5.2.1. Coolant System Suction of water from the pool is provided by an inlet which extends no more than 2 meters below the top of the reactor tank. The coolant water is drawn through the coolant pump and forced through the heat excharger. Return of cooled water to the reactor pool is provided by single or multiple discharge outlets above the reactor core or an outlet near the tank bottom. A diffused water jet is created at the outlets above the reactor core by a no::zle. Delay and diffusion of the reactor core convective coolant column is enhanced by the action of the coolant discharge nozzle. Accidental siphoning of reactor pool water is prevented by the presence of suction brsaks on both suction and discharge lines of the coolant system. Siphon breaks are created by holes located in the lines approximately half a meter below the normal water level. 5-6
SAR 5/91 O n
}
X
& ld OB e se J JZ D O ef
)' @s l _
A 9 - O
- eses ***
X ! 5 O j gl 1
$x 2 ! 00 X 1
bo E 0
$ b
' gly1 1
XX l
$ ees COOLANT SYSTEM LAYOUT O Figure 5-1 5-7
SAR 5/91 e --_-3 I I
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r1 9 s X 3 g 7 l - 2 6 'N = g 6 t V
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1 PURIFICATION SYSTEM 1AYOUT Figure 5-2 5-8
SAR 5/91 The heat exchanger and pump, the major components of the cooling system, are located in room 1.104b at about the same vertical level as the reactor core. Valves are provided in the coolant loop for control and isolation of the cooling system function. Specifications of cooling system components are listed in Table 5-1. A positive pressura difference of 1 psi (7 kilopascals) between the shell side outlet and tube side inlet of the heat exchanger is designed to prevent leakage of primary pool coolant into the secondary chilled water system. The pressure difference is maintained under varying flow conditions by a differential pres "re controller which regulates the position of a throttle valve in the heat exchanger shell side outlet pipe. Coolant water supply temperature is regulsted by a temperature controller coupled to a mixing valve in the chilled water supply line. Table 5-1 REACTOR COOLANT SYSTEM DESIGN
SUMMARY
Reactor Tank Material Al plate (6061) Thickness 1/4 in.. Volume (maximum) 11000 gal.(0.635(41.64 cm)3) m Coolant Lines Pipe Aluminum (6061) Valves Iron Plastic liner, 316 ss. Ball and Stem Fittings Aluminum (Victaulic) Coolant Pump Type Centrifugal Material Stainless steel Capacity 250 gpm (15.8 liter /sec) Heat Exchanger Type Shell and tube Materials: shell Carbon steel tubes 304 stainless steel Heat Duty 1000 kW Flowrate: tubes 250 gpm (15.8 liters /sec) shell 400 gpm (25.2 liters /sec) Typical Parameters: Tube inlet 100*F 42 psia Tube outlet 69'F 27 psia Shell inlet 48'F 55 psia Shell outlet 67'F 48 psia 5-9
l SAR 5/91 5.2.2, Purifica11on System Suction of water from the pool is provided by two inlets, neither of which extend more than 2 meters below the top of the reactor tank, Valves at the paol surface allow suction from either a subsurface inlet or from a surface skimmer designed to collect and remove floating debris, Accidental siphoning of reactor pool water is prevented by siphon breaks similar to those on the coolant piping. The purification skid is located in room 1.104b at about the same level as the reactor core. The skid consists of a pump, flovmeter, filter, resin bed, and instrumentation. Normally the purification system is operated continuously to provide removal of suspended particles and soluble ions in the coolant water. The system flow rate is about 10 gpm (0.6 1ps). Purification functions of the loop are generated by two components, a filter for removal of suspended materials and a resin bed for removal of soluble elements. Typical filtration is provided with 25 micron filters. Typical ion exchange is provided by .085 cubic meters of mixed cation and anion resin. Water purity is measured by conductivity cells at the inlet and outlet of the resin beds, Return flow to the pool is thru a subsurface discharge pipe. Valves are provided for isolation of av uction or return lines, and for isolation of system components f or maintenance or resin replacement. 5.2.3. Rater System Instrumentation Several monitoring sensors are installed to allow remote readout of water system parameters in the reactor control room. Other system parameters are indicated by local monitoring devices. Parameter n:onitoring points are illustrated in Figure 5-1 and 5-2. The parameters that are considered part of the water system instrumentation system are presented in Figure 5-3. Indication of the reactor pool status is determined by two sensors located in the pool. Pool level and bulk pool temperature sensors in the pool are monitored in the control room. An annunciator alarm indication is generated in the control room by abnormal pool levels and by high pool temperatures. The cooling system parameters normally available in the control room include coolant temperatures, flowrates, and dif ferential pressure status. Two temperature probes, one in the pool suction line and one in the pool discharge line, allow monitoring of heat exchanger cooling function. Typical temperature probes used are resistance temperature detectora (RTD's). Two flow meters, one in the chilled water line and one in the pool water line provide information on system flow rates. A dif ferential pressure monitor provides an alarm if the pressure at the high pressure point on the heat exchanger tube side is not less than the low pressure point on the shell side. The differential pressure is designed for a difference substantially greater than 7 kilopascals (1 lb/sq. in.). 5-10
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SAR 5/91 Numerous water system parameters are measured by local pressure or temperature sensors in the system lines. Both temperature and pressure probe are located on the inlet and outlet lines of the pool water side and chilled water side of the heat exchanger. A local indication of flow in the coolant loop is provided by the pressure drop across a venturi in the flow path. Purification loop flow is measured by an in line flow meter. Water pressure before and after the filter in the purification loop la measured for indication of filter condition. Water quality is measured by two conductivity cells in the purification loop. The cells are located on inlet and outlet lines of the demineralizer that readout locally in the control room. Typical conductivity cells are composed of two parts, titanium electrodes shielded by ryton for conductivity measurement, and a thermister for temperature compensation. A Wheatstone bridge circuit on the purification skid is connected to the cells. A switch allows selection of either inlet or outlet conductivity. 5.3 WATER SYSTEM DESIGN EVALUATION The water system including the reactor pool and the external cooling and purification loops have similar design features as used in many other operating TRIGA facilities. The demonstrated capability and integrity of this system provides assurance that the coolant system will perform its function properly and safely. Availability of prol water for cooling and vertical shielding is assured by designing the system with siphon breaks on suction lines and discharge lines within 2 meters of the normal pool level. Greater losses of pool water are extremely improbable, although they could conceivably be initiated by rupture of the reactor tank. As shown in the loss of pool water accident analysis, even with complete loss of pool water fuel clad integrity is not threatened. Adequacy of reactor cooling is assured by the large amount of cooling capacity inherent in the reactor pool. volume as well as the capacity of the external cooling circuit which can dissipate heat at a rate equivalent to 1000 kW steady-state operation. If available heat exchanger capacity is diminished to 900 kW and initial pool temperature is 100*F, the reactor can be operated for more than 5 hours before the bulk pool tempe ratt reaches 120*F. The actual time would be considerably longer ,ince as bulk pool temperature increases heat exchanger heat removal capacity increases. Without external cooling or other heat loss the bulk pool temperature will rise about 20.7'C after one hour of operation at a steady state power level of 1000 kW(t). Heat removal capacity and thus pool heat rate is specified by analysis of a tube and shell heat exchanger. At a flow rate of 400 gal / min (25.2 liters /sec) of chilled water at 48'F (8.89'C) a heat removal rate of 1140 kW is expected. The presence of fouling in the heat exchanger is considered minimal based on the purity of the two heat 9 exchanger fluids. of .0004. 5-2. Capacity is reduced to 1070 kW for a fouling factor The heat transfer and hydraulic parameters are shown in Table 5-12
l SAR 5/91 Table 5-? IlEAT EXCllANGER: IIEAT TRANSFER AND llYDRAUI.lc PARAMETERS Tubes: Outside Diatteter 0.750 inch (1.91 cm) Wall Thickness 0.049 inch (0.124 cm.) Theimal Conducttvity 8. 21 Bt u/hr- f t
- F Flow Area:
Tube Side 8.1 inz (52.3 cm2) Shell Side 33.8 in? (218 cm2) lle n t Transfer Surface 346 ft2 (32.1 m2) Average Prandtl Number Tube 5.38 Shell 8.41 Average Kinematic Viscosity Tube 8.63 x 10'6.ft2/sec (8.02 x 10'7 m2 /sec) shell 1.28 x 10-5 ft 2/sec (1.19 x 10 6 m2/see) Reynolds Number 7 Tube 6.19 x 10'0 Shell 2.02 x 10 Correc t ive lleat Trans f er Coef fic ient s Inside Tubes 1710 Btu /hr-ft 2*F (9701 W/m2. C) Outside Tubes 1395 Btu /hr ft 2*F (7922 W/m2 *C) Overall llent Transf er coef ficient clean 520 Btu /hr-ft2.'F (2953 W/m? *C) Fouled 430 Btu /hr-ft2*C (2442 W/m2 *C) Effectiveness (<) Clean 0.60 Fouled 0.56 1.MTD 26.l*F (14.5'C) Corrective Factor F 0.83 Capacity 9 Clean Fouled 1140 kW 10/0 kW 5 13
SAR 5/91 Experience with this purification equipment in other TRICA systems has-shown that coolant conductivity can be easily maintained at levels of less than five micrombos per centimeter using the materials contained l- in the coolant system design. Furthermore, this experience has shown th at'no apparent corrosi on of fuel clad or other components will occur if the conductivity of the water does not - exceed five micrombos per centimeter when averaged over a 30 day period. Control of radioactivity in the coolant is provided by the purification system, Should radioactivity be released from a clad leak or rupture of an experiment, detection of the release would be signaled by the continuous air acnitor or by the reactor room area monitors.
' Based on coolant transport time calculations in the safety analysis section, these monitors should register an increase in coolant radioactivity within approximately 60 seconds of the time of radioactivity release. The transport time is estimated from the time for the coolant exposed in the core to reach the surface of the water where the continuous air monitor will detect a release of radioactivity 4
from the pool water. An alternate indication of radioactive release is provided if a water activity monitor is installed or by a CM detector area monitor.
)
5-14
SAR $/91 Chapter 5 References
- 1. llolman , J . I' . , "lic a t Transfer", Mc C r aw - }{ill , Fourth Edition, 1976, pp 386 391,
- 2. Dittus, F.W. and Boelter, L.M.K., " Univ. California (Berkeley) l'ub . Eng.", vol. 2, pp 443, 1930.
3, Colburn, A.P. , "A Method of Correlating Forced Convec tion llent Transfer Data and Comparison with Fluid Friction". Trans. AlchE vol. 29, pp 174 210, 1933.
- 4. White, F.M., "lient Transfer" Addison Wesley, 1984, pp 512 513.
- 5. Bowman, R.A., Mueller, A.C., and Nagle , W.M., "Mean Temperature Difference in Design". Trans. ASME, vol. 62 (1940), pp 283 294.
- 6. Tubular Exchanger Manufacturers Association, " Standards TEMA/3rd Ed.", New York, 1952.
- 7. Keys, W. and London, A.L., " Compact lleat Exchangers", McGraw-liill ,
Second Edition, 1964. 5-15
l SAR 5/91 Chapter 6 INSTRUMENTATION AND CONTR01. SYSTEM Design of the instrumentation and control system is intended for new TRICA reactor facilities and replacement of old reactor consoles. Initial verification and testing of the design by the manufacturer is a requirement prior to installatton at The University of Texas at Austin. An evaluation by the University of the instrument and control console for the TRICA is part of the initial installation of the console by the vendor. The system development, installation and initial testing is the responsibility of the vendor, General Atomics. Control of research reactors, including the TRICA family of reactors, has been accomplished using instrumentation and control The logic systems based on solid-state, hardwired, analog circuitry. for this control circuitry has been developed through many years of experience and includes attention to matters such as (1) necessary redundancy in the single inputs, signal processing, and controls; and (2) prevention of single-mode failure mechanisms. Present day developments of personal computers (PC), such as the f amily of IBM PCs, provide many opportunities to replace the earlier analog circuitry by compact and powerful computer-based control systems. The system described in this document is a microprocessor-based instrumentation and control system developed by the General Atomics (CA) TRIGA Reactor Division. This system incorporates (1) a digital wide-range neutron power monitor, (2) analog power safety channels, (3) a variety of state-of-the-art signal conditioners and process controllers, plus (4) a digital data acquisition and control system incorporating IBM PC/AT or compatible computers in industrial versions. 6.1 DESIGN BASES The design and manufacture of this system complies with the guidance given in American Nuclear Society and the American National Standards Institute Guide Criteria for the Reactor Safety Systems of Research Reactors (ANSI /ANS 15.15 1978) [1,2}. This standard serves the research reactor community in lieu of the ad hoc application of similar standards for power reactors. Even if single-failure criteria for plant protective actions - not deemed mandatory by ANSI /ANS 15.15 for negligible risk reactors were applied, the standard allows the use of simple redundancy, i.e., the monitoring of the same reactor parameter using independent, redundant equipment, to satisfy the single failure criteria for the reactor safety system. There are several advantages in a microprocessor-based system which enhat.ce system safety, reliability, and maintainability:
- 1. The use of powerful microcomputers allows data (operator input as well as output) to be more efficiently and systematically processed and recorded than ever before.
6-1
__- . - n - ._. . . . - - . _- -~--_- - . - . .- - ~ _ - - - . _ - 9 SAR 5/91
- 2. Several data reductions not previously possible (such as on-line calculation of the prompt period during a pulse) can be done in near real-time.
- 3. On line self-diagnostics can be performed which determines the state of the system at all times.
- 4. Operational surveillance and operations data are accommodated as never before with all information gathering and ptocessing done routinely and regularly by the console computers.
The Instrumentation and Control System for the TRIGA reactor [3) is a computer-based design incorporating the use of one multifunction, NH-1000 microprocessor neutron flux monitoring channel and two companion _ NP-1000 current mode neutron monitoring safety channels. The , combination of these two systems provides an independent operating channel and the redundant safety function of percent power with scram. The NM 1000 provides wide range log power and multi range linear power from source level to full power. The control system logic is contained in a separate control system computer (CSC) with graphics and text displays which are the interface between the operator and the reactor. Another system for data acquisition and control (DAC) functions as the interface point for interface circuitry, process signals and conununications . The multifunction NM 1000, two NP 1000 units, and two g system microprocessors, the control system computer (CSC) and data acquisition and control system (DAC) are development products of General Atomics. The basic system configuration is shown in Figure 6 1. Information from the NH 1000 channel is processed and displayed by the_CSC. The two NP-1000 are -independent channels that deliver power level- data to the safety system scram circuit, hardwired analog indicators, and to the CSC for processing and display. Operating ranges for the neutron channels are shown in Figure 6-2. The NM 1000 digital neutron monitor channel was developed for the nuclear power industry and is fully qualified for. use in the demanding and restrictive conditions of a nuclear power generating plant. Its design is based on a special CA designed fission chamber, and low noise ultra-fast pulse amplifier. The NP.1000 was - developed specifically for
- use with research reactor safety systems and includes several features not usually found in this type of application.
The CSC and its acquisition system, the DAC, manages all control rod movements, accounting for such things as interlocks and choice of l. particular operating modes. It also processes and displays information on control rod position, power level, fuel and water temperature, and can display pulse characteristics. The CSC also performs many other l functions, such as calibrating control rods, monitoring reactor usage, i and historical operating data can be saved for replay at a later date. l A computer based control' system has many advantages over an analog
\ system: speed, accuracy, reliability, and the ability for graphic
displays self-calibration, improved diagnostics, and logging of vital information.
62
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SAR 5/91-6.1.1, NM-1000 Neutron Channel The NM-1000 nuclear channel has multifunction capability to provide neutron monitoring over a wide -power range from a single detector. The selectable funrtions are any or all of the following:
- a. Percent power,
- b. Wide range log power.
- c. Power rate of change.
- d. Multi-range l'near power.
For the TRICA ICS, one NM-1000 system is designated to provide the wide range log' power function and the multirange linear power function. The wide range log power function is a digital version of the patented CA 10-deende log power system to cover the reactor power range from below surface level to 150% power and provide a period signal. For the log power function, the chamber signal from startup (pulse counting) range through the campballing (root mean square (RMS) signal processing) range covers in excess of 10 decades of power level, The self contained microprocessor combines these signals and derives the power rate of change (period) through the full range of power. The microprocessor automatically tests the system to ensure that the upper decades are ( operable while the reactor is operating in the lower decades and vice versa when the reactor'is at high power. For the multirange function, the NM 1000 uses the same s'ignal source as for the log function, However, instead of the microprocessor ; converting the signal into a log function, it converts it into 10 linear power ranges. This feature provides for a more precise reading of liacar power level over the entire range of reactor power. The same el'-checking- features are included for the log function. The mult range function is either auto-range or slave to a position switch on tue operator's console via the control system computer. A linear power level signal is available for the percent power safety function for 1 to 125%. The NM-1000 system is contained in two - National Electrical Manufacturers Association (NEMA) enclosures, one for the amplifier and-one for the processor assemblies. The amplifier assembly contains modular plug in subassemblies for pulse preamplifier electronics, bandpass filter and RMS electronics, signal conditioning circuits, low voltage power supplies, detector high voltage power supply, and digital diagnostics and communication electronics. The processor assembly is made up of modular plug in subassemblies for communication electronics (between amplifier and processor), the microprocessor, a control / display module, low voltage power. supplies , isolated 4 to 20 mA outputs, and isolated alarm. outputs. Outputs are Class lE as specified by IEEE. Communication between the amplifier and processor assemblics 1s vla two twisted-pair shielded cables. 6-5
l SAR 5/91 The amplifier / microprocessor circuit design employs the latest concepts in automatic on-line self diagnostics and calibration verification. Detection of unacceptable circuit performance is automatically alarmed. The system is automatically calibrated and checked (including the testing of trip levels) prior to operation. The checkout data is recorded for future use, and operation cannot proceed without a satisfactorily completed checkout. The accuracy of the channels is 13% of full scale, and ttip settings are repeatable within 1% of full scale input. The neutron detector uses the standard 0.2 counts per "nv" fission chamber that has provided reliable service in the past. It has, however, been improved by additional shielding to provide a greater signal-to-noise ratio. The low noise construction of the chamber assembly allows the system to respond to a low reactor shutdown level which is subject to being masked by noise. 6.1.2. [iPflp00 Power Sdety Channel The NP-1000 Power Safety Channel is a complete linear percent power monitaring system mounted within one compact enclosure which contains current to voltage conversion signal conditioning, power supplies, trip circuits, isolation devices, and computer interface circuitry. The power level tiip circuit in normally hardwired irto the scram system and the isolated analog outputs are monitcred by the CSC as well as being hardwired to a bargraph indicat.or. A special version of the safety channel, the NPP-1000, provides measurement functions for peak pulse power, total pulse energy, automatic gain change and related trip points. The control system automatically selects proper gain setting for steady-state or pulse mode when the operator determines the reactor operating mode. Peak pulse power and total pulse energy are also set by the pulse operation mode. Both safety channels, the Nr iOOO and the NPP-1000, are identical except for the peak and energy circuits. The detector for each safety channel is either an ionization chamber or self-powered in-core detector. 6.1.3. Reactor Control Console A conceptual layout of the control console is shown in Figure 6-3. The reactor control console contains several components needed by the operator for reactor control. Included are the following: a, Reactor control panels.
- b. Control System computer (CSC),
- c. Two color graphics CRT monitors (one is high resolution).
- d. Power and temperature meter panels.
- e. Two disc drives and a graphics printer.
6-6
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SAR $/91 A keyboard interface to the system computer is provided for operator control of several system functions. As previously mentioned,
; the power and period information from the NM 1000 channel and power i level from the Np.1000 channels are processed and displayed by the CSC.
lloweve r , several wide range channel parameters are also present on
- linear bargraph rneter displays at the console. Each Np.1000 safety system is independent, has its own output displays, and connects
- directly to the control system se rain circuit. Thus, wide range log power, period. multirange linear power and both percent power channels, have their output displayed on reters as well as on the color CRT. This i is also true of fuel temperature. Typical layouts of the console panels ,
and video dispisys are shown in Figures 6 4 and 6 5. l Functions of the rod control panel are represented in Figure 66 and are presented as:
- a. Key switch for rod magnet power (also operates " Reactor On" lights),
- b. Rod cont rol switches and Annunciators. ,
- c. SCRAM switch for safety function.
- d. Annunciation is also provided for reset of the audio channel, as well as for reset of the alarrn indicator >
following alarm clearance. The CSC provides all of the logic functions needed to control the , reactor and augments the r.afety system by reonitoring for undesirable operating characteristics. It displays teactor operational information in a color format on a high resolution CRT- inonitor for ease of comprehension. Essentially all of the control systems logic contained in previous- TRIGA reactor control systwas is incorporated into the CSC. lloweve r , .instead of using electronic circuits and electrical relay circuits, the logic programmed into the computer. The availability of the computer allows great versatility and flexibility in operationally related activities aside from the direct control of rod movements. Many.other functions can also be performed by the CSC, such as monitoring reactor usage, storing pulse data, reactor operating history and logging operator usage. ; Two auxiliary cabinets can be provided to the consolo for che addition of process _ instrument readout. 6.1.4. Renetor-O m ring Modes - There are four standard operating inodes: manual and automatic, pulse, and square wave. The manual and automatic modes apply to the steady state reactor condition; the pulse and square wave rnodes are- the conditions implied by ! their names and require a pulse rod drive. l 68
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SAR $/91 The manual and automatic reactor control modes are used for reactor operation from source level to 1001 power. These two modes are used for manual reactor startup, change in power level, and steady state operation. The pulse mode generates high power levels for very short periods of time. High power and low-power pulse mode options are available. The square wave operation allows the power level to be rained quickly to a desired power level. Manual rod control is accomplished by the lighted push buttons on the rod control panel. The top row of annunciators, when 111tuninat ed, indicates magnet contact with the armature and snagnet current. Depressin6 any one of the AIR MAGNET push buttons will interrupt the current to that magnet and extinguish the magnet current on indication. If the rod is above the down limit, the rod will fall back into the cort and the AIR MAGNET light will remain extinguished until the magnet is driven to the down limit where it again contacts the armature. The middle row of pushbuttons (UP) and the bottom row (D0k'N) are used to position the control rods. Depressing the pushbuttons causes the control rod to move in the direction indicated. Several interlocks prevent the movement of the rods in the M2 direction under conditions such as the following:
- a. Scrams not reset.
- b. Magnet not coupled to armature,
- c. Source level below minimum count,
- d. Two UP switches depressed at the same time,
- e. Mode switch in one of the pulse positions,
- f. Mode switch in AUTO position (regulating rod only).
There is no interlock inhibiting the slpyn direction of the control rods except in the case of the regulating rod while in the AUTO mode. Automatic (servo) power control can be obtained by switching frorr manual operation to automatic operation. All the instrumentation, safety, and interlock circuitry described above applies and is in operation in this mode. However, the regulating rod is now controlled automatically in response to a power level and period signal. The reactor power level is compared with the demand level set by the operator and is used to bring the reactor power to the demand level on a fixed preset portod. Logic for- the - automatic control operation by proportional and - integral control is contained within the digital algorithms of the . control system. The purpose of this - feature is to automatically maintain the preset power level during long term powet runs. The function of automatic control is provided by the regulating rod with a stepping motor drive. 6 12
1 i SAR 5/91 l l V Reactor control in the pulsing mode consists of establishing criticality at a flux level below 1 kW in the MANUAL mode. This is accomplished by the use of the motor driven control rods, leaving the transient rod either fully or partially inserted. The mode selector switch is then depressed. The MODE selection switches automatically l 4 connect the pulsing chamber to monitor and record peak flux (nv) and energy release (nyt). pulsing can be initiated from either the critical or suberitical reactor state. In a square wave operation, the reactor is first brought to critleality below I kW, 1 caving the transient rod partially in the core, i All of the manual instrumentation is in operation. The transient rod is ejected from he core by means of the transient rod FIRE pushbutton. ' Vhen the power level reaches the demand level, it is maintained much the same as in the automatic mode except that two rods are used to maintain power af ter the pulse rod is ejected. 6.1.$. Reactor Scram and Shutdovn System A reactor protective action [4] interrupts the magnet current and result.s in the immediate insertion of all rods under any of the following:
- a. liigh neutron fluxes from either NP.1000 or NPP1000.
- b. liigh voltage failure on the NH.1000, NP 1000, or NP1000.
' c. liigh fuel temperature (one out of two),
- d. Manual scram.
- e. Peak neutron flux or energy (pulse mode),
- f. Minimum period (availabic for use as desired),
- g. External safety switches (for experiments).
h .- less of electrical power to the control console l 1. Watchdog circuits for each computer to monitor computer l status by updating timers. I l All scram conditions..are automatically indicated on the CRT monitor. A manual scram will also insert the control rods and may be l used for a normal fast shutdown of the reactor. The scram circuit safety: function is an -independent system- that--depends on- wiring independent of the digital control system functions. Several conditior. of the digital- processing system will cause the
-scram mode condition. Among these are the loss of-communication between the two computers, a database timeout condition or failure of a digital input scanner. By updating dual programmable timers, watchdog circuits at periodic intervals, determine the execution status of key elements 01 the computer digital program.
6 13
SAR $/91 Two options for reliable operation performance may be installed as
-necessary. One option for conditions requirinS loos teria, high power steady state operation, is configuration of the three safety channels l with 2 out of 3 logic, allowing one channel to be out of service without requiring reactor shutdown. Another option is an uninterruptable power supply as auxiliary power for the reactor control and monitoring systems for interisittent power failures of periods upto 15 minutes.
6.1.6. Logie Functigns 6 A sitmplified cor. trol system logic diagram is shown in l'igure 6 7. The two separate flux monitoring safety channels ensure safe operat. ion of the reactor by monitoring the power level and act independently to shut the reactor down if a potentially dangerous condition exists. They provide information to the control system, which consista.of three major parts: a reactor control console (RCC), Control Systern Computer (CSC)
- and Data Acquisition Cornputer (DAC). In addition, there are two color graphics CRT monitors and a graphics printer. The high resolution display monitor contains basic reactor operation control data. A medium resolution display monitor provides infor: nation on annunciatoro and special control features. Data from both displays may be sent to the printer for a los record.
The CSC provides the operator with inunediat e information concerning reactor conditions visually on the monitors. At the same t irne , the DAC is collecting data f roin the reactor system and concentrating it into a permanent data base, which is transreitted to the CSC on roquest. and maintained for historical purposes. During operation of the reactor, the operator's commands to adjust control rod positions are transmitted frorn the CSC to the DAC to the drive mechanterns. In the automatic mode the DAC controls the position of t.be rods. The rod control program for automatic operation applies l proportional. integral differential control logic. Digital rod position indication is shown in inches, with a resolution of $ 0.1 in, and an accuracy equal to or better than 1 0.2% of indicated position, i The control rod interface accepts the digital commands from the data acquisition and control system (DAC) to operate the control rod motors. It contains the opto. isolation circuits which send the up down limits and loss of contact signals to_the control rod logic systers. An excitation power supply provides a stable reference voltago for the rod i position indicator system. The magnet supply furnishes the required 200 mA needed for the rod magnets--to hold control rods in contact with the armature. An opto. isolator detects the absence of magnet current to each drive magnet. A gamma chamber provides the signal for peak power (nv) and energy release (nyt) in the pulse mode. The nv/nyt amplifter provides the high impedance interface, ht S h voltage and calibration circuits for the pulsing detector. 1 6 14 E ,._ .... - . u _ . _ _ __. _ _ - ,-- - . . ~ , _ . - . . . . . _ _ - . . , _ - , - _ _ _ _ . _ _ _ _ _ . _ . ._
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asaus ietstrecs ania saquisette ase Ctarftn gattiggjggg a,. m.. ,- s mveman su we - vu..uw. n'on= ou u. ..iv. _ u, i. a, i. LOGIC DIAGRAM FOR CONTROL SYSTEM Figure 6-7 6 15
SAR $/91 All of the analog signals and digital signals are routed to the I DAC chassis. However, the prime reactor operating signals are also sent directly to the control roorn. These signals include log power, period, percent power (2), fuel temperature (2), and pulse mode signals for peak and energy. The DAC system converts the anslog signals to a digital equivalent for transtnitting along with the digital signals to the CSC in the l control room. The DAC chassis receives control instructions from the CSC, via the communication link, which in turn moves the control rods as requested by the operator and causes the individual subsystems to go to the calibrate mode when commanded by the system or operator. The fuel temperature transmitters are accurate, highly stable units which convert the 0 600*C fuel temperature into a 4-10 mA output signal. A level comparator is included which provides scram capability through an isolated contact state change when the preset level is exceeded. The water temperature transmitters are standard Resistance Temperaturo Detector (RTD) t ranstai t te rs which convert the O to 100'C temperature into a 4 20 mA signal. The transmitters have a self-contained power supply. External switches are provided with terminal strips to terminate and connect various switches to the DAC chassis (beam port open close, etc.) 6.1.7. Mechanical Hardware Typical reactor installation will be contained in two NEMA enclosure junction boxes, one electronic equipment cabinet, separate stepping motor power supplies installed in the reactor bay, and reactor operator console components installed in the reactor control room. The control console consists of the components needed by the. operator for reactor control . These components include rod control switches and annunciators, the digital rod position indicators, on line reactor status meters (power and temperature), the control systetr computer (CSC), reactor operating mode switch panel, color CRT monitor, printer, disc drives (2) and external switch annunciators (beam port open close, reactor access, etc.). Enclosure 1 contains NM-1000 high and low voltage power supplies, a pulse _ pre-amp with discriminator, an RMS Campbell convertor and a communications module. Enclosure 2 contains the NM 1000 microprocessor selected to _ provide the 10 decade log signal and the multi range linear function from the information provided by the circuits in enclosure 1. The information processed by the microprocessor is 10 decades of los power, rate of power change (period), multi range linear function, linear percent power from 1 to 125%, level trips from the log and linear percent power, calibrate and failure signals, 6 16 _ . . _ ~ _ , _ _ , . _ _ _ _ . _ . _ _ . _ . _ . . _ . . . _ _ _ . - . _ . _ - . _ . _ _ . _ . _ _ _ _ _ _ _ _ _ _
SAR 5/91 Enclosure 3 is a standard rack type equipment enclosure for electronic components. Space in the enclosure provides the terminal strips for connections to the various signal detection systems and the communications to the Rcc. The cabinet enclosure includes eight shelves with functional separation between shelves. Power supplies for subsystems are on shelf 1. Shelves 2 and 3 contain, respectively, ac digital and de digital circuits for processing input or output circuits. Shelf 4 provides several special modules for signal processing, The two power safety channels are positioned on shelf 5. Shelf 6 contains the regulating rod drive control signal module and signal terminations. The ' l 2 computer and its expansion chassis comprise the final two shelves 7 and
- 8. The regulating rod drive translator for the stepping motor drives is contained in a separate, fourth enclosure.
