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Updated Pages for Ref 106 of PSAR Section 1.6,CRBRP-3,Vol 2, Hypothetical Core Disruptive Accident Considerations in Crbr:Assessment of Thermal Margin Beyond Design Base
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Issue date:	06/30/1982
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CRBRP-3 O

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HYPOTHETICA.L CORE DISRUPTIVE ACCIDENT CONSIDERATIONS IN CRBRP VOLUME 2:

O ASSESSMENT OF THERMAL MARGIN BEYOND THE DESIGN BASE CLINCH RIVER BREEDER REACTOR PLANT

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llTLE Hypothetical Core DOCUMENT NO.~

Disruptive Accident O- Considerations in CRBRP CRBRP-3 CHANGE CONTROL RECORD volume 2 volume 2

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Assessment of Thermal Margin Beyond the Design Base CHANGE REV NOlDATE RELEASE PAGES AFFECTED REMARKS DOCUMENT 0/ March 80 All First formal issue of document Rev 1/5-81 1, 2-4, 2-6, 2-7, 2-10 Pages added 2-16A thru 2-16, 2-24, 2-25, thru 2-16G, 2-34A 2-26, 2-32, 2-34, 2-43, 2-44, 2-76 Rev 2/10-81 2-16B, 2-16C, 2-16F, 2-16G Rev 3/2-82 i, 2-4, 2-18 Pa,ge added 2-18A

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Rev 4/6-82 ii, iii, vii, xi, xiii, xxiii, 2-7, 2-8, 2-10, Pages added 2-10A, 2-10B Revised Appendix A 2-76, 2-78, 3-24, 4-10, Added Section H.3 4-19, 4-24, A-1 to A-25, C.1-3, C.1-12, C.1-18, C.3-19, G.2-1, G.2-3, G.4-4 to G.4-6, H.0-1, H.1-4, H.1-5, H.1-9, H.3-1 to H.3-7 01142

o CRBRP-3 Vol. 2. Rev. 3 TABLE OF CONTENTS

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1.0 INTRODUCTION

1-1

1.1 REFERENCES

1-5 2.0 DESIGN FEATURES PROVIDING TERMAL MARGIN BEYOND TE DESIGN BASE 2-1 2.1 TMBDB FEATURES REQUIREENTS 2-2 2.1.1 General 2-2 2.1.2 Feature Requirements 2-3 2.1.3 Maintenance and Testing Requirements for 2-16A TMBDB Features 1

2.2 DESCRIPTION

OF DESIGN FEATURES 2-17 2.2.1 Reactor Cavity to Containment Barrier 2-17 2.2.2 Reactor Cavity Penetrations and Recirculating i 2-18 Gas Cooling System 3 2.2.3 Guard Vessel Support 2-18 2.2.4 Reactor Cavity and Pipeway Cell Liners 2-18A l 3 2.2.5 Reactor Cavity and Pipeway Cell Liner Vent 2-19

.# System 2.2.6 Reactor Cavity Vent System 2-22

( s 2.2.7 Containment Purge Capability 2-24 2.2.8 Containment Vent Capability 2-25 2.2.9 Containment Cleanup System 2-26 2.2.10 Annulus Air Cooling System 2-27 2.2.11 Containment System Leakage Barrier 2-29 2.2.12 TMBDB Instrumentation System 2-30 2.2.13 Electrical Power System 2-34 2.2.14 Containment Structures 2-34 2.2.15 Control Room Habitability 2-35 2.3 OPERATOR ACTION SEQUEhCE 2-38 2.4 RE FERENCE S 2-41 3.0 ASSESSFENT OF TKRMAL MARGIN 3-1 3.1 TERMAL MARGIN WITHIN TE REACTOR VESSEL 3-2 3.1.1 Core Debris Distribution 3-2 3.1.2 In-Vessel Debris Retention Capability 3-10 3.1.3 Secondary Criticality Considerations 3-15 3.1.4 Penetration of the Reactor and Guard Vessels 3-15 3.2 TKRMAL MARGIN EXTERNAL TO TE REACTOR VESSEL 3-18 3.2.1 Scenario 3-18 3.2.2 Containment Transients Prior to Sodium Boildry 3-26 C 3.2.3 Containment Transients After Sodium Boildry 3-60 3.2.4 Secondary Criticality Considerations (Ex-Vessel) 3-70 i

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CRBRP-3 Vol . 2, Rev. 4 O

TABLE OF CONTENTS (Continued)

Page

3.3 CONCLUSION

S ON THERMAL MARGIN BEYOND THE DESIGN BASE 3-72

3.4 REFERENCES

3-73 4.0 ASSESSMENT OF RADIOLOGICAL CONSEQUENCES 4-1 4.1 HCDA RADIOLOGICAL SOURCE TERM 4-1 4.1.1 Non-Energetic Core Meltdown 4-2 4.1.2 Energetic HCDA 4-5 4.2 RADIOLOGICAL DOSES FROM ATMOSPHERIC RELEASES 4-8 4.2.1 Methods and Data Base 4-8 4.2.2 Radiological Doses 4-10 4.3 GROUNDWATER CONSIDERATIONS 4-11 4.4 HETEROGENE0US CORE CONSIDERATIONS , 4-13

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4.5 CONCLUSION

S ON RADIOLOGICAL CONSEQUENCES 4-14

4. REFERENCES 4-15 5.0 SHMARY AND CONCLUSIONS 5-1 APPENDICES: A.0-1 INDEX TO SENSITIVITY STUDIES IN APPENDICES A.0-2 A. DEVELOPMENT PROGRAMS SUPPORTING THERMAL MARGIN ASSESSMENTS A-1 A.1 S0DIUM-CONCRETE INTERACTIONS DEVELOPMENT PROGRAM A-1 A.l.1 Purpose A-1 A.l.2 Program A-1 A.l.3 Schedule A-2 A.l.4 Criteria of Success . A-3 A.1.5 Fallback Position A-3 A.2 HYDROGEN AUT0-CATALYTIC RECOMBINATION A-4 A.2.1 Purpose A-4 A.2.2 Program A-4 A.2.3 Schedule A-5 A.2.4 Conclusions A-5 l4 ii

s CRBRP-3 Vol . 2, Rev. 4 OV T/SLE OF CONTENTS (Continued) .

Page,

'A.3 FURTHER VALIDATION OF THE CACECO COMPUTER CODE A-6 A.3.1 Purpose A-6 A.3.2 Program A-6 A.3.3 Schedule A-6 A.3.4 Conclusions A-7 A.4 COMPREHENSIVE TESTING PROGRAM FOR CONCRETE AT ELEVATED A-8 TEMPERATURES A.4.1 Purpose A-8 A.4.2 Program A-8 A.4.3 Schedule A-11 A.4.4 Conclusions A-12 A.5 BASE MATERIALS TEST FOR LINER STEEL A-13 A.5.1 Purpose A-13 A.S.2 Program A-13 A.S.3 Schedule A-15 t,

A.S.4 Criteria of Success A-15 A.S.5 Fallback Position A-16 A.6 S0DIUM SPILL DESIGN QUALIFICATION TEST (LT-1) A-17 A.7 TMBDB AIR CLEANING SYSTEM PERFORMANCE TESTS A-18 A.7.1 Purpose A-18 A.7.2 Program A-18 A.7.3 Schedule A-20 A.7.4 Conclusions A-21 A.8 REFERENCES A-22 4 B. REACTOR AND GUARD VESSEL PENETRATION ANALYSIS B-1 B.1 INTRODUCTION B-1 B.2 ASSUMPTIONS B-3 B.3 DESCRIPTION OF PENETRATION ANALYTICAL MODEL B-5 B.4 DESCRIPTION OF CASES B-6 B.5 RESULTS B-9 B.6 CONCLUSIONS B-12 B.7 REFERENCES B-13 4

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4 CRBRP-3

Vol.2, Rev.0 TABLE OF CONTENTS (Continued)

Page C. ANALYTIC THERMAL MODELS AND DATA BASE C.0-1 C.1 PRE-S0DIUM BOILDRY MODEL AND DATA BASE C.1-1 C.I.1 Physical Description C.1-1 C.1.2 Heat Sources C.1-3 C.1.3 Heat and Mass Transfer C.1-5 C.1.4 Material Properties C.1-8 C.1.5 References C.1-ll C.2 POST S0DIUM BOILDRY MODEL AND DATA BASE C.2-1 C.2.1 An'a' lytic'al Model C.2-1 C.2.2 Data Base C.2-3 .

C.2.3 References C.2-5 C.3 STRUCTURAL MATERIAL PROPERTIES AT ELEVATED TEMPERATURES C.3-1 C.3.1 Introduction C.3-1 C.3.2 Concrete Properties C.3-1 C.3.3 Properties of Reinforcing Steel C.3-11 C.3.4 Properties of Liner Steel C.3-12 C.3.5 References C.3-16 ,

D. AEROSOL BEHAVIOR MODEL D-1

0.1 INTRODUCTION

0-1 0.2 AEROSOL MODELING AND BASES D-1 0.3 SENSITIVITY TO RCB TEMPERATURE D-4

0.4 REFERENCES

D-6 E. RELEASE OF PLUT0NIUM AND FISSION PRODUCTS E-1

E.1 INTRODUCTION E-1 E.2 PLUT0NIUM RELEASE FROM S0DIUM-0XYGEN REACTIONS IN THE E-1 INERTED CELLS E.3 PLUTONIUM RELEASE FROM A BOILING POOL OF SODIUM E-2 E.4 PLUT0NIUM RELEASE AFTER S0DIUM BOILDRY E-4

, E.5 CONCLUSIONS ON PLUT0NIUM RELEASE E-6 E.6 FISSION PRODUCT RELEASE TO THE RCB E-6 i E.7 EFFECT OF CO2 REACTION WITH NaOH ON FISSION PRODUCT E-8 RELEASE FROM THE RCB -

l E.8 REFERENCES E-10 F. EVALUATION OF ALTERNATE SCENARIOS F.0-1 l

l F.1 EFFECT OF DECAY HEAT ON THE BASE TMBDB SCENARIO F.1-1 i F.1.1 Introduction F.1-1 F.1.2 Model Modifications F.1-1 ,

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Vol . 2, Rev. 4 O

1 Q

TABLE OF CONTENTS (Continued)

.P.,aSe H.2 FLAME LENGTHS FOR TMBDB SCENARIO H.2-1 H.2.1 Introduction H.2-1 l H.2.2 Flame Length Calculation H.2-2 H.2.3 Effects on Containment H.2-6 H.2.4 Conclusions H.2-8 H.2.5 References H.2-9 H.2.A Flame Length Correlation H.2-16 H.3 POTENTIAL FOR HYDROGEN STRATIFICATION H.3-1 i

H.3.1 Introduction H.3-1 II.3.2 Pre-Hydrogen Ignition Tinie Period H.3-1 H.3.3 Hydrogen Ignition to Sodium Boildry Time H.3-4 H.3.4 Post Sodium Boildry H.3-4 H.3.5 Conclusions H.3-5 H.3.6 References H.3-5 4 I. ASSESSMENT OF CONSEQUENCES OF FUEL IN PHTS PIPING FOLLOWING I-1 REACTOR VESSEL DRAINING

> I.1 INTRODUCTION 1-1 1.2 EFFECT ON PHTS PIPING I-3 I.3 EFFECT ON CELL LINERS I-5 I.4 SENSITIVITY ASSESSMENT OF FUEL IN PHTS PIPING I-7 J. NRC'S REQUESTS FOR ADDITIONAL INFORMATION -J-1 (RAI's) ON CORE MELTDOWN CONSIDERATIONS

O vii

o CRBRP-3 Vol.2 Rev.0 LIST OF TABLES Table Number P_aSe 2-1 CRBRP FEATURES PROVIDING THERMAL MARGIN BEYOND THE 2-42 DESIGN BASE 3-1 PUMP C0ASTDOWN DATA 3-76 3-2 PARTICULE DISTRIBUTION FOLLOWING IN-VESSEL 3-77 SETTLING e 3-3 DISTRIBUTION OF UPWARD EJECTED DEBRIS 3-78 3-4 PARTICLE DRAG COEFFICIENTS 3-79 3-5 MAXIMUM CRBR FULL EQUILIBRIUM CYCLE DECAY POWER BY 3-80 REGION 3-6 STABLE DEBRIS BED DFPTHS 3-81 3-7 COMPARISON OF FUEL RETENTION CAPABILITY WITH 3-82 PREDICTED FUEL DISTRIBUTION 3-8 SENSITIVITY OF CONTAINMENT PARAMETERS AT 24 HOURS 3-83 AFTER AN HCDA TO VESSEL PENETRATION TIME 3-9 ANNULUS COOLING SYSTEM ANALYSIS DATA 3-84 ,

3-10 CONTAINMENT STRUCTURAL CAPABILITY 3-85 3-11 Keffective 0F VARIOUS GE0METRIES 3-86 3-12 REACTOR CAVITY AND PIPEWAY CELL LINER FAILURE TIMES 3-87 3-13 HEAT TRANSFER COEFFICIENT USED IN THERMAL ANALYSIS 3-88 0F THE REACTOR CAVITY LEDGE 3-14

