ML20008E429

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Final Hazards Summary Rept,Vols 1 & 2,for Yankee Rowe
ML20008E429
Person / Time
Site: Yankee Rowe
Issue date: 07/19/1963
From:
YANKEE ATOMIC ELECTRIC CO.
To:
References
NUDOCS 8101070160
Download: ML20008E429 (882)
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YANKEE'NUCLEM1 POWER STATIOII YAIIKEE ATOMIC ELECTRIC COMPANY l-1 PART B, LICEIISE APPLICATI0!!

AEC DOCKET NO. 50-29 TECHNICAL IITFORMATION A!ID FII!AL HAZARDS SUIC!ARY REPORT

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  • Q' VOLUME I 7/O/076 to a

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ACUCTlDGE2E:.'r This report was prepared as a joint effort 'cy the per-sonnel of Yankee Atc=ic Electric Conpany, Westirfncune Electric Corporation, and Stone Ec We'oster Engineering Corporation. 3pecial technical assistance vu provided by Professor James M. Austin, Dr. Thecs J. L anpson and Mr. Samel Levin of Macsachusetts Institute of Technology.

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\\s 207 CORROSION CC?: TROL SYSTDI, PRI?'ARY PLANT 207:1 208 SAMPLI?D SYST91 208:1 209 RADI0 ACTIVE WASTE DISPOSAL SYSTDI 209:1 210 SHUTDOWN C00LI?D SYSTEM 210:1 211 VENT AND DRAIN SYSTEM, PRIMARY PLANT 211:1 t

212 SAFETY INJECTION SYSTEM 21221 213 REACTOR CONTROL SfST:24 2131 21h NUCLEAR INSTRUENTATION AND REACTOR PROT:CTICN SYSTEM 21h:1 215 RADIATION MONITORIID SYSTEM 215:1 216 VAPCR CONTAINER AT'403PHERE CONTRDL SYSTEMS 216:1 217 DECONTAMINATION SYSTE4 217:1 218 FUEL HANDLIN3 SYSTEM 218:1 219 MAIN AND AUXILIARY STEAM SYSTB4 21931 220 CCNDENSATE AND FEED WATER SYSTEM 220:1

' 's 221 CIRCULATING WATER SYSTE4 221:1

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222 WATER SUPPLY SYSTDIS 222:1 223 VENT AND DRAIM SYSTEM 223:1 22h COMPRESSED AIR SYSTE!S 22h:1 225 LUBRICATI!D OIL SYSTEM 22$:1

_226 ELECTRICAL SYSTEM 226:1 227 HEATIN3 SYSTEIS 227:1 228 VENIILATION SYSTEMS 228:1 229 FIRE PP.0TECTION SYSTE4 229:1 230 REACTOR VESSEL 230:1 231 VAPOR CONTAIN'ENT 231:1 232 RADIATION SHIELDI!D 232:1 233 TURBINE GENERA'FR 233:1 23h CONDENSER 23h:1 235 ARCHITECTURAL AND STRUCTURAL FEATURES 235:1 9

B:1 7/19/63 TABLE OF CONTENTS

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VOLUME I Page 1 NUCLEAR REACTOR DESIGN 100:1

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/100 GENERAL 100:1

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' 101 CORE MECHANICAL DESIGN 101:1 s,)

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General 101:1 YI Fuel Assenblies 101:1 4/

Control Rods 101:7

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Core Structure 101:9 NC?

Control Rod Drive Mechanism 201:10

' l Neutron Source 101:12

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102 CORE THERMAL AND HYDRAULIC DESIGN 102:1 u

General 102:1

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Core Thornal Design 102:1

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Core Hydraulic Design 102:5

'O Supplenental Core Thermal Analysis 102211 I

Thermal Analysis of Core II $h0 MdT 102:15 Thermal Analysis of Core III 102:23

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103 CCRE NUCLEAR DESIGN 103:1 Core Steady State Characteristics 103:1 General 103:1

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Criticality and Core Lifetine 103:1 Control of Core 103:8 i

General 103 8 Control Rods 103:8 Chemical Centrol 103:10 j

Control Effects on Power Distribution 103:10 Core Kinetic Characteristics 103:1h Reactivity Coefficient 103:lh Doppler Temperature Coefficient 103:15 i

Moderator Tempe:nture Coefficient 103:16 Pressure Coefficient, of Reactivity 103:19 q["

Delayed Neutron Fraction 103:22 Void Coefficient of Reactivity 103:20 L

Prompt Neutron Lifetime 103:22

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Nonuniform Plutonium Distribution 103:23 I

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Xenon Effects 103:2h j

Core Power Distribution 103:27 i

Analytical Evaluation 103:27 Experimental Evaluation 103:31

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Post Operational Nuclear Analysis of Core I 103:36

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/10h FULL CORE !!UCLEAR EXFERIME iTS 10h:1 Ij General 10h:1 Initial Core Loading 10h:1 Initial Criticality 10h:2

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Plant testing 10h:2 10$ REACTOR CORE CAPABILITY 105:1 General 10$31 Fuel Bearing Control Rod Followers 10$32 Control of Reactivity at P)wer with Chemical Poison 105:3 Multi-region Loadirg 105:3 Summary 10$33 Supplemental Analysis 10$8h pore II Capability 10$:$

Core III Capability 10$:6 106 REACTOR COOLANT CHE?!I3TRY 106:1 General 106:1 Main Coolant Water Specification 106:2 Corrosion Behavior of Yankee Materials.

106:3 Neutral pH Chemistry 106:3 Galvanic Corrosion 106th

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Water Chemistry Stability 106:6 Coolant Radioactivity 106:7 Summary 106:8 107 CORE INSTRUMENTATION 107:1 General 107:1 Mechanical Description 107:1 Flux Wire Control and Readout 107:6 Effects on Plant Oaeration 107t7 Mechanical Reliability 107 7 Evaluations by Core Instrunentatior.

107:7 1

2 PLANT DESIGN 200:1 200 GENERAL 200:1 201 MAIN COOLANT SYSTEM 2(1:1 202 PRESSURE CONTROL AND RELIEF SYSTEM 202:1 203 CHARJIN3 AND VOLUME CONTROL SYSTEM 203:1 20h CHEMICAL SHUTDn'N SYSTEM 20h:1

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20$ PURIFICATION SYSTEM 205:1 206 COMPO'NENT C00LIN3 SYSTEM 206:1 L. _

B:h 7/19/63 Page b;

v' 3 SITE 300:1 300 GENE?JLL 300:1 Location 300:1 Access 300:1 Population 300:1 Land Use 300:h Public Water Supnlics 300:h Site Layout 300:5 301 METEOROLOGY 301:1 Pollution Climatology of the Deerfield River Site -

Report by Professor James M. Austin 301:1 302 HYDROLOGY 302:1 303 GEOLOGY 303:1 30h SEISMOIDGY 30h:1 305 ENVIRONMENTAL RADICA1TIVITY SURVEY 305:1 Preoperational Survey 305:1 Type of Samples 305:1 n

Postoperational Survey 305:2 r]

