INEEL/EXT-04-02530 Feasibility Study of Supercritical Light Water Cooled Reactors for Electric Power Production Nuclear Energy Research Initiative Project 2001-001 Westinghouse Electric Co. Award Number: DE-FG07-02SF22533 Final Report 12th Quarterly Report Principal Investigators: Philip MacDonald, Dr. Jacopo Buongiorno, Dr. James W. Sterbentz, Cliff Davis, and Prof. Robert Witt Telephone: 208-526-9634 Fax: 208-526-2930 Email: [email protected] Collaborating Organizations: University of Michigan Principal Investigators: Prof. Gary Was, J. McKinley, and S. Teysseyre Westinghouse Electric Company Principal Investigators: Dr. Luca Oriani, Dr. Vefa Kucukboyaci, Lawrence Conway, N. Jonsson, and Dr. Bin Liu Executive Summary The supercritical water-cooled reactor (SCWR) is one of the six reactor technologies selected for research and development under the Generation-IV program. SCWRs are promising advanced nuclear systems because of their high thermal efficiency (i.e., about 45% vs. about 33% efficiency for current Light Water Reactors, LWRs) and considerable plant simplification. SCWRs are basically LWRs operating at higher pressure and temperatures with a direct once-through cycle. Operation above the critical pressure eliminates coolant boiling, so the coolant remains single-phase throughout the system. Thus the need for a pressurizer, steam generators, steam separators and dryers is eliminated. The main mission of the SCWR is generation of low-cost electricity. It is built upon two proven technologies, LWRs, which are the most commonly deployed power generating reactors in the world, and supercritical fossil-fired boilers, a large number of which is also in use around the world. The reference SCWR design for the U.S. program is a direct cycle system operating at 25.0 MPa with core inlet and outlet temperatures of 280 and 500 °C, respectively. The coolant density decreases from about 760 kg/m3 at the core inlet to about 90 kg/m3 at the core outlet. The inlet flow splits with about 10% of the inlet flow going down the space between the core barrel and the reactor pressure vessel (the downcomer) and about 90% of the inlet flow going to the plenum at the top of the rector pressure vessel to then flow downward through the core in special water rods to the inlet plenum. Here it mixes with the feedwater from the downcomer and flows upward to remove the heat in the fuel channels. This strategy is employed to provide good moderation at the top of the core. The coolant is heated to about 500 °C and delivered to the turbine. The purpose of this NERI project was to assess the reference U.S. Generation IV SCWR design and explore alternatives to determine feasibility. The project was organized into three tasks. • Task 1. Fuel-cycle Neutronic Analysis and Reactor Core Design. • Task 2. Fuel Cladding and Structural Material Corrosion and Stress Corrosion Cracking. • Task 3. Plant Engineering and Reactor Safety Analysis. Task 1. Fuel-cycle Neutronic Analysis and Reactor Core Design Metallic and oxide fertile fuels in a fast-spectrum SCWR were investigated during Year 1 to evaluate the void and Doppler reactivity coefficients, actinide burn rate, and reactivity swing throughout the irradiation cycle. These results were reported in the 1st Quarterly. A variety of other core arrangements and moderator types for a thermal-spectrum SCWR were also assessed during the three years of this project. Detailed results from the solid moderator studies were presented in the 3rd Quarterly and 4th Quarterly Reports and in two papers by Buongiorno and MacDonald (2003a & 2003b). Results from an analysis of an alternative thermal-spectrum SCWR design based on vertical power channels, hexagonal fuel assemblies, and water moderation between the fuel assemblies were reported in the 2nd Annual Report. Also reported in the 2nd Annual Report were the results of the steady-state thermal-hydraulic analyses for two other thermal spectrum SCWRs, one design with solid moderator rods, and one design with water filled moderator rods. This report presents in Chapter 3 the results of a two neutronic evaluations for two different SCWR fuel assembly designs. The first evaluation is for a 25x25 fuel assembly that used MA956 oxide dispersion steel for the fuel rod cladding, water rod duct, and assembly duct materials. The second is for a 21x21 fuel assembly that used silicon carbide (SiC) for the fuel rod cladding, water rod duct, and assembly duct materials. ii The 25x25 fuel assembly with MA956 cladding and duct material contains a 6x6 array of square water rods interspersed uniformly within the UO fuel pin array to increase neutron moderation and assembly 2 reactivity. The assembly exhibits many desirable neutronic characteristics that include sufficient reactivity to achieve burnups of at least 31.0 GWD/MTU, a strongly negative Doppler coefficient, and a negative void worth for both the coolant and water rods. The assembly also exhibits some characteristics that may complicate the design, such as a wide spread in the required radial enrichments (3.2-12.4 wt% U-235) to flatten the radial power profile, and relatively lower reactivity than in LWRs because of the high parasitic neutron absorption of the MA956 cladding and duct material. In addition, the assembly axial power profile exhibits a strong sensitivity to small changes in axial enrichments that may lead to power oscillations under normal operation, if not properly controlled with burnable poisons and control rods. The sensitivity is believed to be primarily due to the interplay between the non-uniform axial water density profiles that affect neutron moderation and the time-dependent axial burnup of the fissile heavy metal. The 21x21 fuel assembly with the duplex SiC/SiC