6,2 DESIGN EVALUATION The TRICA reactor console 16,7,8] has developed through the i successful operation of many installed facilities throughout the world. Design of the new ICS unit incorporates similar bacic logic functions proven effective in prior designs. Incorporation of digital electronic techniques in the design to replace analogue circuits is justified by improved performance. Functional self-checks, circuit calibrations, and automated data logging are implemented effectively and efficiently. Installation and verification of the original design is planned for a reactor operated by the ICS manufacturer, CA Technologies. Subsequent installation of the ICS unit at The University of Texas facility is planned with appropriate design changes, inclusion of facility specific parameters, and completion of an acceptance evalua11on. A multiphase design, development and installation program by the
' system manufacturer provides t.he initial demonstration of the system acceptance by analysis and review. No license modification was found necessary to implement the new digital system in place of the old analog system. An evaluation of the system relative to the facility safety analysis report was the basis for system evaluation. The . final test phase and subsequent operation will provide information on- system performance . and reliability. The analysis and test program determined '
- 1) that there was no increase of-the probahtlity of occurrence or the consequences of an- accident or malfunction of equipment important to sa fe ty , 2) that the system does not create the possfoility of an accident or malfun_ction of a dif ferent. type and 3) no reduction occurs in: the margin of safety as defined in the basis for any technical specification.
6-17
CAR 5/91 c References Chapter 6
- 1. " Criteria for the Reactor Safety Systems of Research Reactors",
American Nuclear Society, American National Standard, ANSI /ANS-15.15 1978.
- 2. "Hieroprocessor Based Research Reactor Instrumentation and Control System", INS 27, Rev. A., CA Technologies, August 1987.
- 3. General Atornien, private conununications , 1988.
- 4. " Safety Analysis of Microprocessor Reactor Control and Instrumentation Systern", The University of Texas at Austin, 1989
- 5. " Guidelines for the Verification and Validation of Scientific anc Engineering Computer Programs for the Nuclear Industry", America:
Nuclear Society, American National Standard, ANSI /ANS 10,4-1987.
- 6. " Operation and Maintenance Manual Microprocessor Based Instrumentation System for the University of TRICA Texas Reactor",
E117-1004, General Atomics 1989
- 7. " Operation and Maintenance Manual NM1000 Neutron Monitoring Channel", E117 1000, General Atomics 1989.
- 8. " Operation and Maintenance Manual NP1000/NPP1000 Percent Power Channel", E117 1010, General Atomics 1989.
6-18
SAR $/91
+
Chapter 7 DESIGN FEATURES AND AUXILI ARY SYSTEMS Features and systems of the facility Jesign provide several i functions that determine safety conditions and auxiliary support of operations. Engineering features include the reactor shield system and the air confinernent system for the reactor bay room and fuel storage systems. Auxiliary systems consist of standard systems for conditioned air, fire protection, communication equipment, and other systems. Engineering safety. features of the TRICA reactor are a property of the fuel material and element design. These engineering safety features include two features of the reactor core and two features of t ht. instrumentatf.on control and safety systern. Engineering safety features of the reactor core are the fuel material type and the fuel eMent cladding. Instrurnentation, control, and safety systern featutes are t ht-design of the " scram" safety circuit and the design of the actuators for rod drives and control rods. The description of each of these design features are. found in other chapters of this report. Building design features important to safety of the reactor are the coactor pool and shield structure, and the air confinement system for : tie reactor room. These systems are not cornplete engineering safety features but may require specification of specific conditions or parameters to assure the appropriate safety function. 7.1 DESIGN BASES The design of a structure to contain the TRICA reactor depends on the protection requirernents for . the fuel elements and the control of exposures to radioactive materials. Fuel elements and other special nuclear -materials are protected- by physical confinement and surveillance. The physical confinernent sill also control the release of radioactive materials during routine operation or potential accident conditions. Release of airborne radioactivity consists mostly of air activation products from routine operation or fission product materials from a non routine fuel element failure, Liquid and solid radioactive-material are also controlled to assure cornpliance with appropriate release criteria-standards. Other potential releases may be associated with specific types of experiments that require special equipment to provide sufficient control of material releases. The reactor rocm confinement is designed to control the exposure of operation personnel and the public from radioactive material or its release caused by reactor operation. Release criterion are based on Title 10 Chapter 20 of the U.S. Code-of Federal Regulations [1).
- 7. 2 - DESIGN FEATURES
' Two design conditions provide the primary safety features of the facility. One feature is the design of the reactor pool and shield system. The other feature is an air confinement system. These facility design features supplement safety features of the fuel element design, reactor core assernbly and fuel storage systems. 7-1
SAR 5/91 7.2.1. Reactor Pool and Shield Structure pool water system and shield structure design combine to control the effective radiation levels f roin the operation of the reactor. One goal of the design is a radiological exposure constraint of 1 rnrein/ hour for accessible areas of the pool and shield systein. Dose levels asstume a f ull power operation level of 1.500 inegawatts (thermal). Radiation doses above the pool and at specific penetrat. ions into or through the shield may exceed the design goal. Figure 7 1 displays the basic design dimensions of the pool and s.hield system with some of it's features. Representativo 1 terem/ hour dose curves for a reference case design are shown in Figure 7 2. The reference case design is a solid structure without any systein penetrations. Design of the reactor pool was of 1/2 inch (1.27 cm) base plate and 1/4 inch (9.64 crn) wall plate of 6061 altunintun alloy. Tank assembly is by shop f abrication. A protectivo layer of epoxy paint and bit.umen coal tar pitch with paper provide < barrier between the alurninurn pool tank and the reactor shield concrete. A four foot (1.22 meter) thick foundation pad supports the reactor pool and shield structure. Standard weight concrete, 150 lb/fts (2.33 g/cin8), coinprises the foundation pad. !!!E h density concrete, 180 lb/ft) ' (2.89 g/cin3), with a magnetite aggregate is the shield raaterial of the firat level of the shield structure. A transition from high density to standard density concrete is present about 4.5 feet (1.4 in) above the inid level platforra of the shield. The top part of the shield st ein and the top level platform are standard density concrete. The total shield weight is 2.03 x 106 lbs. (920 metric tons). Approximately 24,400 lbs. (11,100 kgs.) of structural steel, 56 conduits for sl 6nal and electrical lines with diameters of 1/2 ' inch to 3 inch, three central junction boxes and ntunetous local junction boxes are part of the shield systein. Five beam tubes at the level of the reactor provide experimental access to- reactor neutron and - gamma radiations. Two of the tubes combine to penetrate the coruplete reactor pool and shield structure from one side to the other side. Special design features of the beam tubes are_ beam plugs, sliding lead shutters, bolted cover plates, and gasket seal for protection against reactor radiation and coolant leakage when the tubes are not in use. 7.2.2. histor Bay Ventilation Desirn Ventilation design is specified to control air confincinent and to isolate the reactor bay in the event a - radioactive release is detected in the reactor areas.- The ventilation system-is-designed to maintain a negative pressure within the reactor bay with respect to the building exterior and other building areas. Confinement and isolation is achieved by air control dampers and leakage prevention tnaterial at doors
- and other room penetration points. A separate sys t ein is designed to exhaust air from several locations within the reactor bay that could contain airborne radionuclides such as . argon 41. Manual operation ot start /stop controls of both main and purge air systems will be available in the reactor control roorn, 7r2 m __ _._ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ - _ _ . . _ __ _. ,. _ -_. _ _ _ -_ _
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SAR 5/91 i Ventilation of the reactor bay is provided by two modes of system operation. One mode is for standard operation with recirculation of air. The other mode is an exhaust operation with high volume flow that has no air recirculation. Design during exhaust Total mode operation is a volume for the rate of air exchange in excess of two pet hour. room exceeds 4120 cubic rne t e r s . Normal operation of the ventilation system uses a roof stack for the exhaust of air frotn the reactor bay. Air f iltrat ion in the vent ilation syst ern is to be of typical design for normal HVAC operation with no special provisions. Schematics of the ventilation system for the reactor bay area and a logic diagram of the ventilation control system sensots and controls are provided in Figure 1 3. Control of air confinement within the reactor bay is provided tiy pressure control between the reactor bay and a differential representative ambient external m" surement point. Additional measurement points in ventilation rones adjacent to the reactor ba) reaintain the ditferential pressure between the reactor bay and adjacent access areas. The differential pressure control is intended to function in both standard and exhaust operation modes of the ventilation system. Isolation of the reactor bay is provided by ventilation dampers. These dampers will shut in response to either manual or automatic signal actuation. An automatic signal will initiato shutdown of the vent ilation system by closure of the dampers if a set point for airborne particulate radioactivity exceeds a setpoint. Protective switches within the ventilation system will cause the air fans to respond to the position change of the damperc. Damper design is for fail-safe operation so that loss of control power will isolate the reactor hay. Dampers locations are in the vicinity of the duct penetrations into the reactor bay. An isolation damper is in each of two supply air ducts. One return air duct with two sections contains two isolation dampers, one in each section. A pair of return air ducts also contain a dampet in each duct. The separate air purge system is designed to exhaust air that un> contain radionuclide products by a low volume system. The primary nuclide of interest is argon-41. Figure 7 4 shows a schematic of the argon purge system and its control logic. Air from potential sources of neutron activation such as beam tubes, sample transfer systems, and releases from exchanges at the pool water surface are subject to ! confinement and isolation by tho system. Filtration of air in the system will iw.lude prefilter and high ef ficiency particulate filter. Design provisions allow for the addition of charcoal filters Sample if experiment conditions should require the additional protection. parts in the turbulent flow stream of the purge system exhaust provide for measurement of exhaus'. activities. . .c t ua t i o n of the isolation damper in the argon pury system is by manual operation of the fan control switch. A schematic of each ventilation system is shown in Figure 7-5. 7-5
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l SAR $/91 3 7.2.3. Fuel Materials Design of reactor fuel elements and reactor core assembly are specified for conditions that exceed the normal operation of the system, t Special provisions are necessary for the storage of fuel elements that are not in the core assembly. The design of fuel storage systems requires consideration of the geometry, cooling, shielding, and the ability to account for each of the fuel elements. These storage systems will be racks for use both inside the reactor pool and outside the reactor shield. Inside the reactor pool several racks will deterinine the total capacity of the storage. The racks are aluminum, suspended from the pool edge by connecting rods, Elements are stored six per rack in a ' linear array. Each rack is 24 inches long (60.96 cm.) by 12 inches wide (30.48 cm.) by 3 inches (7.62 cm.) deep and is generally located below more than 8 feet of water shleiding. To facilitate extra storage, 2 racks may be attached to the same connecting rods by locating one rack at a different vertical level and offsetting the horizontal position slightly. If a storage requirement of 80% of the core grid capacity is specified, then 16 racks with a total of 96 positions would be necessary. Outside the reactor pool, rack design is intended to fit in special storage wells. The wells are pits in the reactor floor that are [ fabricated of 10 inches (25.4 cm.) diameter stainless steel pipe. Six wells are designed each 15 feet (4.57 m.) deep and located 3 feet (0.91 m.) f rom the adjacent well. Each storage well pit will contain a shield plug and features to control access and to circulate water. Nineteen elements may be stored in each well and water added for shielding or cooling. Spaces in the rack will provide a storage array for the fuel equivalent tc, the innermost 3 rings of the reactor core. This includes the placement of one element in the center. Most. routine fuel storage is -- intended to be inside the reactor pool. Outside the reactor pool, supplemental fuel storage is planned for temporary storage of elements transferred to or from the facility,- for isolation of fuel elements with clad damage, emergency storage of elements'from the reactor pool and core assembly and routine storage of other radioactive materialt Temporary storage for some reactor components or experiments may also use the fuel storago racks' in the reactor pool. Other locations not in the pool will also provide storage ( for radioactive non fuel materials. 7.2.4. Safety Feature Evaluation Design and features of the reactor pool and shield structure systems, air' confinement system, and fuel storage systems are similar to those . of other TRICA installations. Operation at there installations demonstrates the performance of the system, Evaluation of the performance-indicates that the basic and typical design characteristics O provide safe and reliable performance. 7 10
SAR $/91 AUXILIARY SYSTEMS 7,3 Several systems that supplernent the purpose of reactor safety are standard applications of the applicable construction codes or typical of university specifications. In a few cases, specifications particular to this installation or its systems will be applicable. A few of the systems of importance will be those for fire protection, user comfort, communications, and utility services. These systems are building systems and not specific to the reactor bay. 7.3.1. 1.ife safety and Fire Protection Design specifications are to meet life-safety requirceents appropriate for the conditions. These specifications will set requirernents for emergency lighting, stairwells and railings, exit doors, and other building safety features. An emergency shower and eye wash are available in the hallway adjacent to laboratory areas. The goal of fire protection shall be to provide reasonable assurance that safety reinted systems perform as intended and that other defined loss criteria are met [1,2]. For the purpose of fire protection, loss criteria should include protection of safety related systems, prevention of radioactive releases, personnel protection, minimization of property damage, and maintenance of operation continuity. Three components shall be applied to the fire protection objective. The three components are passive and active fire protection, and fire prevention. Each of the three components of the fire protection program shall be applied to the design, operation and modiilcat.lon of the reactor facility and it's c ottponent s . Fire prevention is primarily a function of operation rather than design. 7.3.2. f{tglive Fire Protection Elements passive fire protection s,hould provide a potential for fire safety that does not require physical operation or personal response to achieve the intended function. Passive elements to be considered should include inherent design features, building physical layout, safety related systems layout, fire barriers, and construction or component materials. Other passive elements that may be considered for special conditions will be frangible walls for overpressure relief, curbs for containtment of hazardous liquids, and drainage for control of fire protection runoff water. Penetrations in fire barriers shall have fire resistant ratingr compatible with the purpose of the fire barrier. Safety-related systems shall incorporate passive fire protection features that provide protection necessary for the design functions of the system. Separation of redundant system components, if applicable, and protection of distribution systems should be examined. Materials of noncombustible or limited combustion properties should be used when proctical. l l 7-11
l l l SAR $/91 i l I
?
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Materials such as sealing materials, electrical insulation, st.ructural finishes, adhesives, and linings should be selected so as to a ininimize fire hazards. Material selection may consider characteristics ! " such as calorie content, ignition properties, flame retardation, and rate of heat release. Eternents of the passive fire protection include the structurni construction system and the architectural separation into two separate l buildings. Building structural materials are concrete cast in place for l foundation, support columns and roof. Steel beam, metal and concrete i deck comprise the reactor bay roof. A built up composition roof with fire barrier materials complete the roof system. Pre cast panels that , are cast at the construction site cover 75% of the external perirneter. Met al paneling covers the other 25% of the perimeter. Design and installation of systems and components are subject to the applicable building codes.- The common vall between the acadernic wing and the reactor bay wing _of the building is to be a fire barrier. Doors between these two building sections and other penetrations such as llVAC chases will conform to appitcabic codes. Although a few metal stud and plaster , board walls will be in the reactor bay wing, the typical vall system is of concrete block construction. Basic _ design features . of the reactor assembly, pool and shield systern, and the instrumentation, control - and safety system represent passive fire protection elements. These basic features are sufficient passive protection to protect safety related systems. 7.3.3. Active Fire Protection Elements i Active fire protection elements differ from passive elements in that active features require automatic operation, manual response, or
- personnel action for the intended function. Active elements to be considered r,hould - be automatic- fire detection, automatic fire suppression, fire information transmission, manual fire suppression and other manual-fire control or loss control measures.
Manual _ protection shall _ consist of sanual fire fighting actions : l and the systems- necessary to support those actions such as ! extinguishers, pumps, valves, hoses, and the inspection, maintenance and l testing of equiprnent to assure reliability and proper operation. . Other
- L manual actions that are elements of active fire protection shall be l utility control, personnel control, and evacuation. Preplanning shall g
be an additional- element applied to active protection by --training i facility' personnel and emergency personnel and the appropriate actions I in response of fire and the possibic hazards involved. 1 7 12
- , _. . , _ _ . _ . _ . _ . , ~ . . _ _ . _ . _ _ _ . __
1 l SAR $/91 l I 1
)
Automatic and manual protection systems in the building include ; several different type systems. In the academic wing of the building automatic protective actions are provided by a sprinkler systein with i heat sensitive discharge nozzles, detectors for heat and smoke, and ; ventilation systems dampers. The reactor bay wing will use detectors for smoke and heat for areas that exclude the reactor bay. The reactor bay vontilation systern will have smoke detectors that provite i a warning of problems within the reactor bay. Al thotigh not a strict safety 1 requirement, a gaseous extinguisher system may be f unctional to protect the reactor ins t rtueenta t ion , control and safety sys t ero. Manual protection equipment will include a c'ry stand pipe system in each building stairwell, portable extinguishers such as CO2 , halon and dry chernical will be in specific locations throughout the building. 7.3.4. HVAC System personnel comfort conditions in the building use air handling units that have cold and hot water coils for temperature control. Two ; separate ilVAC systems consisting of a total of three air handling units provide all the air ventilation and comfort. One unit with both cold and hot water ec11s is a single duct systern that serves the reactor bay. The other two unit s are the cold and hot deck components of a double duct system that conditions air in all building zones which exclude the reactor bay. - Both air handling syatems are adjacent to the reactor bay on the fourth level of the reactor bay wing. Water t erope ra ture s of the heating and cooling coils in the air handling units are controlled by both on site and off sito systemt. The heating system is an on site boiler unit with a design capacity set by local building (HVAC) requirements. The cooling system is an of f site treatment plant with design capacity set by overall research center requirements. Thermostats control zone or room terope ra ture s . An instrument air system will provide control air for ilVAC systems. Controls and air balancing of the two air handling systems are to provide user cornfort and pressure differentials between the reactor bay and adjacent zones, and between the adjacent gones and the academic wing of the building. Relative to ambient atmospheric pressure the design goals for the reactor bay, adj acent zones and academic wing of the building is a negative pressure difference. The di f ferential pressures are 0.06: 0.04: 0.03 inches (0.15:_0.10: 0.80 cms.) of water. Air balance values relate typical or standard values except in the case of the reactor bay, 7.3.5. Communications and Security A communication system of typical telephone equipment will provide basic' services between tho' building and othar off site points. Supplernental- features to this system, such as intercom - lines between terminals or points within the building and zone speakers for general announcements are to provide additional communication within the building. The telephone system is installed and maintained by the university. Connection of the main university telephone system is to standard commercial telephone network. 7 13
SAR $/91 , Telephones with intercom features are to be located at several locations in_ the building. Arnong these -locations will be the ' receptionist office, the reactor control rootn and the reactor bay. By use of the intercom feature, each of these locations will be able to access public address speakers in one of several building zones. A video camera sys t ern and a separate itit e rcorn sys tern inay supplement the building telecommunication network. These two sys terns are intended to contribute to safe operation by enhancement of vir.ual , and audio communication between the operator and an experinnenter. Each system should have a central station in the control room with other remote stations in experiment areas. 7.3.6. Corporessed Air and Delenized Nattr The compressed air system will provide air for general applications and the operation of the transient control rod sys t ern . Design _- of . the system will consist of a dual cornpressor with filter, dryer and regulator for supply air, One air supply line with valves, regulator and sur6e tank will supply air to the transient control rod , d r iv.2. The dolonized water system will provide water for laboratory applications and replacement water for the reactor pool sys t aro . A system with replaceable resin containers, pipes, valves, and
} conductivity indication performs treatment of the don.estic water supply.
At periodic intervals either batch quantitles or continuous replenishment of the pool water will be made from the deionized water system. 7.3.7. tit il i t i e s Design of systems for electric power, cold water, - hot water, chilling water, waste water, and sanitary sewer are sirnilar to other university installations.- Specifications for these systems are the responsibility-of physical plant organization for operation, maintenancc-
-and repairs. Other systems in the building include a natural Sas supply ,
2 and vacuurn systm. Computer networks may be available for experiment systems. 7 3.8. jhtgatd Liould Waste Drains from several sinks and the emergency shower w'.11 be part of a liquid hazardous waste sys t ern. The system includes drains, traps, stainless steel pipe, valves, vent, and three waste storage- tanks;
~
Design will allow isolation of each tank, a dou' ale valve to drain any tank and a samplo outlet line. Rupture of any tank will be contained by a sump that-is part of the floor design. To drain any tank or the sump requires a pump (portable) and a conner. tion to the sanitary drain. The ' design intent is to prevent accidental releases by elimination of the hazard waste to sanitary waste pathway. Only if analysis and authorization allows, would the connection and transfer be made. 7 14 _ . _ _ . ~ . ____.. _ _ . _ _ . _ . _ . . _ _ . - . _ _ _ . _ _ _ , _ . - - _ - . _ ,
SAR-5/91 7.4 CONplNEMENT DES 1CN EVAL.UATION Confinement evaluation depends on the quantity of aliborne radioactivity release possible frort the air and water that are in the region of the reactor during operation, Calculation, measurement, and experience of similar research reactors support the evaluation. Evaluation is limited to routine effluenp and should be supplemented for experiment conditions that present specific release problems. ) Analysis of fission product icleasen are treated in another chapter. ! The most signi ficant radiological ef f luent s of tho reactor are argon-41 l' and nitrogen 16. Measurement and experience of other facilities have shown that for a facility of this type the most significant routine radiological l contributions are caused by argon 41 generated by the exposure of air in [ experimental f aellitJes and by nitrogen-16 transported in the coolant i from the reactor core region. Argon 41, a noble gas, is contained for i doeny or eventually released to the atmosphere for decay. Nitrogen-16 a I- dissolved gas is contained in the coolant and is dissipated by radioactivo decay. 7.4.1. Itelease of Nit regen 16 and Argen-41 The nit rogen 16 produced through the (n.p) reaction with the oxygen in the water moleculo has a very short half life (7 see) to only a very small fraction _of that produced in the core will find its way to the pool surface. The principal radiological effect of the nitrogen 16 is as a contriinitor to the radiation level at the pool' surface. Calculations of the nitrogen 16 transported to the pool surface estimate radiation dose rates of between 16 to 400 mrem / hour with the reactor operating at 1000 kW.
- Arson 41 is produced in the reactor pool as a result of the (n,y) reaction with argon 40 dissolved in the pool _ water. Most of this argon.
41 remains in solut.lon'but some of it is transferred to the reactor room t air at the pool surface. Calculat ions of the argonreleased{ rom the [ pool surface estimate a concentration'in the room of 1,6 x 10' pCi/cc with-the reactor operating at 1000 kW. [ 7,4,1,1. Hitrogen 16 Activity in Reactor Ronn The cross section threshold for the _ oxygen 16 (n.p) nitrogen-16 ! reactions ir 9.4 McV; however, the - minimum energy of the incident ! neutrons must he about 10.2 MeV because of conter of mass corrections, ( Thi s _ . hi gh threshold limits _ the production of nitrogen 16 since only l about 0,11 of all fission neutrons have an energy tu excess 'of 10 MeV. i Moreover, a single hydrogen scat t ering event will reduce the energy of [ theso high energy neutrons to below the threshold. The offcctive cross- l section of oxygen 16 (n.p) nitrogen-16 reaction averagcd over the TRICA spectrum is 0.0?! millibarns, This value agrees wall with the value } obtaleed from integrating the effective cross section over the fission
- spectrum, e
7-15 ; _J
j SAR 5/91 I 3 l The concentration of nitrogen 16 atoms per em of water as it leaves the reactor core is 6 ven 1 by N1 og (y l N2- ( 1 - c'A2 ' ) , (2) l A2 where 3 N2 - nitrogen 16 atorns per em of water, dy - neutron ilux - 1.0 x 10 I3 n/cm 2 -sec, (0.6 15 MeV), 1 3 N1 - oxygen atoms per em of water - 3.3 x 1022 gje,3, f a1 - (n.p) cross sectton of oxygen - 2.1 x 10'29 cm 2 l (averaged over 0.6 - 15 MeV). A 2 - nit rogen 16 decay constant - 9.35 x 10'2 s. e c ' I , i f t - average' time of exposure in reactor. The average exposure time in the core (2.3 sec), was derived in l the discussion on-argon activity. Solving for Np in the egnation above, one obtains N2 - 7,41 x 107 (1 e 9.35 x 10 2 x 2.3) , 7 3
- 1.43 x 10 atoms /cm ,
as the density of nitrogen 16 in thb water leaving the core. If it is assumed that the water continues to flow at the same velocity to _the surface, a distance of -640 cm, the transit time fron: core to surface is trise - 640/16.8 - 38.1 sec , (3)' where the flow velocity. 16.8 cm/sec (Table 4-7), was eteen in the discussion on' heat transfer. This assumption is quite conservative: as energy losses from the fluid stream resulting from turbulent -mixing will reduce the velocity significantly. _ Furthermore, delays in t.rans i t time resulting from operation of the diffuser pump are siceable. Measurements made of the dose - rates at ' t.be pool surface of several TRIGA reactors show that. the
. operation of the diffuser-pump-reduces the_ nitrogen 16 contribution to the surface dose rate by an order of magnitude of more depending on the sir.e of the pool.
O l.16
-.r. ,*._. .<. . . - . . . _ - . . . _ . _ . _ _ _ . . .,i. .
SAR 5/91
@ In 38 seconds the nitrogen-16 decays to 2.86 x 10 ' ,' times the value of the activity leaving the core. Thus the concentration of nitrogen-16 atoms that reach the region near the surface of the pool is estimated at about 410,000 atoms /em) per cubic lentimeter.
Only a small proportion of the nitrogen-16 atoms present near the pool surface are transferred into the air of the reactor room. When a nit rol,en- 16 at om is formed, it appears as a re(oll atom with various degrees of ionization. For high-purity water (approximately 2 p n.ho ) practically all of the nitrogen-16 combines with oxygen and hydrogen 3 atoms of the water. Most of it combines in an anion form, which has a l tendency to rnmati, in the water [4}. It is assumed that at least one-half of all ions formed are anions lic c aus e of its 7.1-see half life, M{ the nitrogen-16 decays before reaching a uniform concentration in the tank water. The activity will be dispersed over the surface area of the y pool and rauch of it will decay daring the lateral movement. , For the purpose of the analysis it is postulated that the water- [ bearing nitrogen-16 rises f r<>m the core to the surface and then spreads across a disk source with an equivalent radius of 125 cm. For a ( constant velocity of 16.8 cm/see the cycle time for distributing the nitrogen 16 over the pool surface would be ts - 125 cm/16.8 cm/sec - 7.4 see The average concentration during t his time is
- "t s
- A t-N - 1/t 3 Ne o dt , (4)
*o N 4.10 x 10
(1 - e'Als) - (1 - e-0.69)/7.4 , (5) Ats 9.35 x 10-, 3 3
- 3.0 x 10 4._ oms /cm The interest from the point of safety is then the number of nitrogen-16 atoms escaping into the air from the diffusing source above the core. The number escaping to the air would be about is estimated from the iscape velocity, 0.009 cm/sec, from Dorsey [5] as (3.0 x 10 5atoms /cc) (0.9 x 10-2 cm/sec) 2 2700 atoms /cm -sec In the room, the activity is af fected by dilution, ventilation, and occay. Thus the rate of accumulation of nitrogen-16 in the room ar a whole is given by d(VNi6)
-S- (A4 g/V) VN 16 , (6) dt I _ __ _ __- __ _ _ __ _ _
7-17
. ;l
l l SAR S/91 l l l vhere S - number of nitrogen-16 atoms entering the room from the pool per second, 4 1.36 x 10 8atoms /sec, (2700) (5.05 x 10 ) - 9 V - volwne of the reactor room, 4.12 x 10 cm3 , (effective), q - volume flow rate, 2.29 x 10 6 c,3/sec, (reactor room exhaust). I For saturation conditions l S 1.36 x 10 0 VN16 , ,
,, r
, (7)
(AN + q/V) 9.35 x 10 ' '- + 2.8 x 10 9
- 1.4 x 10 nuclei.
This corresponds 3 an activity concentration of 8.8 x 10'7 pCi/cc. The gamma dose rate f rom nit rogen-16 of this concentration in the air is 3 3.7 x 10 4photons /sec-pci x 8.8 x 10'2 pC1 /cm x 1000 cm D-2 x 1.6 x 105 (photons /sec-cm2 / rad-br)
- 1.0 x 10"* rad /hr - 103 prad/hr ,
when the effective rc L ; of t.he room, taken to be a hemisphere with a volume of 4120 cubic meters 11 10 m. The thickness of the layer of nitrogen-16 bearing water is vity 8.0 x 10 3 x 7,4 h - - - 1.17 cm , (8) 3 0 A3 5.05 x 10 3 where the volume flow rate 8.0 x 10 cm /see 3 was given in the discussion on heat transfer. The dose rate at the pool surface arising irom the nitrogen-16 near the surface is AII D - - [1 - E (ph)] 2 , (9) 2pK where p - attenuatior coefficient for 6 MeV photons in watet (0.0275cm'g),
?
7-18
- - _ _ . . - _ m_ _ _ . ~ . _ - . _ - _ _ _ _ _ _ . - _ - . _ . . _______.
SAR $/91 K - 1.6 x 10 5photons /cm'-sec per rad /hr flux to dose
- tate conversion factor, E2- second exponential integral.
This yields, approximately, D - 400 rnr/hr . 4 This value is larger than those extrapolated from measurements inade on other TRIGA reactors. Transport times from the reactor core to the pool surface .in excess of those estimated will lower the calculated dose substantially. A delay time twice as long as 38 sec. will generate a calculated dose rate twenty-five times less, 7.4.1.2. Relense of Argon-41 from Reactor Pool Water 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 at which argon-41 escapes from the reactor pool water into the reactor room. The calculations show.that most of the argon-41 decays while in the water. The changes in argon-41 concentration in the core region, in the pool water external to the reactor, and in the air of the reactor room, are
,/ calculated.using the vattables as follows:
i N atomic density (atoms /cm3 ) A decay constant (sec*1), 1.06 x 10'4 o -absorption cross section (cm 2), 0.61 b q volume flow rate, reactor room exhaust (cm3/sec) V volume of region (cm 3) p- density (gm/cm3 ) 2 Ag channel free-flow area (cm ) te channel length (cm) w- mass flow rate (gm/sec) 3
-v- volume flow rate through the core (cm /sec)
P 2 l: h average thermal neutron flux in the core (n/cm ,3gc) l l-l- 7-19
SAR 5/91 The volume flow rate through the core is
. w 8000 g/sec v - - - - 8.0 x 10 3 c,373pc ,
p 1 g/cin 3 From the flow channel volume, Aff c, the exposure time in the core is
. . 485.0 x 38.1 t - V/v - Art /c v - - 2.3 see it remains to find the atom density N for dissolved argon 40 in the reactor pool water.