L h ACCIDENTS AND HAZARDS h00 GENERAL h00:1 h01 REACTIVITY ACCI1ENTS h01:1 Startup Accident h01:1 Continuous Rod Withdrawl at Power holth Cold Water Accidents h01:5 Boron Concentration Accidents h01:8 Reactivity Accidents with Core II at 5ho MWT h01:12 h02 hdb fbbbkb

ht Loss of Coolant Flow Accident h02
1 Loss of Load Accident h02810 Loss of Coolant Accident h02:16 Mechanism of Blowdown h02:16 Criticality of Core During Blowdown h02:21 Core Meltdown h02:23B Chemical Accident h02:25 Vapor Containment h02:27 Missile Protection h02:32 Loss of Coolant Flow Accident with Core II at 5ho M h02:3h Loss of Ioad Accident with Core II at 5ho m h02:38 (4

Ioss of Coolant Flow - Core III h02:38A t/

Loss of Icad'- Core III h02:38A Ioss of Coolant. Accident at 600 WT h02:39

3:5 7/19/63 hge (ns LJ LO3 EAZARDS ?RCM FliCTC?. ACCIDEhTS E03:1 Maxin r. Cradible Accident LO3:1 Eyrethetical Accident LO3:1 Direct Radiation On-site LO3:3 Direct Radiation Off-site LO3:3 Vaper Centainer Icakage and Air-berne ?adiation h03:6 Hycothetical Accident at600 MiT LO3:9 h0h CCNCLUSICNS LOL:1 70LUME II 5 PLUIT OPE:.ATION 503:1 500 3D E.AL 500:1 501 FLANT C5.3ANIZATICN 501:1 502 PERSONNEI T?.ADiDG 502:1 503 INITIAL FLUiT D;SFECTION AND STARTUP TEST PRCGRAM 503A;l 503A EQUI: MENT SPEOIFICATIONS AND MANUFACM' TESTS 503A;l

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(.'j Specification 503A 1 Piping 503A:1 Cleanliness

$O3A:2 Shop Inspection 503A;2 5033 DELIVERED E;UIPMDIT DiSFH, IONS AND INSIALIATICN T

FROCEDURES 5033:1 Receipt of Equipment at the Job Site 5033:1 Cleanliness 5033:1 Erection 5033:2 503C FC.EOFE:.ATIONAL IROCIDURES, SECCEDARY PLANT 5030:1 503D FREDFED.ATIONAL PROCEDURES, NUCLEAR FLtNT

$03D1 Fillis, Cleaaiv and Functional testirg of A M h v Systems 503D1:1 503D2 Main Ccclant System Pre-core Leading Tests 503D2:1 p.

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503E INITIAL CCRE LOADING AND NUCLEAR CCRE TESTS 503El Initial Core Loading 503El:1 503E2 Control Rod Drive ard Plant Scra: Tests 503E2:I 503E3 Initial criticality Instruction 503E3:1 503E4 control Rod and Boron Worth Determinations -

At Low Te=perature 503E4:1 503E5 Te=perature, Pressure and Flow Coefficient Deter =inations - With Increasing Te=perature 503E5:1 503E6 Reactor Startup 503E6:1 503E7 Control Rod and Boron Worth Deter =inations -

At Operating Te=perature 503E7:1 503E8 Temperature, Pressure and Flow Coefficient Determinations - With Decreasing Te=perature 503E8:1 503E9 Main Coolant System Heating Rate DA,ermination 503E9:1 503E10 Unclear Instru=entatien System Response to Asy==etric Control Rod Positioning 503E10:1 503F INITIAL NUCLEAR PLANT PCWER OPERATIONAL TESTS 503F1 Power Coefficient and Loss of Lead Transient Tests 503Fl:1 503F2 Power Calibration of Nuclear Instrumentation 503F2:1 503F3 Variation of Reactivity Due to Change in Fission Product Level Following Reactor O'

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so3r>=1 503F4 Biological Shield Effectiveness Test 503F4:1 503F5 Instrumentation and Control Respense Determination 503F5:1 503F6 Variation of Reactivity to Change in Fission Product Level Following Reactor Shutdown 503F6:1 503F7 E=ergency Cooling by Natural Circulation 503F7:1 504 NORMAL PLANT OPERATING INSTRUCTIq @

5041 PLANT STARTJP 504A1 Reactor Startup from Cold Condition 504A1:1 504A2 Reactor Startup fro = Hot Standby 5C4A2:1 504J'-

Reacter Startup Following Scra: Condition 504A3:1 50~.a4 Turbine Generator Startup frc= Cold Condition 504A4:1 w4A5 Turbine Generator Startup from Hot Condition 504A5:1 t

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B:7 9/15/59 Pace 504B PLANT OPERATICN 504B1 Changing Reactor Load 504Bl:1 504B2 Increasing Turbine Generator Load 504B2:1 504B3 Decreasing Turbine Generator Load 504B3:1 504C PLANT SEUTDOWN 504C1 Scheduled Reactor Shutdcyn 50401:1 504C2 Reactor Cooldcun 50402:1 504C3 Scheduled Turbine Generator Shutdown 504C3:1

'504D MAIN COOLA'iT SYSTEM 504D1 Filling and Venting of Co=plete Systen 504Dl:1 504D2 Filling, Venting and Draining an Isolated Loop 504D2:1 504D3 Hot Leak Test 504D3:1 504D4 Startup of Co=plete Syste:

504D4:1 504D5 startup of Isolated Loop 504D5:1 504D6 Running Operation 504D6:1 504D7 Shutdown of Conplete System 504D7:1 504D8 Shutdown of Individual Loops 504D8:1 f--

504E PRESSURE CONTROL AND RELIEF SYSTEM 504E:1.

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504F CHARGING AND VOLUME CONTROL SYSTEM 503:1 504G CHEMICAL SFUTEGIN SYSTEM 504G1 Boric Acid Preparation 504Gl:1 504G2 Boric Acid Addition 504G2:1 504G3 Boric Acid Removal 504G3:1 504H PURIFICATION SYSTFL' 504H 1 504I CCEPO?iENT COOLING SYSTEM 504I:1 504J PRIMARY PIMIT CCRROSION CONTROL SYSTEM l

5C4J1 Hydreg:r. Addition 504J1:1 504J2 Hydrazine Addition 504J2:1 5C4J3 Lithium Hydroxide Addition 504J3:1

%4K PRIMARY PIRiT SAMPLING SYSTEM 504K1 Main Coolant System 504K1:1 504K2 Auxiliary Systens 5C4K2:1 l

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Bt8 9/15/59 Page 504L RADI0 ACTIVE WASTE DISPOSAL SYSTDI 504L1 Liquid Waste Disposal 504Ll:1 504L2 Gaseous Waste Disposal 504L2:1 504L3 Solid Waste Disposal 504L3:1 504M SHUTDOWN COOLING SYSTEM 504M:1 3

504N REACTOR CONTROL SYSTEM 504N1 5040 NUCLEAR INSTRUMENTATION SYSTEM 5040:1 504P RADIATION MONITORING SYSTEM 504P:1 504Q VAPOR CONTAINER ATMOSPHERE CONTROL SYSTEMS 504Q:1 504R ELECTRICAI. SYSTEM 504R1 120 Volt A-c Vital Bus System 504Rl:1 504R2 125 Volt D-c System 5".,R2:1 504R3 Station Power System 504R3:1 505 EMERGENCY INSTRUCTIONS 505A 1 505A GENERAL 505A 1 505B PRIMARY PLA?['