SUMMARY

OF RESULTS FOR SUBMERGED LINER (WITHOUT 3-89 l' CREEP) 3-Ir

SUMMARY

OF RESULTS FOR LINER AB0VE S0DIUM P0OL 3-90 (WITHOUT CREEP) 4-1 CORE SOURCE TERMS RELEASED TO THE REACTOR 4-17 CONTAINMENT BUILDING FOR HYPOTHETICAL ACCIDENT SCENARIOS CONSIDERED 4-? ATMOSPHERIC DiltlTION FACTORS 4 18 4-3 DOSE

SUMMARY

FOR HYP0THETICAL ACCIDENT SCENARIOS 4-19 CONSIDERED

, O viii i

CRBRP-3 Vol . 2, Rev. 4 LIST OF TABLES (Continued)

Table Number P_ age F.7-1 EFFECTS OF PIPEWAY FLOOR LINER FAILURE ON F.7-4 CONTAINMENT CONDITIONS AT 24 HOURS F 7-2 EFFECTS OF PIPEWAY FLOOR LINER FAILURE ON F.7-5 CONTAINMENT CONDITIONS AT 30 HOURS F.7-3 EFFECTS OF PIPEWAY FLOOR LINER FAILURE ON F.7-6 CONTAINMENT CONDITIONS AT 36 HOURS F.7-4 EFFECTS OF PIPEWAY FLOOR LINER FAILURE ON F.7-7 CONTAINMENT CONDITIONS AT 50 HOURS F.7-5 EFFECTS OF PIPEWAY FLOOR LINER FAILURE ON F.7-8 CONTAINMENT CONDITIONS AT BOILDRY F.7-6 DEPENDENCE OF PEAK CONTAINMENT CONDITIONS TO F.7-9 SODIUM BOILDRY ON PIPEWAY FLOOR LINER FAILURE ASSUMPTIONS F.7-7

SUMMARY

OF CONTAINMENT CONDITIONS FOR "LESS F.7-15 SEVERE" LINER FAILURE SENSITIVITY ANALYSES G.1-1 ASSESSMENT OF POTENTIAL FOR BED DRYOUT IN THE G.1-7 REACTOR CAVITY G.2-1

SUMMARY

OF SENSITIVITY STUDIES ON S0DIUM-CONCRETE G.2-4 REACTIONS G.3-1 SENSITIVITY OF RCB AND RC CONDITIONS AT 24 HOURS G.3-4 TO RC VENT FAILURE MODES G.4-1 EFFECT OF INITIAL FUEL, FISSION PRODUCT, AND G.4-4 S0DIUM RELEASE G.4-2 EFFECT OF INITIAL FUEL AND FISSION PRODUCT RELEASE G.4-5 WITH SODIUM RELEASE FIXED AT 1000 LB.

G.4-3 EFFECT OF SODIUM RELEASE FOR A GIVEN 10% G.4-6 FUEL-FISSION PRODUCT RELEASE H.1-1 CONTAINMENT ATMOSPHERE CONDITIONS AT 24 HOURS H.1-10 H.3-1 ATMOSPHERIC ENTRAINMENT IN HYDROGEN JET H.3-6 4

I-1 DISTRIBUTION OF UPWARD EJECTED DEBRIS FOR LOF HCDA I-8 I-2 HEAT GENERATION AS A FUNCTION OF TIME FOR TOTAL I-9 CORE AND BLANKETS O I-3 PARAMETERS FOR PIPE MELTING ANALYSIS I-10 I-4 FLOOR PENETRATION ANALYSIS PARAMETERS I-11 xi l

. I CRBRP-3 .

Vol.2, Rev.0

)

i LIST OF FIGURES Figure Number Page 1-1 Design Features Providing Thermal Margin Beyond 1-6 the Design Base 2-1 Maximum Pressure Differential Between the Reactor 2-43 Cavity and Head Access Area 2-2 Maximum Reactor Cavity Atmosphere Pressure 2-44 2-3 Maximum Reactor Cavity Atmosphere and Sodium Pool 2-45 Temperature 2-4 Maximum Pressure Differential Between the Reactor 2-46 Cavity and Cell 105 2-5 Maximum Containment Atmosphere Temperature 2-47 2-6 Maximum Containment Atmosphere Pressure 2-48 2-7 Maximum Containment Vent Rates 2-49 2-8 Maximum Heat Load Through the Containment 2-50 Structure Steel Shell 2-9 Maximum Reactor Cavity Submerged Wall Structural 2-51 "

Temperatures 2-10 Maximum Reactor Cavity Floor Concrete Temperatures 2-52 2-11 Maximum Reactor Cavity Non-Submerged Wall 2-53 Structural Temperatures 2-12 Maximum Reactor Cavity-Pipeway Wall Structural 2-54 Temperatures 2-13 Maximum Pireway Cell Wall Structural Temperatures 2-55 (2.5Ft.ThickWall) l 2-14 Maximum Pipeway Floor Structural Temperatures 2-56 i

! 2-15 Maximum Pipeway Cell Wall Structural Temperatures 2-57 l (4.0 ft. Thick Wall) l 2-16 Maximum Pipeway Cell Roof and Head Access Area 2-58 Wall Structural Temperatures 2-17 Reactor Cavity Ledge Radial Temperature Profile 2-59 l E xii l

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CRBRP-3 s

Vol . 2, Rev. 4 LIST OF FIGURES (Continued) 2-18 Reactor Cavity Ledge Axial Temperature Profile 2-60 2-19 Maximum Reactor Cavity Upper Wall Temperatures Prior 2-61 to Wall Failure 2-20 M'aximum Reactor Cavity Lower Wall Temperatures Prior 2-62 to Wall Failure 2-21 Maximum Reactor Cavity Floor and Basemat (Axial 2-63 Centerline) Temperatures 2-22 Maximum Reactor Cavity Floor (Elevation 733.5) 2-64 Temperatures 2-23 Maximum Basemat (Elevation 728.0) Temperatures 2-65 2-24 Maximum Basemat (Elevation 718.0) Temperatures 2-66

.i 2-25 Maximum PHTS Wall (Facing RC Wall) Temperatures 2-67 2-26 Maximum PHTS Wall (Not Facing RC Wall) Temperatures 2-68 2-27 Maximum Containment Wall Temperatures (Below 2-69 O Elevation 816.0) 2-28 Maximum Confinement Wall Temperatures (Below 2-70 Elevation 816.0) 2-29 Maximum Temperatures of Wall Separating Cells 102 2-71 and 105 2-30 Maximum Operating Floor Temperature 2-72 2-31 Location of Post Boildry Temperature Profile Plots 2-73 2-32 Guard vessel Skirt Support Detail 2-74 2-33 Reactor Cavity Vent System 2-75 2-34 Containment Vent, Purge and Cleanup Systems 2-76 2-35 CRBRP Containment TMBDB Instrumentation 2-77

, 2-36 Atmospheric Temperatures for Determining Containment 2-78 Penetrations Environment (Beyond 100 Hours) 4 3-1 10 x 11 Array VARR-II Outlet Plenum Model with 3-91 Complex UIS Geometry i

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CRBRP-3 .

Vol.2, Rev.0 .

LIST OF FIGURES (Continued)

Figure Number Page 3-2 Simplified Drawing of Settling in Horizontal Pipes 3-92 3-3 Particle Size Distributions From Large-Scale 3-93 Out-of-Pile Tests 3-4 Hydrogen Burning Criterion 3-94 3-5 CACECO Code Containment Model 3-95 3-6 Reactor Cavity and Containment Atmosphere and 3-96 Containment Steel Dome Temperatures 3-7 Reactor Cavity and Containment Pressures 3-97 3-8 Sodium Entering Conta;nment From the Reactor 3-98 Cavity 3-9 Integrated Water Release from Concrete Structures 3-99 3-10 Sodium Concentration Entering Containment 3-100 3-11 Containment Building Hydrogen Concentration 3-101 3-12 Containment Oxygen Concentration 3-102 .'

3-13 Gases leaving Containment 3-103 3-14 Air Purge Rate into Containment 3-104 3-15 Total RCB Sodium 0xide and Sodium Hydroxide 3-105 Inventory (Neglecting Aerosol Leakage) 3-16 Reactor Cavity Submerged Wall Concrete 3-106 Temperatures 3-17 Reactor Cavity Floor Concrete Temperatures 3-107 3-18 Reactor Cavity Non-Submerged Concrete Wall 3-108 Temperatures 3-19 Reactor Cavity - Pipeway Wall Concrete 3-109 Temperatures 3-20 Pipeway Cell Wall Structural Temperatures 3-110 (2.5 ft. Thick Wall) 3-21 Pipeway Cell Wall Structural Temperatures 3-111 (4.0 Ft. Thick Wall)

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CRBRP-3

Vol . 2, Rev. 4 D LIST OF FIGURES (Continued)

Figure Number g H.1-6 Flammability Limits for Predicted Scenario and H.1-16 Hydrogen Burning Scenario that Bounds Containment Pressure and Temperature H.1-7 Effect of Hydrogen Burning Assumptions on Containment H.1-17 Atmosphere Temperatures (Without Containment Venting and Purging)

H.1-8 Effect of Hydrogen Burning Assumptions on Containment H.1-18 Atmosphere Pressure (Without Containment Venting and Purging)

H.1-9 Effect of Hydrogen Burning Assumptions on Containment H.1-19 Atmosphere Hydrogen Concentration (Without Containment Venting and Purging)

H.1-10 EffectofHydrogenBurningAssumptionsonContainmeni H.1-20 Conditions with Respect to Flammability Limits ,

(Without Containment Venting and Purging)

H.2-1 Flow Patterns in Laminar and Turbulent Flames H.2-10 H.2-2 TMBDB Mass Flow Rate Through the Reactor Cavity to H.2-ll O

d RCB Vent

.H.2-3 Dependence of Flame Length on Nozzle Diameter H.2-12 H.2-4 Multiple Nozzle Arrangement H.2-13 H.2-5 Dependence of Flame Length on Number of Nozzles H.2-14 (0.7853 FT2 Total Area)

H.2-6 Temperatures on Axis of a Hydrogen Diffusion Flame H.2-15 (Burner Diameter = 4 Inches, Flow Rate = 3.5 CFM)

H.2.A-1 Flame Length Correlation (Figure 255 of Ref. H.2-1) H.2-18 H.2.A-2 Effect of Nozzle Velocity on Flame Length (Figure 246 H.2-19 of Ref. H.2-1)

H.3-1 Flow Patterns for Jet Injection Into a Closed Container H.3-7 4

i I-l PHTS Pipe and Insulation Configuration I-12 I-2 Peak Pipe Temperature versus Fuel Content I-13 1-3 Sketch of Fuel Debris In Pipe I-14 I-4 Cell Liner Thermal Model Schematic I-15 l

I-5 Liner Mid-Plane Peak Temperature versus Fuel Density I-16 xxiii

CRBRP-3 .

Vol.2, Rev.0 ABBREVIATIONS ANL Argonne National Laboratory ARD Advanced Reactors Division (of Westinghouse Electric Corporation)

ASME American Society of Mechanical Engineers ATWS Anticipated Transients Without Scram BOEC Beginning of Equilibrium Cycle BOL Beginning of Life CAM Continuous Air Monitors CFR Code of Federal Regulation CRBR Clinch River Breeder Reactor CRBRP Clinch River Breeder Reactor Plant CRDM Control Rod Drive Mechanism CSS Core Support Structure DHRS Direct Heat Removal Service DOE U. S. Department of Energy E0EC End of Equilibrium Cycle ERDA U.S. Energy Research and Development Administration m

ESF Engineered Safety Features j FMEA Failure Mode and Effects Analysis FFTF Fast Flux Test Facility FWP Feedwater Pump GE General Electric Conpany HAA Head Access Area HCDA Hypothetical Core Disruptive Accident HEDL Hanford Engineering Development Laboratory HEPA High Efficiency Particulate Air HTS Heat Transport System HVAC Heating, Ventilating and Air Conditioning IEEE Institute of Electrical and Electronic Engineers IHTS Intermediate Heat Transport System IHX Intermediate Heat Exchanger j IMAS Impurity Monitoring and Analysis System l IRP Intermediate Rotating Plug IVTM In-Vessel Transfer Machine l

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2. To insure containment atmosphere mixing before venting, the purge air shall be injected into containment below elevation 840'.

3. The purge system shall prevent backflow from containment to the outside atmosphere. ,

4. The purge s'ystem, in combination with the containment vent and cleanup systems, shall maintain containment at a negative pressure after the containment pressure is reduced by the initial venting after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

5. The purge system operations shall be by remote manual actuation from the control room.

1 2.1.2.8 Containment Vent System i .

1. To prevent containment failure by excessive pressure, the vent system t ,/ shall have a capacity between 24,000 and 26,400 acfm with a containment pressure of 30 psia, a containment atmosphere density of 0.07 lb/ft 3 and a 91scosity of 0.06 lb/ft-hr. It shall remain functional if up to 300,000 pounds of aerosol enter the system at a maximum rate of 5,600 lb/hr.