Dissolved gases in the reactor coolant contain the radioactive noble gas argon-41. The release of argon-41 activity f rom the coolant depends on the gaseous exchange rate at the air-water interface and the change in gas solubility as a function of temperature. Accorning to lle nry 's law for gases in contact with liquids the equilibrium concentration in the liquid is proportional to the partial pressure of the gas, The saturated concentration of argon in water at one atmosphere of st.andard air is given in Table 7-1. Concentrations at equilibrium conditions for argon in air that is in contact with the water depends on the air volume, air exchange rate, water surface volume and the water-air exchange rate. Tabic 7-1 SATURATED ARGON CONCENTRATION IN k'ATER (2) Temperature 3 ('C) S(atoms Ar-40)/cm 112 0 10 1.14 x 10 16 16 i' 20 0.94 x 10 l- 30 0.79 x 10 16 40 0.69 x 10 16 l f 50 0.62 x 10 16 60 0.56 x 10 16 70 0.52 x 10 16 80 -0.48 x 10 16 [ a 7-20
. - . . . . . . . - - - - - . . . . ~. . - - - . - . - . . - _ - - - ~ . - . . - .
l SAR 5/91 l
@ Argon 41 activities at conditions of equilibrium concentration are functions of the pool water volume , reactor core volume, water flow rates through the core, and the production and decay - rates for the radioisotope.
The argon-40 concentration in the water at the core inlet temperature (38'C) is 40 15 3 N1
- 7.1 x 10 atoms /cm ,
and the concentration of argon-40 in the water at the core exit temperature (68'C) is 40 - 5.3 x 10 15 atoms /cm 3 N1 , From the reactor power the cycle times for argon exposed in the core and circulated in the pool are estimated. The volume flow rate is given by V - Q/(Pw Cp AT) , (10) where: v - Volume flow rate of water through the core.
@ Q - Reactor power level (10 6 y,gt,),
3 Py - Pool water density (0.96 gm/cm ), Cp - Specific heat of water (4.2 watt sec/gm *C). AT - Temperature rise in core (30*C). Thus: 3 3 v - 106 /(0.96 x 4.2 x 30) - 8.3 x 10 cm /sec. The exposure time, t, and cycle time, T, in the core are calculated from t - Ve/v - Ar Le/v , (11) and T-V p / v - [ nh(w/2)2 + h(1-w)w J / v . . (12)
- whete
2 Ag - flow channel area (485 cm ), Lc - length of flow channel (38.1 cm), Ve - volume of water in the core, 7-21 .
,e ----,--------,-e~we - n.,-w ,n.
SAR 5/91 w - width of reactor pool (200 cm), 1 - length of reactor pool (300 cm), h - height of pool water (750 cm). Exposure time, t, is about 2.2 seconds and cycle time, T, is about 4.6 x 103 seconds. Argon atoms exchanged at the water-air interface depend on a water thickness depth that is small relative to the pool dimensions and, therefore, a small fraction of the available saturated argon is exchanged with the air. During the time required for the pool water to circulate once through the reactor core, about one hour and twenty minutes, the argon equilibrium concentration should deplete to the lowest solubility value for equilibrium concent rat ion. The argon release as a function of temperature and solubility thus approaches zero. This depletion occurs as the activity of the argon radioisotope increases but is substantially complete as the argon-41 activity reaches half the equilibrium value at about 110 minutes. Evaluation of the water air interface exchange rate for argon is related to an air and water thickness depth that depends on the argon atom diffusion coefficient. The total exchange rate then is a function of the pool surface area, A s , and an effective release volume Vt ' The two terms are related by
#1 A s -
f -+j Vi' i ' (13) where
#1 is a surface exchange coefficient (cm/sec), and fi ,j is the fraction of atoms exchanged from volume i to j (sec' ).
Estimates of the surface exchange coefficient (i.e., the gas in a unit volume that is exchanged at the surface per unit time per unit surface area) for argon very considerably. One method of arriving at a value for this parameter is through the diffusion coefficient of the gas in water. The mean square distance traversed by a molecule is
< AX >2 -
2Dt (14) where D - diffusion coefficient (cm'/sec), t - time (sec). The exchange coefficient is assumed to be evaluated for 1 see as ( < AX >2 )l\2 /t ( 2D/t )h2 7-22
SAR S/91 The diffusion coefficient at 40'C is about 1.1 x IO'S cm2 j3,,, and, if_one assumes that only one-half of the argon atoms within one diffusion length of the surface escape,
'~p_- 1\2 (2 x 1.1 x 10 5)1\2 -
2.35 x 10-3 cm/sec Values for the surface er hange coefficient have been reported by Dorsey [3) for air, 02 , and N2 The values for these three gases are all about' equal. Assuming argon c ehaves as do these gases, a value is obtained of
- 5.7 x 10*3 cm/sec.
Measurements have been made of the argon 41 activity in a TRIGA Mark III reactor pool and from the data acquired from these measurements it was possible to construct. a value for Li surface exchange coefficient. This value at 40*C is about 2.9 x 10*fecm/sec. During equilibrium conditions and assuming no difference in the rates of escape fractions for Ar-40 and Ar-41, the number or 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., i fjV'Ni i - fj41 Vj' NJ , (15) where Nj - 2.1 x 10 I7 argon atoms /cm 3 of air ~N 40 , 15 3 /0 4 Ng - 7.1 x 10 argon atoms /cm of water -N . Solving for fj t V' gives fj i Vj' fj-.j Vj ' (N /Nj) 1 (16) The following calculations were performed to evaluate the rate-of Ar 41 escaping from the reactor pool water into the room enclosure. The calculations show that the Ar-41 decays while in the water, and most oi the radiation is safely absorbed in the water. The changes in Ar 41-concentration in the reactor, in the pool' water external to the reactor, and in the air of the room enclosure are given by 41 V1 -Vi&N1 40 ,40 , p 41 (y 4 yl 4 ,41 , 341 yg 4 g y1 dt (17) 41 dN2
- -A 41 N2 41 V2 V2 + v1 '* 1 ( N I *1 -N2 01) dt 2 2' V'2 (f 3 N' -f 342 N'3'l V3 ') (18)
V3 - (f243 N2 0I V' 2 - fM3 N 01 V3 ') -N3 41 0 41 V3+9) dt (19) 7-23
SAR 5/91.
@ where subscript 1 - Reactor region (water region in core) subscript 2 - Reactor tank water region external *.o the reactor subscript 3'- Room enclosure region superscript 40 - Ar-40 superscript 41 - Ar-41 i
3 V - Volume of region (ctn ) N - Atornic density (atoms /cm3) A - Decay constant (sec'I) o - Absorption cross section (cm )2 q - Volume flow rate from room enclosure exhaust (cm3 /s) vi- Volume flow rate throuch region No.1 (cm3 /s) 4 - Average thermal neutron flux in region No.1 (n/cm2 x s) f i.,j V i' - Fraction of Ar.41 atoms in 3region i that escape to region j per unit time (cm /sec) The values of constants in equations (17), (18), and (19) are 0 3 J ~ 1.2 x 1013n/cm 2 cm
-s. ,
V1 - 1.85 x 10 , 7 cm 3 a 40-- 0.47 x 10'2' cm2 , V2 - 4.00 x 10 , 2 9 3 o 41 - 0.060 x 10-24 cm , V3 - 4.12 x 10 cm , A'1 - 1.06 x 10-4 sec'l , p - 2.35 x 10*3 cm/sec ,
.q - 2.29 x 106 c,3/s ,
N1 40 - 7.1 x 10-15 fjfc,3 , i 3 3 vt - 8.13 x 10 cm /s , Equation (1) can be reduced to 40 c40 U (20) V1 -V14N1 - (NI - N2 4l) V1 dt
-vi + Vt4a41 + A41 yt , y1 7-24 l
~,,...,..-.. - , . - , - - - , . . - . . - - . _ _ - , . - - _ - .
SAR 5/91 During equilibrium conditions the three equations reduce to: V14N1 40 ,40 - (Ni 'I -N24l)V1
* (21) p 01 V2+f3 V2 '] - (N1 41 -N2 01) v1 4 f342 N3#1 V3 ',(22) 01 N2 2 N3 01 p 01 V3 4 q+f2 3 V3 ') - f243 N2 l V2 (23)
Combining equations (21) and (22) gives Vi pNi ## o f23 N ^l 3 V3
- t . (24)
N2 A 0I V2 i f243 V'2 A 41 V2+f243 V' 2 0 which inserted into equation (23) for N2 yields (25) A 01 V3+q+f342_ V'3 f 342 V' 3 V 1(N"0 0 t N3 ( f243 V'7 A 01 V2+fN V'. 2 ^41 V2 4 fN V'2 Solving for N3 N yields - 7.4 atoms /cm3 . This corresponds to a concentration of Ar 41 activity of-0l 0I 1.06 x 10'0 x 7.4 A N3 2,12 x 10-8 pCi/cm 3 p A 01 - , C 3.7 x 10,, g where A01 - Ar 41 concentration, pC1/cm 3 , C - Conversion factor from disintegrations to pCi.
-Thisisbelowthereferencelgvellim{trecommendedby10CFRpart 20 for unrestricted areas (4 x 10' pCi/cm ) and takes into account no dilution, which is conservative. -
7.4.1.3. Activation of Air in the Experimental Facilitieg In the TRIGA reactor installation, the following experimental facilities contain air: rotary specimen rack, pneumatic transfer tube,
- and neutron beam - tubes. Of the radioisotopes produced in these air -cavities, argon-41 (with half life of 110 min.) is the most significant with respect to airborne radioactivity hazards. Nitrogen 16 (7.11 sec.
half-life) and oxygen-19 (26.9 sec. half-life) are considerably less significant. Estimated releases of argon-41 from reactor operations indicate an upper limit for the release exposure as 190 prad/hr. . Actual values are expected to be less than 1/50 of the estimated value. The saturated activity of argon-41 in an experimental cavity is f calculated from A - N01A 41 - pCi/cm 3 , (26) 01 4q/V) C(A 7-25
SAR 5/91 where C - 3.7 x 100 disintegrations /sec per 9 01 S - (Ean/cm 3-sec. Ia - 1.59 x 10'7 cm'I The effective air volumes of several experimental facilities are listed in Table 7-2. Also given are conservative estimates of the average thermal neutron fluxes for 1100 kV operation. Table 7-2 VOLUMES AND THERMAL. FLUXES OF FACII.ITIES Effective Average Thermal Air Volume Flux at 1100 kW Region (cm3) (n/cm2 .sec) x 1012 Central thimble 4. 3 x 10 2 23.8 0 7.2 Rotary specimen rack 3.3 x 10 3 Pneumatic tube 1.6 x 10 7.2 165 Thru beam port 2. 7 x 10 5 0.1 0 2 Tangential beam port 7.2 x 10 0.1 3 Radial beam port 1.8 x 10 5 0.1 4 Radial beam port 7.0 x 10 4 0.1 For volume exhaust rates where the decay constant is negtigible, such that A41 << q/V , (27)
.the activity released from each volume is given by A
41 .3 41 (4Ea li V i /C pC1/sec (28) l Total volume of all air cavities without any experiments in place l l is about 0.6 cubic meters. A flow rate of 6.4 x 102 cm3/sec is neccessary to neglect the decay. Actual flow rates are about 3,15 cm3/sec ( 4 cfh ) for each beam port. At this rate the fraction of air removal from the cavities is only 9 percent per hour. The total activity calculated for the air leaving the beam port experiment facilities is therefore, 6 pCi/sec. I It should be emphasized that the air activation and subsequent release activity are predicted for vacant beam ports and conservativt neutron fluxes. Actual release rates depend on the particule-configuration of experiments and the air exchange rate in each facility. ' 7-26
- - - . ..- -. - - - ~ . . . .- . . ~ _ _ . _ _ . . . -. -.~ . . _ _ . - . - SAR 5/91 7.4.2. Evaluation of Arnon 41 Release The release of argon-41 from the facility is diluted by the ventilation exhaust rate, assumed to represent two air changes per hour, and averaged for a 5 day, 8 hour operation schedule at full power. The release concentration from the pool averaged for one year is,
.24-(1.65 x 10-6) - 3.96 x 10*7 pCi/cm 3 Only 20% of the experiment facility argon-41 is assumed to exhaust since experiments will replace some or most of the exposed air, 6 (.20) (.24)/2.29 x 10 6- 1,3 x 10'7 pC1/cm 3 ,
Total estimated release is 4.1 x 10~7 pCi/cm 3 The whole body gamma ray dose rate to a person immersed in a semi-infinite cloud of radioactive gases can be approximated by D - 900 EAp (29) where-E - the photon energy, 1.3 Mov 3 AD - ef fective exposure concentration, Ci/m
.f* The concentration downwind from the point at which the activity is discharged from the building is Ap - Aq p(x), (30) 3 where $ . - the dilution factor at the distance x, (sec/m ),
A - activity concentration in the discharge (Ci/m 3), q - the building exhaust rates (m 3/sec). If it is assumed that the discharge is at the roof line, the dilution' factor in the lee of the building (x - 0), is given (6) by:
$(0) - 1/csu _, (31) where c - a constant (0.5),
s-buildingcrogs-sectionalareanormaltothewind direction (m ), u - wind velocity (m/sec). 2 A minimum cross-sectional area is assumed of-234 m (60 x 42 ft) and, for a wind velocity of 4 m/sec. p(0) - 1/(0.5 x 4. x 234) - 2.I x 10'3 sec/m 3 (32) 7-27
SAR 5/91 The averaged dose rate at the exhaust stack is D - 900 x 1.3 (4.1 x 104) - 4.8 x 10'0 rads /hr, an average of 0.48 mrad /hr in the stack or D - 4.8 x 10'0 ( 2.3 x 2.1 x 10'3) - 2.3 x 10'Crads/hr, an average of 2.3 prad/hr at ground level. At the limiting exhaust rate, 640 cm3/sec, to ignore the argon-41 decay, the source term is 27 pCi/see for the beam ports which would increase the dose rate to 10 prad/hr. 11 core experiment facilities, such as the center tube and rotary specimen rack, are vented at the same exhaust rate the releases would increase by 102 pCi/sec. However, these exhaust rate conditions represent limiting conditions, not actual release rates. For the exhaust manifold rate of 3.15 cm3/sce thc celease rate would be 50 pCl/sec for the two core experiment facilities. The rotary specimen rack is the primary source of the activity. Venting of experiment facilities, especially the rotary specimen rack, will require monitoring of the release rate or replacement of the air with gases such as nitrogen or carbon dioxide to control release concentrations of air activity. However, normal operating conditions do nct vent the rotary specimen rack, although this is an optional operating condition. The pneuro t i c transfer system, by c omp a r i son ., routinely contains nitrogen or carbon dioxide gas to limit the releases within the room that contains the access terminal. Actual dose values for the argon-41 release may vary. The beam port release estimate is less than 4.8 prad/hr which is equivalent to 10 mrad /yr. Lower neutron fluxes, smaller air volumes, shorter operation times and larger dilution factors will assure that releases do not exceed annual release constraints. Monitoring ti.e exhaust will verify that other release points such as the core experiment facilities do not cause the total-to exceed preset limits. i 7-28 l
- m. - - _ _ . _ _ _ - - _ _ _ _ . . - _ _ _ _ . _ _ . ,
SAR 5/91 L ). Chapter 7 References-
- 4. -
- 1. Code of Federal Regulations, Chapter 10 part 20, U.S. Gove rrunent Printing Office, 1982.
i. ! 2. Dorsey, N.E., Properties of Ordinary Water-Substance, Reinhold Publishing Corp., New York p. 537. I J 3. Ibid., p. 554 i
- 4. Mitt 1, R.L., and M.H. Theys, "N-16 Concentration in EBWR,"
! Nucleonics p. 81 (1961), t- - 5. Dorsey, N.E., on cit., p. 554. i=
- 6. Slade, D H., (ed.), " Meteorology and Ate:nic Energy," USAEC Reactor l
Develop, and Tech. Div. Report TID-24190, DFSTI, Springfield, l Virginia, 1968. l i i ? 9 l j -. l
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SAR 5/91 i s Chapter 8 EXPERIMENT AND IRRADIATION FACILITIES The experimental and irradiation facilities of the TRICA Hark 11 reactor are extensive and_ versatile. Physical access and observation of the core are possible at all times through the vertical water shield. Experimental tubes can easily be installed in the core region to provide facilities for high-level irradiations or in core experiments. Areas outside the core and reflector are available for larger experiment. equipment or facilities. 8.1. STANDARD EXPERIMENT 5'ACILITIES 8.1.1. Central Thimble The reactor is equipped with a central thimble for access to the point.of maximum flux in the core. The central thimble consists of'an aluminum tube that fits through the center hole of the top and bottom grid plates. Dimensions of the tube are 1.5 in, o.d. (3. 81 - cm. ) and 1.33 in,_i.d. (3.38 cm.). Holes in the tube assure that it is normally filled with water. Water is expelled from the tube by compressed air. Experiments with the central thimble include irradiations of small samples and the exposure of materials to a collimated beam of neutrons f or gamma rays. 8,1.2, Rotary Specimen Rack
-A rotary, multiple-position (40) specimen rack located in a well in the _ top of the graphite reflector provides for the large scale production of - radioisotopes and for the activation and irradiation of multiple samples. All positions in this rack- are exposed to neutron fluxes of comparable intensity. Specimen positions are 1.23_in. (3,18 cm.) in diameter by 10.80 in. (27.4 cm. ) in depth. Samples are loaded from the . cop of the reactor through a water-tight tube into the rotary rack using a specimen lifting device or pneumatic pressure for insertion and removal of samples from the sample rack positions. The rotary specimen _ rack can - be turned from the ~ top of the reactor by manual operation or by a motor drive. Figure 8-1 shows the rack.
8,1.3. Pneumatic Specimen Tube A pneumatic transfer system permits applications with short lived radioisotopes. The in-core terminus of this system is normally located in the outer ring of fuel element positions, a region of high neutron flux. -The sample capsule (rabbit) is conveyed to a receiver sender station via-1.25 in. o.d. (3.18 cm.) aluminum tubing. Effective space in=the specimen transfer capsules is 0.68 in. (1.7 cm.) diameter by 4,5 in. (11.4 cm. ) height. An optional - transfer box may be employed to permit the sample to be sent and received from up to three different rocciver sender stations. A schematic of the pneumatic irradiation terminal and air flow control valves is shown in Figure 8-2. 8-1
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SAR 5/91 8.1,4. Beam Tube Facilities Access to horizontal neutron beams is created by five beam tubes penetrating the reactor shield structure. All beam tubes are 6 inch diameter tubes originating at or in the reactor reflector. One tangential beam tube is composed of a penetration in the reactor reflector assembly with extensions through both sides _of the reactor shield. A second tangential beam tube penetrates and terminates in the reactor reficctor. The two remaining tubes are oriented radial to the reactor core. _The beam ports (Figure 8-3) provide tubular penetrations through the concrete shield and reactor tank water, making beams of neutrons (or gamma radiation) available for experiments. The beam ports also provide an irradiation facility for large sample specimens in a region close to the core. Beam port diameters near the core are 6 in. (15.2 cm.). There are five beam ports divided into two categories as follows: 8.1.4.1. Tanzential beam ports Two beam ports are oriented tangential to the reactor core, penetrate the graphite re flec t or , the coolant water, and the concrete shield. A hole is drilled in the graphite tangential to the outer edge of the core. One beam port terminates at the tangential point to the core. The other beam tubes extend both directions from the reflector and out opposite sides of the reactor shield. 8.1,4.2. Radial beam port s There are two radial beam port s, each which penetrate the concrete I shield structure and the coolant water. One radial port terminates at the-outer edge of the reflector. The second radial port also terminates
-at the outer edge of the re flec to r. However, a hole drilled in the graphite reflector extends the effective source of the radiations to the reactor core-region.
8.1.4.3. Beam Tube Plugs A step is. incorporated into each beam port to prevent radiation streaming through the gap between the beam tube and shielding plug. The inner section of each beam port is an aluminum pipe 6 in. (15.2 cm.) in diameter. The outer section of beam ports 1. 2 and 4 is a steel pipe 8 in. (20.3 cm.) in diameter. Beam ports 3 and 5 have three outer sections with 8 in., 12 - in. , and 15.25 in, diameters. A . lead shield ring in the shield structure provides a " shadow" shield for the 15.25 in, beam port section. Special shielding reduces the radiation outside the concrete to a safe level when the beam port is not in use. The shielding is provided in four sections as follows:
- a. An inner shield plug,
- b. An outer shield plug.
- c. A lead-filled shutter, and door.
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SAR 5/91 The inner shield plug consists of graphito cylinder, backed with a 0.125 in, (0.32-cm) sheet of boral and 5 in. (12.7 cm) of lead, sandwiched between two 1.25 in. (3.2 cm) thick steel plates. Beam ports 1, 2, and 4 have a section of graphite 6 in. (15,2 cm) in diameter. Beam ports 3 and 5 have the same configuration as the other beam ports, except that the graphite portion is 6 in. (15,2 cm) in diameter, with a change to 8 in, (20.3 cm) in diameter to provide graphite shielding in the 6 in. and 8 in, portions of the tube. Two rollers are provided to facilitate the insertion and removal of the inner shield plugs. To help guide the shield plug over the steps in the beam tube during insertion, the inner end of the plug is cone-shaped. A threaded hole is provided in the outer end of the plug for attaching the beam tube plug. handling tool. The graphite sections are encased in an aluminum cannister. The outer shield plug is wooden and is 8 in. (20,3 cm) in diameter and 42 in, (1,07 m) long for beam ports 1, 2, and 4. Beam ports 3 and 5 have a wooden shield plug for the outer portion of the tube that has a length 'of 48. in, (1.22 m) and diameter of 15 in (38,1 cm) for the outer' portion of the tube. A handle on the-outer end of this plug is provided for manual handling. The plug is equipped with an electrical circuit consisting of a position switch mounted in the front of the plug and an electrical connector at the rear of the plug. The switch can be actuated only by the inner plug when the inner plug is installed in the beam tube. A physical contact between the inner and outer shield plug, and an electrical connection between the outer plug and the beam tube are part of .an installation status circuit. The circuit monitors the plug configuration or other experiment shield conditions, Information on the console for each beam tube indicates t.he plug or beam tube status, The lead-filled shutter and lead-lined door provide limited gamma shieldine when the plugs are removed. The shutter is contained in a rectangut . steel housing recessed in the outer surface of t.he concrete shield, The shutter is ~10 in. (25.4 cm) in diameter and_9.5 in. (24,1 cm) thick for beam ports 1, 2, and 4, Beam ports 3 and 5 have a shutter that is 15.25 in. (38.7 cm) in diameter and 9.5 in. (24,1 cm) thick. The shutter is operated by a removeable push rod on the face of the shield structure and can be moved _ even with the shutter housing door closed. In the open - position, a section of the shutter consisting of pipe _of equal diameter to the outer portion of the beam tube is aligned with the beam port and the outer shield center plug to facilitate insertion or removal of the beam plugs; The shutter housing is equipped with _ a steel cover _ plate lined with 1.25 in. (3.2 cm) of lead for additional shielding. A removeable cover plate provides easy access to the beam port. The plate can.be bolted shut so that the seal would prevent loss of shielding water'if the beam tube should develop a serious leak. F 8-6 _ ._ _ .,_ u._. __ . ., ,_ _ . . - - . - _ _ . _ . . - . _ _ , . . - . _ . _ - - , , _ _ - - . _
SAR 5/91 8.1,5. Evaluation of Materials in Experiment Fac il i t it.R The following information is a guide for evaluating experiment materials in order to prevent the introduction of materials that could damage the reactor or its components. A carciul evaluation [1] of proposed experiment materials shall be performed to classify the experiment as an approved experiment, culdelines for the following types of experiment materials are provided: materials which require double encapsulation, explosive materials and their confinement, fueled experiments, and materials which could be sources of atrhorne radioactivity. The limits referenced in this section are technical specJication requirements to prevent the occurrence of a serious safety hazard. 8.1.5.1. Double Encapsulation Experiments containing materials corrosive to reactor components,
. compounds highly reactive with water, potentially explosive materials, and liquid fissionable materials shall be doubly encapsulated. Chemical Hazard Information in the allandbook of Laboratory Salety," [2] shall be used as a guide for classifying the first three categories of materials.
A table of_ isotopes shall be. used as a guide for the last category. The Chemical llazard Information lists several categories of information about hazardous materials. One category of significant interest has the title " Relative llaza rd to llealth from Concentrated Short-Term - Exposure." This category identifies materials that are highly corrosive. From a conservative standpoint, those materials which are corrosive to human t. i s s ue shall be considered to be corrosive for the-purposes of double encapsulation. Re fe rence s also exist to other toxic or hazardous material listings for details about a particular material. Two categories determine the explosive and flammability properties of materials. One titled "N. F. P. A. Hazard Identification Signals" and another titled-- " Flammable Limits - in Air " specify materials that can represent safety hazards. The table identifies degree of explosive potential, chemical reactivity and flammability limits. Material ratings for "N.F.P.A. Ilazard Identification" consist of a scale of 0 -(low) to 4 (hi gh) , regarding their health, fire, and reactivity hazards while under fire conditions, In this table, fire ratings indicate the degree of susceptibility for a particular material to burn. Further, a reactivity rating indicates the degree of susceptibility for a particular material to release energy. The code rating scale for flammability specifies limits as a function of the air volume percent. 3-Materials having fire ratings greater than four or reactivity _ ratings greater than 2 should be doubly encapsulated. The encapsulation is to protect against the energy release and corrosion properties. Materials'. having . a reactivity rating of 1 or a flammability rating greater than 1 should be evaluated individually to determine if doubli encapsulation is warranted. 8-7
SAR 5/91 The primary hazard attributed to liquid fissionable experiment materials is leakage of the material from the experiment container. To minimize this risk, all experiment materials containing more than 100 ppm of thorium, uranium, or plutonium shall be doubly encapsulated. The type of encapsulation material should be compatible with the encapsulation contents. Although double encapsulation of hazardous experiment materials greatly reduces the likelihood of their release, it is possible for these materials to escape their experiment containers. If an experiment capsule fails and releases material which could damage the reactor fuel, structure, or systems, the capsule shall be promptly removed and inspected in order to determine the consequences and the need for corrective action,
~
8.1.5.2. Explosive Materials A 25 milligram explosive releases approximately 25 calories (104.2 joules) calories of energy with the creation of 25 cm3 of gas. For the explosive TNT, the density is 1.654 gm/cm3 so that 25 mg represents a volume of 0.015 cm). If the assumption is made t h a t. the energy release occurs as an instantaneous change in pressure, the total force on the encapsulation material is the sum of two pressures. For a one cm3 volume the energy release of 104.2 joules represents a pressure of 1032 atmospheres. The instantaneous change in pressure due to gas production in the same volume adds another 25 atmospheres. The total pressure within a 1 cm3 capsule is then 1057 atmospheres for the complete reaction of 25 mg. of explosive. Typical construction materials of capsules are stainless steel, aluminum and polvethylene. Table 8-1 lists the mechanical properties of thase encapsulation materials. Table 8-1 Material Strengths Material Yield Ultimate Density Stainless Steel 35 ksi 85 ksi 7.98 g/cm 3 (type 304) (500 lb/ft3) Aluminum 40 ksi 45ksi 2.739 g/cm 3 (alloy 6061) (171 lb/ft3) Pc, lye thylene 1.7 ksi 1.4ksi 0.923 g/cm 3 i 8-8
SAR 5/91 b- Analysis of the encapsulation material determines the material stress limits that must exist to confine the reactive equivalent of 25 mg of explosive. The stress limit in a cylindrical container with thin walls is one half the pressure times the ratio of the capsule diameter to wall thickness, Pd Omax g (1) where
- maximum stress in container wall, omax p - total pressure within the container, d - diameter of the container, t - wall thickness.
When evaluating an encapsulation material's atility to confine the reactive equivalent of 25 mg of explosive, the maximum stress in the container wall is required to be less than or equal to the yield strength of the material: s pd
- $ "yleld , (2) 2t where ay tetg is the yield strength. Solving this equation fo. d/t provides an easy method of evaluating an encapsulation material:
d 2
" yield (3) t p Table 8-2 Container Diameter to Thickness Ratio Material d/t Stainless Steel 4.5 (type 304)
Aluminum 5.1 (alloy 6061) polyethylene 0.22 h (low density) V 8-9
SAR 5/91 Assuming an internal pressure of 1057 atm (15,538 psi), maximum values of d/t are displayed in Table 8-2 for the encapsulation materials ] of Table 8 1. The figures indicate that a polyethylene vial is not a practical container since its vall thickness must be 4.5 times the diameter. Both the aluminum and the stainless steel made satisfactory containers. 8.1.5.3. Fueled Experiments Each fueled experiment shall be controlled such that the total inventory of iodine isotopes 131 through 135 in the experiment is no greater than 750 mil 11 curies and the maximum strontium inventory is no greater than 2.5 mil 11 curies. These restrictioni are used to limit the fission product release in the event of an experinent container rupture In terms of hazard to humans, the iodine isotopes represent the most harmful short-lived isotopes. The values are justified by making the following assumptions:
- a. Half of the total iodine and strontium inventories are comprised of I-131 and Sr 90, the isotopes of each element which have the lowest permitted concentration levels,
- b. The iodine / strontium inventories are evenly distributed throughout the reactor room, which has a volume of 4.83 x 103m3
- c. At two air changes per hour, the isotope inventory is removed from the room in one-half hour.
- d. Average indoor concentration over an occupational year.
Calculations based on these assumptions result in indoor concentrations below the occupational concentration limits of each isotope, as shown in Table 8-3. Assuming the building wake dilution factor to be 5.3 x 105 sec/ft3 Calculations indicate that ground level concentrations outside the building are below the reference levels (4) of each isotope. Table 8-3 Calculated Isotope Release Values Units I 131 Sr 90 indoor pCi/ml 1 9 x 10-8 6.5 x 10'11 occupational limit pCi/ml 2.0 x 10-8 2.0 x 10 outdoor pCi/ml 2.2 x 10~11 7.5 x 10'16 reference level pC1/ml 2.0 x 10-10 5.0 x 10-12 8-10
SAR 5/91 8,1.5,4. Airborne Experiment Releases Experiment materials, except fuel materials, which could off-Gas, ' sublime x volatize, or produce aerosols under one of the following conditions shall be limited to occupational levels for airborne radioactivity concentration, as specified in IUCFR20, when averaging over a year:
- a. Normal operating conditions of the experiment or reactor,
- b. Crddible accident condit Ions in the reactor.
- c. possible accident conditions in the experiment.
The radioactivity release is based on the assumption that 100% of l the gaseous activity or radioactive acrosols produced escape to the reactor-room or the atmosphere. When considering materials for experimentation, the type of radiation emitted by the airborne radioactive products should be noted. For example , alpha-emitters, while not considerably hazardous outside the body, may cause significant damage to human tissue if they are inhaled, Although concentration limits in 10CFR20 reflect these considerations, the experimenter has a responsibility to know what hazards- may exist during and af ter t.he experiment. I
-The following assumptions shall be used to calculating the airborne reactivity concentration .
1
- a. If the effluent from an experimental facility exhausts through a holdup tank which closes automatically on high radiation level, at least 10% of the gaseous activity or aerosols produced will escape,
- b. If the effluent from an experimental facility exhaust through a filter installation designed for greater that 99%
efficiency for 0.25 micron particles, at least 10% of these ' vapors will escape.