505B1 Emeri mcy Shutdown 505Bl:1 505B2 Loss f Icad Accident 505B2:1 505B3 Fuel ladding Failure 505B3:1 505B4 Cold ater Accident 505B4:1 505B5 Malfunctioning Control Rod Drive (s) 505B5:1 505B6 Failure in Reactor Control Circuit 505B6:1 505B7 Total Loss of Main Coolant Flow 505B7:1 505B8 Partial Loss of Main Coolant Flow 505B8:1 505P9 Chemical Neutron Absorber Accidents 505B9:1 505310 Total Loss of Main Coolant 505B10:1 505Bil Large or Par +,ial-Loss of Main Coolant 505Bil:1 505B12. Neutron Shield Tank Isak 505B12:1 505B13 Excessive Radioactivity Level 505B13:1' 505B14 Malfunction of Pressure Relief and Safety Valves 505B14:1 505B15 Malfunction of Pressurizer Pressure and Level Control 505B1521 505B16 Failure of Regenerative Heat Exchanger 505B16 1 505B17 Loss of Shutdown Cooling 505B17:1 505B18 Failure of Charging and Volume Control System 505B1821 505B19 Loss of component Cooling 505B19:1 O

B:9 9/15/59 Page 505C SECONDARY PLANT - EMERGENCY SHUTDOWN AND TURBINE THROTTLE TRIP 505C:1 505D ELECTRICAL SYSTEM 505D1 Ioss of 120 Volt A-c Vital Bus 505Dl:1 505D2 Loss of A-c Supply 505D2:1 506 PLANT MAINTENANCE INSTRUCTIONS 506A GENERAL 506A:1 506B PRIMARY PLANT 506B1 Opening and Closing Isolated Main Coolant Icop 506Bl:1 506B2 Decontamination System Operation 506B2:1 506B3 Cold Leak Test 506B3:1 506B4 Fuel Transfer Pit Purification System Operation 506B4:1 506B5 Fuel Transfer Pit Cooling System Operation 506B5:1 506B6 Replacement of Ion Exchange Containers 506B6:1 506B7.Replacenent of Pressurizer Heaters 506B7:1 506B8 Neutron Shield Tank Maintenance 506B8:1

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506C SECONDARY PLANT 506C:1 506D' VAPOR CONTAINER ACCESS 506D:1 4

506E REACTOR REFUELING 506El Site Handling and Storage of New fuel and Control Rods 506El:1 506E2 Preparation of Reactor Systems for Refueling 506E2:1 506E3 Shield Tank Cavity-fill and Drain 506E3:1 506E4 Fuel and Control Rod Replacement 506E4:1 506E5 Site Storage and Shipping of Used Fuel and Control Rods 506E5:1 l

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B:10 9/15/59 Page 507 RADIOLOGICAL HEALTH AND SAFETY 507:1 OBJECTIVE 507:1 GENERAL 507:1 RADIATION STANDARDS 507:2 Personnel Radiation Exposure Limits 507:2 Personnel Contamination Limits 507:3 Clean Area Radiation and Contamination Limits 507:4 Potentially Contaminated Area Radiation and Contamination Limits 507:4 Regulated Equipment Radiation and Contamination Limits 507:4 Shipment and Waste Disposal Limits 507:5 PERSONNEL RADIATION EXPOSURE CONTROL PROCEDURES 507:6 General Radiation Safety Pules 507:6 Personne1 Dosimeters and Film Badges 507:10 Protective Clothing 507:11 Respiratory Equipment 507:12 Personnel Monitoring 507:12 Personnel Decontamination 507:12 O

xeasca1 ana 81e-a a7 zx-inatien 507=13 Personnel Radiation Exposure Records 507:13 Access Control 507:13 Area Monitoring 507:15 Barriers, Signs, Tags 507:15 Radiation Work Permit 507:16 Area and Equipment Decontamination 507:17 Radioactive Shipments and Waste Disposal 507:18 Radiation Incidents 507:19 REFERENCES 507:21 508 PLANT SECURITY AND SPECIAL NUCLEAR MATERIALS CONTROL 508:1 Plant Security and Access 508:1 Classified Document Handling 508:1 Control of and Accounting.for Special Nuclear Materials 508:2 509 ROUTINE AND CONTINUING PLANT TESTS 509:1 General 509:1 Routine Nuclear Tests 509:1 Routine Mechanical Tests 509:2

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3:11 6/1/62 Following O

Drawing Page Na 9699-FM-2A Fundamental Flow Diagram 200:1 9699-FM-18A Macaine Location Plan Operating Floor -

Turbine Room 200:5 9699-FM-183 Machine Location Plan Oround Moor -

Turbine Room 200:5 9699-FM-18C Machine Location Sections - Turbine Room 200:5 9699-FM-18D Machine Location Plan Mezzanine Floor -

Turbine Room 200:5 9699-FM-38F Circulating Water System Plan 200:5 9699-FM-38J Circulating Wat-System Sections 200:5 9699-FM-45A Heat Balance Diagram - 157 mv Load M ai m m Capability for Design 200:6 9699-m 45B Heat Balance Diagram - lk5 =v Load 200:6 9699-FM k50 Heat Balance Diagra= - 125 =v Load 200:6 9699-FM 45D Heat Balance Diagram - 72 =w Load 200:6 9699-FM 45E Heat Balance Diagram - 111 =v 7 200:6 9699-FM-h5F Heat Ralance Diagram - 170 nr.

200:6 646-J h21 Main Coolant System 201:1

&6-J k22 Pressure Control and Relief System 202:1 646-J 430 Charging and Volu=e Control System 203:1

& 6-J 426 Chuical Shutdown System 204:1

&6-J-723 Purification Syste=

2 05 :1

& 6-J-424 Cc::ponent Cooling System 206:1

&6-J-431 Corrosion Control System 207:1

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& 6-J-429 Sa=pling System 208:1 V

9699-RM-41F Flow Diagram - Radicactive Waste Disposal 209:1

&6-J-425 Shutdown Cooling System 210:1

& 6-J-428 Vent and Drain System 211:1

& 6-J-&4 Safety Injection System 212:1 ESK12659A

  • Power Supply & Control - Safety Injection System 212:2 517-F-076 Power Supply & Control for Rod Drive 213:6 517-F-069 Nuclear Instru:nentation Block Diagram 214:1 9699-FE-2G Turbine Shutdown and Reactor Scram - Block Diagram 214:1 548-D-638 Operational Radiation Monitoring System 215:1 517-F-417 Vapor Container Air Cooling and Ventilation System 216:1 549-D-295 Vapor Container Leakage Monitoring System 216:3

&6-J-616 Decontamination System 217:1 9699-FM-19B Fuel Handling Arrangement 218:1 9699-FM-21B Fuel Storage Vault 2181 9699-N-21A Fuel Transfer Pit 218:2 9699-FM-51A Arrangement and Details of Liner, Shield Tank Cavity - Sh. 1 218:5 9699-m-3A Flo'v Diagram - Main and Auxiliary Steam Lines and Steam Generator Blowdown 219:1 9699-m-4A Flov Diagram - Condensate and Feed Water Lines 220:1 9699-FM-7A Flow Diagram - Extraction Steam Feed Water Heater Vent & Drain Lines 220:1 B-5030493 Feed Pump Mini =um Flov Recirculation Control 220:6 E-5030434 Three Ele =ent Feed Water Control Equipment &

