Design Of An Actinide Burning, Lead or Lead-Bismuth Cooled Reactor That Produces Low Cost ...

227 Pages·2005·6.72 MB·English

INEEL/EXT-02-01249
Design Of An Actinide Burning,
Lead or Lead-Bismuth Cooled
Reactor That Produces Low
Cost Electricity
October 2002
Idaho National Engineering and Environmental Laboratory
Bechtel BWXT Idaho, LLC
INEEL/EXT-02-01249
MIT-ANP-PR-092
Design of an Actinide Burning, Lead or Lead-Bismuth
Cooled Reactor That Produces Low Cost Electricity
FY-02 Annual Report
October, 2002
Report compiled and edited by
P. E. MacDonald and J. Buongiorno
MIT Principal Investigators INEEL Principal Investigators
Prof. Ron Ballinger -Material Studies Dr. Jacopo Buongiorno, Reactor Design
Prof. Ken Czerwinski –Polonium Technology Cliff Davis -Reactor Design
Richard Herron – Reactor Design Dr. Steve Herring -Neutronics
Chris Larson - Polonium Technology Philip MacDonald - Project Leadership
Vaclav Dostal –Reactor Design Dr. Eric Loewen -Materials Studies
Dr. Pavel Hejzlar -Neutronics Dr. Kevan Weaver -Neutronics
Jeongyoun Lim-Materials
Prof. Mujid Kazimi -Reactor Design
Prof. Neil Todreas - Project Leadership
Table of Contents
TABLE OF CONTENTS.............................................................................................................II
EXECUTIVE SUMMARY.......................................................................................................VII
1. PROJECT OVERVIEW..........................................................................................................1
1.1. BACKGROUND.....................................................................................................................1
1.2. PROJECT OBJECTIVES AND ORGANIZATION........................................................................3
1.3. CURRENT RESEARCH DIRECTION........................................................................................4
2. RESULTS OF FY-02 REACTOR CORE NEUTRONICS STUDIES..................................8
2.1. MIT FY-02 RESULTS...........................................................................................................8
2.1.1. Neutronic Design for Self-Controllability....................................................................8
2.1.1.1. Objective................................................................................................................8
2.1.1.2. Reactivity Feedback Ratios Criteria for Self-Controllable LBE-Cooled Core......8
2.1.1.3. Evaluation of Core Radial Expansion Coefficient...............................................13
Core Radial Expansion Phenomenon and State-of-the Art Modeling...........................14
Simplified Approach for Assessment of Core Radial Expansion in ABR using MCNP15
Verification of the Simplified Approach on an IFR Core Design.................................17
2.1.1.4. Comparisons Of the ABR and IFR Reactivity Feedback Coefficients and S-
Criteria...............................................................................................................................19
2.1.1.5. Modified core design with reduced control rod worth........................................20
2.1.2. Neutronic Performance Of The Modified Reference Core Design.............................21
2.1.2.1. Power Distribution At Beginning-Of-Life..........................................................21
2.1.2.2. Burnup Performance............................................................................................23
Burnup Comparison using MCODE, MOCUP and MONTEBURNS and Code Choice
.......................................................................................................................................25
Uncertainties in Minor Actinide Libraries.....................................................................27
Cycle Length and Transuranics Destruction Rate.........................................................31
Proliferation Considerations..........................................................................................32
2.1.2.3. Reactivity Feedbacks and Control Parameters....................................................32
Doppler Coefficient.......................................................................................................32
Fuel Thermal Expansion Coefficient.............................................................................34
Coolant Void Worth and Coolant Temperature Coefficient..........................................35
Effective Delayed Neutron Fraction..............................................................................35
Control Rod Worth and Driveline Expansion...............................................................37
2.1.2.4. Self-Controllability Characteristics for the Modified Core Design.....................40
2.1.3. Decay Heat Calculation.............................................................................................42
2.1.4. Fuel Cycle Cost Assessment.......................................................................................43
2.1.4.1. Approach Using Accelerator-Driven Facility Data.............................................45
2.1.4.2. Approach using Generation IV Fuel Cycle Cost Guidelines...............................48
2.1.4.3. Comparison With Accelerator-Driven Facility Fuel Cycle Costs.......................50
2.1.5. Conclusions and Future Work....................................................................................50
2.2. FY-02 INEEL RESULTS – A QUALITATIVE ASSESSMENT OF SODIUM AND LEAD-BISMUTH
...................................................................................................................................................52
3. RESULTS OF FY-02 PLANT ENGINEERING AND ECONOMIC STUDIES..............56
3.1. FEASIBILITY OF A GAS-LIFT PUMPING FOR THE PB-BICOOLED REACTOR.......................56
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3.2. ANALYSES OF REACTOR TRANSIENTS...............................................................................58
3.2.1. ATHENA Model Description......................................................................................60
3.2.2. Primary Coolant Pump Trip.......................................................................................64
3.2.3. Station Blackout..........................................................................................................66
3.2.4. Step Reactivity Insertion.............................................................................................67
3.2.5. Heat Exchanger Tube Rupture...................................................................................68