- c. For materials whose boiling point is above 55'C and where vapors formed by boiling this material can escape only j through an undisturbed column -of water above the core, at least 10% of these vapors will escape.
These three assumptions contain phrases stating that at least 10% of the vapors escape from the specified, engineered safety systems that are part of the experiment design. ~ For-the purposes of calculation, the assumptions are intended to provide a conservative estimate of the amount of-radioactiv!ty that will be released. Calculations of the airborne radioactivity concentration depend on the . reactor room effective volume, building wake dilution factor, and the ventilation system flow rates. The reactor room effective volume, 4.83 x_103 m3, the bsilding wake dilution factor, 5. 31 x 105 sec/ft3 .- and the ventilation system fl ow rates are facility design features. 3 Calculations have been done to demonstrate the fraction of airborne release that will occur interior and exterior to the building for different ventilation flow rates. 8-11
SAR 5/91 8.2. SPECIAL EXPERIMENTAL FACILITIES 8.2,1. Reactor Core Facilities Reactor core experirnent facilities are designed for replacement of either single fuel element positions or a special nwltiele.nent position. Access to the core peak flux is provided by a central thimble. The wet central thimble is designed to allow insertion of an encapsulated sample into the core center or extr,.ction of a vertical neutron beam from the core center. Experiment facilities are located in single element positions in the reactor core for insertion of samples into the reactor neutron flux. A pneumatic terminal is one type of single element experiment facility. More 'han one such terminal may be in place for experiment use. Experiments with reactor characteristics may be placed in single element positions. One example of a single posit ion experiment is th< measurement of void or renetivity coefficients of various materials. Another characteristic is demonstrated by a reactivity oscillator apparatus. The oscillator is fabricated with rotating absorbers that are inserted in a single element position. Two types of experiment locations in the core structure provide for removal of multiple element sections. One of these types is a three element section at two different positions of the D and E rings of the core array. The second type is a six element removeable section that removes the elements of the B ring and changes the size of the center hole position. Fabrication of experiments for the multielement positions are projected for future f acility development. Reactor core limits on reactivity for experiments will apply to installation of any experiment facility. 8.2.1.1. Three Elemrpt Feature The three element feature of the core grid structure is a 2.062 in. (5.24 cm) diameter hole. Location of the hole center coincides with the center of a three element subarray. Total area of the hole is 3.34 in.2 (21.5 cm2), but also contains additional area for each of the three elements of the original subarray. If the total available area of the cutout in the subarray were used. the area would exceed the primary hole, but be less than the sum of three element holes and the primary hole. Three single element holes have a total area of 5.34 in.2 (34.4 cm2). One three element subarray cutout in the subarray is one D ring and two E ring locations The other three element subarray is one E ring and two D ring locations. 8.2.1.2. Six Element Feature A six element feature at the center of the core grid structure controls the position of the center hole and adjacent B ring elements, Removal of the A and B ring element array will create a 4.005 in. (10.02 cm) diameter hole. This hole has an area of 12.6 in.2 (81.3 cm2) with it's center at the reactor core center. 8-12
SAR 5/91 O i Additional space is available in the partial element holes that I I rediain in the B ring, This additional space is the result .of the experiment hole diameter, which exceeds the diameter of the center hole, but is less than a diameter that would include all the B ring holes, A limitshall exist on the use of this space such that the total area available to the experiment remains less than 15.8 in.2 (102 cm2). 8.2.2. Camma Irrafdafim Facility Gamma irradiation with cobalt-60 is provided by an irradiation facility in the reactor pool . The irradiator consists of 156 pencil size sources. Each source is contained in a space 0.5 inches in diameter and 11.25 inches in length. Sources are doubly encapsulated with inner aluminum clad and outer stainless steel clad. The source pencils are arranged in a double staggered, circumferential array, with an inside radius of 10.5 inches and outside radius of 12.5 inches (Figure 8 4a). The total source strength when installed was 9260 Curles, A shelf suspended in the pool at the end opposite to the reactor location holds the cobalt-60 irradiation facility below 10 feet of water. Access to the facility is provided by water tight canisters or dry irradiation tubes. Two canisters with 0,625 inch wall thickness, 5.0 inch radius and 11.5 inch height, and several 2 inch diameter vertical beam tubes, provide exposure volumes for routine experimentation. 8.2.2,1. Hazard to the Pool Water System Hazards to the pool water system consider three conditions. The conditions are the physical. location with reference to the reactor system, the possible leakage of a facility source and the radiation in the areas of the pool water system. Separation of the gamma _ irradiation facility and the reactor system is necessary to avoid system interactions. Separation of 0.5 meters provides sufficient distance to assure that_the irradiator or its experiments do not represent a potential reactivity source to the reactor. The distance also ~ controls the radiation impact of one system on the other. Two examples of possible radiation impacts consider both irradiator and reactor radiations. One example is irradiation of the gamma tource by neutrons that induce activation reactions within the source or its experiments. .Another example is the irradiation of reactor instrumentation detectors by gamma ays that could cause undesirable background levels at the detectors. Double encapsulation makes the probability of cobalt-60 escaping into the pool water system extremely remote. Further, the cobalt sluge within the pencils are quite insoluble in water, so that any leakage would occur slowly, allowing time to locate and remove the leaky source, Detection of a source leak will be done by monitoring the concentration of beta-gamma activity _in the pool water. Measurements each two months will monitor the pool water activity. In the event a leak doc a occur the pool water purification system, which operat es continuously, will 4 become a factor controlling cobalt-60 accumulation in the deionizing resins and concentration in the pool water. 8-13
SAR 5/91 mL
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3 ass taas GAMMA IRRADIATOR Figure 8-4 8-14
SAR 5/91 An action icvel for the cobalt-60 concentration in the water is set at 2.5x10's pCl/m3, equivalent to the allowable concentration for the monthly release of sewage waste (3x10'5C1/m)). The pool water system contains approximately 10,500 gallons (40,000 liters) with a water treatment rate of 8 gallons per minute (30 liters / min). This treatment rate processes all the water in about 22 hours. However, no discharges to a waste system occur as a part of normal operation. Equilibrium concentrations in the water balance the source leakage with the cleanup rate, S - (A + q/v)N o, (1) where S = is the production rate (leakage) of the source (uCi/hr) No w equilibrium activity in the pool water (C1) 1 = radioactive decay constant (br-I) 3 q a flow rate f or the purification system (m /hr) v a volume of the pool water system (40 m3) At equilibrium the reference concentration or action level represents a source leak rate of S - (1.3 x W5 + 1.8/40) 40 (2.5 x 10-5) a 4.5 x 10 5 C1/hr. Assumptions of the analysis are complete dispersion of the activity in the pool for a uniform concentration and 100% efficiency for cleanup function of the pool water purification system. Although the assumptions are ideal, none-the-less, the results demonstrate the bounds of poc:sible detectable events. Two s i t.ua t ions are of interest:
- a. If a pulse release occurs with no water cleanup or monitoring occurs immediately following the event the detectable leak is nne millicurie.
- b. As the recirculation of the water introduces a cleanup rate to balance leakage rate the monthly equilibrium detectable activity is thirty millicuries.
8 15
SAR 5/91 l l i Pulse rnlease: (40m 3
)
- 10'S C1/m - 1 x 10'3 C1 3
(2) Equilibrium release (24 Fr/dy) (30 dy/.vu) 4.5 x 10'5 C1/hr - 3 x 10-2 C1 (3) Projection of the dose from the cobalt.-60 isotope in the detontrer resin has been done by calculation with Microshield (2)(version 3.12). Table 84 Microshield Data [2] CASE: Dose at detonizer, 30 mil 1icurles cobalt 60, one meter CEOMETRY: Cylindrical source from sidt - cylindrical shields
.100, cm.
Distance to detector... . .. . ....
- 86. "
Sourc e length . . . . . . . . . . . . .. . ... . . . Source cylinder radius.... . . ...... 23. Dose point height from base . ..... . . 42. Thickness of shield... ... . .. ... . . . 0.480 " Air Cap........... ... . . . . . . 76.520 Source Volume: 142924. cubic centimeters MATERIAL DENSITIES (g/ce': , Mat e rid Sourn Shitld 6)r rne Air .(11i20 Iron 7 . 8 ( ') Water 1.0 BUILDUP FACTOR: based on TAYLOR method. Using the Characteristics of the materials in the source region. INTEGRATION PARAMETERS: Number of lateral angle segments (Ntheta).... .. . 5 Number of ar.imothat angle s.egments (Mpsi)..... 5 Number of radial segmentn (Nradius)..... . .. .. 5 SOURCE NUCL1 DES: Co-60: 3.0000e-02 curies RESULTS: Group Energy Ac t ivi t-y Dose point flux Dose rate
# (Mevi .
(1.hoton61s ec ) MeV/(so em)/ser (mrd1r) . I 1.3359 1.110e+09 /,591e+03 1. 3 70e+ 01 2 1.1797 1.110e+09 6.736e+03 1.252c401 3 .6953 1.811e405 7.741e-01 1.594e 03
*10 TAI.S : 2.220e+09 1.433e+04 2.622e+01 8-16
SAR $/91 At 30 days _ the accumulation of 30 rail 11 curies in the dolonizer resin represents a dose of 26 milliren/ hour at. one meter from the tank axis. Table 8 4 contains the data for the analysis. Measurements each two months will monitor the pool water activity. 8.2.2.2. Llazard_t!LLahnrat ory Pe rsonnel i The cobalt 60 incility location is deep enough in the pool so that dose rates at the surface of the pool water and at other points of the pool water systern. shield will be .lcss than 1 rnrern/hr. Tho location of the source and the geometry of the shield used for these calculations are shown in Figure B 4. Calculations assume a point to obtain the dose contributions f rom j the irradiator at the surface of the pool water and outside the concrete i reactor shie ld ,- Calculations for the irradiation facility at the position shown indicate that the dose rate outside the reactor pool shield structure is just under 1.0 mrem /hr. Ilowever, the done rate at
. the : surface - of : the pool _ will he less than .01 mrern/hr . Due to the i results of these calculations, the Co60 trradiator shall be no less than 4 48 m (14.75 ft) from the pool surf ace deck. At the normal wates elevatton of 8.10 meters this will assure 3.00 meters of water
?
shielding. Dotatis of the c al cul n t.i onn are given in the following pa ra gr4.phs . 8.2.2.3. E91Dt_SJSLLee ShielMur,,Jdilu11at ions A conservative- estiinate of the dose at the surface of the pool ' water due to the presence of a 10,000 curie co'o source assumes a point sourco, Calculations are done for distances in pool water and shleid concrete. locat. ion of the point source aus urnes a water depth of 4.05 rneters and a shleid thickness of 1.2 meters, Tito dose rate from this ! source is given by the expression 9"' SoBe 4sr? g i whott 1 D - dose rate, L So
- source strength (photons /sec),
_B -. butid up factor, p - linear absorption coefficient (cm'I), x - thickness of shield (cm), r - distance from sourco to reception (cin). K - conversion from gamma ray flux to dose rate, 8 17-
SAR $/91 The linear absorption coefficient depends on the inass attenuat ion coefficient and the density, i p - (p/p )p . I'o r water and concrete the respective mass attenuation coefficients and densities are as follows: p/p p water 0.060 cm 2/g 1.0 g/cm 3 concrete 0.0567 cm 2 /g 2.9 g/cm 3 Calculating px for the distances in water and concrete determines the shield attenuation and build up effects. In water, 2 3 px - (0.0600 cm /g) (1.0g/cm ) (40', em) ,
- 2 4 . ') ,
in contrete, px - (0.0567cm 2
/g) (2.9g/cm3 ) (122 cm) ,
- 20.1 The product, px, not only determinos the shie ld attenuation but also relates to the scattering buildup within a shield thick enough to cause multiple scattering.
The build up factor, B, is calculated from the expression F B - Al e'"1H x + A2 e'"2 x , where A 1 , A2, al, and og are constants [3]. For water the buildup factor constants have the following values (E7 1.35): A1
- B.5 A2
- -7.5 at - 0.093 "2 - 0.064 Thus, the build-up factor raay be expressed as B f.5e 40.093(24.3) . 7,3e-0.064(24.3) ,
- 81.5 1.5 ,
- 80 8-18
l . i 1 l SAR 5/91 i t i ' !e 1 For concrete, the butid up factor constants have the following values (E7 1.35 MeV): l i i A1
- 9.9 A2
" *8 9 [
i at . 0.088 a2 - 0.029 l i 1 k l Thus, the build up factor may be expressed as 3 , 9,9, 40.088(20) 8.9c.0.029(20) , ( ; I - 57.5 5.0 , l I
- 52.5 .
l Conversion from gamma. flux to dose rate is l t 2 i K - 4.17 x 105 (photons /cm sec)/(rad /hr) , and the source strength is j 1 10 So
- (10,000 curies) (').7 x 10 dis /rtec C1) x (2 photons / dis)' i i
So
- 7.4 x 10I4 (photons /sec)*
Doso Rate at Pool Surface . i-(7.4 x 10I4) (80) e 24.3 i D - . ' 0 4 m (4.17 X 10 ) (412)2 D - 1.9 X 10 6 rad /hr . Dose Rate at the Shield Surface , (7.4 x 10I4) (52.5) e 20.0 I D- - . D 4m (4.17 X 10 ) (120)2 . D - 1.1 x 10*3 rad /hr In the event of no pool water shielding the dose rate for a 10,000 : Curie : source at the example location will be equivalent to the calculation of a loss of pool water for the reactor core one hour after operation. The dose rate which will be about 850 rad / hour will require special precautions, e ; 8 19
SAR 5/91 8.3. OTilER FXpERIMENT FACll.lTIES 8.3.1. Euberitical Renctor and ModtLnin11 Cylindrical assemblies of graphite and polyethylene are utilized for student laboratory experiments with neutron sources and a suberitical uranium 235 reactor assembly. The plutonium-beryllium neutron sources and uranium dioxide used in the polyethylene suberitical assembly may be stored and used in the room cont aining the reactor , but are licensed separately from the reactor. The subcritical core and moderator astemblies are products of 1,ockheed Nuclear products (rigure 8 5). The ot he r i t ical polyethylene core is a cylinder 10 inches in diameter and 14 inches long. Reflector assemblies can be asserabled with or without the fueled core Dimensions of the cylindrical reflectos assemblies are 30 inch diameter by 34 inch length for the graphitt moderator and 22 inch diameter by 25 inch length for the polyethylene modtrator. An addit ional graphite moderator cylinder 30,5 inches high by 24 inch diameter is available for neutron source moderation. 8.3.2. J 4 MeV NeultmLC eneUtutt An accelerator is maintained for the production of DT reaction neutrons for research measurements and activation experiments. The accelerator is a Cockcroft-Valton type with an acceleration voltage of 150 kilovolts and beam current of 50 mic r o niip s . Application areas of the source of neutronn are planned in neutron dosimetry, neutron activation, and neutron interactions for analystr. of related research problems. 1 8 20 l
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i SAR 5/91 1O Chapter 8 References 1 l
- l.
- Review o l' Experimentre for Research Reactors", ANSI /ANS 15.6 j 1974 (H401).- ]
i
?. MICRQ1tilfLD Users Manual, Version 7, Grove Engineering, Rockdale, )
1 Maryland, 1985. I I 3. A. 11 ChiIton, J. K. Shul t. i s , R. E. raw, y, tin d p h of_ga g g g I j T.1;1pigiat, Prent ice lin11 Inc. , 1984 ,
- 4. " Standards for Protection Against Rndlat ion," Part. 20, Chapter ll'
) U.S. Code of Federal llegulationu, 1991, i 3 l 1-6 I I ! J 9 ; i i w F i E a b U f i d' a 8 22 T-I ",, - - - ,-,_ _ _w-.---,,, _ -v-,~=e-,*--*-
SAR 5/91 Chapter 9 RAbl0 ACTIVE MATIRIA1.S AND RADIATION MI.ASURI. MENT Radioactive materials and todiation control within the Nuclear Eny,ineering Teaching 1.abointory will be subject to industry standards 11,2) and license condltions (3,6} of state and federal a ge nc i e s . The ll . S . Nuclear RegulatotyCommins ton will tegulate the '!RICA reactor, une of special nuclear materials, and r e l a'.ed ac t ivi t i e s . Other materials and activities in the f acility will be regulated by the State of Texas Department of ilealth, Division for Radiation Control. Monitorinr. and sample programs .ill control telease of ef fluent s and waste Effluent pathways from the reactor bay consist of the HVAC exhaust air and the purge system exhaust air. No liquid effluent will occur from thi reactor bay an a result of 'mtmal reactor operation. Eifluent pathways f rom the building consist s of f ume hoods, exhaust vents of vacuum pumps and liquid waste discharge irom storage tanks. Solid waste for t he reactor and facility is packaged as necessaty for shipment and disposal, pigures 9-1,2 show innterial use areas and release pathways, pentures of the building design provide two monitoring points for ground water. One is a cample well in the reactor hay floor. The other point is a sump for water di a i na ge from the teactor shield foundation. Bot h of these points are f or evaluat ion of envi r onn+nt al condit ions. No conditions of normal operation will release ef f luent s to the ground water Some a cas are likely to contain concentrations of radioactive materials for extended periods of t line 9.1. RADIOACTIVE MATERIAL.S CONTROL. Physical control of radioactive materials shall be provided as an essential part of the radiological safety p r o gr ain . 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 radioactivo materials. 9.1.l. licac11'Lh!tl Irradiated reactor fuel shall be maintained in the reactor core, teactor pool storage racks, or reactor hay st orage pits. ~ Fuel elements wi11 bo removed from these iaci1itles only tor transport, measurement, or experinnent at ion. An area of the reactor facility will be de si gnat ed for the storage of a few single fuel elements of fresh fuel prior to irradiation in the reactor. 9.1.7. Reactol_CoFP.onenb Each reactor component removed I roin the reactor pool shall be mea su re ct for activation levels and removable contamination. All c oir pone nt s remaining in the pool shall be a s s ui..e d to be radioactive Components t emoved f rom the pool will be cleaned or covered as necessary to control radioactive contamination, components that contain radioactive materiai will be labeled and st ored in an area designated f or such components 9-1
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RADIOACTIVE MATERIAL. t EFFl.UENT CONTROL SYSTEMS Figare 9-2 9-3 _. . _ . _ _ _ ___. __ _ _ _ _ . _ . _ . _ . _ - . . _ _ . _ _ _ . . . _ . _ . _ __ _ _ . _ . _ . _ _ _ . _ . _ ._ _- - ~ ~
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SAR 5/91 l 9.1.3. jiuT1[ ment Facilities i Experiment incilities shall consist of all tubes or penetrations into the reactor core or reflectos that provide access to the reactor j neutron flux for an experiment application and shall include facilities in which rnat erials are exposed to beams originat lug from the reactor ! core. The cobalt 60 Irradiator shall also be considered an experiment , facility. Removal of experiment facilities from the pool or the beams originatitig f rom the t eactor shall be subject to the same controls as those for reactor component s. 9.1.4. Act ivated jippph.n l 1 Materials that are inserted into reactor experiment facilities or reactor beams -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. Samples exposed in a gamma irradiator (cobalt 60) will not be considered activated as radioactive materials. Specific locations for the storage of 3mples may depend on the analytic s t.a t us . Locations shall be designated and labeled for storage of samples and sample encapsulations, before and af ter analysis. Locations should be designated for storage of sample or encapsulation materials that are decaying and that are to he:
- a. analyzed b, disposed
- c. released to an authorized user
- d. released as non radioactive material e, and retrieved for subsequent use.
9.1.5. Radioactive Waste Cannisters shall be available and labeled for radioactive westo at locations where contamination from sampic procesr.ing or other activities with contaminated materials occur. Locations shall be designated for storage of solid wastes that are to be released for disposal. Liquid
-wastes shall. be maintained in a designated storage location until release critoria are determined, such as decay, dilution, or processing.
Specific sinks _ and drains in the facility that are designated - for radioactive materials shall be identifled. Cascous wastes are to be vented through low volume facility hoods according to allowable release c r i te ria .- Apprcprieste monitoring will be applied as required. 9.1,6. Other Materials Other materials that are to be identified and controlled by identification and location are encapsulated isotopic radiation sources, radiochemical source materials and process equipment, or tools possibly contaminated with radioactive materials. Activity levels of O encapsulated and radiochemical sources are expected to vary widely as 5 will- the handling and storage precautions. Activity levels associated with process equipment or tools shall be identified so that appropriate handling-and storage precautions can be instituted. 9-4 '
t 4 SAR 5/91 i 9.2. RADIATION HONITORING Radiation monitoring shall consist of fixed, portable, or sampling type systems. Monitoring systems will be applied to measurement of radiation areas and high radiat ion areas around the reactor f ac i l i t.y , significant contamination within and adjacent to the reactor facility, and radioactive sna t e ria l s and their concentrations in effluents. Moni t oring shall be considered for routine operations, abnormal ! conditions, and emergency situations. ;
- 9. 2.1 tiittLm. net l'rfeedures Zone i de nt. i f i ca t ion , access control, and protective equipment shall be designated. Zone identification for radioactive materials and radiation areas are designated as specified by 10 CFR, part 20 (Standards for Protection Against Radiation). Access control for zones shall be to control radiation erposures and physical security of the reactor facility and it's material as specified by 10 CPR part s 19 and 73, (Notices, Instructions, and Reports to Workers; inspections and Physical Protection .of Plants and Materials). Protective equipment for routtoe abnormal and emergency conditions shall include at least tape, plastic bags, absorbent paper, Sl oves, shoc covers, coveralls, and half mask air purifying respirators.
Continuous monitoring or control of radiation fields in the restricted area around the reactor shall occur whenever levels greater than 100 rnrem/hr are produced in accessible areas. The radiation levels may be caused by _ normal operation of the reactor or an experiment,
-devintions- from normal operatton, or casily changed shield configurations. Periodic measurement of accessible areas should occur in _ locations with significant radiation levels that do not require continuous monitoring. Personnel shall be informed of high radiation levels and care taken ' to prevent inadvertent increases in the levels.
Continuous monitoring _ may be replaced by periodic monitoring for temp'orary conditions that do not violate applicable rc6ulations or license constraints. 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 1 caves the des i r,na t ed ' area . Measurements shall provide action levels for removable l- activities of 500 disintegrations per minute. Periodic monitori"6 of , l' areas in which contamination is probable shall be of adequate frequency
- l to reveal significant changes in contamination levels. Decontamination l
of personnel, equipment, ~ and surfaces ~ shall be appropriate to ! requirements for control of radiation exposure and control of l radioactive material-containment. Airborne radioactive monitoring shall consist of continuous sampling of air . particulate activity in the reactor area. Warning levels and action levels -will be determined relative to allowable maximum permissible _ concentrations. Measurements should be sensitive to one maximum permissible concentration change in one hour. Monitoring will occur during reactor operation or activities involving fuel, core, 95 l ' -
4 4 s Salt $/91 f i 4 or experiment facilities, and will provide measurements for routine, abnormal, and emergency conditions. Additional airborno rnonitoring i equipment should be provided f or special experiment needs or locations l remote to the reactor area particulate monitor, I i j Effluent monitoring shall be provided for the discharge of the j tadioactive noble gan argon 41. Monitoring will conalst of either the ' une of integrating donimeters at a location of int erent or sampling of a point in the release pnth. M'anurements shall determine that the dose l l a t. a location of interest is either less than ten mrem per year above j i natural background or two pe' cent of the allowable maxirnum permissible ; ] concent rat ion for t he ym. Liquid ef fluent s shall be monitored beforo relear.e by satupling of gross beta-gamma activity, Specific isotopes l identified -and dilutions calculated such that released j should be I concent rations averaged over one year do not exceed 1% of the allowable ; maximuin perminuible concentrations. Other gar.coun or radioactive cilluents are to be examined on a case to case bants. , Personnel donimet ry shall bc sequired for access to reactor areas and some other f acili ty act ivi t ier,, Monitoring devices will typically- - be film badges with pocket dosimeters and thermoluminescent detectors l , for nupplemental measurements. Other personnel monitoring, such as ' I bloassays or whole body count ing, will be applied an deterrnined by the acttvity and conditions of radiation exposure situations. Personnel : f supplemental during activities that deviate ! ! shall use dosimetry [ subst antially f rom routino operations with supplernental dosimetry also i [ provided for persons visit ing areas with potential radiation exposure, j l 9.2.2. Monit orlur. TechnLques Implementation of radiation monitoring to maintain the goal of "as i low as reasonably achievable" should consist of: (a) preoperation l planning, (b) operations technieucu, (c) and post operation analysis. l l 9.2.3. thtniif.xment Survei ilante , A review by management of radiation exposures related to ! opeintions that cause significant radiation exposures compared to ; routino operations will be performed. The review should be applied to determine whether facility modifications or procedures should be implemented to maintain radiation exposuren "as low- as rearonably achievable." , l 9.2.4 Frequency atu,iletuutty Monitoring frequency and accuracy of activit les will be determined ~ by several factors related to personnel access, requirements, probability, and consequences of equipment failure, contamination and adequacy of current i pot ent tal, periodicity of modifications, moni t o ring. Accepted standards for measurernent sensitivity and accuracy should be appropriate to maintain radiation exposures "as low as j j reasonably achievabic." Frequency and accuracy specifications should be specified by procedures or other documents when appropriate. i L 9-6 3
j SAR 5/91 9.3, INSTRUMENTATION Ins trurnentation for the (valuation of radiation exposures from routine, abnormal, and einergency situations shall consist of fixed area monitors, portable survey monitors, and appropriate sampling methods. The ininimum instrumentat. ion available during reactor operation shall consist of fixed area gamma dose rate monitors, continuous air particulate monitor, portable thin window GM tube survey meter, portable neutron sensitive counter, and pocket dosimeters with charger. Other detecting equipment that should be available includes alpha-beta proportional counter, multichannel garnma pulse height analyzer, the rmo turninescent detector with reader, alpha scintillation detector. l l high and low range beta gamma dose rate meters, and GM tube or equivalent. friskers. l 9.3.1. Fixed Area Monitors fixed area gamma toonitors shall have remote 3;eadouts with audible and visual alarrns at the reactor control console. local readouts should , be provided in areas with significant radiation Icvels and routine ! personnel access. A multiple channel area monitoring system with GM type detector probes will be installed. The system should have at least l f our channels funct.ional . Me a surettent should include the dose range of 1 millirem / hour to I rem / hour. The fixed area monitors are designated for six general areas. The exact location within an area may varying depending on the presence of experiments or equipment. Three of the fixed locations, that are not ' likely to change are the cont rol room (room 3.208), the level three access point to the reactor pool, and the level two access point to the reactor shleid structure. The three remaining fixed locations in beam port experiment areas are intended to change if the arrangement of , experimental equipment requires a more appropriate monitoring configuration. Locations of the three experiment area monitors are at beam port 1 and between beam ports 2 and 3, and ports 4 and 5. 9.3.2. Airborne Radioactivity Monitors A continuous air particulate monitor with audible and visual ' alarms shall be func t.t onal in the reactor vicinity during reactor operations. A fixed - filter beta particulate monitor with 70 1pm flow rate capacity or equivalent system will provide air particulate monitoring. Detectors such as thin window GM detectors will monitor activity and provide alert and alarm conditions with visual and audible annunciators. Count _ rate of the instrument should include the range of
$0 to 50,000 counts / minute. ,
A gas monitor system for the noble gas effluent, argon-41, shall also be operable during operation, or sufficient data available to demonstrate a calculated release quantity. Design goal for the argon-41 monitor- is a sensitivity of 50% of 1 mpc (maximurn permissible concentration) for unrestricted areas. Test measurements indicate ~ a sensitivity for a ten minute count of 2 x 10' pC!/cc. 9-7 _ _ ~ - . _ _ - _ __ . . _ _ . , - _ _ . . - . . _ _ _ _ _ _ _ . _ - _ _ . _ . _ _ _
SAR $/91 The airborne monitors are located at two separate locations, for air particulates the sample point and instrument location are both in the vie ndty of the reactor pool observation deck, room 3.206. In contrast the argon.41 gaseous monitor is located in the control room area, room 3.208h, with sampling pipes that allow sampling up stream or down stream of the argon purge s,ystem filter bank. The sample return , line returns the air to the purge system exhaust. By alignment of the purge system sourco valves the gaseous monitor has the ability to sample air I rorn the reactor bay room, the reactor pool access area, or the experiment systems manifold. A filter in the sample line is available for analysis of particulates, 11 abnormal conditions should occur. ! 9.3.3. Survev and Lnboralery_ Instrumentation Portable survey rnoni t ors for alpha, beta, gamma, or neutron radiation shall be maintained for area surveys of laboratory and experiment areas. Survey instruments will consist of the following inst ruments or equivalent s: (1) a pancake style GM probe or alpha beta scint111ator to detect contamination, (2) ionization chamber for radiation field of 0 to 50 R/hr, and (3) neutron detector with spherical moderator to rnonitor neutron radiations. Supplemental rnea sur ement s should be available with alpha beta proport ional coun. ,rs or gamma ray pulse height analyzers. A Tl.D (thermoluminescent) detector should be available for measurement and evaluation of doses. Survey instruments and an alpha heta proport.ional counter are located in roorn 2.208, the llealth Physics Laboratory. Camma ray spectroscopy systems are located in room 3.112. 9.3.4. Llauld Effluent Sampline. Liquid eifluents shall be monitored by sampling methods to determine gross alpha beta act.ivity. Gamma spectral analysis should be applied for identification of isotope mixtures that require substantial dilut lon. for disposal . Liquid ef fluents shall be released in batches af ter storage for decay and dilution determinations. Reactor coolant may be -monitored for radioactivity in the coolant or purification loops as a supplemental indicator of water activity. 98
f f f ! l 4 SAR 5/91 i i l 9.3.5, Range and Spectral Rest.onse Instruments shall be available to measure the various types of ' radiation and the presence of low and high levels of radiation. Several
- types of detectors should be availabic for measurement determinations.
i i l 9.3.6. Calibrations ; r 1 Callbration methods, accuracy, frequency, and functional checks ! i shall be established for radiation monitors. Two classes of monitor i calibration will be applied. One class of calibration will consist of l monitors pplied to routine faellity operation and surveys. , l i Maintenance, calibration, and functional checks will be subject to l reactor ope *.ation specifications. ' The second class of instruments i should hr.-Je functicnal checks at annual intervals, but may be calibrated ;
- . b.hequently or at the time of application. .
I J. l .9.4 RECORDS i i l Records are specified for maintenance of radiological data that
-relate to reactor operation. These records shall include: ,
h a. Personnel dosimetry including bioassays or other special measurements made, : Ij
- b. Radiological specifications, control surveys required by facility [
i e, Caseous and liquid radioactive effluents released to the ! l environment, f i d. Radiation survey records, t
- c. Instrument calibration records,
- f. Radioactive material receipt and transfer records j g, Solid radioactive waste disposal records, l
- h. Leak tests of scaled sources, .
i 1. Data on. radiological incidents. j l f L l
)
I
- ]
h I I O : 9-9
- - - _ ~ . _ _ . _ . _ _ _ - - - - -. .