2. The vent system shall exhaust the containment atmosphere from the containment into the containment cleanup system. 4

3. The containment vent system shall be compatible with the following gases, vapors and aerosols: Ar, Np , H2, H 0, CO, CO , 0 , Na 0, Na 0 '

2 2 2 2 22 ha0H, Na 2 3 00 , fission products, and compounds resulting from fission product reactions. The system must remain functional for inlet gas temperatures and pressures given on Figures 2-5 and 2-6, and, beyond 150 hours0.00174 days 0.0417 hours 2.480159e-4 weeks 5.7075e-5 months for temperatures up to 250 F.

4

4. The vent system operations shall be by remote manual actuation from the control room.

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CRBRP-3 Vol . 2, Rev. 4 0

2.1.2.9 TMBDB Containment Cleanup System

1. The containment cleanup system efficiency shall be a minimum of 99% for vented materials in the solid or liquid state, 97% fo' vapors r (NaI, Se02 , and Sb230 ) subject to condensation in the cleanup system, and 0% for noble gases. These efficiencies shall apply when subjected to the vent rates on Figure 2-7 and containment atmosphere temperatures on Figure 2-5 with a containment atmosphere density of 0.07 lb/ft3; beyond 150 hours0.00174 days 0.0417 hours 2.480159e-4 weeks 5.7075e-5 months , containment atmosphere temperatures up to 250 F shall apply. It shall be capable of performing all of its intended 4 functions in the presence of Ar, N ,2 H , H 0,2 C0, CO , 0 , Na 0, Na 0 '

2 2 2 2 22 Na0H, Na 2 C 3, fission products, and compounds resulting from fission product reactions.

2. The containment cleanup system shall remain functional at an aerosol mass flow rate of up to 5,600 lb/hr and a total mass of 300,000 pounds of aerosol entering the cleanup system. The principal constitutents of the aerosol are Na0H and Na 2 0, the proportions of which can vary from 0 to 100% of the aerosol, and NapC0 3 which can vary from 0 to 8% of the aerosol.

The aerosol particle properties are:

Mass Mean Radius (microns): 5 < r50 < 10 Aerodynamic Equivalent Radius (microns): 2.3 < AER < 4.7 Density (g/cc): 2.1 < p < 2.5 Mass Geometric Standard Deviation: 3.0 < a < 3.5 Aerodynamic equivalent radius is based on AER = r **

50 where p = 2.21 and a = 0.1 0

2-8

CRBRP-3 Vol.2. Rev.0 v 3. The containment cleanup system shall remain functional at fission products power levels in the accumulated filter aerosol of:

Time Fission Product Power (hours) (MW) 0 0 24 ,

3.1 x 10-5 48 0.16*

96 ,

0.16*

240 0.11 720 0.05

4. The containment cleanup system design shall be capable of performing all its intended functions with the following chemical and physical states of the 10 most radiologically significant fission products in the containment atmosphere:

Maximum value.

MAXIMUM PERCENTAGE OF THE FISSION PRODUCTS BY CHEMICAL AND PHYSICAL FORM Elemental Oxide Element Vapor Liquid or Solid Vapor Liquid or Solid Se 1% 1% 100% 100%

Rb 1 1 1 100 Sr 1 1 1 100 Zr 1 1 1 100 Sb 1 1 100 100 Te 1 1 , 1 100 Cs 1 1 1 100 Ba 1 1 1 100 Ce 1 1 1 100 Nal 1 1 1 33 100 0

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CRBRP-3 Vol . 2, Rev. 4

5. The exhaust from the containment cleanup system shall have a temperature compatible with operation of the TMBDB Exhaust-Plant Effluent Radiation Monitoring System.

6. The containment cleanup system operations shall be by remote manual actuation from the control room.

1 2.1.2.10 Annulus Cooling System

1. To insure containment and confinement do not fail from excessive temperatures, the annulus cooling system shall remove the heat load into the containment steel shell on Figure 2-8.

2. Steel containment temperatures shall be below those that cause structural failure or excessive containment leakage.

3. Concrete confinement temperatures shall be below those that cause O

structural failure.

4. The annulus cooling system operations shall be by remote manual actuation from the control room.

1 l

l 2.1.2.11 Containment System Leakage Barrier At any given time, containment out-leakage shall not exceed the greater of:

4

1. The design leakrate (0.1 volume percent per day).

2. The design leakrate adjusted for pressures above the containment design pressure of 10 psig. Leakrate = besign Leakrate x (Actual Pressure (psig)).5/3.2.

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CRBRP-3 Vol . 2, Rev. 4 O

All containment penetrations, including fluid systems piping and containment isolation valves, electrical penetrations, and hatches, shall maintain their integrity such that the above leakages are not exceeded for at least 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months following the initiation of the TNBDB scenario. The limiting TMBDB temperature, pressure and radiation environmental requirements for penetrations during the 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months period shall be as follows:

CONTAINMENT PENETRATIONS'ENVIROM4ENT (0-100 HOURS)

Peak Total Containment Maximum Radiation Accumulated Penetration Temperature Maximum Level Dose Location (OF) Pressure (R/hr) (R)

PHTS Cells 250 Figure 2-6 1 x 106 3 x 106 RCB Above Figure 2-5 Figure 2-6 1 x 106 1 x 107 Operating Floor O. All Other Cells 250 Figure 2-6 1 x 106 3 x 106 After containment venting / purging is initiated (24 to 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months ),

containment in-leakage shall not exceed 1000 scfm at a negative pressure s inside of -3.5 inches water. s 3

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CRBRP-3 Vol . 2, Rev. 4 O

The limiting TNBDB temperature, pressure and radiation environmental requirements for containment penetrations after 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months are as follows:

CONTAINMENT PENETRATIONS ENVIRONMENT (Beyond 100 Hours)

Peak Total Containment Maximum Minimum Radiation Accumulated Penetration Temperature Pressure Level Dose Location (OF)" (In. Water) (R/hr) (R)

PHTS Cells Figure 2-36 -3.5 5 x 103 1 x 107 RCB Above Figure 2-5* -3.5 1 x 105 1 x 108 Operating Floor All Other Cells Figure 2-36 -3.5 5 x 103 1 x 107

Beyond 150 hours0.00174 days 0.0417 hours 2.480159e-4 weeks 5.7075e-5 months , the maximum temperature is 2500F. Figure 2-5 defines the atmosphere temperatures. The designer shall determine the appropriate penetration temperature that would result from the atmosphere temperatures. Figures 3-27 and 3-28 provide selected steel shell temperatures that can also be used for this determination.

Containment penetrations above the operating floor shall perform their function in the presence of the following plated sodium aerosol (including NaOH, Na 0, and Na CO )*

2 2 3 Vertical Surfaces: 0.5 lb/ft 2 Horizontal Surfaces: 80 lb/ft 2 4

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'~ ELEVATION 715' KEY G HYOR0 GEN MEASUREMENT SYSTEM SAMPLE POINTS O REACTOR CONTAINMENT VESSEL TEMPERATURE a REACTOR CONTAINMENT ATMOSPHERE TEMPERATURE O REACTOR CONTAINMENT PRESSURE Figure 2-35 CRBRP Containment TMBDB Instrumentation 3379-4 O

2-77

CRBRP-3 '

Vol . 2, Rev. 4 .

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Figure 2-36. Atmospheric Temperatures for Determining Containment Penetrations Environment (Beyond 100 Hours) 7184-1 0

2-78

CRBRP-3 Vol.2, Rev.0 (v ). Criteria for hydrogen burning:

i

a. The hydrogen-nitrogen mixture entering containment is above 14500F.

b. The hydrogen-sodium-nitrogen mixture entering containment contains at least 6 g/m3 of sodium at temperatures above 5000F.

c. The oxygen concentration is above 8%. With the oxygen concentration above 5% and the hydrogen concentration above 4%, the hydrogen in excess of 4% would burn. Figure 3-4 illustrates this '

burning criterion.

Criterion (a) is not met for the analyses reported herein. The sodium concentration entering the reactor containment building would satisfy criterion (b) after about ten hours following vessel penetration. At the end of this time the hydrogen concentration in containment would reach approximately 4.5%. This hydrogen would burn while the burning criteria are satisfied as the natural circulation in containment moves the hydrogen through the flame. The oxygen concentration is predicted to remain above 8% (satisfying criterion c) for 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months ; thus, no ,

hydrogen accumulation would occur during this time (i.e.10 to 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months ).

The flame characteristics are shown in Appendix H.2 to be such that excessive local containment steel temperatures would not occur.

When hydrogen burning criteria are no longer satisfied, hydrogen would accumulate in the containment. The hydrogen concentration would be controlled to less than 6% by purging (see item 17). Since the v'

3-23

CRBRP-3 Vol . 2, Rev. 4 O

hydrogen concentration is maintained well below explosive levels, the containment integrity would not be challenged by hydrogen reactions.*

14. Water vapor from concrete in the operating floor and head access area would enter the containment atmosphere.

The sodium oxide created by the sodium-oxygen, sodium-carbon dioxide, and sodium-water reactions in containment would react with the excess water vapor in the containment atmosphere to form sodium-hydroxide.

15. The maximum containment pressure of 22 psig (peak pressure results from assumption that accumulated hydrogen burns instantly) is well below the failure pressure. Scoping calculations indicate that the failure pressure is in excess of 30 psig for these conoitions (see Table 3-10).

16. At 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months the annulus cooling system is assumed to be activated to maintain the containment temperature at an acceptable level and the containment is vented to avoid excessive long term pressure and to allow a purge of fresh air to be initiated.

Appendix H.3 considers the potential for a high concentration of the lighter hydrogen collecting at the top of the RCB due to buoyancy forces (stratification). It is concluded that in the TMBDB scenario the mixing forces would be more than adequate to counteract buoyancy forces and to prevent any significant collection (stratification) of hydrogen at the top of the RCB. 4 O

3-24

d CRBRP-3 Vol.2, Rev.0 Dose Factors Dose conversion factors (rem /ci) used in the COMRADEX code to calculate specific organ doses were taken from References 4-8 and 4-9 where possible. Factors for isotopes not given in these References are from Reference 4-10.

Containment Modeling The time dependent radiological source term is released directly to the RCB. The release rate from the RCB is that calculated by the CACECO code.

For the first 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months of the scenario the RCB atmosphere leaks at a low rate (based on 0.1%/ day at 10 psig) to the annulus filter system (described in Section 6.2.5 of the PSAR). During this 36 hour4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months period unfiltered bypass leakage at the rate of 1% of the filtered leakage is considered.

After 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months the RCB is vented and subsequently purged (Figures 3-13 and 3-14) to maintain the hydrogen concentration at an acceptable level.

During this phase filtering is by the TMBDB filter system which is designed for the higher vent rates. The efficiency of the TMBDB filter system is

{ 99% for solid fission products and fuel and 97% for condensible fission products (halogens, Se, and Sb). Noble gases are assumed to pass through

, the filter system unattenuated. (There is some question of the effectiveness of the filter system to remove Na 2CO3 and the fission products which may be tied up with'this aerosol component. This subject is addressed in Appendix E.) Because the bypass leakage rate is expected to be so small relative to the high vent rate af ter 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months , the bypass leakage is not expected to make a significant contribution to the released l radioactivity and is therefore not considered beyond 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

l The direct ganna contribution to the whole body dose considers the l

shielding provided by the steel RCB and the concrete confinement building.

Figure 4-1 shows the dose rate inside the reactor containment building for Case 2.

4-9 l

CRBRP-3 Vol. 2, Rev. 4 0

4.2.2 Radiological Doses Using the methods described in Section 4.2.1 the radiological coses at the Exclusion Boundary (0.42 miles) and the Low Population Zone (2.5 miles) were calculated for the four different source terms described in Section 4.1. These doses are summarized in Table 4-3. The 30 day LPZ doses include the plutonium released after boil-dry to 30 days. Plutonium release beyona 30 days could result in an additional 10 rem to the LPZ bone dose. Control room doses were provided in Section 2.2.15.

The dose consequences of the four cases that assumed varying degrees'of severity of the hypothetical accident are all quite low for accidents beyond the design base. For example, the maximum whole body dose is predicted to be about 3.5 rem and the maximum thyroid dose would be about 100 rem. Bone doses are about 55 rem. 4 The results also show that the consequences are not strongly sensitive to O

the degree of severity of the initial release source term. As the initial release to the RCB increases, the rate of aerosol depletion increases which acts as an inverse feedback to limit the release from the RCB.

Consequently, so long as the initial release does not result in failure of the containment barrier, the radiological consequences are relatively insensitive to the magnitude of the release. For the full range of releases considered in Cases 1 through 4, the RCB pressure and temperatures woula not result in failure of the containment barrier.

l l Table 4-4 compares the consequences, in terms of curies released, of a comparable scenario (core meltdown with enough containment leakage to prevent containment failure by overpressure) for CRBRP and light water reactors (LWR). The CRBRP values are for the worst of the above four cases. The LWR releases are for the accident scenarios PWR-6 and BWR-4 described in Sectico 2 of Appendix VI of WASH-1400. This comparison shows the atmospheric releases for CRBRP to be comparable to those for LWRs.

Figure 4-2 shows the integrated radioactivity released to the environment for Case 2.

