Static and dynamic neutronic analysis of the uranium tetra-fluoride, ultrahigh temperature, vapor core reactor system

396 Pages·1991·8.2 MB·English

by  Kahook, Samer Dakhlallah, 1961-

STATIC AND DYNAMIC NEUTRONIC ANALYSIS
OF THE URANIUM TETRA-FLUORIDE, ULTRAHIGH
TEMPERATURE, VAPOR CORE REACTOR SYSTEM
By
SAMER DAKHLALLAH KAHOOK
A DISSERTATION PRESENTED TO THE GRADUATE SCHOOL
OF THE UNIVERSITY OF FLORIDA IN PARTIAL FULFILLMENT
OF THE REQUIREMENTS FOR THE DEGREE OF
DOCTOR OF PHILOSOPHY
UNIVERSITY OF FLORIDA
1991
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Dedicated to
my parents,
Mr. and Mrs.
Dakh1a1lah Kahook,
asking A11ah
to reward them,
have mercy on them,
and grant them paradise
as they raised and cherished me in my childhood.
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ACKNOWLEDGEMENTS
The author would like to express his appreciation and sincere
thanks to the members of his supervisory committee, Dr. Edward T. Dugan,
Dr. Nils J. Diaz, Dr. Alan M. Jacobs, Dr. Samim Anghaie, Dr. William E.
Lear, Jr., and Dr. Willis B. Person for their guidance and assistance
during the course of this research.
Special thanks are extended to Dr. Dugan, chairman of the author's
supervisory committee for his patience and enduring support. The author
recognizes that much of his knowledge in reactor physics and computer
programming was realized while researching under the guidance and
direction of Dr. Dugan.
Support for this research has been provided, in part, by the Air
Force Wright Aeronautical Laboratories (AFWAL), the Frederick Hauck
Fund, and the University of Florida. The AFWAL work was performed for
the Innovative Science and Technology Directorate of the Strategic
Defense Initiative within the Innovative Nuclear Space Power Institute
(INSPI). This support is greatly appreciated.
Funding for the computer analysis was provided for by the National
Science Foundation at the San Diego Supercomputer Center and the
University of Florida and the International Business Machines (IBM)
Corporation through their Research Computing Initiative at the North
East Regional Data Center. The author is grateful for these funds.
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Thanks are also due to the fellow students whose friendships,
comments, and encouragements have also facilitated in this research.
The author would like to express his love and respect to his
parents Mr. and Mrs. Dakhlallah Kahook, to his brothers Nofal and
Mohammed, and to his sisters for their love, understanding, and patience
throughout the author's stay at the University of Florida. The
financial support provided to the author by his family is gratefully
acknowledged.
Finally, the author would like to express his love and deepest
appreciation to his wife, Layali, whose understanding, patience, and
support provided the motivation needed to finish this research.
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TABLE OF CONTENTS
Page
i
ACKNOWLEDGEMENTS
i i i
LIST OF TABLES x
LIST OF FIGURES xiv
ABSTRACT xvi
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CHAPTER
INTRODUCTION
I 1
Introduction
1
Description of the Ultrahigh Temperature Vapor Core
Reactor 2
Dissertation Objectives 6
Dissertation Organization 7
II PREVIOUS RESEARCH ON RELATED CONCEPTS 11
Introduction 11
Previous Research on Gas Core Reactors 12
Previous Research on Coupled Core Reactors 13
Previous Research on Circulating Fuel Reactors 14
Remarks 18
III DESIGN OF THE URANIUM TETRA-FLUORIDE, ULTRAHIGH
TEMPERATURE VAPOR CORE REACTOR 19
Introduction 19
Prel iminary Design Considerations 20
Choice of Materials 21
The Moderator-Reflector Material 21
The Fissioning Fuel Material 24
The Working Fluid Material 29
Description of a Uranium Tetra-Fluoride, UTVR/Disk
MHD-Rankine Power Cycle 29
Neutronic Analysis of the Ultrahigh Temperature Vapor
Core Reactor 37
Static Neutronic Calculations 37
Dynamic Neutronic Calculations 39
CHAPTER Piae
IV STATIC, ONE-DIMENSIONAL, UTVR NUCLEAR CHARACTERIZATION
AND CONFIGURATION OPTIMIZATION 40
Introduction 40
Scoping Calculations 43
Geometric Variations 43
UTVC radius 43
Inner BeO moderator-reflector region thickness 50
Outer BeO moderator-reflector region thickness 55
UF^ boiler region thickness 57
UF? boiler core volume 59
Fuel Density Variations 61
UF^ partial pressure and mole fraction (UF^rNaF)
Jlin the UTVC 61
U"5 enrichment in UF« 62
u""* as the fissile isotope 62
Average density of the UF^ in the boiler region 66
Material Variations 68
Choice of metal fluoride in UTVC 68
Wall cooling region 70
Other metal fluoride working fluids 70
NaF mass flow rate to the boiler region 73
UF^/NaF inlet velocity to the boiler 76
Addition of Li F poison to the boiler 76
BeO in the annular boiler region 79
Reactivity effects of liner materials 81
One-Dimensional Results 84
The Neutron Multiplication Factor 86
Power Sharing Factor 87
Spherical "Mock-up" Comments 90
V STATIC, TWO-DIMENSIONAL, UTVR NUCLEAR CHARACTERIZATION
AND CONFIGURATION OPTIMIZATION 94
Introduction 94
Scoping Calculations in R-^ Geometry 96
Geometric Variations 98