3.2.6. Turbine Stop Valve Closure........................................................................................71
3.2.7. Rupture of the Steam Line Piping without Scram.......................................................73
3.2.8. Loss of Preheating Without Scram.............................................................................74
3.2.9. Cleanup System LOCA Without Scram......................................................................75
3.2.10. Summary of Transient Results..................................................................................78
3.3. IMPROVEMENT AND BENCHMARK OF THE METAL-FUEL PERFORMANCE MODEL............80
3.4. PUMP SELECTION...............................................................................................................81
3.4.1. Introduction................................................................................................................81
3.4.2. Current Pumping Requirements of the LBE Reactor.................................................82
3.4.3. Similar Liquid Metal Cooled Reactors.......................................................................83
3.4.3.1. Electromagnetic (EM) Pumps.............................................................................84
3.4.3.2. Centrifugal Pumps...............................................................................................85
3.4.4. Pump Selection for the LBE Reactor..........................................................................86
3.4.4.1. EM Pump Limitations.........................................................................................86
3.4.4.2. Centrifugal Pump Selection.................................................................................88
Affinity Laws................................................................................................................89
Scaling a Lead-Bismuth Pump......................................................................................89
Scaling and Correcting a Water Pump..........................................................................91
3.4.5. Pump Cavitation.........................................................................................................93
3.4.5.1. Cavitation Theory................................................................................................94
3.4.5.2. Cavitation Prediction...........................................................................................95
3.4.6. Pump Erosion.............................................................................................................97
3.4.7. Pump Selection Conclusions......................................................................................98
3.5. HEAT EXCHANGER DESIGN...............................................................................................98
3.5.1. Introduction................................................................................................................98
3.5.2. Shell Side Heat Transfer.............................................................................................99
3.5.3. Tube Side Heat Transfer...........................................................................................100
3.5.3.1. Superheated Steam Generator...........................................................................100
Sub-Cooled Region.....................................................................................................100
Nucleate Boiling Region.............................................................................................101
Post Critical Heat Flux Region....................................................................................101
Superheated Region.....................................................................................................102
Critical Quality............................................................................................................102
3.5.3.2. Supercritical Steam Generator...........................................................................103
3.5.3.3. CO Heat Exchanger..........................................................................................104
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3.5.4. Overall Heat Transfer..............................................................................................104
3.5.5. Heat Exchanger Sizing.............................................................................................105
3.5.5.1. Energy Balance..................................................................................................105
3.5.5.2. The (cid:304)-NTU Method............................................................................................105
3.5.5.3. Heat Transfer Area............................................................................................107
3.5.5.4. Axial Variation in the Shell Side Convection Coefficient.................................107
3.5.6. Shell Side Pressure Drop..........................................................................................108
3.5.7. Design Constraints...................................................................................................110
3.5.7.1. Tube Bank Vibrations........................................................................................110
Fluid-Elastic Instability...............................................................................................110
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Vortex Shedding..........................................................................................................111
Turbulent Buffeting.....................................................................................................112
3.5.7.2. Velocity Limit...................................................................................................112
3.5.7.3. Tube Structural Analysis...................................................................................113
3.5.8. Application of Design Principles..............................................................................114
3.5.8.1. Pitch-to-Diameter Ratio.....................................................................................114
3.5.8.2. Superheated Steam Generator...........................................................................115
3.5.8.3. Supercritical Steam Generator...........................................................................116