4-10

Os CRBRP-3 Vol. 2, Rev. 4 TABLE 4-3 00SE SINMARY FOR HYPOTHETICAL ACCIDENT SCENARIOS CONSIDERED Doses in REM -

~0rgan Case 1 Case 2 Case 3 Case 4 Bone 0.0043 0.028 0.93 3.83 2 Hour Lung 0.0035 0.0055 0.15 0.39 Exclusion Thyroid 0.0067 0.0096 11.3 9.51 Boundary W. Body 0.16 0.16 0.24 0.32 30 Day Bone 55.1 55.1 55.7 56.2 Low Lung 3.95 3.96 3.02 3.02 Population Thyroid 99.2 99.2 5.31 1.72 Zone W. Body 3.51 3.50 3.07 2.94 4

l O

4-19.

CRBRP-3 Vol.2, Rev.0 TABLE 4-4 COMPARISON OF RADIONUCLIDE RELEASES TO ATMOSPHERE FOR CRBRP WITH LWR'S FOR A COMPARABLE MELTDOWN SCENARIO Radioactivity Released (curies)

Element CRBRP PWR(3) BWR(3)

Xe-Kr 2.4 x 107 1.0 x 108 2.1 x 108 I 1.6 x 105 2.0 x 106 1,1 x 106 Cs, Rb 5.4 x 100 1.2 x 104 7.6 x 104 Te, Sb 3.5 x 104 2.2 x 105 8.6 x 105 Ba, Sr 6.5 x 102 3.3 x 104 2.2 x 105 Ru(1) 1.5 x 103 3.9 x 104 3.3 x 105 La(2) 3.7 x 103 2.9 x 104 2.9 x 105 (1) Includes: Ru, Rh, Co, Mo, Te I2) Includes: Y, La, Zr, Nb, Ce, Pr, Nd, Np, Pu, Am, Cm (3)From WASH-1400 Appendix VI, Calculation of Reactor Accident Consequences, October 1975. The LWR scenarios used for comparison here are PWR-6 and BWR-4 described in Section 2 of WASH-1400 Appendix VI.

I 1

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O 4-20

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Figure 4-2. Integrated Radioactivity Released to Environment for Case 2 7308-1 4-24

, 1 CRBRP-3 !

Vol. 2, Rev. 4 ;

O APPENDIX A DEVELOPMENT PROGRAMS SUPPORTING THERMAL MARGIN ASSESSMENTS This appendix describes the development programs completed and the remaining experimental work scheduled to complete support of TMBDB analyses. 4 A.1 S0DIUM-CONCRETE INTERACTIONS DEVELOPMENT PROGRAM A.l.1 Purpose The large number of sodium-concrete interaction experiments performed at HEDL and Sandia (Appendix C.1, and References A-1 through A-3) resulted in the model of this phenomenon used in the analysis 'in Section 3.2.2. These tests included both bare concrete and simulated faulted liners. The. sodium-concrete experimental program objective has been to increase the experimental data O

t base, to confirm the self-limiting nature of sodium-limestone concrete interactions and to further understand the underlying phenomena involved. 4 A.l.2 Program Two additional large scale sodium-concrete reaction tests have been performed.

4 A large Sodium-Concrete Test (designated LSC-2) has been performed with bare concrete. The test specimen had an interaction surface of 3 feet by 3 feet and used limestone concrete prototypic of CRBRP. The specimen was approximately 2 feet thick. Approximately 1000 pounds of hot (s1100 F) sodium was poured onto the test specimen and heated to approximately 1600 F and maintained near this temperature for about 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months .

O A-1

CRBRP-3 Vol. 2, Rev. 4 O

A Large-Scale Faulted Liner Feature Test (designated LFT-6) has also been performed with a steel liner and a layer of Mg0 gravel above the concrete specimen. The liner contained a 6 inch diameter centered hole. The concrete specimen was similar to that described above for the LSC-2 test.

Approximately 1000 pounds of hot (*1100 F) sodium was poured onto the test specimen and heated to approximately 1600 F and maintained near this temperature for about 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months .

Monitoring during the tests and post-test examinations provided information on temperature histories, gas release rate and composition, depth of sodium penetration into the concrete and sodium-concrete reaction product composition. Details of the LSC-2 and LFT-6 tests are reported in Reference A-8.

2 Several additional intermediate-scale tests (1 ft concrete surface area) ,

have been completed. Reference A-9 presents a comprehensive summary of all the sodium-concrete reaction tests completed as of December 1981, both in this development program and tests from other programs at U.S. labs and abroad.

Reference A-16 is an executive summary of the sodium-concrete reaction test data and its application to limestone concrete.

A.l.3 Schedule Additional tests scheduled in this development program are as follows:

Scheduled Tests Description Completion

1. Dolomite Aggregate This series of tests will determine July 1982 Characterizations differences, if any, in sodium reactions with calcitic limestone and dolomitic limestone aggregates.

l l

2. AB/AA Comparison This 4 test series will assess July 1982 Tests differences, if any, in results at HEDL and Sandia labs with both dolomitic and calcitic limestone l concretes.

4 A-2

CRBRP-3 Vol. 2, Rev. 4 O

Scheduled Tests Description Completion

3. LCT-Series This 11 test series will investigate December 1982 the effects on sodium-concrete reactions of CRBRP design features such as surface orientation (horizontal or vertical), presence of insulating concrete, restraint, external pressure, embedments in the concrete, and pre-dehydration (simulating delayed liner failures). 4 A.l.4 Criteria of Success In order to confirm the scenario for the Thermal Margin Beyond the Design Bar,e evaluation, this program is required to evaluate the penetration and interaction of sodium with the reactor cavity concrete floor including the effects of a potential floor liner failure. "The criteria of success are that the test program confirms th'ta the sodium-concrete reactions are approximately as modeled and are self-terminating because of the buildup of reaction products, and that spalling and mechanical breakup of the concrete will not enable the reaction front to proceed through the concrete structure.

A.l.5 Fallback Position In the event that the tests do not substantiate the current models that show that the sodium-concrete reactions are self terminating, the experimental results will'be factored into the analyses along with other updated information and thermal margins will be provided, l

O A-3

9 CRBRP-3 Vol .2, Rev.0 A.2 HYDR 0 GEN AUT0-CATALYTIC RECOMBINATION

)

A.2.1 Purpose l'ydrogen auto-catalytic recombination is important in assessing the Thermal Margin Beyond the Design Base as indicated in Section 3.2.2. The burning .

criteria are based on extensive experimentation (Reference A-4) in cluding the following ignition tests:

o Ignition of Hydrogen-Nitrogen Jets o Ignition of Hydrogen-Nitrogen-Sodium Jets o Ignition of Hydrogen-Nitrogen-Sodium-Water Jets o Effects of Oxygen Oepletion on Hydrogen Burning o Hydrogen Formation in Sodium-Water-Air Atmospheres o Effects of Jet Velocity and the following extinguishment tests:

~

o Effects of Oxygen Depletion on Hydrogen Jet Burning Efficiency .,

o Effects of Jet Sodium Concentration j o Effects of Jet Velocity o Effects of Jet Temperature o Effects of Atmosphere Water Vapor Concentration These tests were performed in a simulated containment vessel having a volume of 3.5 ft .3 The remaining tests are aimed at confirming the validity of the burning criteria in a larger simulated test vessel. These tests employed a vessel having a volume of 3800 ft3 , which provided a scaleup of a factor of more than 10 3, A.2.2 Program To provide more prototypic conditions to assess hydrogen auto-catalytic recombination, three large scale tests involving sodium-concrete interactions were performed and hydrogen ignition characteristics were determined. A O

A-4

CRBRP-3 Vol. 2, Rev. 4 simulated containment vessel having a volume of 3800 ft3 was attached to the sodium-concrete reaction test components. The oxygen concentration in the containment vessel can be controlled. The three tests run were:

o LSC-2 (See Section A.l.2 for description) o LFT-6 (See Section A.I.2 for description) o LFT-5. This is similar to LFT-6 except that faulted liner conditions for FFTF were being simulated; below the faulted liner are firebrick, insulating brick, and refractory mortar above the basalt concrete test specimen.

For the naturally generated hydrogen from these tests, ignition conditions were determined in terms of the jet gas temperature, jet sodium concentration and the oxygen concentration of the simulated containment vessel.

A.2.3 Schedule This development program has been completed and the final results are reported in Reference A-8. 4 A.2.4 Conclusiors The test results for the larger-scale tests were consistent with the small-scale results reported in Reference A-4, and gave further confirmation of the hydrogen burning criteria of Section 3.2.1 and Appendix C.l. No further testing is required.

4 O

A-5

CRBRP-3 Vol. 2, Rev. 4 A.3 FURTHER VALIDATION OF THE CACECO COMPUTER CODE A.3.1 Purpose The objectives of this activity were to further validate the CACECO computer l4 code used in many of the analyses in this report. Initial validation is reported in References A-5 and A-6. In these references, the code and input data were validated using experimental results available through 1976 and some results from 1977. With the additional experimental data that subsequently became available, further confirmatory validation was performed under this development program. 4 A.3.2 Program The further validation of the CACECO code included the following items: 4

l. A report has been issued, Reference A-10, which describes the analytical l4 validation of the code. This included a comparison of the code with analytical solutions, other validated codes and hand calculations.

2. A revised users guide has been issued (Reference A-5) to provide updated 4

user information.

3. A report has been issued (Reference A-ll) which summarizes the 4 experimental information used in the code and/or code input. The code and input data were compared with data from appropriate sodium-concrete and heated concrete tests, the HEDL hydrogen auto-ignition experiments, and appropriate concrete water release experiments.

A.3.3 Schedule This development program has been completed and the final results reported in References A-10 and A-ll. 4 A-6

l CRBRP-3 Vol . 2, Rev. 4 l

, A.3.4 Conclusions The analytical and experimental validations of the CACECO code reported in References A-10 and A-ll provide confirmation that CACECO is a valid tool for assessing beyond design base thermal margins for the CRBRP. No further work

is necessary under this development program. 4 f

O I !

l ,

lO A-7 l

CRBRP-3 .

Vol .2, Rev.0 A.4 COMPREHENSIVE TESTING PROGRAM FOR CONCRETE AT ELEVATED TEMPERATURES A.4.1 Purpose The purpose of this testing program was to establish a data base of the analysis and design of concrete exposed to elevated temperatures under conditions comensurate with nuclear power plant applications.

The specific objectives of this testing program were:

a. To define the variations in the physical (thermal) properties of limestone aggregate concrete and lightweight insulating concrete exposed to elevated temperatures resulting from a postulated large sodium spill.

b. To develop thermal relationships for use in the analysis and design of reinforced concrete components under high temperature conditions resulting from postulated large sodium spills in equipment cells.

c. To define the variations in the mechanical (strength) properties of limestone aggregate concrete and lightweight insulating concrete exposed to elevated temperatures resulting from a postulatea large sodium spill.

d. To develop strength relationships for use in the analysis and design of reinforced concrete components under high temperature conditions resulting from postulated large sodium spills in equipment cells.

A.4.2 Program The program of research specified in the Comprehensive Testing Program for Concrete at Elevated Temperatures defined the variations in the physical (thermal) and mechanical (strength) properties of prototypic CRBRP limestone aggregate concrete and lightweight insulating concrete exposed to elevated temperatures.

O A-8

CRBRP-3 Vol.2, Rev.0 0 The Comprehensive Testing Program for Concrete at Elevated Temperatures ,

consisted of two major phases. The scope of the testing program phases consisted of the following:

Phase I: Confirmation of Mechanical Properties This phase consisted of testing to determine the effect of elevated temperature exposure on the strength properties of structural concrete and lightweight insulating concrete. A limited number of tests were performed on the lightwight concrete to determine its load response characteristics at penetration locations where localized crushing due to thermal expansion is likely to impact the penetration design. All testing were performed ~1n an open moisture migration state while the concrete is at test tettperature (open-hot condition). Each test sample was heated to test temperature a,t a rate of 30 0F/hr and was heat soaked for 336 hours0.00389 days 0.0933 hours 5.555556e-4 weeks 1.27848e-4 months , unless otherwise noted, prior to mechanical testing. All samples were a minimum of 60 days old at the time of initial heat-up. The tests conducted are shown in Table A-1.

p Part 1 j Concrete cylinders 6" x 12" were tested for each temperature in the designated quantities in Table A-1 for the following parameters:

a. Compressive Strength (f'c)

b. Modulus of Elasticity (Ec I

c. Stress-Strain Relationship (a versus c)

d. Moisture and Weight Loss

e. Poissons Ratio (p) (Standard Weight Concrete Only)

A-e l

1

. 1 CRBRP-3 Vol.2, Rev.0 TABLE A-1 OPEN-HOT M0ISTURE MIGRATION STATE Test Temperature Standard Weight Concrete Lightweight Concrete (DF ) (No. of Cylinders) (No. of Cylinders) 720F (Control 3 (min) per each concrete Cylinders) batch

1500F 6* 3

2250F 6* 3

3500F 6* 3 5000F 3 3 7000F 3 3 900 F 3 3 11500F 3 0 Total 30 18 Three cylinders - (standard weight concrete only) for each of the above designated temperatures were heat soaked for approximately 672 hours0.00778 days 0.187 hours 0.00111 weeks 2.55696e-4 months prior to mechanical testing, to evaluate the long term heating effect on the

( mechanical properties of standard weight concrete.

