Thesis: Ph. D., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2017. / This electronic version was submitted by the student author. The certified thesis is available in the Institute Archives and Special Collections. / Cataloged from student-submitted PDF version of thesis. / Includes bibliographical references (pages 197-208). / The restructuring of the power sector that began in the 1990s created acute competition for nuclear power plants. While merchant reactors benefited from this change in the early years, recently they started to retire for economic reasons well before the expiration of their operating licenses. This unexpected wave of premature shutdowns has severe implications for energy and climate policy. It invites us to re-assess the viability and role of nuclear in a transitioning energy sector. The thesis first develops two new tools aimed at measuring nuclear competitiveness and informing retirement strategies: (1) a structural model of electricity markets based on supply and demand equilibrium and (2) a long-term asset valuation framework accounting for stochastic price dynamics and flexible retirement options. We employ these tools to analyze the challenges facing nuclear energy in two countries: the United States and Japan. After evaluating the drivers and likelihood of premature retirements, we discuss a range of technological innovations and regulatory options that could help nuclear bring value to future competitive markets. We show that low natural gas prices and stagnant electricity demand have been responsible for the drop in nuclear plant revenue in the United States. We measure that renewable wind and solar PV impact nuclear operations only for penetration levels above 15% and 30% respectively. We also find that spot price volatility, a feature of competitive markets, defers nuclear retirement decisions rather than precipitates them. In this context, nuclear must adapt. Greater operational flexibility can prevent financial losses in areas where renewables are being deployed on a large scale. In the medium-term, heat storage technologies would protect plants' profitability while enabling deep decarbonization of the energy sector. Finally, a few plants may be able to reach niche markets by diversifying their output beyond electricity. We recognize that the carbon-free attributes of nuclear energy are not valued in competitive markets. Yet, even a moderate price on carbon would save most reactors. If not possible, states may adopt nuclear subsidies to meet their policy objectives. As a last resort, the exercise of a new mothballing license could prevent the irreversible loss of nuclear assets. / by Geoffrey Haratyk. / Ph. D.

Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2006. / Includes bibliographical references (p. 105-107). / Sobering electricity supply and demand projections, coupled with the current volatility of energy prices, have underscored the seriousness of the challenges which lay ahead for the utility industry. This research addresses the impending global need for electricity through the development of advanced annular fuel designs with both internal and external cooling which can achieve higher power densities and hence, higher electricity output from the same basic reactor vessel and containment. Therefore the objectives of this project are to determine the optimal geometrical design parameters of an annular fuel assembly for both PWRs and BWRs for the purpose of achieving maximum power density. It is theorized that utility companies can utilize this design through either retrofitting of their existing reactor facilities or incorporation of the fuel design into new plant concepts. For the case of annular fuel for PWRs, a high performance uranium nitride fuel assembly concept capable of achieving a 50% higher power density was successfully developed. It is shown that a 5% enriched UN annular-fuel assembly can operate at 150% power density for about 50 effective-full-power-days more than that of the nominal 17xl7 solid-fuel-pin assembly operating at 100% power density. Furthermore, neutronic simulation times of this assembly was reduced from approximately 2 days per simulation for a Monte Carlo based analysis to approximately 2 minutes for a deterministic based simulation via the development of an appropriate correction factor for the CASMO-4 neutron transport code. It was shown that a 25% increase in U238 number density for the un-poisoned pins and a 35% increase for the 10 weight percent gadolinium nitride poisoned pins produced the optimal plutonium tracking and infinite multiplication factor simulation. / (cont.) Finally, the 13x13 annular fuel assembly was shown to have a smaller reactivity swing over the fuel lifetime. Thus it was concluded that an annular uranium nitride assembly at 150% power density can be designed for PWRs so as not to require enrichments above 5% in order to reach the desirable cycle length of 18 months. For the case of annular fuel for BWRs, thermal hydraulic simulations were carried out for a 9x9 solid fuel reference assembly and three different annular assemblies with 5x5, 6x6 and 7x7 fuel pin geometries. Prior research had utilized the Hench-Gillis CPR correlation for all thermal hydraulic simulations and determined that as much as an 11% uprate for 5x5 annular geometries and an 18% uprate for 6x6 annular geometries might be achievable. However, since Hench-Gillis uses bundle average conditions for its calculations, it was theorized that this treatment was not appropriate for annular fuel. A benchmarking analysis against experimental critical power data for a 9x9 assembly confirmed this is a more appropriate heat balance correlation, the EPRI-1 Reddy Fighetti, which was adopted in our simulation of the critical power using the subchannel analysis code VIPRE. Several different strategies were pursued in order to improve the minimum critical heat flux ratio of the three different annular fuel assemblies including optimization of the fuel pin dimensions, fuel pin gap, and orifice loss coefficients. However it was concluded that annular fuel is not a promising strategy for increasing the power density. This can be due to the fact that the CHFR margin gained from the increase in heat transfer surface area is being lost due to the need for increased flow velocity, which retards the CHF for BWR conditions. This is exacerbated by the inability for the coolant in the inner channels to mix with the surrounding subchannels. / by Tyler Shawn Ellis. / S.M.