3.5.8.4. Supercritical CO Heat Exchanger....................................................................117
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3.5.9. Summary of Heat Exchanger Design Results...........................................................117
3.6. HEAT EXCHANGER ACCIDENT ANALYSIS.......................................................................119
3.6.1. Introduction..............................................................................................................119
3.6.2. Standard Scenario....................................................................................................119
3.6.3. Steam Explosion.......................................................................................................120
3.6.4. Lead Oxide Formation.............................................................................................121
3.6.4.2. Possible Reactions.............................................................................................121
3.6.4.2. Gibbs Free Energy.............................................................................................122
3.6.5. Calculating Choked Flow.........................................................................................123
3.6.5.1. Single Phase Critical Flow (CO )......................................................................123
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3.6.5.2. Two-Phase Critical Flow (H O)........................................................................123
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3.7. SUPERCRITICAL STEAM CYCLE.......................................................................................125
3.7.1. Introduction..............................................................................................................125
3.7.2. Supercritical Steam Generator.................................................................................125
3.7.3. Power Cycle..............................................................................................................125
3.7.4. Existing Supercritical Steam Power Cycles.............................................................126
3.7.5. Supercritical Steam Cycle Conclusions....................................................................127
3.8. UNDER-LBE VIEWER......................................................................................................127
3.8.1. Introduction..............................................................................................................127
3.8.2. Operating Conditions...............................................................................................128
3.8.2.1. Radiation Conditions.........................................................................................128
3.8.2.2. Temperature Conditions....................................................................................128
3.8.3. Sound Transmission Properties................................................................................128
3.8.3.1. Speed of Sound in Coolant................................................................................128
3.8.3.2. Attenuation Losses............................................................................................129
3.8.4. Conclusions Regarding Under-LBE Viewing...........................................................131
3.9. SUPERCRITICAL RECOMPRESSION CO BRAYTON CYCLE..............................................131
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3.9.1. Recompression Cycle................................................................................................131
3.9.2. Design Case Studies of the Recompression Brayton Cycle......................................133
3.9.3. Component Design...................................................................................................135
3.9.3.1. Supercritical CO Recompression Brayton Cycle Heat Exchanger Design......135
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3.9.3.2. Turbo-Machinery Design..................................................................................137
Compressor Design.....................................................................................................137
Turbine Design............................................................................................................139
Comparison with other turbines..................................................................................139
3.9.4. Summary...................................................................................................................140
3.9.5. Future Work..............................................................................................................141
3.10. CAPITAL COST ANALYSIS..............................................................................................141
3.10.1. Introduction............................................................................................................141
3.10.2. Capital Cost Analysis Methods...............................................................................141
3.10.2.1. Scaling Relationships......................................................................................141
3.10.2.2. Interest During Construction...........................................................................142
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Cash Flow....................................................................................................................142
3.10.2.3. Contingency Calculations................................................................................144
3.10.2.4. Constant to Current Dollar Conversion...........................................................144
3.10.3. Application of Scaling Relationships......................................................................145
3.10.3.1. Nuclear Steam Supply System (NSSS)...........................................................145
Vessels.........................................................................................................................145
Heat Exchangers and Main Coolant Pumps................................................................147
Intermediate Heat Transport System and Steam Generators.......................................147
3.10.4. Sensitivity Analysis.................................................................................................148
3.10.4.1. Variation of heat exchanger to EM Pump Cost Ratio, F.................................148