Instruments 220:6

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B-12 6/1/62 Following Cr Drawing Page 9699- % 10A Flow Diagram - Circulating Vater Screen Wash

& Air Offtake Lines 221:1

%99-FM-38F Circulating Water System Plan 221:1

%99-FM-27A Flow Diagram - Cooling Water Lines 222:1

% 99- % 341 Machine Location Plan - Water Treatment Room 222:3 9699-FM-37A Flow Diagram - Water Treatment 222:3 9699-FW47A Flow Diagram - Gland Steam Laak-off Lines &

Vents to Stacks 223:1 9699- % 261 Flow Diagram - Compressed Air Lines 224:1 9699-N36A Flow Diagram - Lubricating 011 System 225:1

%99-FE-1B Schematic Diagram - Meters, Relays &

Synchronizing 226:1

%99-FE-lO 2,400 V - One Line Diagram 226:1 9699-FE-ID 480 V - One Line Diagram - Sh. 1 226:1

%99-FE-10 480 V - One Line Diagram - Sh. 3 226:1 9699-RE-1F 480 V - One Line Diagram - Sh. 2 226:1 9699-FE-1H 125 V - DC - Ona Line Diagram 226:1 9690-FB-7A Flow Diagram - steam & Condensate - Sh.1 22.7:1

%99-FB-7B ~

Flow Diagram - Steam & Condensate - Sh. 2 227:1 9699-FB-3A Ventilation Arrangement - Ground & Mezzanine Floors 228:1 9699-FB-3B Ventilation Arrangement - Operating Floor &

Sections 228:1

%99-FB-3E Ventilation Arrangement - Office A Service O"

Buildings 228:1 9699- % 10 Machine Iocation - Sections - Vapor Container 231:1

%99-FY-11 Details of Vapor container Piping Penetrations -

231:1 Sh. 1

%99-FV-1Y Details of Vapor Container Piping Penetrations -

Sh. 2 23 1:1

%99-FE-290 Elect-

,al Penetraticns - Sh. 3 231:2

%99-%1A Machine Location Plans - Sh.1 - Vapor Container 232:1 9699-FM-190 Arrangement and Detaile - Fuel Chute - Vapor Container to Transfer Pit 232:2 631-J-915 145 mw Turbine-Tandem Compound Double Exhaust 1,800 Rpm - 450 Psi Cage - 460 F - 1.5 In. Hg Abs 233:1

'%99-FY-5A Exclusion Area Plan 508:1 4

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10 6:1 9/15/59 106 RFACTOR COOIANT CHFJ4ISTRY 0

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Generq The Yankee reactor is of the pressurized water type cooled and moderated with light water. The major materials of construction are AISI Types 304 and 348 stainless steel. Type 304 stainless steel is used for reactor vessel cladding, main coolant piping and tubing for the steam generators; Type 348 stainless steel is used for fuel cladding. Nickel-phosphorus type braze material is utilized for fabrication of fuel tube bundles. An alloy of silver-indium-cadmium (80-15-5 vt %) is the control rod material with Zirealoy-2 as the follover material. The control rod alloy is plated with a thin diffusion bonded coating of electrodeposited nickel. Materials such as AISI Type 410 stainless steel, Inconel and Inconel-X are used in various reactor components.

The materials of construction are exposed during reactor opera-tion to two major types of primary coolant composition. A soluble chemical neutron absorber is used for supplementary control of the reactor during cold shutdown.

Boric acid has been selected for this purpose on the basis of results obtained at several AEC facilitiesl,2,3 A concentration of approximately.55% boric acid (equivalent to 950 ppm natural boron) is required for this purpose.* During normal operation, the primary coolant is essentially boron-free; the boric acid having been removed initially by feed and bleed dilution and finally by ion exchange.

3 Two major concepts of primary coolant chemistry have been con-

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sidered in the Yankee Research and Development Program.

" Neutral" vater chemistry of the type now in use in APPR-1, based on the use of high purity primary coolant without additions for the control of the pH value, and "high pH" vater chemistry of the type now in use in the Shippingport re-actor, which utilizes lithium hydroxide to maintain the pH v21ue of the primary coohnt at about 10.

Both neutral and high pH vater chemistries have teen reported to provide certain advantages for system operation.

For nedral vater chem-istry these are:

1.

Simplicity of coolant chemistry.

2.

No tritium production.

3 High ion exchange decontamination efficiency.

  • The actual shutdown boron concentration vill be determined during reactor startup tests. A nominal concentration of 1,600 ppm was used for testing purposes.

1ANL-5147 (Rev.) Breden & Abers, " Corrosion and Stability Tests on ChemicalPoisonsinHighTemperatureWater",9/1/53 2WAPD-C(PC)-31, Shapiro, "A Study of Chemical Control for PWR",

4/8/55

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3ANL-5244, Breden, Brown & Sivety, " Soluble Poisons in Reactor Control",11/55

106:2 6/1/6h h.

Acceptable corr sion behavior particularly after boric acid preconditioning.

For high pH water chemistry these are:

1.

Reduction of corrosion product release from in-pile and out-of-pile stainless steel systems.

2.

Increased filterability of insoluble corrosion products.

3 Inhibition of crevice corrosion.

h.

Possible increase in resistance to wear.

At the inception of the Yankee Research and Development Program, no operating power reactor experience was available for guidance in the selection of a water chemistry. Even at present, no operating reactor

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conforms precisely to Yankee conditions. The APPR-1 is an all-stainless steel system operating with "neutrcl" water chemistry. The Shippingport reactor is a stainless steel system with a Zircaloy clad core operating in high pH water chenistry. Neither system utilizes boric acid for routinc supplementary shutdown control of the reactor.

Parallel programs were initiated to evaluate both neutral and high pH water chemistry with regard to their effect on Yankee primary f

system materials. Heutral water chemistry is, at present, the reference Q

chemistry for initial operation of the Yankee reactor. High pH water chemistry represents a back-up system in the event that its use becomes desirable. Before it will be used in the reactor, the stability of high pH water chemistry, through LiCH addition, in conjunction with the use of boric acid is to be demonstrated by means of an in-pile loop test, as it has already been demonstrated out-of-pile 5 n

Main Coolant Water Reference Conditions V

Shutdown Cherdstry Boron (as boric acid) as required to fulfill the shutdown margins established in the Techrdcal Specifica-tions unadjusted ( ~ 5.2) pH Chloride less than 0.1 ppm Oxygen saturated kAEC-153,Cytron " Dynamic Test Loop Corrosion Studies of Yankee Plant Primary System Materials" (to be issued).

kAEC-116,Krieg,

" Dynamic Screening Corrosion Tests of Yankee Materials

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in High Temperature Boronated Water."