Thesis (S.B.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2008. / Includes bibliographical references (p. 46-47). / U.S. nuclear weapons force structure accounts for the number and types of strategic and nonstrategic weapon systems in various locations that comprise the nuclear arsenal. While exact numbers, locations, and detailed designs remain classified, motivations for the current and future of the nuclear arsenal is presented as a unique integration of logical technical and political information. The dynamic that results from military requirements, physical design limitations, and congressional response to balance deterrence with stockpile reductions has not produced the necessary level of change in the post-Cold War environment of the 21st century. As such, a stagnant position on nuclear weapons reductions diminishes the effect of U.S. global nonproliferation efforts. / by Amanda Baker. / S.B.

Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, February 2007. / Includes bibliographical references (p. 138-140). / An evolutionary BWR fuel assembly design was studied as a means to increase the power density of current and future BWR cores. The new assembly concept is based on replacing four traditional assemblies and large water gap regions with a single large assembly. The traditional BWR cylindrical UO2-fuelled Zr-clad fuel pin design is retained, but the pins are arranged on a 22x22 square lattice. There are 384 fuel pins with 9.6 mm diameter within a large assembly. Twenty-five water rods with 27 mm diameter maintain the moderating power and accommodate as many finger-type control rods. The total number and positions of the control rod drive mechanisms are not changed, so existing BWRs can be retrofitted with the new fuel assembly. The technical characteristics of the large fuel assembly were evaluated through a systematic comparison with a traditional 9x9 fuel assembly. The pressure, inlet subcooling and average exit quality of the new core were kept equal to the reference values. Thus the power uprate is accommodated by an increase of the core mass flow rate. The findings are as follows: - VIPRE subchannel analysis suggests that, due to its higher fuel to coolant heat transfer area and coolant flow area, the large assembly can operate at a power density 20% higher than the traditional assembly while maintaining the same margin to dryout. - CASMO 2D neutronic analysis indicates that the large assembly can sustain an 18-month irradiation cycle (at uprated power) with 3-batch refueling, <5wt% enrichment with <60 MWD/kg average discharge burnup. Also, the void and fuel temperature reactivity coefficients are both negative and close to those of the traditional BWR core. - The susceptibility of the large assembly core to thermalhydraulic/neutronic oscillations of the density-wave type was explored with an in-house code. / (cont.) It was found that, while well within regulatory limits, the flow oscillation decay ratio of the large assembly core is higher than that of the traditional assembly core. The higher core wide decay ratio of the large assembly core is due to its somewhat higher (more negative) void reactivity coefficient. The pressure drop in the uprated core is 17 %Vo higher than in the reference core, and the flow is 20% higher; therefore, larger pumps will be needed. FRAPCON analysis suggests that the thermo-mechanical performance (e.g., fuel temperature, fission gas release, hoop stress and strain, clad oxidation) of the fuel pins in the large assembly is similar to that of the reference assembly fuel pins. A conceptual mechanical design of the large fuel assembly and its supporting structure was developed. It was found that the water rods and lower tie plate can be used as the main structural element of the assembly, with horizontal support being provided by the top fuel guide plate and core plate assembly, and vertical support being provided by the fuel support duct, which also supports the finger-type control rods. / by Aydin Karahan. / S.M.

