Samuel E. Bays

HTGR Technology Family Assessment for a Range of Fuel Cycle Missions

This report examines how the HTGR technology family can provide options for the once through, modified open cycle (MOC), or full recycle fuel cycle strategies. The HTGR can serve all the fuel cycle missions that an LWR can; both are thermal reactors. Additional analyses are warranted to determine if HTGR “full recycle” service could provide improved consumption of transuranic (TRU) material than LWRs (as expected), to analyze the unique proliferation resistance issues associated with the “pebble bed” approach, and to further test and analyze methods to separate TRISO-coated fuel particles from graphite and/or to separate used HTGR fuel meat from its TRISO coating. The feasibility of these two separation issues is not in doubt, but further R&D could clarify and reduce the cost and enable options not adequately explored at present. The analyses here and the now-demonstrated higher fuel burnup tests (after the illustrative designs studied here) should enable future MOC and full recycle HTGR concepts to more rapidly consume TRU, thereby offering waste management advantages. Interest in “limited separation” or “minimum fuel treatment” separation approaches motivates study of impurity-tolerant fuel fabrication.

Thermal analysis of proposed transport cask for three advanced burner reactor used fuel assemblies

Abstract Preliminary studies of used fuel generated in the US Department of Energys Advanced Fuel Cycle Initiative have indicated that current used fuel transport casks may be insufficient for the transportation of said fuel. This work considers transport of three 5-year-cooled oxide advanced burner reactor used fuel assemblies with a burn-up of 160 MWD kg–1. A transport cask designed to carry these assemblies is proposed. This design employs a 7-cm-thick lead gamma shield and a 20-cm-thick NS-4-FR composite neutron shield. The temperature profile within the cask, from its centre to its exterior surface, is determined by two-dimensional computational fluid dynamics simulations of conduction, convection and radiation within the cask. Simulations are performed for a cask with a smooth external surface and various neutron shield thicknesses. Separate simulations are performed for a cask with a corrugated external surface and a neutron shield thickness that satisfies shielding constraints. Resulting temperature profiles indicate that a three-assembly cask with a smooth external surface will meet fuel cladding temperature requirements but will cause outer surface temperatures to exceed the regulatory limit. A cask with a corrugated external surface will not exceed the limits for both the fuel cladding and outer surface temperatures.

ATR Compendium: Irradiation Test Capabilities

Abstract The Advanced Test Reactor (ATR) celebrated 50 years of operation in 2017. Even after this much time, the general four-leaf clover design by Deslonde de Boisblanc is still deserving of the title “advanced.” This paper provides a high-level overview of the current irradiation capabilities of the ATR. The goal of this paper is to illustrate the types of irradiation facilities that are currently available within the ATR, the current irradiation missions that make use of these capabilities, and their connection to advancing nuclear technology.

An Axially Heterogeneous Sodium-Cooled Fast Reactor Designed to Transmute Minor Actinides

Abstract An axially heterogeneous sodium-cooled fast reactor design is developed for converting minor actinide waste isotopes into plutonium fuel. The reactor design incorporates zirconium hydride moderating rods in an axial blanket above the active core. The blanket design traps the active core’s axial leakage for the purpose of transmuting 241Am into 238Pu. This 238Pu is then co-recycled with the spent driver fuel to make new driver fuel. Because 238Pu is significantly more fissionable than 241Am in a fast neutron spectrum, the fissile worth of the initial minor actinide material is upgraded by its preconditioning via transmutation in the axial targets. Because the 241Am neutron capture worth is significantly greater in a moderated epithermal spectrum than the fast spectrum, the axial targets serve as a neutron trap that recovers some of the axial leakage lost by the active core. A low transuranic conversion ratio is achieved by a degree of core flattening that increases axial leakage. Unlike a traditional “pancake” design, neutron leakage is recovered by the axial target/blanket system. This heterogeneous core design is constrained to have sodium void and Doppler reactivity worth similar to that of an equivalent homogeneous design. Contrary to a homogeneous design, concentrating minor actinides (MAs) in an axial blanket mitigates the problem of above-threshold multiplication during sodium voiding. Because minor actinides are irradiated only once in the axial target region, elemental partitioning of the minor actinides from plutonium is not required. This fact enables the use of metal targets with pyroprocessing. After reprocessing, the target’s newly bred 238Pu and remaining unburned MAs become the feedstock for the next batch of driver fuel.

Impacts of Heterogeneous Recycle in Fast Reactors on Overall Fuel Cycle

A study in the United States has evaluated the attributes of the heterogeneous recycle approach for plutonium and minor actinide transmutation in fast reactor fuel cycles, with comparison to the homogeneous recycle approach, where pertinent. The work investigated the characteristics, advantages, and disadvantages of the approach in the overall fuel cycle, including reactor transmutation, systems and safety impacts, fuel separation and fabrication issues, and proliferation risk and transportation impacts. For this evaluation, data from previous and ongoing national studies on heterogeneous recycle were reviewed and synthesized. Where useful, information from international sources was included in the findings. The intent of the work was to provide a comprehensive assessment of the heterogeneous recycle approach at the current time.