3.10.4.2. Variation of Net Cycle Efficiency...................................................................149
3.10.5. Summary of the Capital Cost Analysis...................................................................149
4. RESULTS OF FY-02 MATERIAL STUDIES...................................................................151
4.1. INEEL FY-02 MATERIAL STUDIES.................................................................................151
4.1.1. Description of the Experimental Apparatus.............................................................153
4.1.2. Inductively Coupled Plasma (ICP) Results..............................................................156
4.1.3. SEM Results..............................................................................................................163
4.1.4. Discussion of the Coolant Chemistry Control, Effects of Zirconium Addition, and
Relative Reaction Rates of Commercial Alloys...................................................................169
4.2. FY-02 MIT MATERIALS STUDIES...................................................................................171
4.2.1. Introduction..............................................................................................................171
4.2.2. Experimental System Description & Test Matrix.....................................................174
4.2.2.1. System Description............................................................................................174
4.2.2.2. Test Matrix and Materials.................................................................................177
4.2.3. Test Results-Refractory Metals.................................................................................178
4.2.3.1. Molybdenum (Mo)............................................................................................178
4.2.3.2. Tungsten (W).....................................................................................................179
4.2.3.3. Tantalum (Ta)....................................................................................................179
4.2.3.4. Conclusions-Refractory Metal Exposure...........................................................179
4.2.4. Iron-Based Program.................................................................................................179
4.2.4.1. Results: Pb Exposure.........................................................................................180
4.2.4.2. Tested in Pb-Bi Eutectic....................................................................................185
4.2.4.3. Discussion..........................................................................................................186
4.2.5. Future Work-Alloy Development..............................................................................186
5. COOLANT ACTIVATION STUDIES...............................................................................188
5.1. POLONIUM HYDRIDESTRIPPING AND FORMATION OF RARE-EARTH POLONIDES..........189
5.1.1. Kinetics of H Po Formation.....................................................................................189
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5.1.2. Interactions of PbPo and Pr.....................................................................................192
5.1.3. Small-Scale Design of Extraction Systems...............................................................193
5.2. ALKALINE EXTRACTION..................................................................................................194
5.2.1. Justification of Tellurium as a Polonium Surrogate.................................................195
5.2.2. Reaction Cell and Preliminary Results....................................................................197
5.2.3. Crucible Materials....................................................................................................198
5.2.4. Conclusions..............................................................................................................199
7. REFERENCES.....................................................................................................................200
APPENDIX A. PUBLICATION HISTORY OF THE LEAD-COOLED ACTINIDE
BURNING REACTOR PROJECT..........................................................................................206
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APPENDIX B. COST ANALYSIS TERMINOLOGY.........................................................212
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Executive Summary
The purpose of this collaborative Idaho National Engineering and Environmental Laboratory
(INEEL) and Massachusetts Institute of Technology (MIT) Laboratory Directed Research and
Development (LDRD) project is to investigate the suitability of lead or lead-bismuth cooled fast
reactors for producing low-cost electricity as well as for actinide burning. The goal is to identify
and analyze the key technical issues in core neutronics, materials, thermal-hydraulics, fuels, and
economics associated with the development of this reactor concept. Work has been accomplished
in four major areas of research: core neutronic design, plant engineering, material compatibility
studies, and coolant activation. The publications derived from work on this project (since project
inception) are listed in Appendix A. This is the third in a series of Annual Reports for this
project, the others are also listed in Appendix A as FY-00 and FY-01 Annual Reports.
NEUTRONIC DESIGN
The major focus of the neutronic analyses performed at MIT in FY-02 was on the design of a core
that can achieve excellent safety through self-controllability (as in the Integral Fast Reactor, IFR)
and high transuranic destruction. The analyses were performed for metallic thorium-based fuel
(Th-U-Pu-MA-Zr) in a once-through cycle assuming that the discharged fuel from the Actinide
Burner Reactor (ABR) remains in temporary storage before multi-recycling is introduced. The
major conclusions of these studies can be summarized as follows:
(cid:120)(cid:3) Using thorium as the prime fertile material is an effective means to reduce the large
reactivity swing occurring in fertile-free cores while still allowing a high net actinide
destruction rate per MWth. In addition, thorium use increases the Doppler feedback in
comparison with fertile-free fuels and reduces the coolant density reactivity coefficient.
Both the Doppler and fuel thermal expansion feedbacks are negative and their values are
comparable to those for the IFR fuel.