Part 2 A sufficient number of tests (estimated below) were perfomed to determine with a high degree of accuracy the following properties for the temperature range 720 F to 11500 F for standard weight limestone aggregate concrete in the open-hot moisture migration system:

a. The variation of concrete shear strength (vc) with temperature (Approximately 24 specimens).

b. The variation of concrete /rebar bond strength (u) with temperature

( Approximately 24 specimens).

c. The variation of sustained load (creep) characteristics with temperature (Approximately 15 specimens).

O A-10

7 $

CRBRP-3 Vol. 2, Rev. 4 Phase II: Confirmation of Physical Properties This phase consisted of testing to determine the effect of elevated temperature exposure on thermal properties of structural concrete and lightweight insulating concrete. The concrete properties investigated included: the instantaneous and average coefficients of thermal expansion, conductivity, specific heat, density, moisture migration rate and weight loss.

A sufficient number of tests (estimated below) were performed to detercine with a high degree of accuracy the coefficients listed below over the 3 temperature range of 720 F to 1150 F for standard weight and lightweight !

insulating concrete. The concrete was tested in an open moisture migration ;

environment while at test temperature (open-hot condition). All samples were a minimum of 60 days old at the time of initial heat-up.

O

a. Instantaneous and Average Coefficient of Thermal Expansion (E j ) and (I) of concrete at elevated temperatures.+

b. Thermal Conductivity (k) "

c. Specific Heat (Cp ) **

d. Density at Elevated Temperatures (p)*

e. Moisture and Weight loss **

A.4.3 Schedule Phase I of this program has been completed, and the final results are reported in Reference A-12. The Phase II tests have been completed and the final report is being prepared.

4

+ Approximately 3 specimens for each property

++ 2 specimans for each material

/^\ +++ 6 specimans for each material l

\J

Natural outfall of Phase I

- Natural outfall of Phase I and Phase II, item d.

A-11

CRBRP-3 Vol. 2, ROV. 4 A.4.4 Conclusions The mechanical properties of limestone concrete (Phase I tests) were determined and the results have been factored into the structural analyses.

The Phase II tests have been completed and the results are being evaluated.

4 O

. CRBRP-3 Vol.2, Rev.0 A.5 BASE MATERIALS TEST FOR LINER STEEL A.S.1 Purpose To obtain physical material data, under strictly controlled conditions, for j the cell liner steel and weldnent material.

A.5.2 Program The experimental program was designed to produce tensile, creep and thennal expansion data on materials for the cell liners. The materials of interest are the cell liner steel, and the weldment material.

The base material and weldment materials underwent mechanical properties testing in the as-received condition. All tests were performed in an air atmosphere.

A.5.2.1 Uniaxial Tensile Tests O Uniaxial tensile tests were performed on the base material and the weldment materials to determine their temperature and strain rate dependency. All tests were performed in an air atmosphere. The number of tests for each material tested and the test condition are presented in the test matrix on the following page.

i l

lO A-13

CRBRP-3 Vol. 2, Rev. 4 Number of Uniaxial Tension Tests Per Material Type and Condition Type 1 (Fig. A-1, A-2) Type 2 (Fig. A-2) Type 3 (Fig. A-2)

Base Material & Weldment Weldment Weldment (See Note 1) (See Note 2) (See Note 2)

Test Temp. ( F) A B C Room Temp. 2 1 1 1 1 600 2 1 1 1 -

800 2 1 1 1 -

1000 2 2 2 1 -

1200 2 2 2 1 -

1400 2 2 2 1 -

1500 2 - - - -

1600 2 2 2 - -

1700 2 - - - -

Totals: Weldment 18 11 11 6 1 Base Material 18 11 11 0 0 Note 1 - Column A at 10-4 in/in/sec strain rate Column B at 5 x 10-3 in/in/sec strain rate Column C at 10-I in/in/sec strain rate Note 2 - All type 2 and type 3 uniaxial tension tests were performed at 4

the 10-4 in/in/sec strain rate.

A.5.2.2 Creep and Stress-Rupture Tests in Air Uniaxial creep tests in air were run over the temperature range 800 F to 0

1600 F, at 200 F intervals. Four stress levels were investigated at each 4 A-14

4 s.

. m. .

^ ' '

CRBRP-3 s .

Voll' 2, Rev. 4 . - \

'1 s s _ ,

s s

, _ w temperature, in order to obtain meaningful creep and rupture relationships.

The stress levels at each temperature were selected so that theirupture times would not exceed 500 h'ours. ,

x, _.

A.5.2.3 Thermal Expansion Tests s

, s In addition to the mechanical property tests described abovh sean and '

instantaneous coefficients of thermal expansion were run to supplement.

4 existing data, limited to temperatures below 800 F. These supplementary tests were performed over the temperature range 700 F to 1700 F, at 100 F intervals. Two tests were run at each temperature for the ASME SA-516, Grade 55 material only. s A.5.3 Schedule ' -

s g The tests originally planned for this development program have'been V

completed. The test resbits for the base liner material were satisfactory,' x

~

(Reference A-13) and have been factored into structural an'alyses. The' test results for the weldment material did not satisfy ductflity requirements for >

temperatures above 1000 F. An alternate weldment material possessing greater ductility has been proposed. Properties of the alternate weldment material -

and the need for additional testing are being assessed. . _

4 A.5.4 Criteria of Success -

\ ,

N Since the purpose of the test is to e'stablish materials properties, it is"not ,

possible to state a specific, quantitative success criterion.' The tests hill-be successful when materials properties have been defined, with reasonable .s accuracy, over the range of interest. It is expected that the. result will "

s v

substantiate the data used in the pres'ent analysis. ,

\

s N \

% s

.g I

'h

^

s A-15 s

_.\_ _ - _ --

~

'{'~

x; N WARD-D-0187 Vol. 2, Rev. O A.5'.5 _ Fallback Position In the event that the test results do not confirm the adequacy of the data used in the present analysis, the results of this program will be factored in the analysis along with other updated TMBDB data.

.s h

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A-16

CRBRP-3 Vol. 2, Rev. 4 O

I A.6 S0DIUM SPILL DESIGN QUALIFICATION TEST (LT-1)

This test program has been completed and information gained from the test has been factored into.the scenario presented. Details of the test and the test results are presented in Reference A-14. 4 O

I l

O .

A-17

CRBRP-3 Vol. 2, Rev. 4 A.7 TMBDB AIR CLEANING SYSTEM PERFORMANCE TESTS A.7.1 Purpose Although the components which make up the TMBDB air cleaning system (quench spray chamber jet venturi scrubber and high efficiency fibrous bed scrubber) 4 are commercially available and have been used in a variety of industrial applications, the performance of these components to the requirements of Section 2 in removing sodium and other reaction products generated during the CRBRP TMBDB scenario had to be demonstrated. This testing program confirmed l4 the adequacy of the'TMBDB air cleaning system.

The specific objectives of this program were: l4

a. Confirm the performance of the TMBDB air cleaning system for conditions characteristic of the CRBRP TMBDB scenario.

b. Provide data in support of the environmental qualification of the TMBDB air cleaning system equipment (quencher, venturi scrubber, high efficiency fiber bed scrubber and associated valves).

A.7.2 Program A sodium aerosol was generated and aged in a test facility to simulate the l4 in-containment conditions predicted from the CRBRP TMBDB scenario. These i products were vented through the air cleaning system. The TMBDB in-containment conditions and air cleaning system flow rate were simulated. 4 The parameters measured both upstream and downstream of each component are listed on the following page.

A state-of-the art report has been prepared to demonstrate the capability of the air cleaning system to remove fission products in the form of Nal, Se02 '

i and Sb 02 3 (Reference A-7).

i A-18 i

CRBRP-3 Vol. 2, Rev. 4 v

Data Requirements The following in-containment data were obtained as a function of time during 4 the conduct of the test:

o atmosphere temperature o containment absolute pressure o Na0 concentration o Na2 0 concentration o NaOH concentration o particle size o Na2 00 3 concentration o total mass concentration o Metallic Sodium concentration o atmosphere concentration

(% N2

, 02 , CO2 , H )

2 o Particle electrical charge o relative humidity o determination of fallout %

from appropriate data above The following data were obtained outside of the containment atmosphere at V

l4 these locations:

(a)Betweenthequenchunitandventuriscrubber(b)betweentheventuri scrubber and high efficiency fibrous bed scrubber (c) between the high efficiency fibrous bed and the HEPA filter,* and (d) downstream of the HEPA filter.

o temperature (both liquid & o flow rate ((a) and (c) only) gas) (liquid and gas) o Na2 0 concentration o water solution concentration of Na0H, Na CO 2 3 o Na0H concentration o particle size distribution

The HEPA filter was a specific characteristic of the HEDL test, and is not part of the TMBDB design features. 4 O

l A-19

CRBRP-3 Vol . 2, Rev. 4 o Na2 00 3 concentration o pressure differential across each component o removal efficiency of each component for particle size distribution Additionally, the following specific information was obtained.

4 o Quench unit humidity o Determination of % of sodium reaction products generated in the test facility which are vented through the air cleaning system.

o Determination of % of duct and air cleaning equipment which becomes plugged by sodium reaction products. -

o Visual inspection of air cleaning system components at intervals of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of operation for any indication of degradation of performance or conditions which might be expected to preclude long term operation.

o Evaluation of effect of increasing sodium reaction product (Na0H and Na2 00 3 ) concentration in the water used for air cleaning equipment (separate water supplies should be provided for each unit (quencher, i

venturi scrubber, and high efficiency fibrous bed scrubber) in order that each component can be evaluated separately, and also so that an additional I

1 means for determining component removal efficiency can be provided).

A.7.3 Schedule i

l This development program has been completed and the final results are reported in Reference A-15. , 4 0

A-20

. . . = . .. . . - . - .- ._ _ _ _ - . _ _ _ - - - . . .. . _ - .. _ _ ...

f .

j CRBRP-3

Vol. 2, Rev. 4 iO A.7.4 Conclusions The test results showed. satisfactory air cleaning system performance. The removal efficiencies exceeded those used in the radiological assessments

, (Chapter 4). No further testing is required.

4 r

i E

O i

l I

i i

l

.i i

O A-21

.I

CRBRP-3 Vol. 2, Rev. 4 O

A.8 REFERENCES A-l. J. A. Hassberger, " Intermediate Scale Sodium-Concrete Reaction Tests,"

4 HEDL-TME-77-99, August 1977.

A-2. R. K. Hilliard and W. D. Boehmer, " Concrete Protection from Sodium Spills by Intentionally Defected Liners, Small Scale Tests 59 and S10,"

HEDL-TME 75-75, July 1975.

A-3 J. A. Hassberger, R. K. Hilliard and L. D. Muhlestein, " Sodium-Concrete Reaction Tests," HEDL-TNE-74-36, June 1974.

A-4. R. W. Wierman, " Experimental Study of Hydrogen Jet Ignition and Jet Extinguishment," HEDL-TME 78-80, April 1979.

A-5. R. D. Peak, " Users Guide to CACECO Containment Analysis Code," HEDL-TME. -

79-22, June 1979.

A-6 R. D. Peak. "CACECO Code Verification," in Fast Reactor Safety Technical Progress Report, April-June 1977, HEDL-TNE-77-67.

A-7 A. K. Postma and R. K. Hilliard, " Nucleation and Capture of Condensible Airborne Contaminants in an Aqueous Scrubbing System," HEDL-TME 78-82, September 1978.

A-8 L. D. Muhlestein and M. W. McCormick, "Large-Scale Sodium-Limestone Concrete Reaction Tests LSC-2 and LFT-6," HEDL-TME 80-82, August 1981.

A-9 A. K. Postma, L. D. Muhlestain and R. D. Colburn, "A Review of Sodium-Concrete Reactions," HEDL-TNE 81-7, Decamber 1981.

A-10 R. D. Peak, " Analytical Validation of the CACEC0 Containment Analysis Code," HEDL-TME 79-2, August 1979.

O A-22

s CRBRP-3 Vol. 2, Rev. 4 O

V A-ll R. D. Peak, " Experimental Validation of the CACECO Containment Analysis Code," HEDL-TME 80-13, April 1980.

A-12 C. B. Oland, D. J. Naus and G. C. Robinson, " Final Report of Comprehensive Testing Program for Concrete at Elevated Temperatures,"

ORNL/BRP-80-5, October 1980.

A-13 M. C. Cowgill, B. C. Gowda and K. C. Thomas, "CRBRP Final Report on Base Material Tests for Cell Liner Steels," CRBPP-ARD-0252, January 1980.

A-14 L. D. Simmons and R. W. Wierman, "Large-Scale Liner Sodium Spill Test, LT-1," HEDL-TME 79-35, December 1980. i A-15 R. K. Hilliard, J. D. McCormack, R. K. Owen, and A. K. Postma, " Aqueous Scrubber Air Cleaning System Demonstration for Containment Venting and Purging with Sodium Aerosols--CSTF Tests ACl-AC4," HEDL-TME 81-1, March b)

I

1981.

A-16 L. D. Muhlestein and A. K. Postma, " Sodium-Concrete Reaction Executive Summary Report: Application to Limestone Concrete," HEDL-TME-82-15, 4 June 1982.