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106:3 1/28/66 Normal Operating Chemistry Boron (as boric acid) -- O to 1300 ppm, as required to fulfill the shutdown margins established in the Technical Specifications Total Solids (other than boric acid) approximately 2 ppm Hydrogen 25 to h5 ml (STP) per Kg 4

pH 5.0 to 10 5 (The pH of the main coolant water shall always be within range from $.0 to 10.5 and no chemical addetive other than up to 20 ppm of NH3 shall be used to control pH, when more than 5 ppm of boron are present in the water).

A 0xygen O

normally less than 0.1 ppm (0 5 ppm maximum) i Chlorida normally less than 0.1 ppm (0 5 ppm maximum)

Primary Grade Make-Up Water Demineralized water of greater than 1 megohm-cm specific resistivity.

I Corrosion Behavior of Reactor Materials l

Neutral pH Chemistry - The austenitic stainless steels are highly resistant to gross corrosion attack in high temperature water, either in the presence or absence of boric acid. The rate of corrosion of Type 30lt stainless steel has been studied in high temperature water containing approximately ho ppm boron (as

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boric acid) at 7 fps velocity and 6000F6 After an initial, i

relatively rapid attack, the rate of descaled weight change (total metal corroded) after the first 200 hr to the end of the test,.

about 1,100 hr, is conservatively estimated at h.$ mg/dm2 per month.

l A loss of 10 mg/dm2 corresponds to an average penetration of O.00$ mils in stainless steel. There is clear indication that towards the end of the test, the rate of descaled weight change

,g has decreased even further to a negligible value. For comparison

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YAEE-68, Kreig & Cytron,

" Corrosion of. AISI Type 30h Stainless Steel in High Temperature Borated Water," 12/58

4 106:3a 6/1/6h O

with the results of other tests, the total descaled weight loss in 700 hr (one month) is approximately 27 mg/dm2 The descaled wei6ht loss of Type 30h stainlass steel in neutral water ag 6004 in the absence of boric acid is given as i

5 mg/dm at velocities of 1/60 and total exposure ttme of 2,000 hr 7. 30 fps for an averageExposure of Type 30h stainless steel with various surface pretreatments at 38 fps velocity and 6000F for one month in water containing 3 ppm boron (as boric acid), to siralate normal peration, yielded a total 2

descaled weight loss of h8 mg/dm.

This was a relatively short term test (707 hr) in which the O

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7TID-7006, DePaul, " Corrosion ar.d Wear Handbook,tr 3/57 i

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106:4 9/15/59 initial period of rapid attack represents a large proportion of the total attack on the material. A similar test in water containing 1,590 ppm boron A

(as boric acid) to simulate shutdown conditions yielded a descaled change 2

in one month of about h4 mg/dm. Thus, the extent of corrosion in one month appears to be independent of boric acid concentration.

Under condi-tions of cycled boric acid concentration, the extent of corrosion of Type 304 stainless steel in one month was found to decrease to about 32 mg/dm2 descaled weight loss. A marked decrease in the excent of cor-rosion of Type 304 stainless steel to approximately 15 mg/d=2 veight loss in 700 hr and low corrosion values for Types 316 and 348 stainless steels in high temperature neutral water were found to result from pre-egposure to high temperature water containing a high boric acid concentration. This effect is being investigated further.

No susceptibility towards stress corrosion cracking of Type 304 stainless steel was found under simulated Yankee operating conditions.

The corrosion behavior of the AISI 400 series stainless steel, Inconel, Zircaloy-2 and both plated and unplated silver-indium-cadmium alloy was investigated under simulated Yankee conditions 5-The extent of corrosion of Type 410 stainless steel in one month decreases slightly with decreasing boric acid concentration. Unt.er cycled conditions of boric acid concentration, the corrosion increases slightly. The corrosion of Inconel also decreases with decreasing Soric acid concentration.

Cycled boric acid concentration does not increase the corrc. ion of Inconel beyond that suffered in water of lov boric r.cid concentration.

Zircaloy-2 corre-sion increases slightly with decreeaing boric acid concentration.

Cyclea boric acid concentration does not increase Zircaloy-2 corrosion above that s_)s experienced in high boric acid concentration water. The base control rod material, silver-indium-cadmium alloy, shows a sharp increase in weight gain as the boric acid concentration decreases.

Under cycled conditions, the weight gain is 1cver than at high or low boric acid concentration.

&cidence of internal oxidation of the alloy was found. Protection against internal oxidation of the control material is provided by electrodeposited diffusion bonded nickel plate. The braze material, either Nicrobraz 10 or Kanigen, shows high resistance to general corrosion in high temperature water, with or without the presence of boric acid.

The general effect of trace lithium hydroxide addition has been to reduce the extent of corrosion of most of the materials tested.

Zircaloy-2 corrosion is not affected by the pH value of the medium.

In-creased weight gains were obtained with the unplated silver-indium-cadmium alloy. The table on page 106:5 gives some experimentally determined cor-rosion results for various materials of the primary system exposed to a variety of water chemistries.

Galvanic Corrosion - The base control rod material, silver-indium-cadmium alloy, was found to be susceptible to internal oxidation in high temperature, deaerated water.

For this reason the control rod material is plated with electrodeposited nickel which provides excellent protection against this type of attack. Proper heat treatment (approximately 4 hr at l

600 C) produces a diffusion bond between electrodeposited nickel and the silver-indium-cadmium alloy. This bond is highly resistant to the galvanic

\\~,b l

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SUMMARY

OF CORROSION RESULTS YANKEE PRIMARY SYSTEM MATERIAIS (As determined in Yankee Research and Development Program)

TEMPERATURE, &)o F - FIDW, 38 I'un Concentration. ppm Boron Lithium Weicht Chanre. MF/dm2 (In the First Mnnth)

(as Boric (as Lithium 304 S/S 410 S/S Inconel Zircalov-2 Acid)

Hydroxide)

' Exmned Descaled Exponed Dencaled Exposed Deccaled Exmsed 1,590 0

-(15-21)

-(40 47)

-65

-121

-77

-141

+17 1,561-1

-(7-10)

-( 23-26)

-33

.-63

-27

-71

+19 3

0

-(15-20)

-(40-62)

-32

-106

-13

-43

+14 189 2

-( 6-8)

-(20-27)

-36

-104

+9

-19

+22 12 25

-(7-15)

-(19-23)

-27

-58 '

+1

-17

+14 1,629-<5*

o

-(10-14)

-( 27-34)

-75

-143

-41

-81

+14 1,790-12*-

1 3-o.8*

-(10-22)

-(23 26)

-26

-73

-16

-35

+16

  • Cycled at weekly intervals R

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106:6 1)l0/60 corrosie. which severely attacks untreated nickel plate under conditions (3

of shutdown chemistry with oxygen present in the coolant. This environ-ment is anticipated daring initial startup and refueling operations.

It has been found that, by using a very thin (about C.0003 in.) diffusion bonded nickel plate, galvanic corrosion is reduced to an acceptable amount.

Tin is produced in the silver-indium-cadmium material by trans-mutation of indium under neutron irradiation. The silver-indium-cadmium alloy containing o 5 % tin is highly resistant to internal oxidation.

Thus, it is expected that, should defects develop in the nickel plate, the alloy itself vill become resistant to this type of attack after exposure to neutron irradiation.

The following experimental results have been obtained for spec-imens grounded to a stainless steel container in 140 F vater containing 1,600 ppm of boron (as boric acid) and 3 5 ppm of dissolved oxygen. The duration of the test was 14 days.