(cid:120)(cid:3) The ABR destruction rate of actinides per MWth-yr is ~35% less than the destruction rate
in the fertile-free critical ABR but only 20% less than in an accelerator-driven facility
(due to the higher capacity factor in the ABR). This is a very appealing feature
considering the simplicity of the proposed reactor versus the more complex accelerator-
driven system.
(cid:120)(cid:3) The discharged fuel from the ABR satisfies proliferation constraints for both the
plutonium and uranium compositions. Plutonium isotopics are significantly degraded
from that of the PWR spent fuel vector making it virtually weapons unusable. Also, the
fraction of in-bred U-233 remains below 12% if depleted uranium is mixed with the
thorium (~30wt% of uranium in the U+Th mixture).
(cid:120)(cid:3) The high coolant void worth typical of liquid metal cooled fast reactors (especially those
with minor actinide fuel) can be effectively mitigated by the employment of streaming
fuel assemblies yielding a negative coolant void worth and a very small positive coolant
temperature coefficient.
(cid:120)(cid:3) The combination of reactivity coefficients satisfies the requirements of self-control
throughout the cycle so that, in concert with the passive decay heat removal design, all
transients without scram lead to an inherent shutdown without exceeding safe fuel and
structural temperature limits.
(cid:120)(cid:3) The fuel cycle cost of the once-through cycle is significantly higher than that of the
multi-recycle scheme (about 30 versus 15 mills/kW-hr, best estimate). Therefore,
considering fuel cycle economics and the small potential for reduction of long-term
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radiotoxicity and heat load on the repository from the wastes of the LWR/ABR system,
the once-through fuel cycle has to be discarded from future considerations.
(cid:120)(cid:3) In comparison with an accelerator-driven facility, the fuel cycle cost of the ABR in the
multi-recycle scheme is slightly smaller, but both the accelerator-driven facility and ABR
fuel cycle costs are well above current LWR fuel cycle costs, even if lower bound values
are used [the lower bound value for both the accelerator-driven facility and the ABR is
about 9 mills/kW-hr versus about 3 mills/kW-hr for light water reactors (LWRs)].
In summary, the proposed design of the 7-year life core for burning transuranics from spent LWR
fuel appears to be very promising and deserving of future refined analyses and optimization
because it offers high consumption of actinides, excellent safety characteristics, and has the
potential to have low electricity generation cost due to its modularity, simplicity, and high
capacity factor. The only drawback is a high fuel cycle cost, which is inherent to all actinide
burning systems because of the currently high fuel reprocessing costs.
PLANT ENGINEERING
The plant engineering work reported in this Annual Report includes: an evaluation of gas-lift
pumping for the ABR, analyses of various reactor transients, improvement of the metal-fuel
performance modeling, pump selection, heat exchanger design and accident analyses, an
assessment of the supercritical steam cycle, an assessment of under-lead-bismuth eutectic (LBE)
viewing technology, and a capital cost analysis.
Evaluation of gas-lift pumping. The feasibility of a gas-lift pump approach for the ABR was
assessed. Gas-lift pumping of the LBE coolant in our case will require generation of a 53% void
fraction in the chimney resulting in 2.3 m swelling of the liquid level. Such a large swelling of
the liquid level would require design of a much longer vessel. The gas flow rate required to
sustain a 53% void fraction in the chimney is 1.6 kg/s. The gas would be injected at the bottom
of the chimney at a pressure of about 800 kPa (corresponding to the weight of the LBE column in
the chimney). The pumping power would be about 5.1 MW versus 3.8 MW for a mechanical
pump. Finally, the 1.6 kg/s helium mass flow rate corresponds to a 28 m3/s volumetric flow and
to a 3.8 m/s helium superficial velocity at the free liquid surface, enough to entrain considerable
amounts of LBE, which would then have to be removed before the compressor inlet. These
results clearly demonstrate that a gas-lift pump approach is not feasible for the ABR.