Thesis (S.M. and S.B.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2012. / Cataloged from PDF version of thesis. / Includes bibliographical references (p. 119-122). / The effects of orientation angle, subcooling, heat flux, mass flux, and pressure on bubble growth and detachment in subcooled flow boiling were studied using a high-speed video camera in conjunction with a two-phase flow loop that can accommodate a wide range of flow conditions. Specifically, orientation angles of 0' (downward-facing horizontal), 30°, 45°, 60°, and 90° (vertical); mass flux values of 250, 300, 350, and 400 kg/m²s, with corresponding Froude numbers in the range of 0.42 to 1.06; pressures of 101 (atmospheric), 202, and 505 kPa; two values of subcooling (10°C to 20°C); and two heat fluxes (0.05 to 0.10 MW/m²) were explored. The combination of the test section design, high-speed video camera, and LED lighting results in high accuracy (order of 20 microns) in the determination of bubble departure diameter. The data indicate that bubble departure diameter increases with increasing heat flux, decreasing mass flux, decreasing levels of subcooling, and decreasing pressure. Also, bubble departure diameter increases with decreasing orientation angle, i.e. the largest bubbles are found to detach from a downward-facing horizontal surface. The mechanistic bubble departure model of Klausner et al. and its recent modification by Yun et al. were found to correctly predict all the observed parametric trends, but with large average errors and standard deviation: 35.7+/-24.3% for Klausner's and 16.6±11.6% for Yun's. Since the cube of the bubble departure diameter is used in subcooled flow boiling heat transfer models, such large errors are clearly unacceptable, and underscore the need for more accurate bubble departure diameter models to be used in CFD. / by Rosemary M. Sugrue. / S.M.and S.B.

Thesis: Ph. D., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2016. / This electronic version was submitted by the student author. The certified thesis is available in the Institute Archives and Special Collections. / Cataloged from student-submitted PDF version of thesis. / Includes bibliographical references (pages 139-145). / The recent development of REBCO superconducting tapes, cabling methods and joint concepts could be a revolutionary development for magnetic fusion. REBCO has significantly better performance at high magnetic fields than traditional low temperature superconductors (LTS), and can be operated at a higher temperature than LTS for reduced thermodynamic cost of cooling. Use of REBCO superconductors in the magnet systems of tokamaks allows building demountable toroidal field (TF) coils, greatly simplifying reactor construction and maintenance. A demountable TF coil system with REBCO superconductors for a fusion reactor has been conceptually designed. The coil system operates at 20 K, with a maximum magnetic field of 20 T. The magnets are divided into two coil segments and can be detached and remounted to allow the internal components of the reactor to be removed vertically as one piece. Operating at 20 T and 20 K, the stress in most of the coils is acceptable (less than 2/3 the yield strength and less than 1/2 the ultimate tensile strength of the structural materials). The strain in the superconductors is lower than the reversible degradation limit. The electrical resistance in each conductor joint is 10 n [Omega]. The total heat generation in the reactor superconducting TF magnets is approximately 1.9 MW, of which about 25 % is nuclear heating and 75 % joint heating. 71 MW of electricity are required for cooling the coils at 20 K, about 7 % of the electric energy the reactor generates. The expected time to warm-up the magnets from the operation temperature to room temperature is 7 days, and approximately the same for cool-down back to the operation temperature. The analysis of the conceptual magnet design is encouraging, as no insuperable problems have been identified. This conceptual design can be used as a starting point for a full engineering design of demountable fusion reactors magnets. / by Franco Julio Mangiarotti. / Ph. D.

Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2010. / Cataloged from PDF version of thesis. / Includes bibliographical references (p. 115-116). / Sodium cooled fast reactors (SFRs) are considered as a novel example to exercise the Technology Neutral Framework (TNF) proposed in NUREG- 1860. One reason for considering SFRs is that they have historically had a licensing problem due to postulated core disruptive accidents. Two SFR designs are considered, and both meet the goals of the TNF that LWRs typically would not. Considering these goals have been met, a method for improving economics is proposed where systems of low risk-importance are identified as candidates for removal, simplification, or removal from safety grade. Seismic risk dominates these designs and is found to be a limiting factor when applying the TNF. The contributions of this thesis are the following: 1. Functional event trees are developed as a tool to allow different designs to be compared on an equal basis. Functional event trees are useful within the TNF as a method for the selection of Licensing Basis Events (LBEs) which take the place of traditional Design Basis Accidents. 2. A new importance measure, Limit Exceedance Factor (LEF), is introduced that measures the margin in system failure probability. It can be used directly with the TNF where standard importance measures cannot. It also reveals that some systems that appear to be of high risk-importance with standard importance measures may have significant margin. 3. The seismic risk dominates these designs. It is shown that even under optimistic assumptions, the goals of the TNF cannot be met by a typical reactor. The effect of seismic isolation to reduce the frequency of seismically initiated large releases is also analyzed and found to be insufficient to reach the goals of the TNF. A limit on initiating event frequency that is consistent with current practices is proposed. / by Brian C. Johnson. / Ph.D.

Thesis: S.M., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2014. / Cataloged from PDF version of thesis. / Includes bibliographical references (pages 110-111). / The effect of sulfur on the corrosion fatigue crack growth of austenitic stainless steel was evaluated under Light Water Reactor (LWR) conditions of 288°C deaerated (less than 5ppb O₂) water, to shed light on the accelerating effect of the LWR environment and to explore the effect of high sulfur content on the retardation of fatigue crack growth rates. Fatigue tests were performed using a trapezoidal loading pattern with rise times of 5.1, 51, 510, and 5100 seconds (fall time of 0.9, 9, 90, and 900 seconds), with Kmzx of 28.6 or 31.9 MPa[mathematical symbol]m and stress ratios (R, Pmin/Pmax) of 0.4 or 0.7. Two test materials were used to evaluate the effect of sulfur: (1) a low sulfur (<0.0025 wt%) stainless steel and, (2) a high sulfur (0.032 wt% stainless steel. The low sulfur stainless steel exhibited increasing crack growth rates from 9.4 x10-5 mm/cycle to 1.2x1 0-⁴ mm/cycle as rise times were increased from 5.1 to 5100 seconds with a stress ratio of 0.7. The high sulfur stainless steel exhibited decreasing crack growth rates from 1.4 x10-⁴ mm/cycle to 7.9 x10-⁵ mm/cycle as rise times were increased for a stress ratio of 0.4, and crack growth rates from 6.4 x10 5 mm/cycle to 3.6 x10-⁵ mm/cycle with increasing rise time at a stress ratio of 0.7. Evaluation of the crack growth rates showed environmental enhancement of the crack growth rates for the low sulfur stainless steel, while the high sulfur stainless steel showed retardation of environmental crack growth rates, likely due to the increased corrosion at the crack tip associated with the high sulfur content. The crack surfaces were characterized using Scanning Electron Microscopy (SEM). The low sulfur material showed a light layer of corrosion product that decreased in thickness as the testing progressed, and faceting on the surface was highly crystallographic. Faceting ran both perpendicular and parallel to the crack for the short rise time steps of the test, but fewer perpendicular facets were evident at the longer rise times. The high sulfur material was heavily corroded throughout the fracture surface, and crystallographic faceting was seen for stages of the test with R=0.4 For R=0.7, the heavy oxidation on the surface made the facets hard to resolve. Striations were apparent during the 5100 second rise time for the low sulfur material (where corrosion was almost nonexistent) and throughout the entirety of the crack surface for the high sulfur material. Materials were also characterized by optical microscopy. The low sulfur material showed pitting along the grain boundaries, due to the boron concentration in this material, which resulted in boron precipitates, while the high sulfur material showed pitting throughout the surface, due to the MnS inclusions. Electrochemical tests were also performed at room temperature on both materials in pH 4 (using H₂SO₄), 7, and 10 (using NH₄OH). Peaks in the passive region of the high sulfur material were seen at potentials of 160, 630, and 1400 mVSHE, due to dissolution of the MnS inclusions. The results suggest that the high sulfur material provides an increase in corrosion when exposed to the environment, which leads to the retardation of crack growth rates at the longer rise times due to prolonged exposure of the crack tip to the environment. At low stress ratios, the proposed mechanism for retardation of crack growth rates is crack tip closure, due to a buildup of corrosion product at the fracture surface, which lowers the effective load that the crack tip experiences. At high stress ratios, the proposed mechanism for retardation is an increased in injected vacancies and enhanced creep, which disrupt the slip bands ahead of the crack tip, reducing the crack tip stresses. Fractography of the fracture surface and crack growth rate comparisons of the low and high sulfur material provide supportive evidence for the proposed mechanisms, and further work is proposed to examine the effect of increased corrosion ahead of the crack tip. / by Lindsay Beth O'Brien. / S.M.