Materini WeiFht Chance 304S/S 0mg/dm2 348s/S 0 mg/d=2 410 S/S

-8 mg/dm2 Inconel

-6=g/dm2 g;

Zircaloy-2

+4mg/dm2 Nicrobraz-10 and Kanigen

-3mg/dm2 (Diffusion heat treated on stainless steel)

Water Chenistry Stability Boric acid has previously been used in reactor systems principally in the various borax experiments for control purposes and safety systems.

There is no evidence in the. borax operations that significagt quantities of boric acid deposit in an insoluble form in the reactor core. The borax experiments were conducted in water which, aside from boric acid additions, was essentially neutral water chemistry.

Hydrazine is commonly used to scavenge oxygen from water at tem-peratures above 200 F.

No adverse effects have been reported as a result of its use. The use of hydrazine as an oxygen scavenger is compatible with 9

the presence of boric acid.

ONuclear Science and Engineering 1-420 43, Zinn ET AL, " Operational Experience with the Borax Power Plant", 1956 9YAEC-ll3 "Q,uarterly Progress Report for the Period October to b)

December 1958", page 53 v

_. =.

106:7 10/2/59 Coolant Radimetivity As in any circulating vater syste=, so=e corrosion of the surfaces in contact with the prira*y coolant vill occur. In the Yankee pri=ary system =ost of the surface exposed to the pri=ary coolant is stainless steel and the corrosion products vill, thezefore, be =ade up largely of the con-stituents of stainless steel. Two sources of activated corrosion products exist in the primary syste=. The = ore i=portant is the surface area located in regions of high neutron flux, such as the fuel cladding.

In addition, the surfaces located outside the high flux region such as the prira y coolant piping, steam generators.nd pu=ps contribute corrosion products which become radioactive as they pass through the high flux region of the core.

l l

Yankee radioactive corrosion product esti=ates are based on an overall cor-l rosion release rrta of 10 =g/dm2 per =onth.

The accu =ulation of fission products resulting from the fission process in the fuel represents a source of activity in the pri=ary coolant.

If sc=e of the fuel rods develop =inor defects, such as pinholes or s=all cracks, then fission products my leak into the pri=ary coolant. As a basis for design of the auxiliary syste=s it was assu=ed that 1% of the fuel rods vould develop such defects.

The following is a list of all radioactive isotopes which have an equilibriu: activity level in the pri=ary coolant g eater than 0.01 =icro-4 curie per =illiliter. These equilibriu= activities assu=e a constant lo gp:

flov to the purification syste=.

i Nonvolatile Fissien Products Vohtile Fission Prodnets

Activity, Activity, Isotores Microcurie ter r.1 Isotores Micromirie ter =1 Rb-88
0. 25 Kr-85 (h.h hr) 0.096 Rb-89 0.032 Kr-85 (lo.3 yr) 15 sr-89 0.o36 Kr-87 0.19 Mo-99 0.029 Kr-88 0.024 Te-lol o.033 Xe-133 2k.2 I-131
1. 6 Xe-135 15 Te-132
2. 2 Xe-138 0.046 I-132 2.1 I-133 2.1 cor osion Products Te-134 0.15
Activity, I-134 0 31 Isotooes Microcurie ter el I-135 0 94 Cs-137 0.o88 Mn-56 0 52 Cs-138 0.1k Co-60 (5 3 yr) 0.077 Cs-139 0.026 Fe-59 0.052 Ba-139 0.025 Ns-24 0.15 Cr-51 0.8 Mn-Sh o.116 Fe-55 0.12 Co-58 0.84 Cu-6h o.019 j

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A 106:8 9/15/59 Sumetry Proposed Yankee materials of construction show relatively minor attack in the med's to which they may be exposed in the reactor.

" Neutral" vater chemistry, including boric acid use for supplementary reactor shut-down, has been selected as the reference chemistry for the system.

"High pH" vater chemistry offers certain possible advantages with regard to corrosive attack and corrosion pro <' et properties.

However, the question of boron " hideout" through deposit.,n of insoluble lithium metaborate is being investigated further under in-pile conditions. Preconditioning of the ujor r:aterials of construction (austenitic stainless steels) vita a high concentration boric acid solution appears to eliminate the difference 1

in extent of corrosive attack between " neutral" and "high pH" water chemistry.

The base control rod material suffers internal oxidation under simulated Yankee operating conditions. Electrodeposited nickel plate provides protection against this type of attack but unless diffusion heat treated is subject to galvanic attack in borated, oxygenated water at ambient temperatures. Heat treated, thinly nickel plated control rod mate-rial is resistant to galvanic attac...

The formation of tin through trans-mutation of indium yields an alloy which is resistant to Internal oxidation and which thereby complements the protective action of the nickel coating.

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106:9 12/23/63 Boron Shin at Power on Core III Based on the results of tests performed dttring operation on Cores I and II, use of a boric acid shin at power is specified for initial oper-ation of Core III.

(See page 103:hh). The hutdown margin provided by boron is restricted to an anotnt equivalent to equilibriun xenon at 5h0 mit.

This will permit complete removal of the boron after only a few days oper-ation at ower, if desired. Reboration would be necessary only in the event of an extended shutdown occurring early in core life.

The calculated xenon equivalent of boron in Core III is approzi-mately h00 ppm. Yankee has operated with this concentration of boron for 7 days on Core I and 18 days on Core II with no problems or adverse effects.

These tests are discussed in detail in Operation Reports No.10 (October 1961) and No. 26 (February 1963),

The tests demonstrate the feasibility of operation with a chemical shin at power. The plant syatens have been shown to be capable of achieving and maintaining the prescribed boron concentrations. No significant increase in primary coolant activity levels was noted. Crud levels were simiHarly unaffected.

Unexplained reactivity variations were experienced during the tests, as they were experienced at other times during operation on Cores I and II. It is not possible to attribute these variatione to tnron. In q

fact, a loss of reactivity of about 0.6% was noted in the first test, while V

a gain in reactivity of 0.5% was recorded in the second test. The two tests were performed under virtually the same conditions, except that the purifi-cation system was operated during the first test only.

J Si=ilie reactivity changes have been noted when the pH of the system is changed. On several occasions hydrazine has been added to the main coolant to increase pH. In every case, a reactivity gain was noted, followed by a comparable loss of reactivity when the hydrazine is removed.

It is clear that reactivity is affected by water chemistry, with or without boron. In all cases reactivity changes have been -11 and have occurred slowly. Such changes present no operating problems.

However, there has been speculation in the industry as to the effect of a crud burst on reactivity, where the crud could contain appreciable amounts of boron. The potential magnitude of this problem has been investigated.

T v

=

106:10 12/23/63 Ater chenistry data taken awr three years of Yankee aperation d

reveals that main coolant crud levels hwe been unifornly 1cw (~1 ppn) except en a very few occasions wnen red arcp tastim has been p. rfarned, or when reactor bypass lines have wen flushed. The reactor was, of course, shut down when these operations tcok place.