Upgrading Limiting Peak-Power Analysis Techniques with Modern Validation and Uncertainty Quantification for the Advanced Test Reactor

Abstract This work supports the acceptability of the two-dimensional deterministic transport code HELIOS to replace the legacy diffusion code PDQ for computing the peak-power performance parameters of the Advanced Test Reactor (ATR). The 95% Confidence Rule, commonly used in the commercial reactor sector, is explored to develop the so-called reliability factors that provide statistical confidence that the peak-power limits within the hottest location along a fuel plate, referred to as the hot stripe, will not be exceeded. Additionally, an alternative “legacy” methodology was explored that attempts to mimic the exact PDQ analysis process used for defining the peak-power limits. The legacy methodology involves interpolating power between regions at azimuthal boundaries subtending the regions of interest. In order to apply the 95% Confidence Rule, a statistically significant calculation-to-measurement bias must first be established. Unlike the commercial world, where thousands of power observations can be collected every cycle using online flux-mapping instrumentation, the ATR power distribution must be measured during “depressurized” zero-power measurements using fission wires in polyethylene wands. In 2012, fission wire activation data were collected during a flux run in the Advanced Test Reactor Critical Facility. Also to improve statistical validity, archival data from ATR zero-power flux runs from 1977, 1986, and 1994 were digitized from scanned reports and used to create new benchmark models. Borrowing from least-squares adjustment methods commonly used for neutron activation spectroscopy, adjusted fission wire powers were calculated for all four data sets. The mean and standard deviation of the bias between a priori calculated and adjusted wire powers were then taken as the bias and uncertainty used in the 95% Confidence Rule.

Fuel Cycle System Analysis Implications of Sodium-Cooled Metal-Fueled Fast Reactor Transuranic Conversion Ratio

If advanced fuel cycles are to include a large number of fast reactors (FRs), what should be the transuranic (TRU) conversion ratio (CR)? The nuclear energy era started with the assumption that they should be breeder reactors (CR > 1), but the full range of possible CRs eventually received attention. For example, during the recent U.S. Global Nuclear Energy Partnership program, the proposal was burner reactors (CR < 1). Yet, more recently, Massachusetts Institute of Technology’s “Future of the Nuclear Fuel Cycle” proposed CR [approximately] 1. Meanwhile, the French company EDF remains focused on breeders. At least one of the reasons for the differences of approach is different fuel cycle objectives. To clarify matters, this paper analyzes the impact of TRU CR on many parameters relevant to fuel cycle systems and therefore spans a broad range of topic areas. The analyses are based on a FR physics parameter scan of TRU CR from 0 to [approximately]1.8 in a sodium-cooled metal-fueled FR (SMFR), in which the fuel from uranium-oxide-fueled light water reactors (LWRs) is recycled directly to FRs and FRs displace LWRs in the fleet. In this instance, the FRs are sodium cooled and metal fueled. Generally, it is assumed that all TRU elements are recycled, which maximizes uranium ore utilization for a given TRU CR and waste radiotoxicity reduction and is consistent with the assumption of used metal fuel separated by electrochemical means. In these analyses, the fuel burnup was constrained by imposing a neutron fluence limit to fuel cladding to the same constant value. This paper first presents static, time-independent measures of performance for the LWR [right arrow] FR fuel cycle, including mass, heat, gamma emission, radiotoxicity, and the two figures of merit for materials for weapon attractiveness developed by C. Bathke et al. No new fuel cycle will achieve a static equilibrium in the foreseeable future. Therefore, additional analyses are shown with dynamic, time-dependent measures of performance including uranium usage, TRU inventory, and radiotoxicity to evaluate the complex impacts of transition from the current uranium-fueled LWR system, and other more realistic impacts that may not be intuited from the time-independent steady-state conditions of the end-state fuel cycle. These analyses were performed using the Verifiable Fuel Cycle Simulation Model VISION. Compared with static calculations, dynamic results paint a different picture of option space and the urgency of starting a FR fleet. For example, in a static analysis, there is a sharp increase in uranium utilization as CR exceeds 1.0 (burner versus breeder). However, in dynamic analyses that examine uranium use over the next 1 to 2 centuries, behavior as CR crosses the 1.0 threshold is smooth, and other parameters such as the time required outside of reactors to recycle fuel become important. Overall, we find that there is no unambiguously superior value of TRU CR; preferences depend on the relative importance of different fuel cycle system objectives.