UTVC radius variations 98
Inner BeO moderator-reflector region thickness
variations 104
Variation in the area of the boiler columns 107
Variation in the number of boiler columns 109
Fuel/Working-Fluid Density Variations 110
UF4 partial pressure in the UTVC 112
Average UF^ density in the boiler columns 115
Varying the UF^ average density in the UTVC as a
function of the radial distance from the
center ine 116
1
Scoping Calculations in R-Z Geometry 121
Geometric Variations 124
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CHAPTER Page
V MBEO region height 124
(cont.) TBEO region height 129
First OBEO region height 131
Boiler: subcooled and saturated liquid region
height 134
Material Variation 135
Poisoning the boiler feedline walls 136
Comments on Power Sharing 140
Two-Dimensional Results 144
The Neutron Multiplication Factor 144
The Power Sharing Factor 146
Remarks 148
VI STATIC, THREE-DIMENSIONAL NEUTRONIC ANALYSIS OF THE UTVR.... 151
Introduction 151
Description of the UTVR Geometry in MCNP 152
Description of the Boiler Column 156
Reactivity Worths of the Boiler Feedlines, UTVC Inlet
PIenums and the MHD Duct Regions 158
,
Reducing the Uncertainty in Parameters Associated with
the Boiler Columns in MCNP Calculations 161
Performance of Variance-Reduction Techniques 165
Nuclear and Physical Characteristics of the UTVR 166
, '
r V Energy Cutoff 168
' * Implicit Capture and Weight Cutoff 168
Weight Windows 173
BB0o1illeerr--ttoo--UuTiVvCt Ssyymmmmeettrryy 178
Neuittrroonn TTrraannssppoorrtt CCoouupp!liinngg CCooeeffffiicciieents 185
ir. Obtained Directly from MCNP. 187
€1*'^ Obtained Indirectly from MMCC^NP 189
Isolation of secondary coupling effects 190
,
Neutron Multiplication Factor.of the j Core, k^ff... 197
Reactivity of the j^" Core, p^..., :I1... 198
Prompt Neutron Generation Time, A'^(t) 198
Results of Density Variations in the UTVC and Boiler
Col umns 199
VII KINETIC EQUATIONS OF A FOUR-BOILER COLUMN UTVR SYSTEM 209
Introduction 209
The Four-Boiler Column UTVR System Coupled Core Point
Reactor Kinetics Equations 209
Core-to-Core Fuel -Flow Coupling 211
Core-to-Core Neutron Transport Coupling 214
Steady-State Solution 218
.
The Linearized UTVR CC-PRK Equations 221
Inherent Reactivity Feedbacks of the UTVR « 229
Reactivity Feedback of the Boiler Columns, Sp {t) 233
Reactivity Feedback of the UTVC, 8p^{t) 251
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CHAPTER r^ :-. ^aae
VIII DYNAMIC ANALYSIS OF THE UTVR 264
Introduction 264
The Unperturbed UTVR Configuration 265
Results of the Dynamic Analysis 269
Boiler Column Reactivity Perturbation 269
UTVC Reactivity Perturbation 276
Variations in Core-to-Core Direct Neutron Transport
Delay Times 283
Variations in the Coupling Coefficients 287
Variations in the UTVC Fuel Mass Reactivity Feedback
Coefficient 291
Concluding Remark 296
IX SUMMARY OF RESULTS, CONCLUSIONS, AND RECOMMENDATIONS
FOR FURTHER RESEARCH 300
Introduction 300
Summary of Results 300
Results from the Static Neutronic Analysis 300
Results from the Dynamic Neutronic Analysis 302
Comments and Conclusions 303
-
Recommendations for Further Research 305
Static Neutronic Analysis 305
'
Dynamic Neutronic Analysis 307
APPENDICES .. J ' ; . 4 '< ' ^* :^w,
A DESCRIPTION OF THE COMPUTER CODES 309
Introduction 309
Description of Nuclear Codes 309
AMPX: A Modular Code System for Generating Coupled
Multigroup Neutron-Gamma Libraries from ENDF/B 309
The AMPX-DRIVER module 311
The XLACS module 311
The NITAWL module 312
The XSDRNPM module 312
DOT-4: A One- and Two-Dimensional Neutron/Photon
Transport Code 315
GIP 316
MCNP-A General Monte Carlo Code for Neutron and
Photon Transport 317
Description of the EASY5 Engineering Analysis Program 318
B BENCHMARK CALCULATIONS OF XSDRNPM AND DOT-4 WITH MCNP 320
Comparison of XSDRNPM with MCNP 320
Comparison of DOT-4 with MCNP 324
Conclusion 329
Vl•l•l•
APPENDICES Page
C DESCRIPTION OF THE ISOLATOR OF SECONDARY COUPLING
EFFECTS CODE 331
Introduction 331
Description of the ISCE Code 331
The MAIN Module 331
The REED Module 332
The ERIN Module 332
The NOUT Module 333
The ESTM Module 333
The RITE Module 337
Input Data Format 337
Input Data File 337
List of Input Data Files 339
Comparison of Results Obtained from ISCE with Results
Obtained Directly from MCNP 340
D CIRCULATING-FUEL, COUPLED CORE POINT REACTOR
KINETICS EQUATIONS 345
Description and Definition of Symbols, Parameters, and
Terms used in the Circulating-Fuel, Coupled Core
Point Reactor Kinetics Equations 350
Definition of Superscripts and Subscripts 350
Definition of Integral Parameters 351
Neutron population, N"^{t) 351
Reactivity, /)J(t) 354
Effective delayed neutron fraction, ^(t) 354
Prompt neutron generation time, A'^{t) 358
Effective delayed neutron precursor concentration
for the delayed neutron grouo, tj{t) 358
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Effective coupling coefficient, c:3 (t) 359
Interpretation of Equations (D-l) and (D-4) 361
Equation (D-l) 361
Equation (D-4) 365
LIST OF REFERENCES 366
BIOGRAPHICAL SKETCH 371
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