SAR $/91 , 9.5. EVALUATION OF MONITORl!4G SYSTI'.MS The radiation monitors provide information to operating personnel ! of impending or existing hazards from radiation r.o that there will be sufficient time to take the necessary steps to control the exposure of l personnel and the release of radioactivity or to evacuate the facility, ; Three types of radiation monitors are used: a continuous. air particulate monitor for determining tadiation levels 1ue to particulate radioisotopes suspended in the teactor room air. a continuous air gaseous monitor for determining sadiation levels due to argon 41 in the roorn air, and atea radiat ion monitors for deterrnining the gamma field at several locations in the facility. ; Each type of radiation monitor has a specific radiological purpose. The particulate air monitor is used to detect radioisotopes released due to fuel element failuie (a design basin accident). The gaseous air monitor is used to determine the eliluent radiation releano of ar6on 41. Argon, a component of air (.04% by volume) may be activated to - produce argon 41 in potentially significant quantities. Finally, the area radiation monitors are used to minimize personnel radiation exposures. The radiation monitors in section 9.5,1., 9.5.2. and 9.5.3. are typical instruments at the time of original installatlon. Replacements may'have slightly different characteristics. f 9.5.1. l'alliculttit MI Monitor Set points f or the particulate continuous air monitor warn of the presence of particulate fission product nuclides. Since gaseous and volatile elements such as krypton, xenon, bromine, and iodine have particulate _ decay products, the presence of some of their radioisotopes should also be detected. An alarm set point at 2000 picoeuries/ milliliter detects particulate activity concentrations at the occupational values of 10 crR 20 for 70% of the relevant isotopen in the ranges 84 105 and 129 149. These ranges of-isotopes represent the one percent yield for fission products of urantita 235. Significant fission products as a percent of total release are shown in Table 9 1. The. air monitor in use is n. Ludlum Model 333-2 beta- air ..:onitor, configured for continuous sampiing of airborne beta emitters. It uses two standard pancake G-H tubes, each having a 1. 7 5 -- inch effective diameter. The cwo. detectors are arranged in line so that gamma . ' background-subtraction is performed. This increases the accuracy of the beta count. The 333 2 will accept air flow rates ranging f rom .10 100 liters per minute. Particle accumulat ton on a fixed iilter continues at a constant-rate Activity.on the filter, however, is a function of the air flow-rate,-filter collection effletency, and the decay rates for nuclides that are present on the filter. If one assumes that the source in the , room is _ constant the activity at the filter will be the accumulation term minus the decay term-and will have the same f unctional form as the acttvation equatlon, 9-10
. _.._..__-.--_.-.-a____.,_-.-..__._.~._.___..__ ,, _ . . _ . - _ . - . .
a ! SAR $/91 i
- l. dNg q i -<N r. (I)
! dt V j j with the solution. l q <Nr i Ng(t) - - - (l.c'A'). (2) J V A t, - l Figure 93 represents a plot of equation 2 for the particulate 1- monitor with typical background conditions and the following ! assumptions; j 6,5 x 100 cm / 3 min 4,83 x 109 cm 3 i
- 1.35 x 10 6/ min .
! l ! c - 98 percent . i l A - (In?)/(2 hours) .
.346 .
At equilibrium for a 2 hour half life isotopo, the saturation l condition is .,
- t
, - 9 8Hr
- Ng(ta) -- ,
i V A l l 2000 cpm ! Nr(te) - ,
- .35 !
1
- $700 dpm , l q cN r l Nr. -- ,
a- V A , i N r.,/ 4 Nr * '
- (q/V)A j 5700/ 98 1
.- (1.35 x 10 6/ min) (.346/hr)(1hr/60 min) 11 3
! Nr - 7.5 x 10 #/cm , t 9 11
i SAR 5/91 { t i !
+
u Tabic 91 i Signi f icant Fission l'roduct s l Contr(button to Total Aettvity, Percent j
-,. j 90 Day *; fL* energy (MeV)
Ell'!!Tiit i nolp1T 1 [My.._.11Uhiys ]()L[ lay _s 1_ Year - l St tont lum Sr N9 2.8 6.7 10,5 2.7 1.49 ; 90 1.8 .546 i Sr 90 2.29 Yttrium Y Y 9I 3.4 7.6 12.5 3.9 1,55 l Sttontium Sr 9I 6,7 1.1.1.38,2.0 Y 92 4.2 3.62 i 7 Yttrium 4 93 7.6 2,99 Y 95 8,2 14.7 7.3 ,366,.398,.Lbb I Zi rcont urn Zr 3./ Niobita Nb ' 4.1 18.2 15 .159 Zi rconitun Zr 7 9,0 1.93 ; 97 9.6 1.28 Niobium Nb N9 6.8 1.23 Holybdentun Ho 4.6 103m 2 . 5 '; 5.5 7.0 , Ithoditun Rh Rut henium Ru l03 I 2.65 5.7 7,2 .2?5 Rh l05 1.35 .566. 25,.26 < Rhodium Rh 106 3,$4 l Ruthenium Ru l06 i 2.4 .0394 i lodine I I3I 6.8 3.7 .606,.25,.81 1 132 2,7 5.3 .80 - l 1 133 7,3 1.27 Tellurium Te l32 2,6 5.1 .23 l Xenon. 'XeI33 1,23 11.4 2.6 .346
-lodino I I35 4.1 1.0,.5,1,5 Xenon Xc I35 12.5 .'91 l
i Barium Ita I37" Ceslum Cu l37 1.5 .512,1,173 ! Barium Ba l/+0 1.25 10.6 10.8 1.6 1.01,.'47 . tenthanum La ll.0 12.0 12.5 2.4 1.36,1.25,1.68 l' cer turn Ce I61 6.3 11.2 8.5 .436,.581 Lanthanum le I0I 1.4 2.43
- l'raseodyinium I'r I63 10.0 11.2 1.9 .932 I43 6.8 1.09,.1 39 Certum Cc 4 CcI'* 6 2. 0 6.0 26,5 .316,.182 l'rasendyinium I'r I44 3.00 9 Neodymium Nd- IO2 l't omethium I'm I69 1.45 4.8 4.1 .804,.364 1.072 9-}2 e
's,_...-.-----_,-.. _,w-_ w-
AR 5/91 Activity Release (althorno concentralien) 1 R.9 \ ~~' _ - .
e. l.
\ .-
. \ _ .....
t.7 . I \ ..- v.6 e, 's t ,
'N'
~N r'4 x
{'3 ".1 / N W. Ig ,/
'~ _ -
+
6 li 1. T ll < 1, li l' B Time Coastent Ratte (ventilation reto)/(decay ratel Activity accumulation it bone half-life) - n-R.% / e - i . e.
. 4.
% . 7< /
'w .
l/ -
- y-_-_
'y f /
r . 4- / f
* /
e . 3- f 1 1 12
/
- n. l' f
'o id ad 3d 44 sd ~d t i.. < >. i t- i lv.. )
s ACTIVITY ACCUMULATION ON PARTICULATE l'lLTER Fi r,u re 93 9-13
_ _ _ . . = . _ .
- j I
SAR $/91 f !O If a sudden appearance of an additional nuclide or group of nuclides of air:11ar half life occurs, an addit.lonal curve siinilar to the i j background condition will occur. Cases are shown in rigure 9 3 for 2 j i br, 8 hr, 2 dy and 8 dy half-lives. l
?
l- To determine the count rate i.e t point, a, on the particulate air monitor, the following equation is used; i a - aVtn, (3) i l where, ; y a = alarm IcVel, epm 7 a a reference activity concentration, dis / min crn 3 i q w air sampling rate, cm / 3min ! t. a particulate accumulation time.-min i , n a cf fielency of the detector, cpin/(dis /rnin)- l The reference activity concentration may be expressed as follows: ; 1 ! dis /sec a - (2 x 10'9 pC1/crn3 ) (3.7 x 100 ) @0 sec/ min) , L l pCi i j dis / min
- 4.44 x 10'3 3 cm I
The following data portains to the 1.udium Hodel 333 2 particulate l air monitor, I q - 6.5 x 100 cm 3/d n , l t n - 0.3 . An electronic circuit. rnonitors the detector chamber flow rate to assure. adequate flow rates. par t iculat es in the air flow path accumulate on a filter with II et ficiency for 0.25 micron particles. ? The value of the detector efficiency is a conservative estimate based on the beta energies of intorest. Using a Tc 99 source, the detect.or -
- ef ficiency is rated at 36% for a beta tpectrum with 0.3 MeV energy )
range. Sinco inost of the inutopes of interest have higher bet.a ener6 ies than Tc 99, values of det.ector efficiency sill exceed 30% in most cases. . In fact, efficiencies of 50% may be applicable for matiy isotopes.. thus assuring a conservative detection limit for all except a few fission product isotopes. 9 14 ,
SAR 5/91 Honitor position to sample reactor room air is within 5 meters of the pool at the pool access level. The location will sample air activity in the vicinity of the reactor pool. 1.cakage of fission products from the fuel into the water then into the room would occur at the room air to pool water interface liackground measurement s of air particulate activities between Sept. 40 and Sept. 91 provide a record of t he naturally occurring count rate levels for the 1.udlum Model 333-2 in room 1.104 of the NETL facility. These data indicate that count rates of 4000 cpm to 6000 cpm will occur f.everal times each year as a res. ult of weather conditions that effect vertical air stability such as, f ront al lines, tersperature inversionu and storin systems. A set point at 5000 cpm will provide an alert level with an occasional alert for a natural occurring condition. The count rate alarm act point will assume beta energies of 0.3 and detector ef ficiency of 30%. The set point may now he calculated: 0 3 a - (4.44 x 10^3' dis) (6.5 3
/ min x 10 cm / min) (120 min) (0.3) cm
- 10,400 count s/rnin, Refer to Table 9 1. Most of the isotopes listed in 6e 1 day colurnn hnvc beta energies greater than 0.5 McV. Since the detect efficiency increases with the incident beta energy, a more representative estimate for the detector efficiency may be $0%. The set point may be calculated as follows:
dia/n.in a - (4.44 x 10'3 : ) (6.5 x 10 0cm /3 min) (120 min) (0,5) 3 em
= 17,300 counts / min.
A particle accumulation time, t, of two hours may be considered, as shown in rigure 9 3. 9,5,2. M gon 41 Mon 11er Set points ivr the argon 41 continuous air monitor should warn of
. excessive radiat ion levels for effluent icica>e and occupational exposurn. This radiation monitor will operate whenever the reactor system and the auxiliary air purge system a c operating.
As specified tn 10 CPR 20, the reference concentration of argon 41 is l'x 104 l4Ci/cm 3 Dividing this number by the purge exhaust system flow rate and by the building wake dilution factor yields the averagg at the stack, which i 10' annun1,3 concentration An alarm set point limit for at ten release times thislevel,2-x10'g?x mci /cm ' 3 mCl/cm . In the event of a gaseous will warn of an excessive daily release. iission product release in interference will occur in the argon-41 count due to betas emitted by isotopes of krypton and xenon, refer to Table 9 2. 9-15
.,m,-,,,, ,m- .-4.-~.,w. _,,,.,.,,mr-+,,-wwcww..,..e%-,..- v. ,....,,%,.. ,_w- ,,c.s-..n_ .w._ ,.,_,-=3,p.~rc,9_ , , _
_ -m,-p., __av.=a,-www<
SAR $/91-Monitor position to sample reactor room air is within 5 meters of the pool at the pool access level. The location will sample air activity in the vicinity of the reactor pool. Leakage of fission l 2 products from the fuel into the water then into the room would occur at the room air to pool water interface. Background measurements of air ' particulate activities between Sept. 90 and Sept. 91 provide a record of the naturally occurring count rate levels for the Ludlum Model 333 2 in room 1.104 of the NETL facilit y. These data indicate that count rates of 4000 cpm to 6000 cpm will occur several times each year as a result , of weather conditions that effect vertical air stability such as, frontal lines, temperature inversions and storta syrtems. A set point at 5000 cpm will provide an alert icvel with an occasional alert for a natural occurring condition. The count rate alarm set point will assume beta energiec of 0.3 and detector efficiency of 30%. The set point may now be calculated: dis / min 0 a - (4.44 x 10'3 3
) (6.5 x 10 cm /3 min) (120 min) (0.3) cm m 10,400 counts / min.
Refer to Table 9 1. Most of the isotopesSince listed in the 1 day column the detector efficiency have beta energies greater than 0.5 MeV. increases with the incident beta energy, a more representative estimate for.the detector efficiency may be 50%. The set point rnay be calculated as follows: dis / min a - (4.44 x 10'3 -) (6.5 x 10 0cm /3 min) (120 min) (0,5) 3 em
= 17,300 counts / min.
A particle accumulation time, t, of two hours may be considered, as shown in Figure 9 3. 9.5.2. f non 41 Monitor Set points for the argon 41 continuous air monitor should warn of excessive radiation levels for effluent release and occupational exposure. This radiation monitor will operate whenever the reactor system and the auxiliary air purge system are operating. As specified 1n 10 CFR 20, the reference concentration of argon 41 is 1 x 10'g uCi/cm3 . Dividing this number by the purge exhaust system
-flow rate and by the butiding wake dilution factor yields the averagt annual concentration limit for releese at the stack, which is 2 x 10"' 3 mci /cm . An alarm set point at ten times this level 2 x 10 5 mC1/cm '
3 will warn of an excessive daily release. In the event of a gaseous fission product release in interference will occur in the argon 41 count oue to betas emitted by isotopes of krypton and. xenon, refer to Table 9 2. 9-15
SAR 5/91 1 1 i Tabic 9 2 i Bota Emitt.ing, Gaseous Radionuclides of Interest I Ref erence Level , Concentrations i
# isotopes in Air i' hintupo Yic1d (%) oL! pass fJ IIni f _Li f e P,etn EntIL ien % (JtCi/ml) ]
\
'I Ar 0 1 109m 2.49 1% 1 x 10*8 1.20 99%
85 Kr 1.33 5 10./y 0.687 99.57% 7 x 10 l 85*Kr 1.33 5 4.48h 0.841 78.8% 1 x le l 87 Kr 2.56 5 76m 3.889 31% 2 x 10'd 3.486 40% 1.335 9.2% 3.044 7.1% 1.475 5.7% ' 88 Kr 3.7 5 2.8h 0.521 67% 9 x 10~9 I 2.913 14% i 0.681 9.1% 89 Kr 4.8 5 3.16m 4.93 23% - 2.33 15% 3.24 10% 2.53 5.6% 133 Xe 6.77 6 5.29d 0.346 99.2% 5 x 10' 135 Xe 6.7 6 9.17h 0.909 96% 7 x 10 138 Xe 6.7 4 14.2m 0.803 34% 2 x 10" 2.82 22% 2.38 18% 2.418 12% 0.567 10% tili) f 9 16 i _-.._..__~...a___._._.-_~.___...-__._...____..._,...._....,__. , . . . , _ . . . , . . _ . . , , _ , , - , - . .
SAR '>/ 91 The radioactive gas stio ni t o r in use is the P.R.M. Model AR 1000 argon 41 ruonitor. It uses a 50mm X 0.4mm Carl scintillator to detect the betas emitted by argon 41. Detection chaniber volume is 11.4 l i t e r s. , and it accepts a nominal gas flow tate of 30 liters pe r min'at e The system autottatically perfotms background subtraction. To det ermitm t he coutit rate set politt on t he argoti 41 ai r trotil t or , the f ollowing equat ton it, used: on c a - - (4) W1 where a a alert level on t he AR 1000, cpm 3 a a reference concentration level, pcl/cm n a detector response, cptn/(pC1/cm3 ) 3
@
- building wake dilutton factor, sec/cm 3
V a argon purge system flow sate, em / min f a iraction of 24 hour day the reactor actually operates Refgrence cgocentration level, o, at ground level outside the building is 1x 10' pC1/cm l'or the AR 1000, the detector tesponse, n, has been determined to be g c oun t/niin n
- 1.5 x 10 ,
pCi/cm 3 The building wake dilution, p, may be calculated from the following equatlon: 1 p - , (S) 0.5Av where 2 A a building cross-sectional area, m v w wind speed, m/sec 9-17
f SAR 5/91 The building cross sectional area, A, is conservatively determined from l the smallest side of the reactor building: l A - 234 m2 - 2.34 x 106 c,2 , The wind speed, v. is assumed to be 1 meter per second, also a i conservative value l 1 l ! The building wake dilution may now be calculated: I i i 1 l
- p - -
' O.$ (2.34 x lu cm) (100 cm/sec) - 8,55 x 10'9 sec/cm3, b I i The argon purge system flow rate, V, is 6.14x10b cm 3/sec (1300 cfm), f i Assuming the reactor operates for 8 hours each day, calculate f: l f l j f - 8/24 - 0.333 , 1 The count rate set point on the AR 1000 for an alert may now be 3 calculated: ; 1 (1 x 10'8 pC1/cm3 ) (1.5 x 10 8 counts / min ) , 3 I pC1/cm l a-0 5 3 sec/cm ) (6,14 x 10 cm /sec) (0,333) 3 (8.55 x 10 !
- 858 counts / min. !
Since the alarm concentration level is ten times the value of the alert , concentration Icvel, the count rate set point for an alarm would be at 8580 cpm, , i 9,5,3. 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 determine the Icvels of radiation during reactor operation. When possible, alarm set points for all area ; radiation monitors will be at either 2 mr/hr or 5 mr/hr. The first number is obtained by dividing the maximum desired dose each week by the l number of working hours each week. The second number is obtained from ! the definition of a railiation area in 10 CPR 20, i h 9 18 r
SAR 5/91 A high radiation area, defined in 10 CFR 20 as having a radiation level > 100 ne/hr, may exist above the pool access area during some operations. The area radiation monitor located above the pool access area will have an alarm set point of 100 mr/hr. Although the doses within the pool protection railings may exceed doses of 100 mrem /hr, the do s, c exists only in the immediate area of the pool surface. At other locat ions of the pool shield plat f or m level, the doses are signif icantly less than 100 mrem /hr, but may exceed the 2 to 5 mrem /br range. While the reactor is operating, one area indiation monitor will operate above the pool access area, cs well as at least two additional area radiation monitors located at other positions around the react or shield and at the beam port facilities. The radiation monitor system consists of six units with GM tubes that detect dose rates from 1 mr/hr to 10 kr/hr.
- 9. S 4. tinitilnLMailahllL1y JimidillMa Several factors apply to the acquirements for availability of t h(
reactor radiation monitoring systems. Among these factors are the types of conditions each monitoring system detects. if one of the continuous air monitors (CAM) is out of service, teactor operat ions may continue for a limited period of time, provided the other CAM is operating Reactor operation is permit t ed f or up to one week when the particulate air monitor is inoperable, provided a filter evaluat ton is performed daily, and a s i gni,1 from the argon 41 air monitor is available to provide information for manual shutdown of the HVAC. This is necessaiy to detect the design basis fuel element leak. When t he argon 41 monitor is not available, operating the reactor with auxiliary air purge system shall be limited to a period of ten days. This count raint restricts any effluent release from the reactor building. The particulate air monitor may be inoperable due to either an electronics failure or a pump failure. If an electronics tallure occur . the filter will accumulate particles as usual. Since the flow rate u known, the radioactive particles in the filter may be evaluated daily using a portable detector The expected drop in detector efficiency may be offset by the eight hour accumulation time If a pump failure occurs, particles will not accumulate on the filter of the particulate CAM. Instead, a cally evaluation is performed of the radioactive particles accumul at ed on the filter of the atgon CAM. Evaluation of the radioactive particles on the CAM filters should occur near the end of daily reactor operation. This technique will detect the presence of those particulate fission products having half-lives greater than a few hours. While daily evaluations of the CAM filter will detect a minor, persistent fuel element leak, the signal from the AR 1000 must be monitored for changes resultinn from a major sudden fuel element failure. This method detects gaseous fission products, such as xenon and krypton, which have rather short half-lives, as shown in Table 2.
~1 he reference level concent rat ions for most of the xenon and krypton isotopesofintgrest meet or exceedthe the reference level concentration of argon-41 alert concentration at argon-41 (1x10' mci /ml);powever, the detector is 2x10' mC1/ml. Because occupational level I
concentrations for the isotopes of interest are typically 300 times 9-19
SAR 5/91 their reference values and because their betas should be readily detected by the argon continuous air monitor, the argon CAM alert set : point can identify concentrations below the occupational levels for . these xenon and krypton isotopes, When the argon 41 monitor is inoperable, argon production and release may be calculated. provided the shielding configusaeton (including beam ports) is not altered, argon production and tricase-should not change. The release rates can be calculated fra o 3 measurement data and design flow rates for air through the s act a t-day ilmit is set for the inoperable period to limit - the unoun~ at releOse without direct measurement. This period represet*. t- tu u averaged over a year. The effluent release during a 10 day pe. . wou*i be about 4% of the average annual concentration limit. If the reactor is operating, at least half of the s ex radiation monitors must be operating, one of which must be located i .c the pool access area. This number of monitoring points includint 'e ' pool . area detector is sufficient to warn of unusual operat.no conditions. llowever, some consideration would he made to assure that monitors are operable within areas of experiment and personnel activity.
)
9 20
SAR 5/91 Chapter 9 keferences
- 1. "Radiolcgical Control at Research Reactor Facilities", ANSI /ANS-15.11 1977 (N628),
- 2. " Design Objectives for and Monitoring f Systems Controlling Research Reactor Effluents", ANSI /ANS - 15.12 1977 (N647)
- 3. " Nuclear Regulatory Commission" , Chapter 10, U.S. Code of Federal Regulations, Part 20.
- 4. " Texas Health Department", Texas Regulations for Radiation Control, Bureau of Radiation Control. -
e e 9-21
SAR 5/91 n-Chapter 10 CONDUCT OF OPERATIONS 10,1 TACILITY ADMINISTRATION 10,1,1. Organization 10.1.1.1. Structure F15 vre 10-1 11lustrates the organizational structure that is applied tr .the mana gernent and operation of the reactor facility. Responsibility for the safe operation of the reactor facility is a function of the management structure of Figure 10 1 [1]. These l responsibilities include safeguarding the public and staff from undue radiation exposures and adherence to license or other operation constraints. -Functional organization separates the responsibilities of academic functions and business functions. The office of the President J.dn.ini s te rs these activities and other activities through several vice presidents. Facility operation staff is an organization of a director and at least four full time equivalent persons. This staff of four provides for basic operation requirements. -Four typical staff positions consist of a reactor supervisor, reactor operator, health physicist, and research scientist. The reactor supervisor, health physicist, and one
. cder position are to be full time. One full time equivalent position may consist of several part time persons such an assistants, technicians and secretaries. Faculty, students, and researchers supplement the organization. Titles for stat f positions are descriptive an" may vary from actual designations. Descriptions of key components of the organization follow.
10.1.1.2. Executive Vice President and Provost, Research and educational programs are administered through the Office of the Executive Vice President and Provost. Separate officers assist with the administration of research activities and academic affairs with functions delegated to the Dean of the College of Engineering and Chairman of the Mechanical Engineering Department. 10.1.1.3. Vice President for Business Affairs. Business activities are administered through the Office of the
.Vice President for Business Affairs ~, One responsibility of -- the office is the administration and operation of safety programs.
I 10.1.1.4. Director of Nuclear Engineering Teaching Laboratory Ncelear Engineering Tetching 1.aboratory programs are . directed by an engineering faculty member that teaches courses in nuclear engineering and performs research related to nuclear applications. The
] Director is a rnember of the College of Engineering and Department ot Mechanical Engineering.
10-1
SAR 5/91 O The Univoretty of Tesas at Aust in organisetton Of fice of the Free ndent The Univoretty of Temas et Austin Vice Freeldent for Univoretty Saf ety Dueinese Affaire Dueiness Affatte hadiation Safety Dadiation Safety Coneit t ee O(ficer
- - ~
Esecutave Vios Preeldent and Provoet tusn of the College of Engineering chairman of the Dept. --- Ottector of urTL, Mechanical Engineering Wuclear beactor teactor committee supe rvisor O piwteer Engineering Teaching uboratory orgentsat ton Director of puolear tagtneering hadistion Teaching labortory safety Officer i i {NealthPhyelce -J Administrative Stert Assistant Director Reactor Supervisor Certified operatore lno...rchss.ff l ADMINISTRATION Figure 10-1 l 10 2
SAR $/91 10.1.1.5. Nuclear Reactor Committee The Nuclear Reactor Committee is established through the Office of the Dean of the Co?lege of Engineering of The University of Texas at Austin. Broad responsibilities of the committee include the evaluation, review, and approval of facility standards for safe operation. The Dean shall appoint at least three members to the Cornmi t t e e that represent a broad spectrum of expertise appropriate to reactor technology. The committee will meet at least twice cach calendar year or more frequently as circumstances warrant. The Nuclear Reactor Committee shall be consulted by the Nuclear Engineering Teaching Laboratory concerning unusual or exceptional actions that affect administration of t.he reactor program. 10.1.1.6. Radiation Safety Officer A Radiation Safety Officer acts as the delegated authority of the Radiation Safety Committee in the daily implementation of policies and practices regarding the safe use of radioisotopes and sourecs of radiation as determined by the Radiation Safety Committee. The Radiation Safety Program is administered through the Universi ty Safety Office and a University Safety Engineer. The responsibilities of the Radiation Safety Officer are outlined in The University of Texas at Austin Manual of Radiation Safety. 10.1.1.7. Radiation safety committee The Radiation Safety Committee is established through the Office of the President of The University of Texas at Austin, Responsibilities of the committee are broad and include all policies and practices regarding the license, purchase, shipment, use, monitoring, disposal, and transfer of radioisotopes or sources of ionizing radiation at The University of Texas at Austin. The President shall appoint at 1 cast three members to the Committee and appoint one as Chairperson. The Committee will meet at least once each year on a called basis or as required to approve formally applications to use radioactive materials. The Radiation Safety Committee shall be consulted by the University Safety Office concerning any unusual or e xc e p t.iona l action that affects the administration of the Radiation Safety Progsam. 10.1.1.8. Reactor Suom visor Reactor operation ai the Nuclear Engineering Teaching 1.aboratory is directed by a reactor supervisor. Responsibilities of the reactor supe.rvisor include control of license documentation, reactor operation, equipment maintenance, experiment operation, instruction of persons with access to laboratory areas, ami development of rerearch activities. 10-3
SAR S/91 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 qualified as a senior operator. This person is to be knowledgeable of regulat ory requirements, license conditions, and standard operating practices. A UT TRIGA Operations Manual will be maintained by the reactor supervisor. 10.1.1.9. 11calt h Physicin Radiological safety of the Nuclear Engiacering Teaching Laboratory is monitored by a henith physicist, who will be knowledgeable of the facility radiological hazards. Responsibilities of the health physicist will include calibration of radiation detection instruments, measurements of radiation levels, control of radioactive contamination, mai nt.e nanc e of radiation records, and assistance with other facility monttoring activitles. Activities of the health physicist will depend on two conditions. One condition will be the normal operation responsibilities determined by the director of the facility. A second condition will be communications specified by the radiation unfety officer. This combination of responsibility and communication provides for safety program implementation by the director, but establishes independent review. Ilealth physicist's activities will meet the requirements of the director and the policies of an independent university satety organization. 10.1.1.10, Professional and Classified 5tnLi Professional and classified staff, such as research scientists, research engineers, reactor operators, technicians and secretaries, will supplement. the organization as necessary to support facility programs. 10.1.2. Qualifications Personnel associated with the research reactor facility (2) shall have a combination of academic training, experience, skills, and health commensurate with the responsibility to provide reasonable assurance that decisions and actions during all normal and abnormal conditions will be such that the facility and reactor are operated in a safe manner. 10.1.2.1, Job Descriptions Qualifications for university employment positions are subject to job descriptions that summarize the job function and scope. The typical description includes title, duties, supervision, education, experience, equipment, working conditions, and other special requirements for the job position. University job positior.s are separated into three generic classification types. The three types are academic faculty, professional staff, and classified staff. Typical staff appointments will include personnel from each of these categories. 10-4
SAR 5/91 O' 10.1.2.2. Facility Director A combination of academic training and nuclear experience will fulfill the qualifications for the individual identified as the facility director. A total-of six years experience will be required. Academic training in engineering or science, with completion of a baccalaureate degree, may account for up to four of the six years experience. The director is generally a faculty member with a. Ph.D. in nuclear engineering or a related field. 10,1,2.3. Reactor Supervisor A person with special training to supervise reactor operation and ' related - functiono will be designated as the reactor supervisor. The reactor supervisor vill be qualified by certification as a senior operator as determined by the licensing agency. Additional academic or nuclear experience will be required as necessary for the supervisor to perform adequately the duties associated with facility activities. The supervisor is typically a person with at least one graduate degree in nuclear engineering or a related field. 10.1.2.4. Uealth Physicist A person with a degree related to health, safety, or engineering, or sufficient experience that is appropriate to the job requirements will be assigned the position of health physicist. A degree in health physics or similar field of study and some experience is preferred, certification is not a qualification, but work towards certification should be considered a requirement. 10.1.2.5. Professional and Classifie L$p 1( Qualifications vary substantially for other staff positions. Technical skills and requirements range from low to high. Education, training and experience may vary in a similar manner. Qualifications for operator certification by senior operator permit or reactor operator permit will require special consideration. This consideration will be necessary to determine that the knowledge and skills are sufficient to expect successful completion of training and certification by the licensing agency. 10.1.3. Br. actor Operations operation of the reactor and activities associated with the reactor, control system, instrument system . radiation monitoring system, and engineered safety features will be the function of staff personnel with the appropriate license certifications (2] Operation will include the implementation of required procedures, execution of appropriate
-experiments, actions related to safety, and the preparation of required I~ reports and records.
l 10-5 i
SAR 5/91 10.1.3.1. Staffing All activities that require the presence of license cert i fied operators will also require the presence in the facility complex of a second person capable of performing prescribed written instructions. Unexpected absence of a second person for greater than two hours will be acceptable if immediate action is taken to obtain a replacement. A designated license certified senior operator will be readily available on call during all periods in which activities requiring a certified operat or are being performed, The person on call will be considered available if the time to initlate a call request and respond on site is less than 1 hour. Movement of fuel or control rods and relocation of experiments with greater than one dollar reactivity worth will require the presence of a license certified senior operator. Other activities, such as initial startup, recovery from unscheduled shutdowns and modifications to instrument systems, control systems, safety systems, radiation measurement equipment or engineered safety features, will require concurrence and documentation by a license certified senior operator. Operation of reactor controls, movement of reactor experiments, maintenance of instrument control, safety, and radiation measurement systems will require the presence of a license certified operator. A license certified operator will be present in the control room whenever the reactor is not shut down by more than one dollar of reactivity or the control and system console panel is not secured. The staff required for performing experiments with the reactor will be determined by a classification system specified for the experiments, Requirements will range from the presence of a certified operator for some routine experiments to the presence of a senior operator and the experimenter for other less routine experiments, Some other activities that occur in the area of the reactor will require knowledge of a license certified operator, but not necessarily the presence of the operator. Such activities will include maintenance, handling of radioactive materials and experiment preparation. 10,1.3.2, Procedures Written procedures shall govern many of the activities associated with reactor operation, preparation of the procedures and minor modifications of the procedures will be by certified operators, Substantive changes or major modifications to procedures, and prepared procedures will be submitted to the Nuclear Reactor Committee for review and approval. Temporary deviations from the procedures may be made by t.h e reactor supervisor or designated senior operator provided changes of substance are report ed for review and approval. Activltles subject to written procedures will include routine st artup, shut down and operation of t he reactor; fuel loading, unloadine and movement within the reactor; and routine maintenance of major components of systems that could have an effect on reactor safety. 10-6
SAR 5/91 i Activitles subject to written procedures will also include
^
surveillance tests and calibrations that may effeet reactor safety; administrative cont rols for operation and maintenance that could effect core reactivity or reactor safety; personnel radiation protection and irnplementation of the emergency plan. 10.1.3.3. Expe r iinentJi Proposed experiments will be submitted to the reactor committee for review and approval of the experiment and its safety analysis {3]. Substantive changen to approved experiments will require reapproval while insignificant changes that do not alter experiment safety may be approved by the reactor supervisor or designated senior operator. Experiments will be approved first as proposed experiments for one time appliention, and subsequently, as approved experiments for repeated applications following a review of the results r.nd experience of the intttal experiment implementatlon. Each expe r iment. will be designated as one of three classes. One class will consist of experiments such as routine reactor operation for calibration or instruction, and routine irradiations such as neutron activation analysis. This class of experiment will require only the reactor operator during the reac t or ope ration or expe riment. set up. A few experiments may require the presence of both a certified operator and the experimenter, and will be designated as a separac class of exper iment . Another class of experiments will be specified fcr experiments that require large reactivity changes, such as experiment - facility movement, fuel or control rod movement, or significant changes - to shielding of core radiation. This class will require the supervision , of a senior operator 10.1.4. ArL[9as and Reports 10.1.4.1. Ope ra t ing Repor_t.fi Routine annual reports covering the activities of the reactor incility during the previous calendar year shall be submitted to licensing aut horities within three months following the end of each prescribed year. Each annual operating report shall include the fol1owlng informatlon:
- a. A narrative summary of reactor operating experience including the energy produced by the reactor or the hours the reactor was critical, or both,
- b. The unscheduled shutdowns including, chere applicable, corrective action taken to preclude recurrence.