O .

A-23 l

Y l

4.380 REF. :

1.2 50 : : 1.565 TYP. :

0.50 R TYP. A .004 0.250 1 0.002 o u D [D l [h <

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$ DEFINED AS A b.250,,0o3 o

--* 0.35 +-- 0.6 50 + PLANE THROUGH 0 TYP. TYP. TRUE CENTERS OF TWO INNER HOLES MINhMUM)

ALL DIMENSIONS IN INCHES TENSION SPECIMEN THICKNESS = 0.250 INCHES CREEP RUPTURE SPECIMEN THICKNESS = 0.150 INCHES NOTE: SPECIMENS MEET THE DIMENSIONAL REQUIREMENTS OF ASTM E-8.

Figure A-l. Tension and Creep Rupture Test Specimens - Base Staterial e ;6 e~

(

l 3 TYPE 1 SPECIMENS TYPE 2 SPECIMENS ' TYPE 3 SPECIMENS s j =3 ! /

1

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APPROX. SIZE APPROX. SIZE APPROX. SIZE GAGE LENGTH 1" GAGE LENGTH 2" GAGE LENGTH 8" l WIDTH 1/4" WIDTH 1/2" WIDTH 11/2" SPECIMEN LENGTH 4 3/8" SPECIMEN LENGTH 8" MIN. SPECIMEN LENGTH 18" MIN. <

i THICKNESS 1/4" THICKNESS ~ 3/8" THICKNESS ~ 3/8" *

.n 2= (TENSION) ro E h 15" CONFIGURATION ASTM E 8 CONFIGURATION ASTM E-8 ,y

! (CREEP Qw l

RUPTURE) NUMBER OF TESTS NUMBER OF TESTS 'o i CONFIGURATION TENSION - 6 TENSION - 1 ASTM E-8 SPECIMENS SHALL BE TESTED IN GRIP ENDS SHALL BE OVERSIZED (l&w)

NUMBER OF TESTS THE AS-WELDED CONDITION WITH. TO ASSURE SUFFICIENT GRIP AREA TENSION - 40 OUT GRINDING OR FINISHING DURING TESTING. THE GRIP END AREA i

CREEP RUPTURE - 20 (IE: BACKING BAR SHALL BE LEFT ONLY SHALL BE GROUND SMOOTH.

IN PLACE). THE REMAINING AREA SHALL BE LEFT DESIGN AS FOR PLATE MATERIAL IN THE AS-WELDED CONDITION (lE:

BACKING BAR SHALL BE LEFT IN PL ACE).

l Figure A-2. Tension and Creep Rupture Tests - Weldment .\taterial

CRBRP-3 p Vol. 2, Rev. 4 V

C.1.2 Heat Sources C.l.2.1 Decay Heat The decay heat is based on steady state operation at 975 MW without uncertainties - the nominal heat. The nominal decay heat is based on the homogeneous core design. The heterogeneous core design will have little or no impact on the scenario since the integrated decay heat over the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is within 95% of the previous homogeneous design. Table C.1-3 gives the decay 4 power for the vario.us classes of fission products. The code uses a log-log interpolation technique to determine the power levels at intermediate times.

The decay heat associated with the noble gases is input to the containment j atmosphere at penetration of the reactor vessel and guard vessel. The heat associated with the halogens and volatiles is assumed to be contained in and carried with the sodium. The remainder of the decay heat is released at the bottom of the reactor cavity. This is a conservative approach, since some of the volatiles may boil away from the sodium pool before sodium boiling begins. Less decay heat in the reactor cavity would reduce the sodium boiloff rate and therefore would result in less severe containment conditions.

C.).2.2 Sodium Activity The energy associated with the decay of Na-24 was added to the sodium in both the reactor cavity sodium pool and the reactor cavity and containment atmospheres. The initial activity would be 25 milli-curies per cubic centimeter. The energy would decay with a 15 hour1.736111e-4 days 0.00417 hours 2.480159e-5 weeks 5.7075e-6 months half-life.

C.l.2.3 Sodium-Concrete Interactions The sodium-concrete reaction parameters used in this study are based on small and intermediate scale tests. These values are:

O 1. A penetration attack rate of 1/2 inch / hour (References C.1-2, C.1-3 and C.1-4).

C.1-3

CRBRP-3 -

Vol.2, Rev.0

2. A total penetretion of 2 inches. The accumulation of reaction products limits the penetration depth.

(Reference C.1-4). g

3. A chemical heat release of 331 Btu /lb of concrete. (Reference C.1-2)

This sodium-concrete reaction energy (Na-H 2O and Na-C02 reactions are considered in the next section) is only 5% of the decay power during the 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months reaction period, therefore, it is not a significant energy source and the conclusions of the analyses would not be impacted by considering the range of experimental uncertainty.

These reaction parameters are represented in the code as an energy input. The analysis assumed that the liner failure permitted sodium to attack the concrete across the full extent of the floor, or 1257 square feet, in the cavity.

C.l.2.4 Chemical Reactions At the outset of the analysis, sodium vapor reacts with all of the oxygen in the cavity to form sodium oxide.

Sodium pool reactions occur next. When both water and carbon dioxide (from exposed heated concrete) are directed into the sodium pool, sodium reacts according to the molar ratio until either sodium or the carbon dioxide-water is consumed, producing sodium hydroxide or sodium oxide, sodium carbonate, carbon and hydrogen. The production of sodium hydroxide or sodium oxide is dependent on the hydrogen partial pressure and system temperature as presented in Figure C.1-2 (Reference C.1-18).

Upon reaching containment, the sodium vapor would react with the oxygen, carbon dioxide and water vapor according to their molar concentrations. The reaction with water vapor would take place if the water vapor concentration is greater than the oxygen concentration. If the water vapor concentration is less than the oxygen concentration the sodium vapcr would react with oxygen and carbon dioxide according to their molar ratios.

Hydrogen reacting with oxygen is considered in the analysis, in the reactor containmerit building. The criteria for auto-ignition, are if either conditions (a) or (b) and condition (c) are met (see Appendix H.1).

O C.1-4

CRBRP-3 Vol.2, Rev.0 O( .

C.1.5 References C.1-1 R. D. Peak, " Users Guide to CACECO Containment Analysis Code,"

HEDL-TME 79-22, June 1979.

C.1-2 S. A. Meacham, "The Interaction of Tennessee Limestone Aggregate Concrete with Liquid Sodium," WARD-D-0141, December 1976.

C.1-3 J. A. Hassberger, R. K. Hilliard and L. D. Muhlestein, " Sodium -

Concrete Reaction Tests," HFDL-TME-74-36, June 1974.

C.1-4 J. A. Hassberger, "Intemediate Scale Sodium-Concrete Reaction Tests,"

HEDL-TME-77-99, March 1978. (Availability: U.S. DOE Technical Infomation Center.)

C.1-5 A. J. Ede, " Advances in Free Convection," in Advances in Heat Transfer, Vol. 4, pp.1-64, Academic Press, New York,1967.

C.1-6 A. P. Colburn and O. A. Hougen, " Studies in Heat Transmission," Ind. '

Eng. Chem. 22, No. 5 pp. 522-539 (1930).

C.1-7 J. G. Yevick and A. Amorosi, Fast Reactor Technology Plant Design, p.

413, MIT Press, Cambridge, MA, 1966.

(

( C.1-8 T. Z. Harmathy and L. W. Allen, "Themal Properties of Selected Masonry l

Unit Concretes," J. Amer. Concrete Inst. 70, No. 2, pp.132-142 (1973).

C.1-9 E. Crispino, " Studies on the Technology of Concretes Under Themal Conditions," Paper SP 34-25, Concrete for~ Nuclear Reactors, Vol. I, ACI

~

SP-34, pp. 443-479, American Concrete Institute, Detroit, Michigan, 1972.

C.1-10 T. Harada, J. Takeda, S. Yamane and F. Furumura, " Strength, Elasticity and Thermal Properties of Concrete Subjected to Elevated Temperatures,"

Paper SP 34-21, Concrete for Nuclear Reactors, Vol. I, ACI SP-34, pp.

377-406, American Concrete Institute, Detroit, Michigan,1972.

C.1-ll l

l

. l CRBRP-3 Vol . 2, Rev. 4 C.1-ll L. Baker, Jr., et al, "Postaccident Heat Removal Technology,"

ANL/ RAS-77-2,Section VII, January 1977. 4 C.1-12 N. G. Zoldners, "Effect of High Temperatures on Concretes Incorporating Different Aggregates," Amer. Soc. Test. Mater., Proc.

60, pp. 1087-1108 (1960).

C.1-13 A. K. Postma, J. D. McCormack and J. A Schur, "A Study of Water and Gas Release from Heated Concrete," HEDL-TC-996, December 1977.

C.1-14 J. D. McCormack and A. K. Postma, " Water and Gas Release from Heated Concrete," HEDL-SA-lll7, 1976. 4 C.1-15 F. Kreith, Principles of Heat Transfer, Third Edition, Intex Educational Publishers, New York, 1973.

C.1-16 G. H. Golden and J. V. Tokar, "Thermophysical Properties of Sodium,"

ANL-7323, August 1967.

C.1-17 " Advanced Safety Analysis Fourth Quarterly Report June - August 1975,"

GEAP-14038-4, September 1975. 4 C.1-18 E. M. Mitkevich and B. A. Shikhov, "The Metallic Sodium-Sodium Hydroxide System," Russ J Inorg Chem, II, No. 3, pg. 343 (March 1966).

C.1-19 L. D. Simmons and R. W. Wierman, "Large Scale Liner Sodium Spill Test, I LT-1," HEDL-TME 79-35, December 1980. 4 i

C.1-20 D. G. Kroger and W. M. Rohsenow, " Condensation Heat Transfer in the Presence of a Non-condensable Gas," Int J_ Heat Mass Trans, II, pp.

l 15-26, 1968.

l I

L O

l C.1-12

' CR8RP-3 1 Vol.2, Rev.0 l

TABLE C.1-4 CHEMICAL REACTIONS Reaction Equation Heat of Reaction Type of Reaction

2Na + H2O = Na20 + H2 1,600 Btu /lb Na Pool

2Na + 2 H 2O = 2 Na0H + H2 4,514 Btu /lb Na Pool 4,326 Btu /lb Na Pool 4 Na + 3 CO2 = 2 Nap C03+C Na + Concrete 331 Btu /lb Concrete Pool 3,800 Stu/lb Na Atmosphere 4 Na + CO2 " 2 .Na 20 + C 2 Na + H O = Na 0 + H 3,400 Btu /lb Na Atmosphere 2 2 2 5,700 Btu /lb Na Atmosphere 4 Na + 02 = 2 Na20 2H2+02 =2HO2 54,425 Btu /lb H 2

Atmosphere 2 Na + H2.= 2 NaH 1,050 Btu /lb Na Atmosphere

~ H O + Na 0 = 2 NaOH 1,500 Btu /lb Na Atmosphere 2 2 s

O(

The production of sodium hydroxide or sodium oxide is dependent on the hydrogen partial pressure and system temperature as indicated in Figure C.1-2.

O C.1 -17

CRBRP-3 Vol. 2, Rev. 4 TABLE C.1-5 CONVECTIVE HEAT TRANSFER COEFFICIENTS TypicalVa}ues (Btu /hr-ft OF) Description 1.21 at 7000F AT Atmosphere of nitrogen (49%), hydrogen (49%) and sodium vapor (2%) inside the reactor cavity.

6.11 at 1000F AT Atmosphere of hydrogen (21%) and sodium vapor (79%)

inside the reactor cavity.

200.0 Natural convection film coefficients for sodium submerged surfaces.

1.13 at 750F AT Containment building and cell 105 atmospheres.

1.74 at 3000F AT 4

O O

C.1-18

CRBRP-3 Vol. 2, Rev. 4 O

V i C.3-28 M. G. Cowgill, B. C. Gowda and K. C. Thomas, "CRBRP Final Report on Base Materials Tests for Cell Liner Steels," CRBRP-ARD-0252, January 4

i 1980.*

C.3-29 H. Kupfer H. Hilsdorf and H. Rusch, " Behavior of Concrete Under Biaxial Stresses," J. Amer. Concr. Inst. 66, No. 8, pp. 656-666 (1969).

C.3-30. S. Freeoman, " Properties of Materials for Reinforced Concrete " in M. Fintel, ed., Handbook of Concrete Engineering, pp. 141-211, Van Nostrand Reinhold Co., 1974.

C.3-31 Letter from P. S. VanNort to R. S. Boyd, Transmittal of " Third Level Thermal Margins Report," Docket 50-537, S:L:977, April'22, 1976.

(Availability: Nuclear Regulatory Commission, Public Document Room.)

O

The material concerning weldments in Reference C.3-28 became available after i this Appendix was prepared and the structural evaluations described in Section 3.2 of this' report were completed. The weldment properties will be taken into account in the final evaluation.