Flushirg of tne bypass lines raises crud levels substantially (52 ppm nax), because these lines are so lo:ated that they act as excellent crud traps during normal operation. However, tnese increases in crud con-centration cannot be considered as crud bursts in the nornal acuse, because they are predictable, repeatable, and under no circumstances are they related to the core.

On the other hand, the crad bursts resulting from rod drops do come from the core, although the high nickel content and low activity of this crud indicates that it comes from control rods rather than from fuel rods. The maximum crud level noted after rod drop testing was 1.9 ppm. If every bit of this crud were boron and it all came from the core, the largest reactivity addition that could occur would be 0.33% d k/k.

"'his could be accepted with-out any control problems or adverse effects, even if it occurred at full power.

Actually, the highest boron concentration that has been measured in Yankee crud is 0.}2%. This means that the worst crud burst (1.9 ppm) would only release h x 10-o g k/k. A burst 1000 times as large as this would not present a problem, regardless of the power level at which it occurred.

Experimants have been performed out of pile by the Westichouse Atomic Power Division which further substantiate these conclusions. The experiments irdicate that at least 0.h mils of crud must be present on heat transfer surfaces at reactor operating conditions before concentrations of trace elements such as boron are found in the crud. While Yankee fuel has been remarkably free from crud,if 0.h mils were present, a crud burst of 20 pp., which is more than 10 times as large as the worst burst that has ever been experienced at Yankee, and larger than anything that has been found in operating data from Shippingport, SM-1 or Saxton,would amount to only 5%

of the crud present on the core. This means, assumir,g the boron to be uniformly distributed, that 10% g k/k would have to be tied up in adsorbed boron in order for a release of 0.5% A k/k to occur. And even this release would not be.a problem.

The 10% figure is, of course, very unrealistics all of the excess reactivity contained in Core III amounts to only 6%. Neverthelesa, these numbers do illustrate the magnitude of the reactivity loss due to boron which would be required before a control problem could exist. This knowledge, com-bined with the experience gained with chemical shim at power during operation on Cores I and II, form an adequate basis for the conclusion that the Yankee reactor can be operated safely and conservatively with boric acid in the primary coolant.

O v

106:11 8/1/67 Boron Shim at Power on Core IV b)

A history of satisfactory operation witn boron in the main coolant has been generated. The first two Yankee cores were operated for several weeks with boron in concentrations up to h00 ppm. Core III was likewise operated for two months. The Saxton reactor has recently completed a nine-month run with boron to 1200 ppm. At no time has the presence of boron in the coolant presented a problem, either from the standpoint of water chemis-try or of reactivity c<mtrol. Based on these entirely satisfactory results, chemical shim operation has been specified for Core IV.

As outlined on page 103:53, it is intended to go critical with only rod groups 1-h in the core. The calculated boron concentration required for criticality with rod groups 1-h fully in and the other rods fully withdrawn is 850 ppm. Normally, xenon will be followed with rods and the maximam boron concentration at power should not differ from the calculated value by a sig-nificant amount. However, in the initial startup it is intended to measure the boron equivalent of equilibrium xenon. This will require that up to 1300 ppm B be carried in the coolant at the start of-life with dilution to approximately 8$0 ppm over the first few days as xenon builds in.

The ability to follow xenon with boron beyond the initial startup will also provide opera-tional flexibility, although it is unlikely that this mode of operation will be used very of ten due to the added load on the waste disposal system.

Ecron Shim in Subsequenj_ Cores Based on entirely satisfactory experience with bcron shim control

()

at power, this mode of operation is contemplated in all future cores.

A maximum concentration of 1700 ppm B ts presently specified when the plant is producing more than 15 MR electric.

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107s1 12/23/63 107 CORE INSTRUFETATION n

General U

The heat output of the first core for the Yankee reactor was conservatively set at 392 ma for initial operation. Although the capability of the core was expected to be greater than this through most of core life, the initial power rating was defined at a value which could be achieved even at times when control rod positioning resulted in unf avorable power distribution. In order to obtain a better understanding of the effects of fuel burnup and control rod progranming on power capability throughout core life, in-core instrumentation is provided in the Yankee reactor. It is expected that this instrumentation will yield information which will permit both the initial core and future cores to be operated closer to their actual thermal limits. Two types of measuring devices are provided as part of this instrunentation: incere wires to measure neutron flux, and core outlet water thermocouples.

Flux wires can be inserted into the core at various locations to obtain " flux-maps" at intervals during core life.

Each flux wire yields an axial distribution along the center of a fuel assembly. Various radial positions of flux wires are then compared to obtain a flux map in a region of the core.

The therrocouples are positicned to measure the outlet water temperature over selected nozzles in the upper core support plate.

The thermocouple locations are concentrated principally above one quadrant of the core, and the renaining locations are spotted above the other three (3

~'j quadrants. These thermocouple measurments, cocined with a knowledge of the coolant flow in the core, yield an integrated power output at each of these positions. These data are used as a running check, in conjunction with the flux wire data, of the fuel burmip distribution at several points in time. They form a valuable complement to the flux wire data in that the information will be continuously available during reactor operation.

Mechanical Description The mechanical configuration of the in-core instrumentation assembly is shown conceptually on page 107:2.

l As original 4 installed twenty-two flux wire thimbles were inserted in the center of selected fuel assemblies. Subsequently two thimbles were damaged and cut off leaving a total of twenty thimbles in locations indicated on page 107:3. Each thimble consists of a 3/8" 0.D.

stainless steel tube running from above the head adaptor down into the core. The thimbles are closed at the bottom and are open above the head l

adaptor, so that the flux wires can be inserted and withdrawn without depressurizing the reactor.

In each fuel assembly, one fuel rod near the center is omitted in order to provide channels for the flux thimbles. Stainless steel spacer tubes are inserted in the channels of the assemblies which accommodate flux thimbles. The spacer tubes provide a continuous lateral support and bearing n

surf ace for the thimbles. There is no significant reduction in either heat V

transfer surface area or mass of fissionable material as a result of omitting one fuel rod from each assembly.

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1 107:5 3/3V61 The 27 thermocouples are stainless steel 3/16" o.D. ungrounded sheathed chrceel and alumel wires. They are insulated with aluminum oxide O

and are fusion velded to form the hot junctions, with end plugs covering the l

hot junctions. The thermocouple sheaths run from above the head adaptor to positions over the specified nozzles. Each sheath serves as a pressure barrier, so that the individual thermocouple wires do not require high pressure insulat-ing seals.

The internal support structure consists of a series of frames and columns located within the reactor vessel in the region above the upper core support plate. The lower and upper frames of the support structure are located just above the upper core support plate and provide the principal support for the thermocouple, thimbles and telescoping columns. Each frame is constructed i

of flat stainless steel bars velded together to fom a rectangular lattice.

l The control rod guide tubes occupy the open regions of each frame lattice.

Four of the instrumentation ports in the reactor vessel head above the "Eixed shim" control rods are used for the flux wire thimbles and thermo-couple sheaths. Four corner blocks of the lower frame rest on four shock ab-l sorber stops on the upper core support plate. Two thimble and two thermo-j couple telescoping columns extend above the upper and 1 N;r frames in line with these ports. The functions of the telescoping columns are to guide the upper frame with respect to the lower frame, to guide the thermocouples and thimbles down to the main support frame, and to support the seal and junction assemblies.