Analysis of Pu-Only Partitioning Strategies in LMFBR Fuel Cycles

Sodium cooled Fast Reactors (SFR) have been under consideration for production of electricity, fissile material production, and for destruction of transuranics for decades. The neutron economy of a SFR can be operated in one of two ways. One possibility is to operate the reactor in a transuranic burner mode which has been the focus of active R&D in the last 15 years. However, prior to that the focus was on breeding transuranics. This later mode of managing the neutron economy relies on ensuring the maximum fuel utilization possible in such a way as to maximize the amount of plutonium produced per unit of fission energy in the reactor core. The goal of maximizing plutonium production in this study is as fissile feed stock for the production of MOX fuel to be used in Light Water Reactors (LWR). Throughout the l970’s, this fuel cycle scenario was the focus of much research by the Atomic Energy Commission in the event that uranium supplies would be scarce. To date, there has been sufficient uranium to supply the once through nuclear fuel cycle. However, interest in a synergistic relationship Liquid Metal Fast Breeder Reactors (LMFBR) and a consumer LWR fleet persists, prompting this study. This study considered LMFBR concepts with varying additions of axial and radial reflectors. Three scenarios were considered in collaboration with a companion study on the LWR-MOX designs based on plutonium nuclide vectors produced by this study. The first scenario is a LMFBR providing fissile material to make MOX fuel where the MOX part of the fuel cycle is operated in a once-through-then-out mode. The second scenario is the same as the first but with the MOX part of the fuel cycle multi-recycling its own plutonium with LMFBR being used for the make-up feed. In these first two scenarios, plutonium partitioning from the minor actinides (MA) was assumed. Also, the plutonium management strategy of the LMFBR ensured that only the high fissile purity plutonium bred from blankets was sold to the MOX LWRs. The third scenario considered a LMFBR fuel cycle in an expansionary mode where excess bred transuranic material is accumulated for spinning off additional LMFBR cores. In this latter scenario, no plutonium partitioning was considered. After every cycle, transuranic from both driver and blankets is sold to the MOX LWRs. The MA production from LMFBR operated in a Pu-only fuel cycle is roughly only 1% that of the transuranic production rate. This is in contrast to LWR fuel cycles where the MA content in TRU is closer to 10% or more. If such a LMFBR were operated to provide fissile material to a fleet of MOX reactors, then 1 GWe of LMFBR could support between approximately 0.11 and 0.43 GWe of LWR-MOX reactors for a LMFBR conversion ratio between 1.1 and 1.5, if the MOX reactors were operated in a once-through-then out mode. If the plutonium is continuously recycled in the MOX reactors then the support ratio is approximately 1 GWe of LMFBR for between 0.13 and 0.65 GWe of LWR-MOX reactors depending on the LMFBR conversion ratio. Also, it was found that if the LMFBR fleet were operated in a purely expansionary mode, the smallest doubling time achievable would be seven years.

Multi-Reactor Transmutation Analysis Utility (MRTAU,alpha1): Verification

Multi-Reactor Transmutation Utility (MRTAU) is a general depletion/decay algorithm under development at INL to support quick assessment of off-normal fuel cycle scenarios of similar nature to well studied reactor and fuel cycle concepts for which isotopic and cross-section data exists. MRTAU has been used in the past for scoping calculations to determine actinide composition evolution over the course of multiple recycles in Light Water Reactor Mixed Oxide and Sodium cooled Fast Reactor. In these applications, various actinide partitioning scenarios of interest were considered. The code has recently been expanded to include fission product generation, depletion and isotopic evolution over multiple recycles. The capability was added to investigate potential partial separations and/or limited recycling technologies such as Melt-Refining, AIROX, DUPIC or other fuel recycle technology where the recycled fuel stream is not completely decontaminated of fission products prior to being re-irradiated in a subsequent reactor pass. This report documents the codes solution methodology and algorithm as well as its solution accuracy compared to the SCALE6.0 software suite.

Reactor Physics Characterization of Transmutation Targeting Options in a Sodium Fast Reactor

In sodium fast reactor designs, the fuel related inherent negative reactivity feedback is accomplished mainly through parasitic capture in U-238. However for an efficient minor actinide burning system, it is desirable to reduce or eliminate U-238 entirely to suppress further transuranic actinide generation. Consequently, reactivity feedback is accomplished by enhancing axial neutron streaming during a loss of coolant void situation. This is done by flattening “pancake” the active core geometry. Flattening the reactor also increases axial leakage which removes neutrons that could otherwise be used to destroy minor actinides. Therefore, it is important to tailor the neutron spectrum in the core for optimized feedback and minor actinide destruction simultaneously by using minor actinide and fission product targets.The threshold-fission (fertile) nature of Am-241 is used to destroy this minor actinide by capitalizing upon neutron capture instead of fission within a sodium fast reactor. This neutron-capture and its subsequent decay chain leads to the breeding of even mass number plutonium isotopes. A slightly moderated target design is proposed for breeding plutonium in an axial blanket located above the active “fast reactor” driver fuel region. A parametric study on the core height and fuel pin diameter-to-pitch ratio is used to explore the reactor and fuel cycle aspects of this design. This study resulted in both a non-flattened and a pancake core geometry. Both of these designs demonstrated a high capacity for removing americium from the fuel cycle. A reactivity coefficient analysis revealed that this heterogeneous design will have comparable safety aspects to a homogeneous reactor of the same size.