C.3-19

CRBRP-3 Vol.2, Rev.0 a

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C.3-20

CRBRP-3 Vol. 2, Rev. 4 O

U G.2 CONSEQUENCES OF INCREASED SODIUM-CONCRETE REACTIONS G.2.1 Introduction Experiments (References G.2-1 through G.2-4) have indicated that 4

sodium-concrete reactions are self-limiting for conditions occurring during the TMBDB scenario. These reactions have been represented in the reference case CACECO analysis as a penetration attack of 0.5 inch per hour for a period of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . This reaction rate is used in Section 3.2.2 as described in Appendix C.l. The sensitivity of the consequences to increased sodium-concrete interactions is examined here.

a G.2.2 Model The CACECO code model defined in Appendix C.1 was modified to perform these sensitivity studies. The model used in the reference case analysis in Section n\~/ 3.2.2 includes a sodium-concrete reaction rate of 0.5 inch per hour for 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , starting at the time of reactor cavity liner failure (assumed to be at the time of penetration of the reactor vessel and guard vessel). To assess the potential impact of more severe sodium-concrete reactions, the CACEC0 model was modified to consider reaction rates of 0.5 inch per hour for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and 1.0 inch per hour for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The reaction energy was maintained constant at 331 Btu /lb of concrete. The sensitivity analyses were run for 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months without venting and the results were compared to the reference case.

G.2.3 Results The two cases in which more severe sodium-concrete reactions were assumed are compared to the reference case described in Section 3.2.2. The additional energy from the more severe sodium-concrete reactions causes the sodium pool to heat up faster. Sodium boiling begins at about 9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months in the reference case; this is reduced to about 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months with a sodium-concrete reaction rate of O.5 inch /hr for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , and to about 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months with a sodium-concrete reaction 10

, G.2-1

WARD-D-0'87 1

Vol. 2, Rev. O rate of 1.0 inch /hr for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . This additional energy source is reflected in slightly higher atmosphere temperatures in the reactor cavity and containment building and a higher containment steel temperature. However, as indicated in Figure G.2-1, these differences are minor and would not impact conclusions on the integrity of the containment.

Because of the additional energy source, the pressure in containment would be slightly higher (except for an early spike due to the treatment of hydrogen burning, discussed below). Again, the differences are minor as indicated in Figure G.2-2, and conclusions on containment integrity would not be impacted.

The hydrogen concentration in containment for the three cases is shown in Figure G.2-3. In all cases, the predicted hydrogen concentration at 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and at 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months is zero. However, the short term (up to $10 hours) concentrations vary slightly with the sodium-concrete reaction assumptions.

With more severe reactions, the sodium pool heatup would be more rapid and the sodium vapor from the pool would be increased. This would result in the hydrogen burning criteria (described in Section 3.2.1) being met earlier ( 4 or 9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months compared to $10 hours in the reference case). The maximum hydrogen concentration prior to the hydrogen burning would be decreased slightly (from 4.5% to $4%). The analytic assumption of instantaneous burning when the criteria are met results in a predicted pressure spike as indicated in Figure G.2-2. The predicted spike is slightly less severe with increased sodium-concrete reactions but the variations are small enough to be of no consequence.

G.2.4 Conclusions Table G.2-1 summarizes the results of these sensitivity studies on sodium-concrete reactions. The reference case considers a total depth of reaction of two inches of concrete. The sensitivity studies considered total depths of 6 inches and 12 inches. This provides factors of 3 and 6 on total energy from sodium-concrete reactions. The results show that the predicted G.2-2

CRBRP-3 Vol. 2, Rev. 4 containment conditions are not very sensitive to the sodium-concrete reaction assumptions. This is due to the fact that the reaction energy from sodium-concrete reactions is a small part of the total energy involved in the scenario from decay heat, chemical reactions in containment and other chemical reactions in the sodium pool.

Since the containment conditions are similar for the range of sodium-concrete reactions considered, it is concluded that sufficient margin exists to cover the uncertainties in sodium-concrete reactions.

G.2.5 References G.2-1 J. A. Hassberger, R. K. Hilliard and L. D. Muhlestein, " Sodium Concrete Reaction Tests," HEDL-TME-74-36, June 1974.

G.2-2 J. A. Hassberger, " Intermediate Scale Sodium-Concrete Reaction Os Tests," HEDL-TME-77-99, March 1978.

G.2-3 L. D. Muhlestein and M. W. McCormick, "Large-Scale Sodium-Limestone Concrete Reaction Tests LSC-2 and LFT-6 " HEDL-TME 80-82, August 1981.

G.2-4 A. K. Postma, L. D. Muhlestein and R. P. Colburn, "A Review of Sodium-Concrete Reactions," HEDL-TME 81-7, December 1981.

4 G.2-3

CRBRP-3 l Vol.2 Rev.0 !

TABLE G.2-1 l

SUMMARY

OF SENSITIVITY STUDIES ON SODIUM-CONCRETE REACTIONS Reference Case 0.5 in/hr 0.5 in/hr. 1.0 in/hr Sodimn-Concrete Reactions for 4 hrs for 12 hrs for 12 hrs PRE-24 HOUR RCB CONDITIONS Peak Hydrogen Concentration (%) 4.5 4.2 3.8 Time for Peak Concentration (hrs) 10.0 9.2. 8.3 24 HOURS RCB CONDITIONS Atmosphere Temperature (oF) 450.0 480.0 520.0 Pressure (psig) 11.1 11.8 12.6 Steel Temperature (OF) 270.0 295.0 320.0 HydrogenConcentration(%) 0.0 0.0 0.0 0xygen Concentration (%) 13.7 13.3 12.7 .

t I 30 HOURS RCB CONDITIONS Atmosphere Temperature (OF) 520.0 550.0 580.0 Pressure (psig) 12.5 12.7 13.4 Steel Temperature (DF) 315.0 335.0 360.0 Hydrogen Concentration (%) 0.0 0.0 0.0 0xygen Concentration (%) 11.7 11.2 10.4 9

G.2-4

CRBRP-3 Vol.2, Rev.0 i Results Using the revised CACECO model, the containment conditions computed are listed in Tables G.4-1, 2 and 3 for the various cases. Tables G.4-2 and 3 indicate containment atmosphere temperature and pressure are more sensitive to the initial sodium release assumed than to the initial fuel release. The most severe results were found when 7,000 pounds of sodium and 7.5% of the fuel and solid fission products were asstaned initially released to containment. .The containment atmosphere temperature and pressure were found to be 10300F and 24.4 psig. The containment metal temperature is not significantly affected by the sharp transient in atmosphere temperature because of its large heat capacity (at 1000 seconds the metal temperature is only 1400F ). The containment could withstand all of the short terin transients resulting from the initial head releases considered.

Conclusions The calculations indicate that containment integrity after the initial release of f.uel and fission products and sodium is not challenged for a wide range of assumptions. Considering the worst assumption (7,000 pounds of sodium and 7.5% fuel and fission products) containment integrity would be maintained following the initial head release.The general conclusion is that the Tie 0B scenario has sufficient margin with respect to the amount of sodium and fuel assumed to be initially released.

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G.4-3

CRBRP-3 Vol. 2, Rev. 4 l

TABLE G.4-1 EFFECT OF INITIAL FUEL, FISSION PRODUCT, AND SODIUM RELEASE

Dose (rem) 1% Fuel & F.P. 5% Fuel & F.P. 7.5% Fuel.& F.P.

1000 lb. Na 3300 lb. ha 7000 lb. ha Bone 0.93 3.83 3.12 2 Hour Lung 0.15 0.39 0.30 EB Thyrcid 11.3 9.51 5.19 Whole Body 0.24 0.32 0.28 Bone 55.7 56.2 55.8 30 Day Lung 3.02 3.02 2.91 LPZ Thyroid 5.31 1.72 0.83 Whole Body 3.07 2.94 2.90 4 RCB Atmosphere Conditions **

Temperature (OF) 270 580 1030

Pressure (psig) 4.6 12.7 24.4 l

Initial release of noble gases, halogens, and volatile fission products to RCB = 100%.

- Peak values for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

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4 CRBRP-3 Vol. 2, Rev. 4 O

TABLE G.4-2 EFFECT OF INITIAL FUEL AND FISSION PRODUCT RELEASE WITH SODIUM RELEASE FIXED AT 1000 LB.

Dose (rem)

Initial Release of Fuel and Fission Products

1% 5% 10% 50%

i Bone 0.93 4.56 9.10 45.4 2 Hour Lung 0.15 0.46 0.85 3.96 EB Thyroid 11.3 11.3 11.3 11.3 Whole Body 0.24 0.35 0.48 1.53 Bone 55.7 58.5 60.6 83.7 j 30 Day Lung 3.02 3.25- 3.35 5.37 l LPZ Thyroid 5.31 4.43 3.53 3.53 Whole Body 3.07 3.10 3.11 3.68 4 RCB Atmosphere Conditions **

j Temperature (OF) 270 290 310 450 i Pressure (psig) 4.6 5.0 5.7 9.5

Initial release of noble gases, halogens, and volatile fission products to RCB = 100%.

- Peak values for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

l l

O G.4-5

(

CRBRP-3 Vol. 2, Rev. 4 TABLE G.4-3 EFFECT OF S0DIUM RELEASE FOR A GIVEN 10% FUEL-FISSION PRODUCT RELEASE

Dose (rem)

Pounds of Sodium in Initial Release 0 '1000 Bone 9.10 9.10 2 Hour Lung 0.84 0.85 EB Thyroid 11.2 11.3 Whole Body 0.48 0.48 Bone 63.5 60.6 -

30 Day Lung 3.63 3.35 LPZ Thyroid 5.28 3.53 Whole Body 3.26 3.11 4

RCB Atmosphere Conditions **

Temperature (OF) 260 310 Pressure (psig) 4.6 5.7

Initial release of noble gases, halogens, and volatile fission' products to RCB = 100%.

- Peak value for I hour.

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CRBRP-3 Vol. 2, Rev. 4 1 (v

APPENDIX H HYDROGEN BURhlNG CHARACTERISTICS This Appendix provides the hydrogen burning characteristics used in the various analyses. Specifically, the hydrogen burning criteria are addressed in Appendix H.1, flame length considerations are addressed in Appendix H.2, and the potential for hydrogen stratification is addressed in Appendix H.3.

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O H.0-1

_ s CRBERP-3 ,[

Vol.2d Rev.0 s

7-()

appropriate. It is noted that Regulatory Guide 1.7 applies the 4% limit to water reactor containments where the temperature, pressure, and water content are considerably above normal.

s i H.I.2.2 Detonation Limit for Hydrogen in Air N

.g Reference H.1-1 gives the lower hydrogen concentration detonation limit in air at 18.3% at ambient conditions. Other experimenters (Reference H.1-4) observed the lower hydrogen concentratioh detonation limit to be as low as 14%. Shapiro and Moffette (Reference H.1-5), in their evaluation of the safety aspects for LWRs, use 19% as the lower hydrogen concentration limit for '

detonation.

The effect of TMBDB conditions on the lower hydrogen concentration detonation limit is not known; however, the magnitude of tho effects would not be expected to be significant compared to' the margin between the flamaDility and ,

detonation limits. Thisissup$orted6ythehydrogenburningexperimentsat HEDLthathaveindicateddetonationsoccubuSderconditionssimilartothose in TMBDB only at hydrogen concentrations above 10%. The exact concentration was not determined from those experiments.

As Figure H.1-3 indicates, it would not be possible in the TMBDB scenario to reach a hydrogen concentration in which a detonation and resulting high pressures (7-15 atmospheres) would occur. This is because the large ignition source would cause the hydrogen to burn at the flamability limit (#4%

hydrogen) which is considerably below the detonation limit (#14 to 19%

hydrogen). It is noted that the large quantity ofs sodium entering containment would provide a continuous ignition source (satisfying criterion (b)), even if for some unknown reason the sodium would not burn, because . sodium at

17000 F would be red hot.

Regulatory Guide 1.7 states that in water reactors with the oxygen content above 5%, tne hydrogen concentration must not exceed 6% to insure that explosive concentrations are avoided, n

H.1-3

L ,

CRBRP-3 Vol. 2, Rev. 4 H.l.2.3 Experiments Under Simulated TMBDB Conditions Experiments have been performed at HEDL to study hydrogen burning under simulated TMBDB conditions (References H.1-6 and H.1-7). A conservative 4 interpretation of the experiments indicated that hydrogen would burn when entering the containment if either criterion (a) or (b) is satisfied in combination with criterion (c).

a. The hydrogen-nitrogen mixture entering containment is above 1450 F.

b. The hydrogen-sodium-nitrogen mixture entering containment contains at least 6 g/m 3of sodium at temperatures above 5000F.

c. The oxygen concentration is above 8%.

Criterion, (c) above, was found to be highly conservative. Figure H.1-4 reproduces the experimental results under the conditions that approach those of the TMBDB event. The two symbols indicate tne oxygen content of the inlet and outlet gases in the simulated containment vessel. The average vessel oxygen content is between the inlet and outlet values - probably closer to the outlet value. As the figure indicates, hydrogen burning continues until oxygen levels decrease to 4 %.

H.1.3 Hydrogen Burning Scenario for TMBDB The predicted hydrogen burning scenario will be presented in this section followed by a scenario that bounds the containment pressure and temperature conditions.

O H.1-4

~

CRBRP-3 (m)3 H.l.3.1 Predicted Hydrogen Burning Scenario With respect to ignition of hydrogen entering the reactor containment building through the reactor cavity vent'in the TMBDB scenario, the criteria used in the Section 3.2.2 analysis are appropriate and supported by the HEDL data (References H.1-6 and H.1-7).