- c. Tabulation o f. major preventive and corrective maintenance operations having safety significance s
SAR 5/91
- d. Tabulation of maj or changes in the reactor facility and procedures, and tabulation of new tests or experiments, or
] both, that are significantly different from those performed l previously, including conclusions that no unt aviewed safety questions were involved.
- e. A emury of the nature and amount of radioactive effluents released or discharged to the environs beyond the effective control of the owner-operator as determined at or bnfore the point of such release or discharp,e. The summary shall include, to the extent practicable, an estimate of individual radionuclides present in the effluent. If the estimated average release after dilution or diffusion is less than 25% of the concentration allowed or recommended, a statement to this effect is sufficient.
f, A summarized result of environmental surveys performed outside the facility.
- g. A summary of exposures received by f acility personnel and visitors where such exposures are greater than 25% of that allowed or recommended.
10.1.4.7. Safety Liml LY.L91nthn Actions tv !.c taken in the case of safety limit violation sball include cessation of reactor oper aticta unt il a resumption is authorized by the licensing authority, a prompt report of violation to 11a nsa authorities and management, and a subsequent follow- up report reviewed by the reactor committee and submitted to t.he license authority. The follow-up report shall dcarrthe applicable circumstances leading to the violation including causes and conttibcting factors that are known, effect of the violation upon reactor f acility components , sptoms or structures, healu. and safety of personnel and the public, and corrective action to prevent recurrence. Prompt reporting of the event shall be by telephone and confirmed by written correspondence within 24 hours. A written report is to be submitted within 14 days. 10.1.4.3. Rel-ase of Radioactivit_v Actions to be taken in ti.a case of release of radioactivity f rom the site above allowable limits shall include a return to r.ormal operation or reactor shutdown until authorized by management if necessary to correct the occurrence a report to management and license authority, and a review of the event by the reactor committee at the next scheduled meeting. Prompt reporting of the event shall be by telephone and confirmed by written correspondence within 24 hourn. A written report is to be submitted within 14 days. O 10-8
.. _ . _ . _ _ . m__. _ _ _ _ . . - . _ . . - . _ _ . . _ _ . _ . . __ _ _ _ . - . _ . . _._ _ _
SAR S/91 10.1.4,4, Other Rerortable Occurrences Other events that will be considered reportable events are listed in this section. A return to normal operation or curtailed operation , untti authorized by management will occur. Appropriate reports shall be submitted to 1icense authorities. (Note: Where cornponents or systems are provided in addition to those required by the technical specifications, the failure of components or systems is not considered reportable provided that the rninimum number of components or sys t erns specified or required perform their intended reactor safety function.)
- a. Operation with - actual safety system settings for required systems less - conservative than the limiting safety system settings specified in the technical specifications.
- b. Operation in violation of limiting conditions for operation established in the technical specifications unless prompt remedial action is taken,
- c. A reactor safety system component malfunction which renders or could render the reactor safety system incapable of performing its intended safety function, unless ~the malfunction or condition is discovered during maintenance tests or periods of reactor shutdowns.
d; Abnormal and significant degradation in reactor fuel, or cladding, or both, coolant boundary, or confinement boundary (excluding minor leaks) where applicable which could result in exceeding prescribed radiation exposure limits of personnel or environment, or both. j- e, An obsetved inadequacy in implementation of administrative
- or procedural controls such that the inadequacy causes, or l could have caused, the existence or development of an unsafe condition with regard to reactor operations.
10.1. 4 . 5 '. Other Reports A written report within 30 days to the chartering or licensing
- authorities of:
- a. Permanent changes in the facility organization involving Director or Supervisor.
-b. Significant changes in the transient or accident. analysis as
' described in the Safety Analysis Report.
10,1.5. Records Records of the following activities shall be maintained and
. retained-for the periods specified below-[4). The records may be in the form of logs, data sheets, or other suitable forms. The required l information may be contained in single or multiple records, or a combination thereof.
10-9 ,
- . . _ _ - - =
- - _ - ~ -
- SAR $/91 ,
p-l 10.1.5.1. Lifetime Records
- Lifetime records are records to be retained for the lifetime of the reactor facility. (Note: Applicable annual reports, if they contain all of the required information, may be used as records in this- .
sectlon.) l a. Gaseous and liquid radioactive effluents released to the g environs. l t
- b. Offsite environmental monitoring surveys required by {
Technical Specifications. .
- c. Radiat. ion oxposure for all personnel monitored. ,
- d. Updated drawings of the reactor facility.
10.1.5.2, Five Year Perlo_4 . Records ca be retained for a period of at least five years or for , the life of the component involved whichever is shorter.
- a. Normal' reactor facility operation (supporting documents such as checklists, log sheets, etc. shall be maintained for a ,
period of at least one year), i !, b. Principal maintenance operations. I i
- c. Reportable occurrences, 4
I1
- d. Surveillance activities required by technical i specifications, j
- e. Reactor facility radiation and contamination surreys where i- required by applicable regulations. I f. Experiments performed with the reactor,
- g. Fuel inventories, receipts, and shipments.
- b. Approved changes in operating procedures, 1 1. Records of meeting and audit reports of the review and audit
. group, i 10.1.5,3. One Training Cvele i r
} Training cycle records to be retained for at least one training
- cycle are the requalification records of certified operations personnel, j
Records of the most recent complete cycle shall be maintained at all , Lises the individual is employed. j-l 4 l l- < j 10-10 J_.._-_;.__-.._-____..-__,_ _ . . . _ _ . _ . _ . . . . _ . . . _ . _ _ - _ _ . - - . _ . - . - _ _ _ _ . _ . _ . - _ . _ _ .
SAR 5/91 10.2 OPERATOR REQUALIFICATION 10.2.1. Introduction Reactor operator requalification applies to all the controls and some features of the TRICA reactor at The University of Texas at Austin (UT), Balcones Research Center (BRC). The purpose of this plan is to provide training of each individual that is to qualify for a license to operate or direct the operation of the TRICA reactor. There are two license classes, one is an operator and the other is a senior operator. License qualification by written and operating test, and license issuance or removal, are the responsibility of the U.S. Nuclear Regulatory Commission. No r i ght s of the license may be assigned or otherwise transferred and the Itcensee is subject to and shall observe all rules, regulations and orders of the Commission. Requalification training maintains the skills and knowledge of operators and senior operators during the period of the license Training also provides for the initial license qualification. 10.2.2. Operator License Status Active status of any licensee shall requit e the performance of the functions of an operator or senior operator for a minimum of four hours each calender quarter. If the condition of an active license status is not met, the Director of the facility shall certify, (1) that the qualifications and status of the licensee are current and valid, and (2) that for recertification a minimum of six hours of license functions have been done. The license functions shall be done with supervision of the appropriate operator or senior operator. Otherwise the license status shall be inactive and no functions of the license shall be done. 10.2.3. Penualification Procram Bases Regulatory requirements and standards provide guidance for requalification training. Specific regulatory requirements are found in 10CFR55 for the licensing of operators and senior operators with regulations for requalification set forth in section 55.59. Standards for the selection and training of facility personnel and reactor operators are available in ANS 15-4. Specific regulations in the form of two sets of license conditions also apply to the facility personnel and reactor operators. One set of conditions for the facility liccuse, 10CFR 50.54, applies to facility personnel. The other set of conditions for individual licenses, 10CFR 55.53 applies to operators and senior operators. The following plan documents the requalification of operators and senior operators for the UT BRC TRICA reactor facility. 10.2.4. Recualification Pronram The requalification progtam consists of training personnel by lectures, instructior., discus.31on and sel f-stu:ly . At times the number of operators with licenses may be as few as 1 or 2. In these circumstances the application of discussion and self-study methods are necessary to accomplish tha traleing process. l 10-11
b SAR 5/91 g 4 g 10.2.5. Schedule , l Lectures from the requalification program topics and on the-job f i training will be dono on a two year cycle for the cornple t ion of: all
- j. requirements, The part of the program done each year will consist of j six-lectures, two on the job training activities, and the performance of I
i sixteen hours of license functions. 1.oc t u re n or instruction on the topics of the requalification
- progrma consist of eight topics shown in sect ion 10.2.6. Three lectures
[ will be- given each six months so that during the year there is an average _of_one topic presentation every two months. Each of the eight j- < topics will occur.during the two year cycle with four topics each year. The other two -lectures each year - are available for special subjects, ' repent subjects or review, i j -On the job training relies on two specific reactor control !. manipulations to be donc cach year. These control manipulations will consist.. of startup, shutdown, operation, coolant loss, loss of control , rod pneumatic air or electrical power events, and other system F malfunctions, The two control manipulationn will require a change in j l system reactivity and will use events from each of the two l i s t.s in i p sec tion 10.2.7. One event irom the list of section .7.1 must be done- t each year and one event from the list of section .7.2 must be done each l-
/" two years. A program of less than two years duration may accelerato the ,
l ( training of pornons for new operator certification. 10.2.6. List of Sulydects ;
- a. Theory and principles of operation. :
l b .- General and specific plant operating characteristics, f
- c. plant instrtunentation and control systems. ,
- d. Plant protectioh systems and Engineered safety systems. '
- c. Normal, abnormal, and emergency operating procedures. !
f, Radiation control and safety,
- g. Technical specifications, l
; h. Applicable portions of Title 10, Chapter 1, Code of rederal !
Regulations. .. l l 10.2.7. On- the- j ob- t raluj.ng: l 10.2.7.1, List of annual training tasks; (one must be done each year):
- a. Plant or reactor startups to include a range that reactivity l feedback from nuclear heat addition is noticeable and heatup j rato is established.
10-12 ; _ _ _ _ _ _ _ _ . - ___ _ . . _ _ _ _ _ .- - - - _ ___.J
SAR $/91 l l l b, Plant shutdown.
- c. Significant (>10 percent) power changes in manual rod control.
- d. Loss of coolant inside or outside primary confinement.
e, Loss of coolant, large or small, including leakrate estimate,
- f. Loss of control air (or inadequate pressure).
- g. Loss of elect rical power (or degraded power sources) .
- h. Loss of core coolant flow / natural circulation.
10.2.7.2. List of training tasks; system malfunctions (one must be done each two years):
- a. Reactor trip.
- h. A nuclear instrumentatton failure,
- c. Loss of protective system channel.
- d. Control rod or drive failure such as rod position error, rod drop or stuck drive, mispositioned control rod or rods (or rod drops), Inability to drive control rods.
e Fuel cladding failure or high activity in reactor coolant or offgas,
- f. Malfunction of an automatic control system that affects reactivity.
10.2.7.3. On-the-job training will perform the following periodic training checks or functions,
- a. Observation at least once each year of a satisfactory understanding of the reactivity control system and knowledge of operating procedures.
- b. Each operator or senior operator will review facility design changes, procedure changes and license changes as they occur or once each 6 to 8 months.
- c. A review of the contents of abnormal and emergency procedures will be done by each operator or senior operator at 6 to 8 month intervals so that at least 3 reviews occur during the two year training cycle 10-13
SAR 5/91 10.2.8. Evaluation Evaluation of license personnel depends on annual examination and periodic observation. The evaluation by annual written examination determines the knowledge level and requirements for retraining by a percentage test score. Other evaluations by visual observation assess the performance and competency with routine procedures and the skill at manipulating the controls of the reactor. The written annual examination will assess .;perator or senior operator knowledge of current training subjects and review requirements. A five part test with objective questions will assess the knowledge of f our of the eight program subject s, and the areas of section .7,3 h and
- c. These two sections pertain to changes in f acility design, normal procedures, reactor license, and abnormal and emergency procedures.
Each of the five parts of the exam will have a 100 point basis with an average of 80% as the acceptance criteria. An overall score of less than 65% shall require an immediate evaluation of license duties. Proficiency by retraining shall demonstrat e acceptance within 4 months or license duties shall suspend until proficiency is acceptable. A person that scores between 65%-80% shall retrain as necessary in those areas that written or oral exams indicate a deficiency. A systematic observation of license activities, by a supervisory senior operator or a level of the facility management to which a supervisory operator is responsible will evaluate operator and senior operator performance. Visual observation of the performance in response to the conditions of sections .7.2.1 and . 7,2.2 will provide the basis for judgement of the operator's skill. In the case of a senior operator the performanen may be either direct actions or the direction of a response by another operator. Judgements of a person's skill or competency is subjective and may include general observations of performance at any time the person is responsible for license functions. 10.2.9 Records Records for each operator or senior operator will consist of the documentation for the requalification activities within the two year training cycle. The records will be kept until the completion of the next training cycle. A record for each operator includes at least the (ollowing information:
- a. Attendance at training lecture or acceptable review of the material, including topic and date,
- b. Completion, satisfactorily, of two on- the -j ob training events with performance evaluation; recording date and performance as excellent, average or poor,
- c. Total # of reactor control system hours and energy production in each calender quarter.
- d. Scores of the written examination and copics of the exam questions, answers, and responses by personnel.
10-14
SAR 5/91 10.3 RADIOLOGICAL PROTECTION PROGRAM Protection of personnel and the general public against hazards of radioactivity and fire is established through the safety programs of the University Safety Office. Safety programs at the reactor facility supplement the university programs so that appropriate safety rneasures : are established for the special characteristics of the facility [5,6] 10.3.1. Manacement and Policy. Radiological management policy shall include a comrnitment to keep occupational exposures as low as is reasonably achievable to facility personnel and the general public. Other elements of the radiological management will_ include:
- a. Instruction of personnel in awareness of- the low as reasonably achievable commitment,
- b. identification of radiation protection personnel and their responsibilities, l
l c. authority of personnel to communicate with management and modify or suspend activities for reasons of -radiation l: p ro t.ec ti on , l
+ d. assurance of sufficient and appropriate training of l
personnel in radiological safety, e, periodic evaluations of the program to determine possibilities for-lower radiation exposures. Suggestions and recommendations for modifications to operating and inalntenance procedures . and to reactor equipment and facilities shall be considered by management to reduce exposure to radiation. Implementation of modifications will occur. if substantial exposure reductions are possible at acceptable cost. 10.3.2. Ersoonsibilities Radiation protection at the reactor facility is the responsibility of-the Reactor Supervisor, llealth Physicist, or a designated senior operator in charge of operation activities. Responsibility shall
-include the authority to act on questions of radiation protection, the arquisition of appropriate training for radiation protection and the reporting to management of problems associated with radiation protection.
10.3,3. Organizational Access The person responsible for radiation protection at the reactor facility will have access to other individuals or groups responsible for radiological safety at the University. Contact with the Radiation Salety of ficer will occur on an as needed basis and contact with the Reactor Committee will-occur on a periodic basis. 10-15
SAR 5/91 10.3,4, Equinment and Sunnlies Equipment and supplies maintained for radiological safety managernent shall include: a, fixed area radiation monitors,
- b. air particulate monitor, c, gaseous ef fluent unitor,
- d. portable radiation monitors,
- c. detectors for contamination measurement, f, maintenance and calibration capacity for equipment,
- g. laboratory counting and analysis equipment,
- h. supplies for storage of contaminated equipment,-
- 1. provisions for radioactive waste disposal, J. decontamination facilities.
- k. protective clothing,
- 1. respiratory protection equipment, m, and emergency response equipment, 10.3.5. Trainitm and Sa[nty
~
Ench person in the restricted area of the reactor facility shall have sufficient radiological safety training for the purpose of access to the area or be escorted by a person with the appropriate training. Training will be appropriate to the activities of persons admitted to the area and will-range from simple instructions of-emergency alarms and
- evacuation -procedures to more complex implement.ation of the area emergency plan,
- Training for f acility personnel shall be specified by the Reactor Supervisor and shall provide sufficient training in radiation safety policies . and procedures, and in the use of radiation safety equipment 4
located in- the- facility to control exposure during normal, abnormal and emergency situations. Training will consist of:
- a. radiological safety policies, plans and procedures, b, Radiation hazards and health-risks,
- c. use of protective clothing and equipment '
d, use of_ portable radiation monitoring equipment. e, and other documents such as the emergency plan and federal and state notices to workers. An evaluation shall occur every two years to determine whether
- ads!-fonal training of personnel is required and that the radiological safety program is functioning adequately, Safety programs, with the exception of reactor operations, are operated as a function of the business administration of the University and includo a radiation safety organization as presented in Figure 10-1.
10-16
SAR 5/91 10.4 FIRE PROTECTION PROGRAM Fire protection consists of 3 factors, passive fire protection, active fire protect. ion, and fire prevention. Management of the Nuclear ' Engineering Teachin6 1.aboratory shall be knowledgeable of fire protection cont.rols. The cont.rols will- consist of actions of equipment, actions of laboratory staff and interactions with University inspection personnel. 10.4.1, Facility Fire Protection Element s Fire protection is recognized as an important element of the safe operation of the TRIGA reactor facility. Commitment by the University to fire protection is provided by the functions of the University Safety Office. The organization for fire protection consist.s of the University Fire Marshall, a member of the University Safety Office and the Reactor Supervisor, a member of the Nuclear Engineering Teaching Laboratory. Responsibilities of the Fire Marshal are the maintenance of fire protection equipment and inspections for itre prevention. Responsibilities of - the Reactor Supervisor are knowledge of potential hazards and implementation of fire protection recommendations. Although fire p ro t.ec t ion is provided for the general safety of ( personnel and preservation of property, special considerations shall be provided for systems designated as . safety related. Primarily special considerations - are applied to prot.ection of the reactor and shield structure, and fuel ' storage wells. Design f eatures _ of these facility _ components provide a maj or factor of the fire protection. Fire protection _ for the instrumentation and control system, and radiation measurement systems are important for the initial reactor shutdown and the . availability in emergency conditions. Fire protection of the reactor hay area -boundary is of importance . to the extent of limiting either internal conditions that would cause the release of hazardous materials or external conditions that would threaten the release of hazardous materials. Loss criteria - for decisions on fire protection at the reactor facility shall consist of preventing any A njuy to personnel, and minimizing - the potential or actual release of radioactivity to the-environment. No injury _or exposure to the public should occur from the adverse ef fects of a fire. Laboratory personnel, particularly certified operators, shall be instructed to continually observe conditions that might represent a risk
-to fire protection. Appropriate assessment of the risk should be provided by the Reactor Supervisor and will include consultation with the Fire Marshal when appropriate.
O 10 17
SAR 5/91
\
Passive fire protection elements effectively protect the reactor core. _ fuel elements and storage wells. Inherent design of the reactor bay and reactor tank structure, construction materials, building layout and fire barriers are all applied to the protection. Instrumentation and control systems and radiation measurement systems primarily are protected by fire detection and alarm information. These systems are important to safety only for the initial shutdown and removal .of personnel. Protection of other equipment and the reactor bay boundary is accomplished in part by building design, but prinarily by detection and alarm. 10.4.2. Facility Fire Protection Control The Reactor Supervisor and the Nuclear Reactor Committee shall consider the inoact of major facility modifications and experiment programs on facility fire protection. The University Fire Marshal will recommend fire protection requirements and provide for inspection and test of fire protection components. Activities such as welding, cutting, open flames, electrical loads,_or other equipment that effect fire protection shall be examined on a case by case basis by the Reactor Supervisor. 1.aboratory staff shall be instructed in fire response actions and notification of responso personnel. A program to familiarize response personnel with laboratory equipment, material hazards and physical layout is considered the major element for response of emergency response organizations. 10.4.3. Fire Safety Assurance At intervals of two years the fire protection program should be examined actively by the Reactor Supervisor, University Fire Marshal and Nuclear Reactor Committee. Evaluations of post inspections, tests or incidents shall be incorporated into an assessment of the fire protection evaluation. Recommendations if any should be identified and appropriate actions taken. 10.5. SECURITY AND EMERGENCY PIANS Plans. for physical security and emergency response shall be established, maintained, and implemented by the Reactor Supervisor. These plans will be separate documents with application procedures. Each plan development will use available reference documents for guidance. Review of the plans will occur at two year intervals. b U 10 18
- - - - - - - - - - - - - - . - - _ . - - - . . - - . - ~
SAR 5/91 [ 10.6, QUALITY ASSURANCE PROGRAM Objectives of quality assurance (QA) may be divided into two major goals. First is the goal of safe operation of equipment and activities to prevent-or mitigate an impact on public health and safety. Second is the reliable operation of equipment and activities associated with education and research functions of the University, The risk or potential release of radioactive materials is the primary impact on public health and safety, and may be divided into direct risks and indirect risks. Direct risks are activities such as waste disposal, fuel transport and decommissioning that introduce radioactive materials into the public - domain. Indirect risks are aceident conditions created , by- normal or abnormal operating conditions that generate the potential or actual release of radioactive materials from the controlled areas of a facility. 10.6.1, Introductinn Characteristics of uranium loaded zirconium hydride fuel used in the TRIGA reactor provide substantial benefits to safe reactor operation. Many accident situations are simulated by normal operation of the fuel _in either pulse mode or steady state mode. Other features, such as fission product retention, stainless steel cladding design, facility _ engineered features, and periodic schedule of operation, ' combine with routine operation procedures to decrease the consequences O of failure of any reactor components. The limited scope of application of formal qualit.y assurance criteria is due to the fact that most parts and procedures associated with operation of the TRICA type reactor are not relevant to public health and safety. Safety-related identifications for quality assurance are
-determined from safety analyses. Although several systems such as the roactor safety and protection system, engineered safety features and.
radiation monitoring systems are important to safety, only one reactor component is identified as safety-related. The quality assurance program is not applied to routine reactor operations and surveillance activities but shall be implemented for non routine activities determined to be safety related in nature or affecting safety-related components. Activities shall include design, construction, testing, modification and maintenance of safety-related items. Other components related to safety limits, limiting conditions for operation and design features, as identified in technical specifications, will apply only those . elements of qualit.y - assurance necessary to establish reliable performance _ of the intended structure, system or component function. The following table lists the - components subject to quality assurance program or selected sections of the program. Two additional conditions remain, however, that are important to the application of at least portions of the quality assurance program. One is the safety to operation personnel and experimenters and the other is continuity of the operations programs. Each of these conditions must
$ be examined objectively relative to operation procedures and program expectations. In general, the appilention of good industry quality assurance practices is sufficient to meet operational program goals.
l 10-19
SAR 5/91 1
\
7s ( Table 10 1 Q List for IMW UT TRIGA QC/QA Structures, systems or components Safety
- Specs Fuel Element Cladding structure 1 manuf.
Shipping package 1 Reactor Core Structural components 2 manuf. Tank structure 2 design Shield structure 2 constr. Experiment Equipment (core reflector) Beam tube components 2 design Rotary rack system design ('~}) ( 2 Experiment Egalpment (core grid) Pneumatic tube-components 2 design Installed core system 2 design Protective Systems Instrumentation system 2 manuf. Control system 2 manuf. Safety system 2 manuf. Auxiliary Systems Pool coolant system 2 design ! Water purification system 2 design Room confinement components 2 const. l 2 const. l Area ventilation components Area radiation monitor system 2 manuf. Air radiation monitor system 2 manuf. l 1 l 1- All sections of quality assurance program shall be considered applicable. 2- Specific sections of quality assurance program should be applied as required to assure reliable performance. l 10-20
SAR 5/91 The quality ass,urance program shall be commensurate with the TRICA type reactor, The University of Texas administrative programs and the goals of quality assurance. This document provides requitements The for establishing, managing, conducting and evaluating the QA Prograin. QA Program applied to items or activities determined to be safety-related follows the guidelines of Reg. Guide 2.5 (77/05) [9,10]. 10.6.1.1. Purpose Quality assurance of cert ain activities associated with the University of Texas TRICA reactor facility is important for the safe and efficient completion of tasks that are identified as safety related. This document outlines the general element s of quality assurance applied to safety related structures, systems or components, and activities. Requirements are documented for establishing, managing, conducting, and evaluating the QA Program. Although aspects of the QA Program may be routinely applied to many facility activities, the formal implementation of the program is limited to specific items or activities related to public health and safety. Table 10-1 lists the quality level and description of key systems and components. 10.6.1.2. Responsthility The University of Texas at Austin as owner and operator of the TRICA reactor facility, shall be responsible for a quality assurance program. The owner-operator shall establish and implement a program consistent with the goals of quality assurance for safety-related activities, structures, systems and components. Identification of safety related items shall be the responsibility of the owner-operator and will include a description of the iteru and the applicable elements of the quality assurance prograra. Special quality provisions, delegated functions of the program, and unresolved quality assurance problems shall also be identified by the owner-operator. The facility supervisor shall have the ultimate responsibility for both the specifications of quality related requirements and the functions of quality related activities. Table 10-2 lists the responsibilities and key personnel participating in the University TRICA QA Program. 10.6.1.3. Organization The organization applied to quality assurance activities shall be part of the normal university administrative structure. The facility Supervisor shall develop and implement the quality assurance program and identify safety related items. Unresolved issues of quality assurance shall be reported to the Director of the facility and the appropriate administrative vice president of the university. Execution of specific elements of the program may be delegated to persons in the University organization or other organizations as appropriate. University peraone shall include committees, faculty, researchers or staff, as required for specific program applications. Non-university organizations or persons may supplement University personnel when specialized qualifications are necessary for specific quality assurance tasks. The Universit" organization applied to reactor safety and quality assurance is the academic administration represented by Figure 10-2. 10-21
SAR 5/91 Table 10-2 RESPONSII51LITIES AND KEY PERSONNEL limsoonsibiltiing Eiy University._ttisennel
- 1. Establish program Director or Supervisor Implement program of TRICA facility Identify Safety related items
- 2. Unresolved issues President. or Executive Vice President and Provost
- 3. Delegated functions Faculty and staff 4, Specialized functions Specified personnel Quattty Amaurence afflee of the Prealdent The Univoretty of Teses et Austin Isocutige vloe Preeldent and Provoet Director of
+- Nuclear Engineering Teaching taboratory Aseletant Otractor peector Supervloor ^
QUALITY ASSURANCE ORGANIZATION Figure 10 2 10-22
SAR S/91 10.6.1.4. Documentation All activities affecting safety related items subject to the quality assurance program shall be ident ified and documented formally. The format of Table 10 3 shall be used to identify applicable elements of the Quality Assurance Program and identify documents, procedures, reviews, inspections, tests, or other quality assurance features that are to be applied to a safety-related activity. The checklist or approval shall be incorporated in the table format for the acceptance of each specified quality assurance element by the faellity supervisor. 10.6.2. Quality Assurance Controls 10.6.2.1. Design Controls. Design controls shall consist of design specifications, references to applicabic codes, standards and regulations, design verifications and document approval. Applicable codes, standards, regulations or other quality requirements will be identified and requirements incorporated into the design documents. Design document approvals shall be part of the design document. Design approval will be by a person, other than the design originator, that is knowledgeabic of the design criteria and is informed of the quality requirements Modifications of safety-related documents shall be subject to the same provisions as the original document. Approval of the design modification will be included with the design document and the modification identified. Verification of design adequacy shall be provided by either design reviews, alternate calculation, test program or other method, determined to be appropriate, Verifications of the design shall check characteristics such as compatibility of materials; suitability of application of inspection, maintenance and repair; proper interfacing of sub-systems, and proper acceptance criteria. Method of verification will be identified and documented by approval of the design docuroent . 10.6.2.2. fIocurement Conrrola Procurement controls shall consist of procurement specifications, re fe re nce s to applicable codes, standards and regulations, procurement. acceptance and document approval. Applicable codes, standards, regulations or other quality requirements will be identified and references incorporated into the procurement documents. Procurement 1' document approvals shall be part of the procurement document. Procurement approval will be by a person, other than procurement originator, that is knowledgeable of the procurement specifications and is informed of the quality requirements. Changes to safety-related procurement documents shall be subject to the same provisions as the original document. Approval of procurement changes will be included with the procurement document and the change identified. 10-23
SAR 5/91 Table 10-3 FORMAT FOR SAFETY RELATED QA CllECKS Each safety- related act ivity st ructure, system, or component will be given a letter symbol, such as A, B, C, and be appended with the following designations (for example, A1.0): 1.0 Title Identification and description of safety-related item 1.1 Participation - supplemental organization and functions 1.2 Documents applicable procedures or special measures 2.1 Design Control 2.1.1 Codes, standards and regulations 2.1.2 Method of verification 2.1.3 Modifications proposed 2.2 Procurement Control 2.2.1 codes, standards and regulations 2.2.2 Quality assurance specifications 2.2.3 Proposed changes enacted 2.2.4 Procurement conformance method 2.3 Document Control 2.4 Material control 2.4.1 Special procedures required 2.4.2 Equipment required 2.4.3 Personnel qualifications 2.5 Process Control 2.5.1 Special procedures 2.5.2 Special equipment 2.5.3 Personnel qualifications 3.1 Inspection Program Description 3.2 Test Program Description 3.3 Measurement Equipment 3.4 Nonconformance Item and Disposition 3.5 Corrective Actions Instituted 4.0 Records List 9 10-24
SAR 5/91 ' Acceptance of procured items or services shall consist of evidence provided by the contractor, evaluation of the procurement source, inspection at the source or inspection upon receipt. Acceptance of the procurement should require measures such as quality assurance by contractor, inspection and test functions, or controls on materials processes and nonconformances. The methods of acceptance will be identified and documented by approval of the procurement document. 10.6.2.3. Docume nt Con tnel Document control consists of monitoring the development, revision, release and use of documents, drawings or specifications affecting safety-related activities. Document control shall include assurance that safety related documents are identified as such, and are completed and maintained properly. The laboratory Supervisor shall provide control of safety related documents that are specified according to the format of Table 10-2. 10.6.2.4 Material Cont rol Procedures shall be written to establish material control when special measures are necessary to assure material quality of safety-related items. Controls shall be applied to activities such as identification, handling, s t o ra ge , shipping, cleaning and preservation. Procedures shall specify equipment and personnel required to accomplish the specified material control. Applicable codes, standards, specifications and personnel qualifications shall be documented. 10.6.2.5. frocess control Procedures shall be written to establish process controt when special measures are necessary to assure process quality of safety-related items. Controls shall be applied to activities such as crimping, soldering, welding, painting, cleaning and heat treating. Procedures shall specify qualifications of equipment and personnel required to perform the appropriate process control. Applicable codes, standards, specifications and personnel qualifications shall be documented. 10.6.3. Inspection and Corrective Actions 10.6.3.1. Inspectton Prorram. An inspection program shall be established for safety related items or activities. The inspection program shall apply to construction, procurements, experiment equipment fabrication, and modifications that effect safety-related structures, systems, or components. Persons delegated to perform inspections shall not be the same person involved in the safety-related activity but may be from the same organization. The inspection program will consist of written procedures that will include, as appropriate, procedures specifying characteristics to be inspected, acceptance criteria and inspection hold points. 10-25 f
SAR 5/91 Procedures should provide for- identification cf inspected and tested items. Provisions shall be made to clear:y identify non conforming items from conforming items. In situations that inspections are not advantageous, a description shall be - provided for monitoring actions. Procedures shall be written for in service inspections of safety-related structures, systems or components. 10.6.3,2. Test Program A test-program shall be established for safety-related items or activities. The test program shall apply to prototype qualifications, installation proofs and functional tests. Testing shall be performed in accordance with _ acceptance criteria derived f rom design or procurement documents. The test program will consist of written procedures that will include, as appropriate, procedures that specify acceptance criteria, monitoring requirements, equipment _ required, personnel qualifications, environmental conditions, data acquisition, and documentation of results. 10.6.3.3. Heasurine and Test Eau toment; Measurement tools, gages, instruments, and other measuring or test
-devices that measure critical parameters of safety-related items shall be identified. Provisions foc identified measuring and test devices shall include availability, adj us tment , calibration and accuracy as required for each application. Test equipment will be identified.