Thesis: S.M., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2016. / Cataloged from student-submitted PDF version of thesis. / Includes bibliographical references (pages 71-75). / Disposal packages containing high heat generating spent nuclear fuels (SNF) require improved heat transfer to keep the peak cladding temperature from going above the tolerance limit. Filling the accessible void spaces between the container and the SNF with a high heat conducting metal is a potential solution. In metal casting, it is well known that a gap forms at the metal-mold interface due to solidification shrinkage and it significantly reduces heat transfer during cooling. This negative heat transfer effect is persistent for a disposal package since the filler stays in the container after solidification. The key to close the gap is to promote metallic bonding by minimizing the oxidation of the container during the required preheating stage of the void filling process. However, direct contact between the container and the molten filler can lead to the growth of intermetallic phases, which can embrittle the container. The contribution of this work is twofold. First, through a down-scaled experiment, it was shown that coating a steel container with Zn and using Zn or Zn-4wt.%Al as a filler and unidirectionally cooling the melt from the bottom successfully suppressed the formation of the gap. Closing the gap increased the effective thermal conductivity of the package by a factor of roughly 6 under the employed experimental conditions. Second, tests showed that using near eutectic Zn-Al and executing the filling process at a temperature below the melting point of Zn suppressed the growth of any intermetallic phases. Specifically, this prevents the growth of Fe-Zn intermetallic phases due to the sufficiently high composition of Al, and it inhibits the dissolution and diffusion of Fe from the container by extending the presence of the ZnO diffusion barrier, which delays the growth of the Fe-Al intermetallic phases. / by Yongsoo Park. / S.M.

Thesis: S.M., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2017. / This electronic version was submitted by the student author. The certified thesis is available in the Institute Archives and Special Collections. / "June 2017." Cataloged from student-submitted PDF version of thesis. / Includes bibliographical references (pages 47-48). / The evolution of fuel assembly, hot-spot formation, and nuclear burn in an Inertial Confinement Fusion (ICF) implosion at the National Ignition Facility (NIF) can be quantified through time-resolved measurements of the neutron spectrum. This information will be obtained with the next-generation Magnetic Recoil Spectrometer (MRSt) that will measure the neutron spectrum (12-16 MeV) with high accuracy (~5%), unprecedented energy resolution (~100 keV) and, for the first time ever, time resolution (~20 ps). To successfully implement the MRSt on the NIF for this measurement, the signal and background distributions at the MRSt detector must be characterized; the detector response to the signal and background must be determined; and the shielding enclosing MRSt must be designed and implemented to reduce the background to the required level. These things have been done, which constitute the main results of this thesis. First, an MCNP model of the MRSt in the NIF target bay was implemented to assess the neutron- and gamma-background fluxes at an unshielded MRSt. Second, models of the MRSt-detector response to the signal protons (or deuterons), and neutron and gamma background were implemented to assess the signal-to-background (S/B) for the unshielded MRSt case. Using these models, it is discussed in this thesis that the combined neutron and gamma background in the MRSt data needs to be reduced 100-400 times. Third, a shielding design, consisting of polyethylene, tungsten, and stainless steel, fully enclosing the MRSt, was developed to reduce the background to the required level. This design reduces the background 100-200 times, and meets the requirement of S/B > 5 for the down-scattered-neutron measurement. Obviously, this design depends on the MRSt-detector response to the signal and background, and some minor adjustments to the design might be applied depending on the results from the upcoming measurements of the MRSt-detector response to signal and background. As the shielding design depends on the engineering design of the MRSt system, which has not been fully defined yet, some adjustments to the design will most likely be made when the MRSt engineering design is finalized. / by Christopher William Wink. / S.M.