With respect to hydrogen burning after ignition, when the oxygen content is reduced, applicable data are more numerous. Figure H.1-5 illustrates the burning limits from the HEDL experiments, the lower flammability limits derived by the U.S. Bureau of Mines and the flamability limit in Regulatory Guide 1.7. Considering that the HEDL data define when hydrogen burns as it passes through the sodium flame at the reactor cavity vent (establishing the 8% oxygen limit for complete hydrogen burning), and the Bureau of Mines data define the burning region in a hydrogen-air atmosphere with an ignition source (establishing the 5% oxygen limit for hydrogen burning above 4%), the predicted hydrogen burning scenario for TMBDB conditions is that shown on Figure H.1-6.

The complete hydrogen burning criteria are that hydrogen will burn when entering containment or within the containment when either criterion (a) or (b) is satisfied in combination with criterion (c).

a. The hydrogen-nitrogen mixture entering containment is above 14500F.

b. The hydrogen-nitrogen-sodium mixture entering containment contains at least 6 g/m 3of sodium at temperatures above 500 F.

c. The containment oxygen concentration is above 8%. With the oxygen concentration above 5% and the hydrogen concentration above 4%, the hydrogen in excess of 4% would burn.

O H.1-5

CRBRP-3 '

Vol.2, Rev.0 H.1.3.2 Hydrogen Burning Scenario to Bound Containment Temperature and Pressure Because of the possibility that hydrogen might burn at lower hydrogen concentrations than 4% in the presence of the large sodium ignition source, a bounding hydrogen burning scenario was developed to maximize the ' containment temperature and pressure effccts. For this bounding case, it was assumed that all hydrogen burns if any oxygen is present in conta'inment. This bound is indicated on Figure H.1-6.

H.I.4 Containment Conditions H.1.4.1 Conditions Prior to Venting and Purging Containment conditions were determined for the predicted and bounding hydrogen burning scenarios using the CACECO code model described in Section 3.2.2 and Appendix C.1. Figures H.1-7, 8 and 9 show the containment atmosphere temperatures, pressures, and hydrogen concentration for these cases as a function of time. Table H.1-1 provides the containment conditions at 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months f or these cases. Figure H.1-10 indicates the containment conditions with respect to the flarmiability limits. Y J At 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , the predicted scenario is identical to the bounding case that assumes complete hydrogen burning since the 8% oxygen cutoff value has not been reached. Consequently, no hydrogen accumulation is predicted in containment at 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Both cases indicate the containment does not require venting and purging until times beyond 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Beyond 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months the pressure, temperature and hydrogen concentration of the predicted hydrogen burning case and the bounding case continue to climb. The pressure differential between the predicted burning case and the bounding case is an insignificant 1 psi. The atmosphere temperature differential is also small. The hydrogen concentration for both cases decreases between 45 and 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . This is a temporary condition due to the formation of sodium hydride in the containment building. The reaction combines free hydrogen and sodium l

H.1-6

CRBRP-3 Vol. 2, Rev. 4 H.1.6 References H.1-1 B. Lewis and G. von Elbe, Combustion, Flames and Explosions of Gases, Second Edition, Academic Press, Inc., New York, 1961.

H.1-2 H. F. Coward and G. W. Jones, Limits of Flammability of Gases and Vapors, Bureau of Mines Bulletin 503, U.S. Bureau of Mines, Washington, D.C.,1952.

H .1 -3 G. W. Keilholtz, " Hydrogen Considerations in Light-Water Power Reactors," ORNL-NSIC-120, February 1976.

H.1-4 W. E. Gordon, A. J. Mooradian and S. A. Harper, " Limit and Spin Effects in Hydrogen, Oxygen Detonations," Seventh Symposium (International) on Combustion Oxford, Aug. 28 - Sept. 3,1958, i pp. 752-759, Academic Press, New York, NY.1959.

H.1-5 Z. M, Shapiro and T. R. Moffette, " Hydrogen Flammability Data and Application to PWR Loss-of-Coolant Accident," WAPD-SC-545, September 1957.

H.1-6 R. N. Wierman, " Experimental Study of Hydrogen Jet Ignition and Jet Extinguishment," HEDL-TME 78-80, April 1979.

H.1-7 L. D. Muhlestein and M. W. McCormick, "Large-Scale Sodium-Limestone Concrete Reaction Tests LSC-2 and LFT-6," HEDL-THE 80-82, August 1981.

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CRBRP-3 .

Vol.2 Rev.0

TABLE H.1-1 CONTAINMENT ATMOSPHERE CONDITIONS AT 24 HOURS' Predicted Hydrogen Bounding Hydrogen Burning Scenario Burning Scenario Atmosphere Temperature (OF) 450 450 AtmospherePressure(psig) 11.1 11.1 Hydrogen Concentration (%) 0.0 0.0 e

H.1-10

CRBRP-3 Vol. 2, Rev. 4 O

H.3 POTENTIAL FOR HYDROGEN STRATIFICATION H.3.1 Introduction During the 1NBDB scenario, the potential for hydrogen collecting at the top of the containment dome in concentrations significantly greater than the average in containment (i.e., stratification) is different at different times in the scenario. There are three time frames it the base case scenario of interest, (1) early in the scenario, before hydrogen ignition (0 to 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months ), (2) hydrogen burning period (10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months to 133 hours0.00154 days 0.0369 hours 2.199074e-4 weeks 5.06065e-5 months ), and (3) post-boildry period.(133 hours0.00154 days 0.0369 hours 2.199074e-4 weeks 5.06065e-5 months to 8000 hours0.0926 days 2.222 hours 0.0132 weeks 0.00304 months ). In all three time periods there would be mixing forces'in operation that are adequate to assure that no significant hydrogen stratification would occur.

In the' initial period (1) the dominant mixing force would be forced convection turbulence due to the high velocity of the hydrogen jet entering containment through the reactor cavity vent. In the hydrogen burning perica (2) the forced convection turbulence would still be in operation and mixing would be still further enhanced by the convection currents induced by the high temperature in the flame. In the post-boildry period mixing would be maintained primarily by the action of the vent / purge system.

It is roted that the hydrogen monitoring sample point is near the top of the containment dome, so that the higher concentration would be measured at all times.

H.3.2 Pre-Hydrogen Ignition Time Period During the initial ten hours of the TMBDB scenario, hydrogen enters the reactor containment building at conditions which do not meet the burning criteria given in H.l.3.1. During this time period the hydrogen jet would have three forces acting to mix it with the RCB air; namely, molecular diffusion, natural convection and forced convection. As shown below, these l

H.3-1

CRBRP-3 Vol. 2, Rev. 4 forces would be more than adequate to counteract buoyancy forces and prevent any significant collection (stratification) of hydrogen at the' top of the RCB.

H.3.2.1 Molecular Diffusion Molecular diffusion is a mass transfer process resulting from a concentration gradient existing between two locations. While the gradient exists, mass will flow from the region of high concentration to the low concentration region, and given sufficient time, this process can bring the system to an equilibrium state. For the TMBDB condition, however, the mass flow into the system, 84 lb/hr average, exceeds the diffusion mass transfer rate, 10-20 lb/hr, and this process would not, of itself, be adequate to prevent stratification even at the higher temperature existing after ignition at 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months . However, the diffusion process would supplement other processes.

H.3.2.2 Natural Convection The hydrogen entering the reactor containment building from the reactor l cavity is at a higher temperature than the containment shell. This j difference in temperature could cause flow in the containment buildin_g which causes mixing. In the initial 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months of the scenario, the temperature l

differences between the incoming hydrogen and containment shell is less than a hundred degrees. Under these conditions, the buoyancy forces resulting from the temperature difference are not sufficient to overcome the buoyancy forces resulting from the concentration difference, and mixing of the streams due to natural convection flow would not be expected. 'After ignition at 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months , the temperature induced buoyancy forces are greater than the concentration induced buoyancy force, and free convective flow would occur, causing mixing of the fluid with the RCB.

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H.3-2

CRBRP-3 Vol. 2, Rev. 4 (D

U H.3.2.3 Forced Convection The major force that would mix the incoming hydrogen with the containment atmosphere is the turbulent action of the hydrogen jet. If the Reynold's number at.the pipe exit exceeds 30, the flow quickly becomes turbulent downstream of the exit, and the turbulent action. entrains fluid from the surrounding environment, Reference H.3-1 (see Figure H.3-1). The Reynold's number for the TMB08 scenario is 1400; thus, turbulent flow and significant entrainment would be expected.

Entrainment, a measure of the mixing, in a turbulent plume can be calculated by the formula Q = .404 & X where:

Q = volumetric flow rate of the plume at location X, ft3 /hr K = J/p, hr 2 Ib-ft J = momentum flux, 2 2 hr -ft p = . density, lb/ft i X = distance from the jet exit, ft It can be seen that as the distance from the jet exit increases, the quantity

! of material flowing in the plume increases; i.e., Q is a function of X. This occurs because of entrainment from the surrounding environment. Thus the difference between the quantity of material flowing at location X and the flow of hydrogen into the RCB is the amount of material from the RCB atmosphere which has been entrained.

4 i

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H.3-3

CRBRP-3 Vol. 2, Rev. 4 Furtnermore, the mean composition at location X is given by Q !O where 0 X the subscripts 0 and X represent respectively, the origin ana a specific location.

Table H.3-1 provices a calculation of the hydrogen concentration at a location 160 feet above the vent pipe outlet, the approximate intersection of the jet with the containment shell. From the table, it can be seen tnat, at this elevation, the hydrogen stream has been diluted to approximately 1.5% of its original concentration. Continued flow within the containment, inducea by the injection of the hydrogen stream would cause further dilution, and in the limit, the concentration in tne containment building would approach that of a well mixed tank.

H.3.3 Hydrogen Ignition to Sodium Boildry Time Af ter the hydrogen stream has been ignited, the surrounding air would continue to be entrained by the turbulent action of the jet (this is the means by whicn the flame gets its oxygen), and convection currents would be induced in the containment due to the large temperature differential between the combustion product gases and the containment building dome.

When venting ana purging are initiated, the mixing action aesignea into the TMBDB vent / purge system woula further enhance mixing even though this feature is not needed until the post-boilary perica.

H.3.4 Post-Sodium Boildry Period Long term (post boilary), the combustible gases expected to be produced by reactions within the reactor cavity are hydrogen ana carbon monoxide which woula be expected to form a mixture having a molecular weight similar to air because of the predominance of carbon monoxide (MW = 28) [see Section 3.2.3].

These gases would enter containment at high temperature and would probably not flow through the reactor cavity vent exclusively, due to the expectation that i

4 H.3-4

l CRBRP-3 l

, Vol. 2, Rev. 4 j l

l many leakage patiis would have opened by the time sodium has boiled dry (the base case scenario assumes leakage through the reactor head seals beginning at 50 hours5.787037e-4 days 0.0139 hours 8.267196e-5 weeks 1.9025e-5 months ). Furthermore, the flow rates are much lower late in the post-boildry period so that forced convection turbulence can no longer be counted on as the major mixing force. Thus, during this period the mixing action from the TMBDB vent / purge system would be the major force for preventing hydrogen stratification (and for breaking up any stratification that might occur late in the scenario if the vent / purge has been turned off for a time).

H.3.5 Conclusions The potential for hydrogen stratification in the top of the containment dome has been assessed for the various conditions existing in containment throughout the TMBDB scenario. It is concluded that concentrations significantly higher than the containment average would not accumulate early in the scenario due to mixing forces inherent in the reactor cavity vent arrangement and the burning at the nozzle. Later in the scenario, stratification is prevented and/or dis,1patec by the mixing action of the containment vent / purge system.

H.3.6 References H.3-1. Schlichting, Hermann, Boundary Layer Theory, McGraw-Hill, 6th Edition, 1968, pp. 170-174, 696-702.

O H.3-5

CRBRP-3 Vol. 2, Rev. 4 TABLE H.3-1 ATMOSPHERIC ENTRAltNENT IN HYDROGEN ilET Hydrogen Flow Rate 84 lb/hr Pipe Diameter 1 foot Pipe Velocity @ STP* 84 lb lb-mole 359 ft 3 4 ,19197 ft hr 2 lb lb mole ft 2 hr Fluid Density @ STP* Ib-mole = 0.0055 lb/ft3 lb le 359 ft 4

K=E= P .0055 8

= 2.9 (10 ) n 2 hr Q @ 160' = (.404) (2.9)1/2 (10 ) (160) = 1,000,000 r

3 ft 3 Q @ origin = 84 lb 15,000 ft /hr

)

I

%H2@160'=I,0bo000 = 1.5%

Note 160' is the approximate vertical distance between the vent pipe outlet and the contaiment shell.

The use of standard conditions gives a lower velocity and a higher density than those that would exist at higher temperatures. This in turn gives a lower (conservative)valueoff=f.

O H.3-6

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CRBRP-3 Vol. 2, Rev. 4 Jf y i y 0 ~f Cb a v v v o"v s -

0 O s -

j Figure H.3-1. Flow Patterns for Jet Injection into a Gosed Container 7184-2 O

H.3-7

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