for the fuel pin cladding and fuel assembly duct material exhibited better neutronic characteristics than the MA956 assembly with steel structures. This assembly showed a significant increase in core reactivity due to the relatively low parasitic neutron absorption of the silicon carbide. This low parasitic neutron absorption in turn translates into significantly higher burnup (41.0 GWD/MTU) when compared to the burnup of the MA956 steel assembly (31.0 GWD/MTU). In addition, the SiC assembly exhibits a strong negative Doppler coefficient ((cid:150)2.5 pcm/°C), and negative void worth for both the coolant and water rods. The assembly does however require again a relatively wide spread in the fuel rod radial enrichment (3.2-12.4 wt% U- 235) to flatten the radial power profile at beginning-of-life conditions, and in addition would require at least a three-zone axial enrichment to flatten and center the unrodded axial power profile about the core midplane at beginning-of-life. As with the MA956 assembly, the fuel assembly axial power profile appears to exhibit a strong sensitivity to small changes in axial enrichments (and therefore burnup) that could lead to power oscillations under normal operation, if not properly controlled with burnable poisons and control rod movement. Task 2. Fuel Cladding and Structural Material Corrosion and Stress Corrosion Cracking The existing data base on the corrosion and stress-corrosion cracking of austenitic stainless steel and nickel-based alloys in supercritical water is very sparse. Therefore, the focus of this work has been corrosion and stress corrosion cracking testing of candidate fuel cladding and structural materials. During Year 1, a high temperature autoclave with carefully controlled chemistry and containing a constant rate mechanical test device was built and tested at the University of Michigan. During Years 2 and 3, a variety of austenitic and ferritic-martensitic alloys were tested. The results of that work are presented in Section 4 of this report and briefly summarized below. The austenitic alloys were tested in deaerated water (dissolved oxygen of the order of a few ppb) at temperatures between 400 and 550 (cid:176)C and they all showed varying degrees of susceptibility to intergranular stress-corrosion cracking (SCC). Susceptibility was determined by examination of the fracture surface, the gage surface, and by analysis of cross-sections of the tensile bars. All these measurements are required to provide a complete description of the cracking behavior. Alloy 625 is the most susceptible, displaying the highest degree of intergranular fracture and some of the deepest cracks along with a very high crack density. The 304L stainless steel is the next most susceptible material, showing the deepest intergranular cracks. Alloys 690 and 316L are the least susceptible austenitic alloys from all measures considered; crack density, crack depth and crack length. iii The degree of intergranular SCC of the austenitic alloys increases with increasing temperature. As the temperature increases, the crack density decreases but the crack length and depth increase, resulting in a net increase in the intergranular cracking severity as measured by the crack length per unit area. There is very strong temperature dependence to the oxidation behavior of the austenitic alloys. The oxidation rate, as measured either by the weight gain or oxide thickness, increases faster with increasing temperature. By 550 (cid:176)C, the austenitic alloy oxide thickness is approaching 10 µm within a few hundred hours. The predominant feature among all of the austenitic alloy oxides was the two-layer structure consisting of an iron-rich outer layer and chromium-rich inner layer. X-ray diffraction has shown that the outer layer was magnetite, Fe O . The outer oxide on the Alloy 690 was probably NiO. 3 4 The ferritic-martensitic alloys do not display any evidence of intergranular SCC as determined by fracture surface and gage surface analysis. They all display strain softening and ductile rupture. However, the oxidation rates of the ferritic-martensitic alloys are very high compared to the austenitic alloys. At 500 (cid:176)C, the ferritic-martensitic oxidation rates are a factor of 10 greater than those for the austenitic alloys at the same temperature. These alloys also display a two-layer structure, in which the outer layer is identified as magnetite, Fe O . The O/M ratio of the inner layer is closer to hematite, but the structure of 3 4 the inner oxide layer was not verified. The addition of 100 ppb oxygen to the water at 500 (cid:176)C resulted in a reduction of the total oxide thickness by about 10% and a slight increase in the O/M ratio. These results are consistent with the objective of combined water chemistry control. Task 3. Plant Engineering and Reactor Safety Analysis SCWR Core Thermal Hydraulic Design Assessment. The Westinghouse Electric Company tasks included an assessment of the reference core thermal hydraulic design. A complete review of the Westinghouse SCWR core assessment activities is provided in Westinghouse Report STD-ES-04-45, while in this report the focus is mostly on the final analyses and the main conclusions of the analyses effort. The first step in performing the core thermal-hydraulic assessment was identification of the design limits that were then used to evaluate the acceptability of the core design. Section 2 of STD-ES-04-45 and the 2nd and 3rd Quarterly reports for this NERI project (MacDonald et al. 2002a and 2002b) provide the considerations used in defining the design limits for the SCWR. Once the boundaries of the analysis were defined, simplified calculations were performed to provide an initial characterization of the design. These