10.6.3.4. Non-Conforminn Material and Parts Non-conforming-materials and parts. associated with safety related _ structures,= systems or components shall be identified. The disposition such as acceptance, repair, rework or rejection of parts from safety-related functions _ w ill be determined by the person - responsible for document _ control. Repair or reworked parts will be removed or labeled until accepted. Rejected parts will be removed and _ labeled. The disposition of-non conforming materials will be documented. 10.6.3.5. Corrective Action
. Documen*;ation of specified quality control or assurance documents shall provide evidence of quality of safety-related items. Significant deviations from acceptable _ quality, repeated quality problems or unresolved quality issues shall be _ noted and reported in writing to administrative management personnel. It should be recognized that a determination of a quality problem may be subjective and should include evaluation of the documented quality requirements relative to the impact on the safety-related nature of the item.
( 10-26
. _ _.___m _ . - -__-___-- - _ _ _ . _ . - . _ . - . _ . _ _ _ _
SAR 5/91 10.6.3.6. . hnerirnental Eauipment Design, construction, modification, inspection, testing and maintenance of experimental equipment shall be subject to this quality assurance program to the extent that these activities are safety-related. 10.6.3.7. Replacements. Modifications. or changes Insofar as possible, the replacement, modification, or change to
- structures, systems or components with a safety-related function shall be documented as meeting the requirements of the original structure, system - or component. Evaluation shculd establish a performance and reliability equivalent or exceeding the original.
10.6.4. Records and Audits 10.6.4.1, Ouality Assurance Records Records that document quality of safety-related i t e.as or activities are identified according to Table 10 3. . The records identified consist of _ inspection and test results, quality assurance reviews, quality assurance procedures and engineering analysis in support of design modifications or changes. The records shall be retained with as-built drawings, manuals and other records of important f facility-and system-information. The retention period is to be the life of- the facility or system for most , if not all, safety-related items. The retention period ..is indicative-of the expectation that items which affect safety related to as TRICA reactor are integrally related to the reactor, instrumentation and facility design and should persist for the system or facility life. 10.6.4.2. Audits An audit shall be conducted to examine the records and function of the quality assurance program, Audits will occur within two years of the QA Program activities by designated persons that were not directly responsible for the audited functions. Written procedures Table 10 4, for the audit will be considered part of the Quality Assurance Program. A report of the audit results, actions to resolve deficiencies and
. evaluation of the program will be made to a facility operations committee and university administrative management, and maintained with other Quality Assurance Program documents.
s 10-27
. . . ~ . . . __. - --- - . . . . . . _ _ ., -
a l l SAR 5/91 ') i .) f Table 10 4 QUALITY ASSURANCE PROGRAM r AUDIT PROCEDURES i
- 1. Designate a person or persons responsible to perform the program audit.
- 2. Determine the date of the previous audit.
3 Review the Quality Assurance Program document. 4, txamine the list of safety related items. ,
- 5. Note additions to the safety related items.
- 6. Identify records applicable to additional items.
- 7. Determine the location of all indicated records.
B. Review records for abnormalities and completeness.
- 9. Prepare stateitent that evaluates functions of Quality Assurance Program.
- 10. Report findings ut and program functions to j operations commitra . inanagement ,
l l l l 10 28
SAR $/91 . t i 10.7. STARTUP PROGRAM t Startup and t >r ng of the Balcones Research Center TRIGA facility shall be performec. oy personnel of The University of Texas with consultation of the reactor manufacturer, General Atomics (GA Technologies). The University of Texas has accumulated nearly 25 years operation expertence with a TRIGA reactor prior to the new facility proposal. More than twenty TRIGA type reactors, eleven in the U.S., , with power levels of one roegawatt or more have been produced by General Atomics (GA Technologies) Training of university personnel associated with startup activities at the new facility will consist of the relicensing of at least two operators from the current facility that have certified senior operator perrnits. Training of an additional operator or retra.ining of a current operator by the manufacturer should occur to provide an effective transfer of the manufacturer's experience to the owner-operator. One or more of the certified operators will have a bachelor's or advanced degree in a field of engineering. A checkout and evaluation plan for the instrumentation, control, and safety system will assure a complete test of the system. Acceptance of the instrumentation, control, and safety system will also depend on the completion of test and acceptance programs at the site of the manufacturer. This test and acceptance at both sites is necessary for proper verification of installation, in this case of substantial technological changes in the system design. A subst.antial change of the system design from analog to digital provides for numerous improvements in perfo;mance, but will also be subject to design limitations that are characteristic of such changes. The startup program is to consist of five phases, beginning with the storage of nuclear fuel on site to the rcporting of observed reactor parameters. At each phase written procedures, check lists and other documents shall be developed for activities or tneasurements that will have significant importance to safety or operation. Documentation shall include information required by the various programs to be implemented at the facility, such. as operntor qualifications, radiological protection, fire protection, and quality assurance, plus operating procedures and other requirements of license authorizations. The startup program is to be divided into the following phases:
- a. Storage of fuel and acquisition of components,
- b. Tests of systetrs before core loading,
- c. Fuel loading and core criticality,
- d. Tests subsequent to core criticality and
- e. Acceptance of core operation.
10 29
- SAR S/91 10.7.1. Storage of ruel and Acqqlgition of CWpanenta provisions for the storage fuel and components for the reactor f acility at the completion of the facility construction shall require the limited ireptement ation of administrative controls. A license authorization for the possession and storage of special nuclear materials and oth r radioactive components, source materials or by.
product tea te rials , will be obtained and materials relocated to the facility. Storage of non radioactive components, storage of other reactor components and instrumentation, and assembly of facility systems will be performed in the initimi startup phase. 10.7.2, Icsts of Systems Before core Luding Facility systems, auxiliary systems, and reactor systema or physical parameters shall be tested for the appropriate opstating conditions prior to fuel transfer into the reactor core. Fuel may be loaded into the pool. during this phase. Systems shall be tested according to designated specifications, when applicable, and acceptable operat. ion shall be establis.hed before core loading proceeds. Facility systems to be tested should include security, fire, communication, and ventilation sys terna . Auxiliary systems to be tested should include radiation monitoring, pool coolant, alarm, and interlock systee. Reactor systems to be tested will include the instrument and control system, and verification of physical specifications for assernbly, and operation of reactor components. Some syste,ms or components that do not meet specifications and are not required for operatio" may be deferred , l for acceptance to a later startup progran phase. 10.7.3. Core Lond For Initid Criticalliy Continuous operation os coolant system, incertion of the neutron source, installation of the cobalt 60 IrrMiator, anti movement of fuel into the core will begin the core bad startup program phase. Certain verifications of instrumentation and control systen functions will be completed before init1 Alization of an approach to critical experiment by standard reciprocal source multiplication factor measurements. Rod worth values shall be estimated from the core loading procedures. 10.7.4 Igsts Subseauent to Core Cru h lily Rod . calibration shall be determined by positive period measurements before reactor operation at power levels affected by the power coef ficient. Next, an intermedfate power calibration shall be made. along with-an evaluation of the fuel temperature as measured by an instrumented fuel element. Last, the fuel loeding of the core should be adjusted for full power operation and operation of the cooling system verified at _ power. Any variation of core parameters significantly l different than predicted by calculations or experience shall be resolveo l-during this startup program phase. 10 30
SAR $/91 1 i
- 10,7.5, M ecotance for OperatiEn l l
I ' 1 The final startup program phase shall consist of the resolution of ! all deviations from specifications. Deviations should be resolved as j specified for quality assurance or other methods determined to be l a c c e,t t abic . Three months alter completion of requisite initial startup j and power esentation testing of the reactor, or nine months after l j initial criticality, a w r i t t .v n report shall be submitted to licensing ; 4 authorities. The report shall include a summary of the following: ; i
- a. Description of measured values of operating conditions or characteristics obtained and comparison of these values with ,
l ] design predictions or specifications. ;
- b. Description of maj or corrective actions taken to obtain )
satisfactory operation.
- c. Re evaluation of salcty analysis where measured values ;
! Indicate substantial variance f rom those values used iti the ; l Safety Analysis Report. , l l Results of the startup program shall becc.mc c supplement to The - l University of Texas T!t1CA Saf ety Analypis Report. Chapter 12 of the 1 ! report will contain results of the sta'. tup program. !O r 4 i t i l- l i t 1 i t t p l l . I I e 10 31 . l
I SAR S/91 i i ! Chapter 10 References l l 1. " Standard for Administrative Controls" ANSI /ANS - 15.18 1979. i 4
- 2. " Selection and Training of Personnel for Research Reactors".
ANSl/ANS - 15.4 1970 (N380).
- 3. " Review of Experiments f or Research Reactors." . ANS1/ANS 15.6 1974 (N401).
- j. 4. " Records and Reports for Rescatch Reactors". ANSI /ANS - l$.3 1974 f (N399). I l
J r 5. " Radiological Cont rol at Research Reactor Facilities", ANSI /ANS. J 15.11 1977 (N628) l . 1 i 6. " Design Objectivet for and Monitoring of Systems Controlling t ! -Research Reactor Effluents". ANSl/ANS - 15.12 1977 (N647) , l
- 7. " Standard for Fire Protection Prograin Critaria for Research Reactors", ANSI /ANS - 15.17 (1981). ,
I 8. Nuclear Reagandi Reac tnrn 1983, Natlonal Fire l'ro t ec t ion Association, Inc., NFPA 802. i , 9. " Quality Ansurance Requirements for Research Heactors", Nuclear l Regulatory Culde 2.5 (77/05).
- 10. " Quality Assurance l'rogram Requirements for Research I Reactors",ANS1/ANS -
15.8 - 1976 (N407). l r
- 11. " Nuclear Regulatory Commission" . Chapter 10 U.S. Code of Federal ,
Regulations ; e L i f 10-32
SAR 5/91 Chapter 11 f SAFETY ANALYSIS in this section an analysis of abnormal operating conditions will be made with conclusions concerning the effects on safety to the reactor, the public, and the operations personnel, as a consequence of any abnormal operations. The abnormal conditions that will be analyzed are: L
- a. Reactivity accident. ,
- b. Loss of reactor coolant.
- c. Fission product release frora clad rupture.
11.1 REACTIVITY ACCIDENT 11.1.1. Summary Rapid insertion of reactivity into a TRICA reactor is a designed l
~
feature of the fuel performance [1]. Thus, most plausible reactivity accidents do not subject the fuel to conditions more severe than normal ' operating situations. postulated accidents for other undetermined scenarios also are predicted not to exceed fuel element safety conditions. The standard TRICA fuel element of U ZrH (ll/Zr; 1.6) is composed of a stabic gamma phase ZrH that does not undergo a phase transition at temperatures less than about 1250'c [2] . pulsing limits for fuel elements clad-in stainless steel are set by the hydrogen equilibrium pressure - within the fuel element. This pressure is a function of temperature and must not exceed the rupture stress of the fuel element cledding. For the stainless ateel cladding (0.02 inch thick), the rupture pressure has been measured to be 1800 psi at 100'C. The fuel temperature at which the equilibrium hydrogen pressure will be 1800 ps! is _ about 1150'C. The average and peak fuel temperatures at 1.5 Mw steady state operation are about 220'C and 400'C, occurring voll below the limit. The average and peak fuel temperatures _ occurring af ter a ' 2.8% reactivity insertion at respective power levels of 1 kW and 880 kW are also less than 1150'C. Values of 376'c and 843*C are predicted for a low power insertion and values of 460'C and 795'C are calculated for a high-power insertlon. Two reactivity accident scenarios are presented. The first is the insertion of ?.8% reactivity at zero power by the sudden removal of the maximum worth control rod. The second is the sudden removal of the same 2,8% reactivity with the reactor operating at a power lesel equivalent to the balance of the core excess reactivity. It is unlikely that movement of reactor fuel or experiments would lead to the _ postulated accidents. Movements'of control rods for the first case are controlled administratively while movements of control rods for the second case are prevented by control circuit design. I 11-1 l 1 -_ ,, _ m __ __ _ . _ _ _ _ _ _ _ . . _ . , _ . _ _ _ _ _ _ _ _ _
i i
\
t SAR $/91 j The analysis of a four dollar (2.8% ok/k) pulse insertion also provides information about accidents with experiment systems. Provided the total worth of reactor experiments are limited to $3.00 no experiment movement could generate the postulated accident. Pulse powers predicted from kinetics formulations based on the Fuchs Nntdheim+Scalletar model are displayed in Figure 11 1. Pulse shape, energy and temperature for $3 and $4 pulse insertions are shown. 11.1.2. Analysis of 2.8% Insertion at 1 kW. l A rapid insert.lon of excess reactivity in the reactor system is I postulated. The method of inserting this reactivity is t'arough the rapid removal of a control rod. This reactivity insertion is the most serious that could occur. i t. is also the normal pulsing condition and the analysis is presented here as a point of information since it is not actually an accident condition. The sequence of events leading to ths postulated reactivity accident is:
- a. The reactor is just critical c st a low ,,over level (less than 1 kW).
- b. Upward force is applied to a h'gh worth control rod causing it to be ejected from the core and to introduce an excess reactivity of $4.00.
The consequences of the above sequence of events are:
- a. Reactor power is increased to a maximum power of approximately 4220 MV.
- b. A maximum energy release of approximately 36 HV sec is reached when the maximum fuel temperature of 843*C is reached.
- c. S t.resse s are predicted in the stainless steel cladding oi !
approximately - 2940 psi. These pressures are caused by expansion of the air and fission product gases and the hydrogen release from the fuel material. Neither of the preceding stress values will cause cladding rupture. The analysis of this accident is conservative in a number of ways, some of whLEa have been indicated in the-reactor design bases (Chapter 4), For example, the equilibrium pressure of hydrogen over the fuel is not achieved during a pulse or _ step insertion of reactivity, i 11-2
1
; SAR 5/91 4
a i 1 O + 10'4 _ ;
- l l
! 10-3__ ! A Tutperature
'C i 4 10'2 .
l 1 - Energy ; j '~~ N-sec i 10^!- : [ [ ' i
~~
twaar
! : ~
N l 10 o..__uu,_p d .L4_i..u s a a_4_u. >>1 4 . i- O .04 00 12 ,16 .2 l l i ) $4 Itine ' Initial Ih er i Kw i { ! ( i 10'4__ 1 - ; 10-3._ _ A Taiperature l * 'C l 10 < - 2 ._ - i : Energy
~
~'
l Ma-sec , l 10^!.[: L Power
~
Ma l l l -10~0 $_u ./. , , . . i 3 i , , , , , q_ , . iy l 0 .04 .00 .12 .16 1 i S4 Pulse Initial Power 880 Kw l CALCULATED PULSE SMAPE, ENERGY nND TEMPERATURE
@ Figure 11-1 f l.
l > 11-3 ,
I SAR $/51 It was assumed that the reactor is just critical at a low power level with a fuel and coolant temperature of 25'C. Additional - J nput parameters are summarized in Table 11.1. ! 5 Table 11 1 - REACTIVITY TRANSIENT INPUT PARAMdERS Reactivity insertion, $ 4.0 Temperaturn coefficient,
. prompt (6k/k)/'C 1.1 x 10-4 Delayed neutron fraction D % 0.70 Neutron lifetime f, psec 41 licat capacity C p, watt.sec/ clement 817 + 1.6 Tf uel The computations leading to these conclusions are deterininrd by the following lumped parameter analysis. The Fuchs Nordheim mo<4e1 for reactor dynamics yields the coupled set of differential equations:
i 6P/6t - (6k - aT) P (1) C 6T/6t - P Po . (2) with .C - Co 4 CT 1 (3) where f - Prompt neutron lifetime, sec, P- Power level, (Po - initial power), watts. ok - Reactivity above prompt critical, 6k/k D. L a- Magnitude of the negative temperature coefficient, 'C'I, T- Temperature (avg. over fuel) above the equilibrium temperature at Po, 'C, C- Heat capacity of the fuel in the core, W-sec/*C, Co - Heat capacity at the equilibrium temperature corresponding to Po, W-sec/'C, 2 C1 - Rate of chan6e of heat capacity with temperature U-sec/'C , l l-l 11 4 l
, 1 - , ._ _.-~.,, . _ , . ,. m. .._,,y y , . . , , m.,, m .-.___-,,._,c__..,_. .._..._,_._.m m ....,_-.m.m.,m __.,.m
I SAR 5/91 9 The abave luroped pararnoter system neglects heat transfer and delayed neutron effects and averages space and neutron energy variations Combining equations so that all coefficients are assumed constant. dP (6k aT) (Co + CIT) P
- - (4) dT f (P - Po)
Integrating, using the condition that T - 0 when P - Po, yields f (( P - Po) - Po In(P/Po) ) - i 2 Tl 6k Co - ( n Co C1 6k )T/2 - a C1 T /3 ) (5) Maximurn (or ininirnum) temperatures occur when the pulse initiates and , after culmination of the pulse such that P - Po - 0 , P - Po , and then, 2 T( 6kCo - (oCo C1 6k)T/2 a C1 T /3 1-0 (6) The roots of this equation are Ta - 3/8 (o 1) 11 + [1 + 16/3 o/(o -1)2)l/2) Tr (7) where a - aco /6kC1, Tt - 2 ok/a, and the positive sign is taken for the square root if a s 1. From the fuel heat capacity the core heat capacity with 90 elernents is l Co - 817 4 1.6 T 3W sec/*C element (8) r i - 857 V sec/*C element x 90 elements / core
- 7.71 x 100 W sec/'C.
im For the insertion of $4,00 of reactivity a value of 6k equal-to-0,021 - ($4.00 - $1.00) _ (0.7%/S), 1.1 x 10 o - (857/1.6) - 2.81 (9) and Tg - 2 (2.1 x 10-2)/1.1 x 10~4 - 382'c , (10)- 11-5
i SAR $/91 Thus t T, - 3/8 (L 81 1) (1 - [1 + 16/3 2.81/(2.81 - 1)2)1/2) Tr (11) 0.92 Tr - 351*C . Therefore, at the conclusion of the pulse the average fuel temperature i will be T,, - 352 4 25 - 376'C . (12) q To.dctermine the maximum temperature in the hottest fuel element, the average energy release is determined and then multiplied by the peak , power ratio to obtain the maximum energy release in the center element. The peak power ratio includes radial, axial and element peaking factors. Then one returns to the energy versus temperature equation to determine the maximum temperature. 1.e t E equal the energy necessary to raise the average core temperature from the temperature at the initial power level to the temperature at the final power icvel- . Then f
. T, . T, ,
E- C dT - [ Co + C1 T ]dT (13) 0 . 0 E - Co Ta+C1 Ta /2 (14) 7 E - 3.57 x 10 watt sec For a peak t.o average power ratio of 2.2 and an element peaking factor of 1.4 the energy release of the element producing the peak power is 3.1E or 110 Hw-sec. A peak temperature is calculated by substituting this energy into the previous equation and solving for the temperature. Tp - Co/C 1 + [(Co/C1 )2 + 2(3.1E/C1 ))1/2 (13) l_ Tp - 818'c l with l l- -Tss - 818 + 25 - 843'C During the time of peak fuel temperature, the stress on the clad ~ l- from the pressure produced by the expansion of air and fission product gases and the hydrogen releasco from the_ fuel is less than the strength of the clad material and therefore there is no loss of clad integrity. l The partial pressure exerted by gases is Pt-Nt RT/V , (16) 11-6 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . - . ..___...._,:.u.__ . . . . _ . . . . _ ,..,_ ,._ _.._ _
. _ . - . ._.___m.___._...______..__
( SAR 5/91 i where initially the volume, V, is taken as a 1/8 in. space between the fuel and reflector end picco. This result is conservative since the ' porosity of the graphite reflector of 201 is neglected.
. The voitune then is V - nr2h a(1,80)2 0,317 - 3.23 cm 3 . (17)
The partial pressurn of the air in the element is RT pair - -- x 103 - 4.46 x 10 5 RT . (18) 2.24 Calculation of the fission product gases in a fuel element is determined- by burnup. For an element operated at three times the 4.5 2 MW days discus. sed in Section 11.2, a total of 0.016 moles of stable and radioactive gases are produced. If the releas.e fraction is taken as 0015% as discussed in Section 11.3, then 3.22 x 10 21
* ~ * "'
Nrp ' 6,02 x 10 23 From this, one obtains, Prp - 7,45 x 10 8 RT , (20) The total pressure exerted by the air and fission products is Pt - (4.46 4 0.007) x 10'b RT - 4.47 x 10 5 M (21) P1-Pair - Also we have Pair - 14.7 T/237 psi . (22) { As an upper limit, assuming an air temperature equal to the peak fuel temperature of 843'c or 1116*K , one obtains P1 - (14.7) 1116/273 - 60,1 psi . (23) The equilibrium hydrogen pressure over Zrti (ll/Zr; 1,6) at 843*C is 20 psi. The total internal pressure then is Pt-Ph+P1 - 80 psi , (24) Assumin6 no expansion of the clad, the stress produced in the clad by this pressure is S- rPt /t - 0.735 Pt /0.020 - 36,75 Pt (25)
- (36,75) (80) - 2940 psi, 11-7
SAR 5/91 For a reactivity insertion of $4.00 the clad surf ace ten.perature would be approximately equal to the saturation temperature of the water j which is 113*C at a pressure of 23.4 psia. At this temperature, the ultimate tensile strength for type 304 stainless steel is greater than 60,000 psi with a yield stress of approximately 36,000 psi. Comparing this strength with the stress applied to the cladding during the reactivity insertion, it is seen that the strength of the material far exceeds the stress which would be produced. Therefore there would be no loss of clad integrity or damage to the fuel as a result of the reactivity accident. 11.1.3. Annivsts of 7.8% Insertion at 880 kW. The reactivit-y accident ceasidered here would take place in the ' following manner. Inittally, the reactor is cold cican with all control ' rods inserted. The reactor is landed with 4.9% 6k/k excess reactivity and the pool coolant. is at a temperature of 42*C. Thic accident requires someone deliberately violating the operating license and several interlocks and scrams. The sequence of events leading to the postulated reactivity accident is:
- a. The operator slowly withdraws all the control rods except the **uimum worth rod, until all the ro?.s are completely out and tne reactor is operating at a high rteady state power.
- b. Upward force is applied to the maximum worth rod ejecting it :
(by some means impossible to conceive) from the hot operating reactor. The consequences of the above sequence of events are:
- a. Reactor power and fuel temperatures -are increased by the compensated reactivity of $3.00 (that l's 4.9% - 2.8% - 2.1%
- $3.00) to levels of 880 kW with fuel temperatures of l 380'C, peak, and 207'C, average.
- . h. A prompt insertton of 2.8% 6k/k results in an average temperature in the core of 460'C and a peak temperature of l
795'C.
- c. Stresses are predicted in the clad of about 2100 psi. Even if the clad were at the maximum fuel temperature this stress is a factor of ten below the ultimate strength of the clad.
The analysis of this accident is conservative as described- in the previous; accident case. The values do not exceed equilibrium element conditions and pressure. Calculations include finite reactivity
-insertion time, delayed neutrons and heat transfer.
11-8_
SAR $/91 It is assurned that the reactor power level is 880 kW ($3.00 of power coefficient); the average fuel temperature is (165 + 42)'C; the peak fuel temperature is (338 + 42)'C; with a pool coolant temperature of 42'C assumed. Values of the input parameters are summarized in Table 11-2. Table 11 2 REACTIVITY TRANSil'.NT INPUT PARAMETERS Reactivity insertion, $ 4.0 Teroperature enef ficient , prompt (6k/k)/'C 1.1 x 10 4 delayed neutron fraction B, % 0.70 neutron lifetime 1, psec 41 ileat capacity, watt sec/ element fuel C p , at O'c 817 fuel C p, temperature dependence 1.6 Tfuel water C p , at 25'C 879 Thermal resistance, 'C/MW: fuel to cooling channel 5.29 x 10 0 cooling to ,iool 1,42 x 10 3 A computer program was used to calculate the energy release in the t.rans ie nt , The program is a one-dimensional combined reactor kinetics heat transfer program that works extremely well for reactor transients in which detailed heat transfer analysis is not required, Delayed neutrons and finite reactivity insertion time are included - in the program. Using the parameters given above it was found that the addition of
$4.00 reactivity (2.8% 6k/k) from an average fuel temperature of 207'C -
the average fuel temperature at 880 kw) produced an energy release of 32 Mw-sec in the 90 element core. The energy density at the axial midplane of the maximum power-density element, E, m is: Em - 1.1 Pr Pa E/" where Pr - relative power jn element,
, Pa - axial peaking factor, E - energy release in the transient, 11-9
SAR $/91 n - number of elements, and the factor of 1.1 acounts for uncertainties. j In Chapter 4 the radial power distribution within a fuel element is abovn. The energy deposited at radius r per unit volume is: Er f(r) E m where f(r) is taken from data in Chapter 4. The after pulse temperature at the radial distance r is 6 t ven by: Tp (r) - Co/C 1 + ((Co/C1 )2 + 2Er /C l y2 , where Co
- 8.17 x 10~0 Mw.sec/'C and C1
- 1.6 x 10 6 gg,,,cj(.C)2, calculation of the final element fuel temperature is accomplished by adding the temperature of the pulse deposited energy to the fuel ,
element temperature prior to the pulse. The radial temperature distribution in the fuel prior to the initiation of the transient is given by: T(r)-Ts - (Tm-T.) [1 r* - r'In(1/r')]/r o' where T(r) - temperature at radial distance from center r, T3 - temperature at fuel surface at axial center, 'C Tm - maximum temperature in element, 'C r,ro - ra:iial position and radius of fuel, cm, r' - (r/r o)2 r,' c - [1 r"-r"In(1/r")) r" - (rt/r o)2 and ri - inside fuel radius, em,
~ In Figure 11-2 there is shown - the before pulse and af ter pulse temperature in the axial midplane of the maximum power density element.
As can be seen the maximem temperature occurs at the periphery of the fuel. The adiabatic value is 795'C. In Chapter 4 a plot is shown of pulse temperature distribution as a function of time. This figuro shows the typical . dependence of temperature as heat flows quickly toward the fuel center and toward the clad. 11-10 . _ . _ . - . _ _ _ _ _ _ _ _ . . _ _ . . . _ _ _ _ _ _ ~ . _ . . . . . . _ _._ _._--_- _ _ _ . , . . _ . _
. - ~ . . _ - - - - _ ~-
SAR 5/91 For a iuel temperature of 800'C the equilibrium hydrogen pressure over the fuel would be less than 15 psi and the pressure exerted by air within the element would be less than 60 psi even if it were at the maximum fuel temperature. The stress imposed on the clad by 75 psi would be about 2800 pai. The ul t irna t e strengtn of the clad is ove 20,000 psi at 800*C, Therefore one can conclude that the clad integrity would not he compromised as a consequence of either of these events. A similar analysis was made in which the reactor was assumed to be operating at 1.5 Mw. In this case only $2.79 (1.95% 6k/k) was available to be inserted. The peak before pulse temperature was 475'C and the reactivity insertion resulted in an energy release of almost 30 Mw sec. The peak af ter pulse adiabatic temperature was 823'C which occurv' at the inner fuel radius because of the high initial temperature at that point. I
- 11. 2. LOSS OF RF. ACTOR C001 ANT 11.2.1. Sim ary The reactor will operate at a calculated maximum power density of 18 kW/ element when the reactor power is 1000 kW and there are 90 e t ernent s in the core, all of which are standard TRICA fuel. If the coolant is lost immediately after reactor shutdown, the fuel temperature, indleated in Figure 11-3, will rise to a maximurn value of 750*C, The stress imposed on the fuel element clad by the internal gas pressure, presented in Figure 11 4 is about 2300 psi when the fuel and clad temperature is 750'C and the yield stress for the clad is about 19,500 psi. Therefore. -it can be concluded that the postulated loss of-coolant accident will not result in any damage to the fuel, will not result in release of fission products to the environment, and will not equire emergency cooling.