Analyses of Reactor Transients. The ATHENA code was used to determine the response of the
ABR to a variety of transients, including pump trip, station blackout, reactivity insertion, heat
exchanger tube rupture, turbine stop valve closure, steam line break, loss of feed-water
preheating, and loss of coolant from the reactor cleanup system. The transients were simulated
without reactor scram to demonstrate the safety margins inherent in the reactor design. The ABR
design successfully met the identified cladding, fuel, and guard vessel temperature limits for each
of the transients analyzed. The cladding temperature was always closer to its limit than the fuel
and guard vessel temperatures. The most limiting transient is initiated by a station blackout. The
station blackout coupled with a failure to scram produced a peak cladding temperature that was
equal to the transient limit. The margin to the temperature limit during a station blackout
increased to a more comfortable 24(cid:113)C when a reactor scram was simulated. A steam line break
does not result in significant overcooling in the actinide-burner reactor. The overcooling
potential is limited by the small water inventory in the heat exchangers compared to the heat
capacity of the lead-bismuth coolant. The transient initiated by a heat exchanger tube rupture
resulted in the highest cover gas pressure of the cases evaluated. Because of the high pressure on
the secondary side of the heat exchangers, a relief valve will be required to protect the reactor
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vessel. The maximum pressure is governed by the relief valve capacity and set-point. It was set
to open at 0.2 MPa in this analysis. Also, a loss-of-coolant accident (LOCA) sequence involving
the coolant cleanup system was analyzed and the ABR was again found to be passively safe.
Metal-fuel Modeling. During FY-02 two major improvements were made to the metal-fuel
model developed in this project during FY-01 and the revised model was benchmarked with
metal-fuel irradiation data from the IFR development program. First, better constitutive
equations for the irradiation creep and thermal creep of the cladding material HT-9 were added to
the code. The second improvement consisted in allowing for non-constant temperature, dose, and
linear heat generation rate during irradiation. The new version of the model can simulate the
behavior of a metal fuel pin with arbitrarily changing operating conditions.
Pump Selection. The work in Section 3.4 illustrated that a centrifugal pump is capable of
meeting the ABR pumping needs. Also, that these needs can be met by two pumps which will
neatly fit within the 1.05 meter annular gap between the reactor vessel and the core chimney.
Heat Exchanger Design and Accident Analyses. The reevaluation of the heat exchangers
reaffirmed that the 700 MWth of core power can be transmitted to the power cycle. The new
design is a baffled, cylindrical shell and tube heat exchanger with a modular design. Eight
cylindrical heat exchangers have replaced the two original kidney-shaped components identified
in the FY-01 design. Each heat exchanger is 9.0 meters in length with an inside shell diameter of
1.0 meters. As before, the LBE coolant circulates on the shell side and the secondary coolant in
the tubes, which have a triangular pitch. Three secondary-side heat exchanger variants are
presented: superheated steam, supercritical steam, and supercritical CO . The worst-case heat
2
exchanger accident with either a steam or CO cycle was assessed. The hazardous result
2
investigated was formation of lead oxide (PbO) in the primary coolant system, which could block
or disrupt flow paths. We determined that reactions which could lead to lead oxide formation
were too slow to result in significant oxide formation in the event of a tube rupture.
Supercritical Steam Cycle. An initial investigation of a supercritical steam cycle was performed
to determine if it is desirable alternative to either the superheated steam or supercritical CO
2
power cycles. The net efficiencies obtainable with the supercritical steam cycle are competitive
with those of the supercritical CO power cycle. However, the balance of plant design with the
2
CO power cycle is much simpler.
2
Under-LBE Viewing. Toshiba has performed significant work in the development of a 3-D
ultrasonic under-sodium viewer. No such work has been done with a lead-based coolant, but an
under-lead ultrasonic viewer could be expected to have higher resolution, but somewhat lower
range.
Supercritical-CO Brayton Cycle. Work was performed during FY-02 to optimize the
2
recuperator and develop preliminary designs for the main components used in the supercritical-
CO Brayton cycle. The need for effective recuperators led to consideration of compact heat
2
exchangers. In order to accommodate the high pressure differential across the recuperator, the
printed circuit heat exchangers manufactured by HEATRIC were selected. The most important
factors for the performance evaluation were the heat exchanger geometry and flow arrangement.
The diameter of the semicircular channels was selected as 1 mm. The heat conduction
characteristic length was assumed to be equal to the plate thickness, even though it is likely to be
smaller.
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