analyses are summarized in Section 3 of STD-ES-04-45. Based on the results of the simplified analysis, it was concluded that the SCWR, due to its very large enthalpy rise along the core, is sensitive to small deviations from nominal conditions, especially variations in the flow to power ratio. Thus, even small effects due to various hot channel factors (coolant flow channel tolerances, operational variations, etc.) might have a large impact on the peak cladding temperature of some fuel rods. This was considered a major feasibility issue for the SCWR, and thus it was decided to perform detailed subchannel analysis of the SCWR core to provide a more in depth assessment of this issue. The W-VIPRE subchannel analyses code was adapted for the analysis of supercritical water, and new correlations that are considered adequate for SCWR analyses were implemented in the code. A complete characterization, including sensitivity studies, of the SCWR with the modified VIPRE core is documented in Section 5.1.2. Based on these results, sufficient information was available for a preliminary thermal- hydraulic optimization of the SCWR core design. Temperature profiles for various core geometries were then analyzed with two different objectives: (1) to identify an optimal geometry that minimizes the temperature differences between core channels, and (2) to confirm and characterize the sensitivity of the iv temperature profile to the local flow to power ratios. The need of maintaining uniform conditions at the exit of the core is dictated by the fact that safety limits need to be verified for the limiting fuel rod, while the overall plant performance depends on the average core exit conditions. Thus, a uniform temperature distribution minimizes the (cid:147)wasted(cid:148) design margin. These analyses are documented in Section 5.1.3 (Section 5 of STD-ES-04-45). Results from the optimization studies suggest that it is possible to obtain a better temperature profile (hence, lower hot channel factors) by employing a more complex assembly configuration. Based on the results of this study, the design should use a geometric configuration with 10mm outside diameter fuel rods for the coolant channels facing the water rods and at the assembly periphery, 9.5mm outside diameter fuel rods for the assembly corners, and 10.2mm outside diameter fuel rods for all other positions. While this study shows a path to obtain an acceptable thermal hydraulic design, it also provides the designer with an important design issue: the flow is clearly extremely sensitive to small variations in the channel flow area. Therefore, rod bowing and even the tolerances in rod dimensions could be crucial in terms of the temperature peaking. This sensitivity, which can be attributed to large channel enthalpy rise coupled with a region of low-density coolant and high exit velocities, renders the design uncertain. Based on the results of this study, it appears that the reference SCWR design is not feasible. Although additional design and analysis might allow the recovery of some margin, it is unlikely that a SCWR assembly and core design can be developed that provides acceptable performance (i.e. low enough hot channel exit temperature). Therefore, the SCWR core design remains a major feasibility issue for which a solution has yet to be achieved at this stage of the program. An Evaluation of an Innovative Safety Concept for the SCWR. Preliminary investigations of the safety characteristics of a SCWR performed by INEEL and the Westinghouse have resulted in the development of a novel safety concept for this Generation IV reactor. Previous analyses have shown that the SCWR can meet transient thermal limits for events initiated by loss of main feedwater only if a large capacity auxiliary feedwater system is actuated rapidly. However, the required rapid initiation of auxiliary feedwater was judged to pose significant technical and economic challenges. Consequently, Westinghouse developed an innovative conceptual design that uses a passive circulation system to mitigate the effects of loss of main feedwater. This safety concept utilizes two, relatively small, feedwater tanks that store water for reactor cooling during normal operation and provide sufficient cooling capacity to mitigate the effects of a loss of main feedwater. Main coolant pumps similar to those utilized in advanced light water reactors provide the head required to circulate the flow in the reactor. Although the proposed concept takes advantage of the SCWR once-through, direct cycle concept during normal operation, it allows the establishment of a recirculation path in the system following containment isolation, with an isolation condenser that provides long-term decay heat removal. The safety characteristics of the design were evaluated for loss-of-flow transients using the RELAP5-3D computer code. The results of these evaluations confirmed the potential of the design. Acceptable short- term results following loss of flow were obtained by adjusting the coastdown characteristics of the main coolant pumps. Acceptable long-term decay heat removal following loss of flow was obtained with 300 or more tubes in the isolation condenser. Preliminary evaluations of loss-of-coolant accidents were also performed. The analysis of an accident initiated by a large cold leg break showed that significantly lower cladding temperatures were obtained after the blowdown peak in the proposed design than in a simple, once-through design due to the recirculation loop and the added coolant inventory provided by the feedwater tanks. The milder evolution v of the accident allows for a significant simplification in the design of an adequate emergency core cooling system. Structural Response of