If the reactor tank is drained of water, the fission product decay heat will be removed through the natural convective flow of air up through the reactor core. If the decay heat production le sufficiently low because of a low fission product inventory or a long interval between reactor shutdown and coolant loss, the _ flow of air will be enough to maintain the fuel at a temperature at which the fuel niements are undamaged. The following analysis shows that;
- a. The maxin. im temperature to which the fuel can increase is 900*C without substantial yleiding of the clad or subsequent release of fission products,
- b. This temperature will never be exceeded under any conditions of coolant loss if the maximum operating power density-is 22 kW/ element or less.
- c. For maximum operating power densities greater than 22 kW/ element, emergency cooling can be provided to ensure that the fuel element temperature does not exceed 900*C.
11-11
SAR 5/91 i 800 - 700 . 600 - 1 vO 500 - g -i w w
- )400 s -
t x W CL r w
300 _
200 - 100 - 1.0 2.0 RAD) AL - DIST ANCE - CM i FUEL TEMPERATURE DISTRIBUTION i BEFORE AND AFTER PUI SE i Figure 11-2 , I l 11-12
, . - - .e,.pf..,--mp., ~.myy .r v w *** + w
'-m*'TM r"wme-m=-=*=='Wr"P=W-"d'=e"'#'"r'*"-'"*'*e"'*"*******' ' * * * " " * * * * - * ' * * * * * ^ ' ^ " " ' ' " * " ' ' * * * * * *
.. _,_- ~ . - - _ - . . . . - - _ _ . . _ - - - . ~ = - , ~ . . . . . _ _ _ _ , , - - - - - . ~ . . ~.~ .._ .... - ._,
SAR 5/91 0 ; 2000 1800 - C00 LING TlHE (SEC 1600 - Maalamas f eel tempes.ture ..r.u. 1100 -
.ei.,i... .e . i.. e.,..,s t.. wor deast t?
es.. 6. m .. ....... . we. - ..... i... t .. i... 3 io 1200 - g ll 1000 - n 800 - 3
-5 600 - /
le00 -
,/'
200 -
- 4 g i i n i t i I 1 0 40 45 5 to 15 20 25 30 35 0
OPERATING POVER DENSITY-KW/ ELEMENT I I IllEL TEMPERATURE AND POWER DENSITY ! FOR EI,EMENT C001,1NG TIMES O Fi gu re 11-3 I i I 11-13 1 l 1 t -- ...---.,,....__-,,-....._..,...,_,..._.,__-.__--,,,,-_,_m. ---.,_.,..-.,.....~.~..__....__,,,,__-,.,,..._.I
SAR 5/91 i l b 10 - ULTIMATE STRENGTH l
'%s
's N N
YlELD STRENGTH \ g , N
\
10 - g N
\
3 \
\
6. N N N
= N b \
\
STRE55 lHP05ED ON CLAD 10 - si....in . 4 .,,it.4 ...... .. . r . .. . i ..r.sur. 9 mi,s3 ru.a. ru.i ew o.4 .
.. i ... ivi.
10 = ' ' ' ' ' ' ' 400 600 600 1000 1200 TEMPERATURE (*C) U-ZR}i (1.6) STRENGT11 AND STRESS VERSUS TEMPERATURE Figure 11-4 I 11-14
1 l SAR $/91 O If power densities exceed the limit for immediate coolant loss, emer6ency cooling vill depend on the power density. The required l l emergency cooling time as a function of maximum operating power density i la s,hown in Figure 11 5. Maxitnum operating power density is to be used I l to determine emergency cooling times. 11.2.2. EMC1112ttratutLittuLElau! I nt e r r l t y l l l, l The st rengt h of the fuel element clad is a function of its i ! t ernpe ra t ure . The strens imposed on the clad is a function of the fuel i , temperature as well as the hydrogen-to zirconium ratio, the fuel burnun, l l and the free can volume within the el einent . In t' , analpis of the ; ! stresses imposed on the clad and s t rengt h of the clad the following l assumptions will be made l I r ' i a. The fuel and clad are at the same tempe ature. I i ! b. The hydrogen to zirconium ratio is 1 65, i
- c. The free volume within the element is represented by a space :
l ) 1/8 in. high within the clad.
- t j d, The reactor contains fuel that han experienced burnup equivalent to only about 4.$ MW days.
The fuel element internal pressure p in given by i i P-Ph iPip + Pair , ! where $ Ph - hydrogen pressure, ; r [ Pfp - pressure exerted by volatile fission product s, L Pair - pressure exerted by t rapped air, t For hydrogen to zirconium ration greater than about 1. 58 the equilibriuin a hydrogen pressure can be approximated by pg - exp [1.767 + 10.3014x 19740.37/(Tg)) (26) { a (atmospheres). l f I where x - ratio of hydrogen atoms to zirconium atoms, and-t Tg - fuel temperature (*K). This expression was derived from least square fits to the data of Dec [ and Simnad (3]. For Zril (II/Zr; 1.65) the hydrogen pressure becomes l 1 Ph - 1.410 x 10 8exp [-19740. 37/(Tg)1 (atmospheres) . 1 2 h 11-15 ,
i l i SAR $/91 J
?
O - : i i t a f 1 10'I. T'a = $00 600 700 00 90 100 a a .t i 8 a 5 , U 2-Q 10 - a ; i W. . . w , ! E [ U f cu..............,.... ! E ..ee... y a. : a ..u.. to. i..p.e.- :. 7 ius. .. ... p . a...ii, i l U jo). I! i
- D
, g-l T = NAI. FUCL TEMPERATURE Arif R VATER LOSS ('C) 2 10 , , y , , , 0 10 20 30 40 50 60 '/o PdWER DE4SITY-rW/[LINikT 9 COOLING TIMES ALTER REACTOR SHUTDOVN TO 1.lMIT MAX 1 HUM FUEL TEMI'ERATURE VERSUS l'0WER DENSITY 9 Figure 11-$ F 11 16
SAR $/91 9 The pressure exerted by the fission product gases is given by . l nR pfp - f - - Tg E , (27) ! EV where f - fission product release fraction, n/E - number of roolcs of gas evolved per unit of energy produced, moles /MW day, R - gas constant, 8.206 x 10 2 liters atmospheres / mole 'K, V - free volume occupied by the gases, liters, and E - total energy produced in the element, MW day. The fission product release fraction (4) is given by ; f-- fn d" (28) n 0 fn - (1.5 x 10*5 + 3.6 x 103 exp [ 1.34 x 10 /TnlI where Tn - fuel temperature in the differential volume of the element durinn normal operation, *K, fn - differential release fraction and , n - fuel volume normalized to 1. The fission product gas production rato n/E is not independent of ; power density _(neutron flux) but varies slightly with the power density, The value n/E - 0.00119 moles /MW+ day is accurate to within a few percent _ over the - range from, a -few kilowatts per element to well over 40 kW/ element. The free volume occupied by the gases is assumed to be a space 1/8 in. (0.3175-cm) high at the top of the fuel so that 2 (29) i V - 0.3175 nrt , i where ri - inside radius of the clad (1.822 cm). For standard TRIGA fuel the maximum burnup is about 4,5 MW. days / clement. Pressure exerted by fission product gases is not significant. The air trapped within the fuel element clad would exert a pressure Pair - RTg/22.4 , (30) l . 11 17 _. . _._ -..--_ __.,.___ _.._._.. _._ .,__a-_ _,_ _ _ _. _ _ _ _ _ . _ _ .
h i ; SAR 5/91 ; 1 l i where it is assumed t hat the initial specific volume of the air (??.4 f } ( j liters / moles) is present at the t inic of the loss of coolant. Actually,
- the air forms oxides and nitrides with the zirconium no that after [
i relatively short operation the air is no longer present in the iree ! l volume inside the fuel element clad. ' i 2 l'or Zrli (ll/Zr; 1.6) f uel hurned up to 4.5 MW days /elenient , with a l l ! f waxiinum ope rat ing tenperatute of 600*C, the int ernal pressure as a f unct ton of maxinitun f uel tetuperat ure Tg is p - 1.410 x 10 0exp (- 19 /40. 3 //Tg) # 3.66 x 10'3 Tg (31) [ (atmosphere) l [ r or l p - 2,0/3 x 10 exp (19/40.37/Tg)
- 9 S.38 x 10'2 Tg (ps!).
i l The strens imposed on t he clad by the gases within the free volume l loside the clad lu l S - (r c /t) p , (32) i { l'
- j. where re - clad outside radius, (1. 8 / 3 cm), a
( i ( t - clad thicl< ness (0.051 cm) . 11 the previous and inittal equation are c on,b i ne d , the stress can bc ; rewritten as S - 36.15 p (33) >
- 7.61 x 10 I" exp (-19/40.37/Tg)
- 1.9 / Tg (psi),
l>lgure 11 4 plots t h i t, imposed stress as a f unct ion of maximum fuel l towperatures. Also plot t ed are the yield and ult imate strength of the type 304 stainless s t.ec t clad, The clad uit Imate strength is not 4 exceeded i f the niaximum f uel t empeiat ure is inatot ained below about 950'c , and the yleid st rength lu not exceeded for any fuel temperatures below about 920*C slightly below the yield point and well holow the rupture
- point I1.2.3. M t e r -llei1L. IlemovallS}} owl.udoo l a!!LI.o s s I t- is assumed that the t eact or operat es cont inuosly at a const ant !
l power density level puso that the maximum inventory of fission products _is available t o produce heat after the reactor is. shut down. The _ po_we r density after reactor 1.hutdown p is given by p - 0.1 p o l(t t 10) # 2 + 0.81 (t t 2 x 107 ) 0.2) (34) i; A {1.3 cos[?,45 (0.26x - 0.5))), 3 where po - operating power density, W/cm , i W ' 18
SAR ';/01 t - timt after reactor shutdown, sec, x - distance from the hottom of the iuel rer,lon, cm. At the time that the coolant in lost irom the core the fuel and its currounding- ate assumed to be at a temperatune of 27'C. This is not necessailly true, for an accident can be postulated in which the coolant loss is the mechautum by which the reactor is shut down. (for the t.tandard non gapped tuel e l e me n t undet normal conditlens, the time to cool down l l oin operating temperatures is a matter of one to two minutes.) Alt hough r.uch an accident does not- appear to be conceivable, calculations indicate that: if it i t, a s <:ume d that the average fuel temperature at the time of coolant loss is equivalent to the operatier average fuel temperature, the n,ax imum t empe rat ure after the coolant low is not. appreciably different (?% 4% hightr) from that c alculat ed assuming 2 7'C f uel initially. The after. beat removal will be accomplished by the flow of air through the core To determlae the Ilow through the core the buoyant forces were equated to the ftiction, end, and acceleration losses in the channel as shown in the e x p re t.s l o n 6Ph - bpg # 6pc
- bpg
- Apa (M)
The buoyant torces ate h i ven by 6Pb - Pol - [ Pdx - P0 l* POI o PI f
- el l t (R) where P. the entrance, nic a n , and exit fluid densities, p0' El -
respectively, I. - the effective length of the channel ( L-L o 1.{ t 1.t ). i 10, l.g , 1t - the lengt h of the channel adjacent to the bottom end reflector, fuel, and top end refl(etor plus ten channel hydraulic diameters, iespectively. The irtetton losses in the llow channel are given by
-1 4; g,2 6pg - in -
(37)
/_, De ? r,P iA' c
where the sunana t i on is over the lower unheated l e ngt h , the heated lengt h , and the upper unheated length, and De - the hydraulic diameter (0.0601 ft), I 11-19
SAR 5/9? fyt - the friction factor (23.46/R e) 15] , Ac - the flow area through the core per element (0.0058 ft2), 0 g - 4.1/ x 10 ft/hr2 The sum of the exit and inlet losses, using appropriate expansion coetficients, in given by 2 ( >:K ) w ope
- 6pl - ,,
(38) 2gf04c' with
):K' -
Ikj (A c/Aj)2) - 1.57 The acceleration losses are given by 6pa - (1/v1 - 1/r0) (w /gA c ) (39) By substituting the appropriate e x p r e s s. l o n in Equation 35, using the deilnition of the iteynolds number, and 1. - 2.40 ft 1.n
- 0.29 ft, Lg -
1.25 ft, and 1t - 0.87 ft, one obtains (O.700/p1 0.149/PO) x 10 w2 (40)
+ (0.153pl/pg 4 0,665p/p *
- 0.l'33pp/pt) x 10~2 w
+ (1.25p 6 0.889p1 - 2.13900) - 0 with t he flow w in units of th/hr and p the visconity in units of Ib/hr-it.
The properties of air for use in 1: quat ion 40 are expressed as pi - 40/Ti (ll>/f c3) (41) alid 7 pg - 5.'/39 x 10'3 6 7.(,01 x 10 ' T1 t
- 1. ? / f. x 10'g T' i (lb/hr It),
where Tg is the appropriate temperature in 'R. The heat transfer coef f icient was calculated through the relationship Nu - 6. 3 Rai 1000 (42)
- 0.806 R 3 0.2976 Ra > 1000 .
where Nu - the Nusselt numhet- - hDc /k, 11 20
SAR 5/91 A2 Ra - the Rayleigh number De p gB6Tep /ukL h - the heat transfer coefficient, Btu /hr-ft? *F, k - the thnrmal conduct ivity of the laminar film, B t ' * ~. . r - f t ?
- F ,
B - the volumet r ic expat.s lon coe f ficient . *F'I , 6T - the temperatute rise over the channel lengt h, L(*F), cp - the speelfic heat. of air ' Btu /lb *F). The expression for the Nusselt number was derived f roin the wor k ef Sparrow, l.oe f fle r , and Hubbard [6] for laminar flow between triangular arrays of heated cylinders. The thermal conduct ivi t y tout spec i f ic he at are given by k - 2.377 x 10'0
- 7.995 x 10~'T- 4.738 x 10'" T2 (Btu /hr ft UF) (43) and c p - 2.413 x 10'l 1./80 x lo 6 7, 3,g3g x ig 8 T 2 (htu/lb *F), (44) where T is the appropriate temperature in*R, These two expressions, as well as that gin n for the dynamic viscosity of air in Equation 41, are least square fits to the data presented by Etherington [7).
TAC 2D [8), a two-dimensional transient-heat transport computer code developed by GA Technologies, was used for calculating the system temperatures after the loss of tank water. The parameters derived above were programmed int o the calculations The maximum temperatures reached by the fuel are plotted as a function of operating power density in Figure 11 3 for several cooling or delay times between reactor shutdown and loss of coolant from the core. For reactor operation with maximum power density of 18 kW/ element, or less, loss of coolant water immediately upon reactor shut down would not cause the maximum fuel temperat ure to exceed 750*C. Operation at inaxiinum power densit ies great er than 18 kW/ element will not result in fuel t emperat ures above /50*C, if the coolant loss occurs sometime after shutdown, or if emergency cooling is provided. (The t irne required between shutdown and the beginning of air cooling depends on power density.) 11 21
SAR 5/91 In Figure 11 $, the data presented in Figure 11-3 were replotted to s.how the time required for natural convective water cooling or este rr,ency cooliny,, after reactor shutdown, to produce temperatures no greater than a r,1ven value Thus, for example, for a reactor in which the max i tnuin operating power density is 2/ kW/eletoent and to limit the lettperature to 950*C, or loss, there inu s t he an interval of at least 3730 sec (or 1.04 hr) between reactor shutdown and either the loas of tank water from the cote or the cessation of emergency cooling, The 65 ininut e delay time applies to the power density of a 90 element core operated a t. 1 . ') MW hut shrinks to a ner,ligible "alue for the power density in a 100 eleinent cote 11.2.4. !hulinilon Leve13 f.v e n though the possibility of the loss of shielding water is believed to be exceedi ng.ly remote , a calculation has been performed to e v,il ua t e the radiological hazard as!.oc iat ed with this type of accident (see Table 11-3). Assuming that the reactor hat, been operating for 10 hours and 1000 hours at 1.5 14W prior to losing all of the shielding water, the radiat ton dose rates at two different locatlons are listed in the table. Time is measured f r om the conclusion of operation at 1.5 MW. Dose rates assume no water in the tank. The first location (direct radiatlon) is 6.4 meters above the unt.hielded rtactor core, near the top of the reactor tank. The second is at the top of the reactor shield; this location is shielded from direct radiation but is subject to scat tered radiat ion f rom a t hick concrete ceiling 4.6 m above the top of the reactor shield. The assumption that there is a thick concrete ceiling maximizes the reflected radiation dose No rtna l roof structures would gi/o considerably less backscattering Table 11 3 Calculated Radiation Dose katas l'o r 1.o s s o f Sh i e l d Wa t e r Direct Radiation Scattcred Radiatloa Decay R/br R/hr Time 10 hour 1000 hour 10 hour 1000 hour 1 minute 3980 4920 1.7 4.6 1 hour 9?9 1820 .87 1.7 1 day 87 681 .08 .64 1 week 10 281 .009 .26 1 month 2 104 .002 .10 I 11-22
SAR S/91 The tabulated data show that if an individual does not expose himself directly to the core he could work for approximat ely 2 hours (3 hours for 1 MW) at the top of the shield tank I day after shutdown without receiving a dose in excess of that permitted by regulations for a calendar quart er. For persons outside the hullding, the radiation from the unshielded core wnuld be col tiinated upward by the shleid structure and, therefore, would n ot give rise to a public hazard, The ite t ho d of calculation follows The core, shut down and drained of water, was t reated as a hare cylindrical source of 1 MeV photons ui uni f orm st r ength. Its diroensions were taken to be equal to those of the active core latLice. The source strength as a tunction of time was determined from Way and Wigner's (9) (Equation 45) data on it!.sion product tscay, No account ing was inade of sources other t han f ission pioduct decay gammas (i.e , ac t ivat ion gannas from the steel cladding and the aluminum gr!d plates.) or of attenuation through the fuel element end pieces and the epper grid plate. The first of these ass.umpt t ons is opt imist ic , the second conservative; the net effect is conservative The conservative assumption of a uniformly distributed source of 1 MeV photonn was balanced by not assuming any buildup in the tore An approximation of the fission product energy release term is taken as: l'(t) - 1,76 t'I'2 , (45) where l'(t) energy release in MeV/sec-fission, t - is the time after iission in seconds By int egrat ton t be total core source term is 39 t+T S(t,T) - 3.1 x 10 Po l'( t ) di (46) l
- 1.95 x 10 lI Po (1 - 11 + T/t1-0.2) t'U 2 (47) where S(t,T) - energy release in MeV/see-watt, Po - reactor power, watts, T - period of tirne at power.
l 1 11-23
- . - - . - - - . ~ - . - - _ . .
l
)
SAR 5/91 ) l' I s i The volumetric source of 1 MeV photons is i S(t2,T) ,
-Sy - - (48) l
, nrc2*c ; i l The direct dose rate at a point outside and on the axis of a I cylindrical source is given by: i
! Sy 'x e "r c ;
' Fez 2.nrdrdx (49)
D d-E 0 0 " AnR2 i where 3 Sy - source strenSth :n photens, 1 MnV)/cm -sec, K - flux-to-dp.econveratonfaccor, J 5.77 x 10 photons /cm2-see per rad /hr, , i 1 l 2Drdrdx - cylindrical volunie element, dV
- rc -core radius, 26 cia !
, xc - core height , 38 cm , ye - core attenuation coeffJcient. 0 '17 cm'l l R - distance from volume element to receiver, cm z - slant penetration in core - xR/(aox), em l a - distance from top to core to rrceiver, 640 cm ; For distan.:es far from the core (i.e. for a >> rc and x)c _ the l:- - above expression reduces to l -S r v c ' Dd
'(1 - e '#c c) (50) 4 Mcn2K l-b The scattered dose rate was calculated t' rom-23 p-Z IC o Qa D, -_6.03 x 10 (51) 2
'A K(E)x I where p
l; p - Density of scattering material, concrete, 2.3 g/cm* I Z - Ratio of average atomic number to atomic-mass l
- A of the scatterer, 0,5-l~
H 11 24 1 i --
- ._-_______.m___ - _
. . _ . _ _ . . . _ . _ . . . . ~ . . . . _ _ , _ , , , , _ , , _ . , , . _ _ , , , _ _ . , . . _ , _ , _ , , . , , , _ _ . , _ , . .,,, _
. - - - . - ~ . - . . . . . - . - - . - - - . . . . . . ~ . . - . . _ . - - - . - . . - - - . - . - - . - - . - - - _ - - . .
SAR 5/91 and loc - Incident current times cross section of beams, photons /sec. K - Photoncugrenttodoserateconversion, 2.75 x 10 photons /cm2 sec per rad /hr A - Energy o' seattered photon, Mev x - Distance from scattering point to detector, 400 cm 1 60 . Qa - - -
#0 4 #1 (cos0 0/C0801 ) 00
#0.#1 - Attenuation coefficient in scatter for incident and scattered photons, cm' , .146, .292 0 0,01- - Incident and scattered angle (mer.sured from the normal to the scatterer), 0, 25 degree 60,60 - Differential Klein Nishina scattering cross e 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 roof (i.e., so - 0) at a point directly over the care, thus
.Ioc - sow (52) where So - Sy arc /#c (53) 2 2 2 '
-l YO (r0 'XO )+ro (ro +xo ) ~ 1 W - <
sin 2
- x/2 - -
, (ro2 + xo2 ) (ro + YO ) . .
2K and r0 - Distance from the core to the top of the pool', ~6.4 m x0 -- . Half width of the pool, ~1 m yo - Half length of the pool, ~1,5 m S y,. r e, pc have already been defined. O 11-25
, . - ,_ - . - _ .__.__._-_._,--_.__..__~..a
_ .. _ . . _ _ . _ . _ _ _ . . . _ _ _ , . _ _ _ _ _ .._. - . _ _ ._ _ _ . _ _ _m. _ _ . _ _ SAR 5/91 " \ ThU energy of the scattered photons is given by EO E- (54) 1 + Eo(1 - cos0)/0.51 where Eo is the incident photon energy (1 Mev) and + is the scattering angle - x - (60 + 0 1), The dif ferential scattering cross section is given by 2 E 6a re 8 F 2 3
- -- -(- sina ) + (-) (55) 60 -2 , Eo Eo Eo
-where r e is the classical elec:ron radius - 7,818 x 10'13 cm, 11.3. FISSION PRODUCT REl. EASE In the analy ;is of fission product releases under ccident conditions, it-is assumed that a fuel element in the region of highest power density falls, The failure is assumed to occur in air af thr operation at full power for an extended period, 11.3.1, Fission Product Inven gry Tabla 11 4 gives the inventory of radioactive noble gases and halogens in the TRICA Mark II after continuous operction at 1,5 MW for four years (i.e. 6MV-y-), The estimated inventory is conservative since actual operation after 4 years is expected to be'less than 5% o f 4 MV-yrs.
11.3,2, Fission Product Release fractions 4 The release of fission products-from U-trH fuel has been studied at some length. A summary report of these. studies -[4] Indicates that . the release..from the U-ZrH (H/Zr; 1,6) fuel meat at the steady-state operating temperatures is principally through recoil into the fuel-clad gap. At high temperatures (above 400'C or 500*C), the release mechanism is through a di f fusion .-process . and is temperature dependent , unlike recoll. . For- the accident considereo here, it- is assumed that a fuel element in the region of highest power density fails in water and that the peak fuel remperature -in the element is less than-400*C, At: this
~
temperature. ttie long term release fraction would be less than 0,0015%, For'the purpose 'of this analysis it is - also assume 6 that-100% of the noble gases and 100% of the halogens'are released from the highest power density fuel element in which 2,22% of the total power is generat'ed, e 11-26
- ~ - - . , - -n- - , . - , . , , , , , . . . _ . ,.,y . , _ _ , _ _ , _ , , _ _ ,, _ _ _
- . .--. -- --- - .~. - .. - . - - _ - . . . _ . _ . - . -
SAR S/91 Table 11-4 NOBLE CAS AND llALOCENS IN Tile REACTOR Isotope Quantity (C1) Br-83 6,120 Br 84m 6,120 Br 84 12,360 Br-85 12,900 Kr-b5m 12,900 Kr-85 678 Kr 87 32,400 Kr-88 46,200 Kr 89 58,500 Kr-90 65,100 Kr 91 44,100 1 131 35,700 Xe-131m 288 1 132 53,100 1-133 86,100 Xe-133m 2,100 Xe-133 86.100 1 134 96,600 1-135 80,400
% Xe-135m 24,300 Xe-135 83,10G I-136 77,700 Xc-137 15,300 Xe-138 70,200 Xe-139 70,800 Xc-140 48,600 It is important to note that the release - f raction in accident conditions is characteristic of the normal operating temperature and not the temperature during the accident conditions. This is because the fission products released as a _ resul t of a fuel clad failure are those that have collected in the fuel clad gap during normal operation.
Other assumptions concerning catimated accident - scenario doses are: a, Assume an element fails in air such that all (100%) noble gases and halogens in the gap are effectively released.
- b. There is no plate-out of any released fission products,
- c. Af ter the failure a ventilation rate of 10 air changes per hour occurs with no air filtration.
- d. Doses are calculated assuming exposure to a semi-infinit cloud.
11-27
_-._ ___.__._ ._ .. _ _ _ _ _ _ _ _ .=. - _ _ _ _ _ . SAR 5/91 l l
-e. Doses ya calculated for release from the total core and a 1 single f .tel element (90 element core with peak to average flux cf 2.0). ,
- f. Doses ext trnal to the building are calculated by iesuming a minimum building dilution factor for releases (1.0 m/sec wind velocity with building cross section of 234 m2),
- g. Doses were also calculated for personnel in the reactot room by dumping rapidly a'small fraction of the total inventory into the room such that the continuous release is equivalent to a constant concentratlon.
The net e f fect of these assumptions is that for the singic element accident condition, the fraction of the noble gases released from the building is: fpc - 2.0 x 10~5 x 1.0 x 2.22 x 10 2
- 4 44 x 10'I , (56) and of~the halogens is:
f}i - 2.0 x 10-5 x 1.0 x 2.22 x 10-2
- 4.44 x 10'7 . (57) where a conservative release fraction of 0.002% is applied.
11.3.3. Downwind Ds_se calculation
.The minimum roof ievel dilution factor was calculated,- assuming a building cross sectional area of 234 square meters. The factor is based on mixing in the lee of the building when the wind velocity is 1 m/sec.
A dilution factor of 0.00854 seconds per cubic meter is applied. The- calculation. of whole body gamma doses and thyroid doses downwind from the point of release was accomplished through the use of the computer code .GADOSE [10]. In this code the set of differential equations describing the rate of production of an isotope through the.
-decay or its precursors and the rate of removal ' thrc' 4.h radioactive
'- decay --and removal by the ventilation system is integrated for .each member of the chain. The release rate qi to the environment for the ith isotope at time ti in hours is: qt(t) - u Q(t) (f/V)/3600- , (58)
- where Qi(t) - the release of the ith isotope in C1, I- 3 l
f/V - the building leakage-rate in (m 3/hr)/m , eg - the filter efficiency for the isotope, gi t1 I' ' 11-28 i , f i
[- ! L SAR 5/91 i , ._ i L ; l The quantity Qi(t) is the amount. of the ith isotope in the ! This quantity is given by i discharged air at the time,-t. 4 '/V)' (59) t-Qt(t) - ft Qt(0) c'(Ai 4 - i ! where Qt(0) - the inventory of the ith isotope as found in l Table 11-4, ; l i At - the decay constant for the Ith isotope, and l ! i
- ft - the releas.e f raction to the reactor-ball, ;
i i-1 l The concent ration dwnwind at- a distance x for the ith isotope is
?
calculated from i , Qt(t,x) - qi(t - t) p(x) c'Al ' , (60) > J where'r = the transit time from the release point to the dose point, I i br, 3 p(x) - the dilution factor at the distance x, sec/m l [ The'whole body gamma ray dose rate for r.he ith 1_otope, Dwt, at i ~ the distance x and time t is calculated, assuming a semi-infinite cloud,
-through the expression:
I e 9 -Dw(t,xi) - 900 Et Qi(t,x) i
, (61)
+
g i -where Ei - the average gamma ray energy per disintegration, MeV, and the , constant includes the- attenuation coefficient for air as well a.i the-L j conversion factors required. .Done rate is in units of rad /hr. j
't Internal dose rates, in this case the dose rate to the thyroid, L _
are calculated by: 1
'Dth(t,x) - 3600 B Kt Qi(t,x) , (62)
- i ;
3 [ where B_- the breathing rate, m /sec, and i Kt - the internal dose effectivity of the ith isotope, . i rem /C1. l .The values for the breathing rate are given in Table 11-5 and are taken - 1 from a published regulatory guide (11), t 5 -> ! The-average gamma ray energy per disintegration and the internal
' dose effectivity for each isotope considered are given in Table 11-6. :
i-t The decay products af these isotopes are also included in -the ; j calculation; however, their contribution to the dose rates are small anu i t he re fore the data for these isotopes were not included in the table. l l 11-29 .
SAR 5/91
\'
11.3,4. D2Wndhd Doses The whole body gamma dose and thyroid do r,e in the lee of the building are shown in Table 11-7. These doses are acceptable relative to the conservative nature of the calculations and likelihood that an accident scenario would actually lead to these rerults. Table 11 5 ASSUMED BREATillNG RATES: 3 Time (br) liteathing Hate (m /sec) O to 8 3.47 x 10'0 8 to 24 1.75 x 10'0 Over 24 2.32 x 10'0 Table 11 6 AVERAGE CAMMA RAY ENERGY AND INTERNAL DOSE EFFFCTIVITY FOR EACH FISSION l'RODUCT ISOTOPE Isotope Eg(HeV) Kt (rem /C1) Br 83 0.92 x 10'2 Br-84 1,87 1-131 0,40 1.486 x 10 6 I-112 1.96 -5.288 x 10 0 T-133 0.56 3;951 x 10 5 1 134 3.02 2.538 x 10 0 1-135 1.77 1.231 x 10 5 1 136 ~2.91 Kr-83m 0.8 x 10'3 Kr-85m 0.16 Kr-85 0.4 x 10'2 Kr 87 1.07 Kr-88 2.05 Kr-89 2.40 Xc 131m 0.82 x 10-2 Xe-133m 0.37 x 10'I Xc-133 0 29 x 10~l Xe-13Sm 0.4b Xe-13S 25 Xc 137 i.22 O Xe-138 1.57 11-30
. _ . _ . . _ . _ _ _ . _ _ _ _ _ . -. ___- - - - - - .- - e - . - - - - -
SAR 5/91 N Table 11-7 DOSES FROM FISSION PRODUCT RELEASES 4 m N pe O @ en en @en 4 m N i l l 1 i l 1 40 6 h M 00 0 4 4 6 4 4 6 4 0 0.e en o e0OO a 0o me m OO 0 C e e . e0me 0 .Oe e.e .-e 0 en 0e e e . O W M M M M M MMM M M M M M MMM w4
*d Ww @hh 46 4 4 h 4 D g g O. O. N 80e m e e o e m en en M
$M em en en me 4
* * * * = e s =4 e en M ** e e h to N @ to N N O .N D}}