I Each telescoping column consists of concentric inner and outer l

cylindrical sections. The inner section passes through a hole in one of four upper frame corner blocks and is fastened to a corresponding corner block of the lover frame. A guide sleeve provides alignment between the two sections.

The outer section is fastened to the corner block of the upper frame.

The thermocouple sheaths extend through the length of the inner sec-i tion of the themocouple telescoping column, make several 90 bends, and are individually fastened to the lower frame. The themocouples terminate three i

inches below the top surface of the upper core support plate, remaining stationary during separation of the upper and lover frames.

i I

i The outer section of each thimble telescoping column consists of two 1

concentric cylindrical columns separated by an annular region. The flux vire thimbles are spiraled within this annulus down to the lower region of the column, are rigidly attached to the upper, retractable frame and extend down-vard through the full length of the respective fuel assemblies. The thimbles l

remain stationary with respect to the upper support frame and the thimble telescoping column outer section, and are thus withirawn from the reactor core during separation of the upper and lower frenes.

Four seal assemblies are provided for four instrumentation ports.

The thermocouple sheaths and flux vire thimbles are brazed along a one-inch length to seal plugs within the assemblies. The seal plugs are welded to seal back-up plugs. Seal assembly adapter insets are threaded into the head i

adapters. Pressure seals are provided between the plugs and the adapter in-serts.

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107:6 12/23/63 Junction assemblies are mounted on support tubes above two seal assemblies, where the thermcouple leads are fusion welded to wires

'O leading to a multipoint recorder in the centrol rcom. For the flux wire thimbles, the junctions are provided between the thimbles and the thimble extensions leading to the flux wire drive system as shown on pages 107:hA and 107:hB.

Plux Wi e Control and Readout The function of the control and readout system is to provide means for insertion of eight flux wires into the reactor, and after a preset time interval, to withdraw the flux wires from the reactor and provide a means for readout and recording of induced radioactivity. The drive system consists of nine gearmotors coupled through adjustable clutches to drive boxes as shown in the block diagram of page 107:h. Cable is stored in spent travel tubes and is fed through a series of conduit paths to selected thimbles within the reactor. The use of solenoid-actuated transfer devices permits a total of 20 pathways to be used by 8 drives, driving 8 cables and wires with the ninth drive utili::ed for remote calibration of the counter with a moveable source. Additional transfer devices allow each wire to be routed in turn past an activation counter of the scintillator-photomultiplier type.

The control system is comprised of two parts -- the section physically mounted with the drive units, and the section contained within the control room.

Limit switches are provided in each drive conduit for semi-automatic positioning of the wires during a flux reading operation.

p Each gear box drives a synchro transmitter for positional information V

during counting and flux plotting.

The control room cabinet contains switches and relays for selection of thimbles to be used, and for the starting, reversing and stopping of the drive motors as required. Readout equipment is comprised of a timer, elapsed time clock, power supply, micro-microammeter, integrator and dual pen plotter. The sequence of operations is as follows:

1.

Seclection of wire routes.

2.

Driving wires at high speed to preset positions above the top of the core where they are automatically and individually stopped.

3.

Insertion of all wires to the bottom of the core where they are irradiated for a preeet time interval.

h.

Removal of wires at high speed back to the storage tubes where they are stopped automatically.

5.

Selection of each wire in turn for recording of activity.

6.

Start of count cycle which drives individual wires to a scintil ' ' c: -

photomultiplier counter, with output proportional to activitt measured directly by a micro-microammeter.

O 7-

" ""*1 =ett1=8 or 8 1==

r =911rier =a 91 tter-

107 7 12/23/63 8.

Slow drive of the selected wire through the counter with dual plotting of the positional r,ctivity and integrated activity.

Effects on Plant Operaticn The in4 ore instrumntation system is used only to gather data on core operation and performs no controlling fuction, and possible functional failures of this system will therefore have no effect on plant operation or safety.

By location of the thermcouples symetrically in each quadrant and throughout one quadrant, erroneous indications of a few thermocouples can be detected by aomparison with indications of other thermocouples.

Calibrations can be checked by circulating water through the reactor at various known temperatures with the reacter shut down. Under this condition, the temperature of the water will be uniform throughout the reactc=, as determined by instrumentation in the primary loop.

Meehanical Reliability The in-core instramentation system includes several features to assure the integrity of the high pressure barrier.

Each thermocouple sheath and thimble will be subjected to reactor operating pressure on the outside and atmospheric pressure on the inside. In the event a leak occurs in any of the thermocouple sheaths, the presence of hard packed aluminum oxide inside the sheath will permit only a small amount of water to escape to the outside of the reactor. If the leakage should become O

exce ive, ta re etor e = he

  • =t ao = =a the 1 a1=8 ther >couv1e sheath can be severed and welded shut above the vessel head.

A leak in one of the thimbles would permit a greater amount of water to escape to the outside of the reactor than a leak in a thermocouple sheath. For this reason, the tubira runs off an 20 flux wire thimbles pass through a common manifold located between the seal assemblies and the drive units and this manifold is provided with a leak detector which sounds an alarm in the control room. Individual valves are installed in each conduit run on the reactor side of the junction box. Originally, it was possible to isolate each thimble in the event of leakage by closing its isolation valve. However, modifications to the two seal assemblies at the head penetrations now makes it nundatory to isolate all the thimbles passing through a penetration in the event of a leak in any one thimble.

Accordingly, a leaking thimble in the East penetration will require isolating all eleven thimbles which pass through the East penetration port.

A leaking thimble in the North penetration will require isolating the nine thimbles passing through the North penetration port.

Gas purge connecticas are also provided to the East and North seal assemblies. Carbon dioxide is employed in purging air from the flux wire thimbles. It is felt that removal of the air and entrained water vapor from the thimbles will reduce the possibility of internal corrosion of the thimbles. Each gas purge connection is provided with an isolation valve which can be closed in the event of thimble leakage.

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107:8 12/23/63 Evaluations by Core Instrumentation N(d The maxinum-to-average heat flux (Fg) which exists in the e re may be evaluated by applying a conversion factor to the relative Mn3 activity of the flux wires to obtain relative fission density along the central axis of the assembly. A seer.nd factor which converts this value to the mrimm power density in the assembly is obtained from two-dimensional diffusion theory calculations. In a similar manner, the outlet temperature indications may be converted to a maximum-to-average enthalpy rise in the core. Although the individual Yankee fuel assemblies are not enclosed, an estimate of the flow distribution in the core can be made by evaluating the ratio of integrated fission density to temperature rise for the 20 flux wire locations. With periodic use of the system it is also possible to determine the integrated power distribution and fuel burnup distribution throughout the life of the core.

Both radial and azimuthal symmetry of power may be evaluated by combining the flux wire and thermocouple information from the fully instrumented quadrant with data obtained from the wires and thermocouples in the other three quadrants. In addition a gross check of power asymmetry is obtained from a comparison of the individual loop TH and TC resistance thermometers.

The flux distributions obtained early in core life have been compared with calculated values at the same point in life to establish the validity of calculated values for later portions of core life. In general, there has been good agreement so far between values obtained n

experimentally through use of the in-core instrumentation system and A )

calculated values arrived at earlier in connection with control rod programming studies.

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