SCWR Reactor Pressure Vessel to Thermal Transients. A process was developed for examining the SCWR pressure vessel structural consequences of thermal transients. This is important because hydraulic transients are often much shorter than the reactor pressure vessel thermal diffusion time. For the case examined here involving a sudden 30 °C (54 °F) drop in main feedwater temperature, the peak Von Mises stress is about 30% higher (65 versus 50 ksi) than that experienced under static conditions at nominal system pressure. Given the low number of anticipated cycles and the relatively small temperature change, the ASME Code indicates the transient is of no consequence with respect to low-cycle fatigue. Design of a Thermal Sleeve for the SCWR Hot Leg. The reference SCWR operates at substantially higher reactor coolant outlet temperature and pressure than existing LWRs. Therefore, the 500 °C reactor coolant outlet water must be isolated from the reactor pressure vessel in order to use conventional vessel materials. We examine hot nozzle isolation and recommend a design for a thermal sleeve. The thermal sleeve consists of a one-inch thick (~ 2.5 cm) steel structural insert surrounded by a generous, two-inch (~ 5 cm) radial water gap between thermal sleeve and reactor pressure vessel outlet nozzle. The outlet nozzles should be positioned above the inlet nozzles so that the isolating cold leg flow fills the annular water gap from below. In this way forced convection reinforces natural convection in the gap and peak vessel temperatures are both low and insensitive to water mass flow rate. Fuel Assembly Conceptual Design. A conceptual design of a fuel assembly for the SCWR was developed to identify any feasibility issues. There are three main difficulties in the SCWR fuel assembly design  the fuel rod spacer, the assembly top structure, and the thermal expansion of the components of the fuel assembly. All of these three aspects were addressed in the preliminary fuel assembly design. vi Table of Contents EXECUTIVE SUMMARY.......................................................................................................................II TASK 1. FUEL-CYCLE NEUTRONIC ANALYSIS AND REACTOR CORE DESIGN...........................................II TASK 2. FUEL CLADDING AND STRUCTURAL MATERIAL CORROSION AND STRESS CORROSION CRACKING................................................................................................................................................III TASK 3. PLANT ENGINEERING AND REACTOR SAFETY ANALYSIS..........................................................IV TABLE OF CONTENTS.......................................................................................................................VII 1. PROJECT DESCRIPTION...............................................................................................................1 2. REFERENCE SCWR DESCRIPTION............................................................................................3 2.1. REFERENCE DESIGN POWER AND COOLANT CONDITIONS...........................................................3 2.2. SCWR REACTOR PRESSURE VESSEL............................................................................................4 2.3. SCWR CORE AND FUEL ASSEMBLY DESIGN...............................................................................4 2.4. REACTOR PRESSURE VESSEL INTERNALS.....................................................................................6 2.5. CONTAINMENT DESIGN.................................................................................................................7 2.6. POWER CONVERSION CYCLE........................................................................................................9 3. TASK 1 RESULTS: FUEL-CYCLE NEUTRONIC ANALYSIS AND REACTOR CORE DESIGN (INEEL, DR. JAMES W. STERBENTZ)...............................................................................12 3.1. NEUTRONIC EVALUATION OF A 25X25 SUPERCRITICAL WATER REACTOR FUEL ASSEMBLY WITH WATER RODS AND MA956 CLAD/DUCT MATERIALS...................................................................12 3.1.1. SCWR Fuel Assembly and MCNP Model Description........................................................12 3.1.2. Computer Codes..................................................................................................................13 3.1.3. Axial Power Profile and Enrichment..................................................................................15 3.1.4. Clad Reactivity Comparison...............................................................................................16 3.1.5. Coolant and Water Rod Void Reactivity.............................................................................17 3.1.6. Beginning-of-Life Doppler Coefficient...............................................................................17 3.1.7. Radial Power Profile..........................................................................................................18 3.1.8. Control Rod Design............................................................................................................19 3.1.9. Fuel Depletion....................................................................................................................21 3.1.9.1. SCWR with MA956 (Depletion Study No. 1)............................................................21 3.1.9.2. SCWR with MA956 (Depletion Study No. 2)............................................................23 3.1.9.3. PWR............................................................................................................................24 3.1.9.4. SCWR with SiC..........................................................................................................24 3.1.10. Neutron Spectra..................................................................................................................25 3.1.11. Re-design of the 25x25 SCWR Fuel Assembly....................................................................26 3.1.12. Conclusions.........................................................................................................................27 3.2. NEUTRONIC EVALUATION OF A 21X21 SUPERCRITICAL WATER REACTOR FUEL ASSEMBLY DESIGN WITH WATER RODS AND SIC CLADDING AND DUCT MATERIALS............................................28 3.2.1. SCWR Fuel Assembly and MCMP Model Description.......................................................28 3.2.2. Material Reactivity Comparison (SiC versus MA956)........................................................30 3.2.3. Coolant and Water Rod Void Reactivity.............................................................................31 3.2.4. Doppler Coefficients...........................................................................................................32 3.2.5. Axial Power Profile and Enrichment..................................................................................33 3.2.6. Radial Power Profile and Enrichment................................................................................34 3.2.7. Conclusions.........................................................................................................................35 vii 4. TASK 2 RESULTS: CORROSION AND STRESS CORROSION CRACKING STUDIES (UNIVERSITY OF MICHIGAN, PROF. GARY WAS).......................................................................37 4.1. CONSTRUCTION AND OPERATION OF THE SCW AUTOCLAVE SYSTEM.....................................37 4.1.1. Water Chemistry Control....................................................................................................37 4.1.2. Supercritical Water Condition............................................................................................39 4.1.3. Mechanical Loading...........................................................................................................39 4.1.4. Safety Features....................................................................................................................39 4.2. PERFORMANCE TEST OF THE SCW SYSTEM..............................................................................39 4.3. ALLOY SELECTION......................................................................................................................43 4.4. SCC AND CORROSION IN THE TEMPERATURE RANGE 400 (cid:150) 550 (cid:176)C IN DEAERATED SCW......44 4.4.1. Stress Corrosion Cracking Data.........................................................................................45 4.4.2. Discussion of Stress Corrosion Cracking Results...............................................................53 4.4.3. Corrosion Data...................................................................................................................54 4.4.4. Discussion of Oxidation Results.........................................................................................59 4.5. EFFECT OF OXYGEN CONTENT ON SCC AND CORROSION OF 304L STAINLESS STEEL............60 4.6. EFFECT OF OXYGEN CONTENT ON SCC AND CORROSION OF FERRITIC-MARTENSITIC ALLOYS AT 500 (cid:176)C................................................................................................................................................61 4.7. SUMMARY...................................................................................................................................65 4.8. CONCLUSIONS.............................................................................................................................66 5. TASK 3 RESULTS: PLANT ENGINEERING AND REACTOR SAFETY ANALYSIS........67 5.1. THERMAL-HYDRAULIC ASSESSMENT OF A SUPERCRITICAL WATER REACTOR (SCWR) CORE (WESTINGHOUSE ELECTRIC CO., DR. LUCA ORIANI AND DR. VEFA N. KUCUKBOYACI)......................67 5.1.1. Introduction.........................................................................................................................67 5.1.2. Initial Supercritical Water Reactor Sub-Channel Analysis................................................68 5.1.2.1. VIPRE-W Model.........................................................................................................68 5.1.2.2. Heat Transfer and Friction Loss Correlations.............................................................74 5.1.2.3. Thermal-Hydraulic Profiles in the Hot and Cold Channels for Different Radial Power Distributions....................................................................................................................................75 5.1.2.4. Preliminary Sensitivity Studies...................................................................................83 5.1.2.5. Summary and Conclusions..........................................................................................84 5.1.3. Thermal-Hydraulic Optimization Studies of the Supercritical Water Reactor Fuel Assembly .............................................................................................................................................86 5.1.3.1. Alternative Assembly Designs....................................................................................86 5.1.3.2. Thermal-Hydraulic Profiles in the Hot and Cold Channels for Different Assembly Configurations:...............................................................................................................................86 5.1.3.3. Core Orifices...............................................................................................................99 5.1.3.4. Cladding Temperatures.............................................................................................100 5.1.3.5. Conclusions...............................................................................................................103 5.2. AN EVALUATION OF AN INNOVATIVE SAFETY CONCEPT FOR THE SCWR (INEEL: C. B. DAVIS, WESTINGHOUSE ELECTRIC CO.: DR. L. ORIANI, L. E. CONWAY, AND N. JONSSON)..........................104 5.2.1. Introduction.......................................................................................................................104 5.2.2. Reactor Core Cooling System Description.......................................................................105 5.2.2.1. Approach Overview..................................................................................................106 5.2.2.2. Anticipated Response Following Loss-of-Flow Events............................................108 5.2.2.3. Preliminary Design and Sizing Basis........................................................................109 5.2.3. Transient Analysis.............................................................................................................111 5.2.3.1. Model Description.....................................................................................................111 5.2.3.2. Short-Term Loss of Flow..........................................................................................116 5.2.3.3. Long-Term Loss of Main Feedwater........................................................................119 viii 5.2.3.4. Loss-of-Coolant Accidents.......................................................................................123 5.2.4. Conclusions.......................................................................................................................130 5.3. STRUCTURAL RESPONSE OF SCWR REACTOR PRESSURE VESSEL TO THERMAL TRANSIENTS (INEEL (cid:150) PROF. ROBERT WITT)...........................................................................................................132 5.3.1. Procedure Definition.........................................................................................................132 5.3.2. SCWR Transients..............................................................................................................133 5.3.3. Limitations Of Finite Element Analysis In Step Change Problems..................................135 5.3.4. ABAQUS Simulation Results.............................................................................................137 5.3.5. Significance of Stresses in Overcooling Transient............................................................140 5.3.6. Summary...........................................................................................................................140 5.4. DESIGN OF THERMAL SLEEVE FOR SCWR (INEEL (cid:150) PROF. ROBERT WITT)...........................142 5.4.1. Introduction.......................................................................................................................142 5.4.2. FLUENT Model................................................................................................................142 5.4.3. FLUENT Results...............................................................................................................147 5.4.3.1. Forced Flow From The Top......................................................................................148 5.4.3.2. Forced Flow From The Bottom................................................................................152 5.4.3.3. Forced Flow From The Side.....................................................................................153 5.4.3.4. Forced Flow From Top Or Bottom With Smaller Annular Gap...............................157 5.4.4. Sensitivity of Results to Modeling Choices.......................................................................158 5.4.5. Summary...........................................................................................................................160 5.5. FUEL ASSEMBLY CONCEPTUAL DESIGN (WESTINGHOUSE - DR. BIN LIU, LAWRENCE CONWAY AND DR. LUCA ORIANI)........................................................................................................................162 5.5.1. Description of the SCWR Preliminary Fuel Assembly Design.........................................162 5.5.2. Spacer Design...................................................................................................................164 5.5.3. Fuel Assembly Top Structure............................................................................................165 5.5.4. No Resistance Thermal Expansion Design.......................................................................165 5.5.5. Assembling of the Fuel Assembly......................................................................................166 6. REFERENCES................................................................................................................................168 7. BUDGET AND ACTUAL COSTS FOR YEAR 3.......................................................................172 7.1. INEEL.......................................................................................................................................172 7.2. UNIVERSITY OF MICHIGAN.......................................................................................................172 7.3. WESTINGHOUSE ELECTRIC CO..................................................................................................173 8. PUBLICATIONS............................................................................................................................174 ix x