Fuel Cycle Costs Research Articles

Mechanistic prediction of Westinghouse TRITON11® BWR fuel critical power with MEFISTO-T subchannel analysis code

The TRITON11® fuel design is the latest Boiling Water Reactor (BWR) fuel product developed by Westinghouse, based on an 11 × 11 optimized fuel rod lattice, including mixing vane spacer grids, three large water rods and 18 part-length rods. The design offers larger fuel cycle cost saving, improved fuel reliability and increased thermal margin over previous Westinghouse fuel products. The critical power performances of the TRITON11 fuel design were assessed at the Westinghouse thermal–hydraulic FRIGG loop using a full-scale test bundle covering a wide range of BWR core conditions (covering normal operation and Anticipated Operational Occurrences) with various radial and two axial power distributions. The resulting FRIGG steady state database was simulated with Westinghouse subchannel analysis code MEFISTO-T based on a two-phase three-field approach of annular two-phase flow, accounting for the drop deposition enhancement provided by the spacer grids. A new model of local film entrainment was introduced due to the liquid “scrapping off” effects provided by the frame and side vanes of the spacer grids along the fuel channel and water rods. After spacer grid effect calibration, the steady state critical power is simulated mechanistically by power iterations up to complete local film dryout. The MEFISTO-T code can successfully predict the critical power performance of the Westinghouse TRITON11 BWR fuel design, including on part-length rods, with almost no bias and trend, within a standard deviation of about 5 %.

Scale effects on core design, fuel costs, and spent fuel volume of pressurized water reactors

The desire to improve the economic competitiveness and deployment pace of nuclear energy through modularization, manufacturing, and series production had led to the development of smaller size reactors. As the standard 17x17 fuel technology is mainly maintained in the pressurized water reactors (PWRs) category, this translates into a lower number of fuel assemblies in the core and sometimes a reduced fuel height. To assess the impact of such scale change in core design on fuel cycle cost and spent fuel volume, a scoping analysis tool is developed based on infinite lattice calculations, leakage, fuel management reduced models, and levelized unit cost of electricity (LCOE) estimate. As such, cost dynamics driven by fuel specific power, burnup, core leakage, feed, cycle length, fuel assembly height as well as uranium market data are captured with consistent set of assumptions and analysis methods. A selection of 5 reactor designs representative of leading PWR developers is assessed and compared. Pursuing higher specific powers and optimal burnups are highlighted as the main fuel cost reduction drivers, nevertheless, practical limitations and opportunities must be evaluated to establish the feasibility of such enhanced fuel operation. In consequence, a detailed core design is performed using SIMULATE3 code for 5 PWR variations including natural and forced coolant circulation modes, two reactor scales, power densities of 73, 112, and 123 kW/l and higher discharge burnups. Design and optimization are performed at the lattice level, for the reflector, and at the core loading level. Satisfactory steady-state operation including power distribution, coolant operating limits, and reactivity requirements are analyzed and reported in this paper. The fuel economics of the detailed core designs confirm the scoping analysis findings. Despite the unlocked power uprates in small PWRs, the achievable burnup for a given fuel specific power requires more enrichment and shorter fuel height results in higher fabrication costs per mass of fuel, which makes scaling down core size a more expensive endeavor on the fuel cycle front. Spent fuel volumes are reported for the PWRs designed in this paper. These volumes are driven by the core average discharge burnup regardless of the scale in consideration. Additional cost and core performance aspects related to heavy reflector gains, fuel-reflector substitution, and disposal cost policy in the U.S. are examined.

Physics-informed Reinforcement Learning optimization of PWR core loading pattern

The nuclear sector represents the largest single source of carbon-free energy in the United States. Nevertheless, keeping existing nuclear assets competitive on the grid is key to their ability to continue to provide dispatchable clean energy along side renewable generation. Optimizing the fuel cycle cost through the optimization of core loading patterns is one approach to addressing this lack of competitiveness. However, this optimization task involves multiple objectives and constraints, resulting in a vast number of candidate solutions that cannot be explicitly solved. While stochastic optimization (SO) methodologies are utilized by various nuclear utilities and vendors for fuel cycle reload design, manual design remains the preferred approach. To advance the state-of-the-art in core reload patterns, we have developed methods based on Deep Reinforcement Learning for single and multi-objective optimization. Previous research has laid the groundwork for these approaches and demonstrated its ability to discover high-quality patterns within a reasonable timeframe. Moreover, RL methods have shown superiority against SO in the multi-objective settings. In this paper, we introduce an innovative paradigm by incorporating multi-objective RL, interpretable AI, and physics knowledge to further improve the performance of the algorithms. When the problem increases in complexity, the classical single-objective sometimes failed to even found feasible solutions. Leveraging physic-information becomes pivotal to find high-quality and more diverse solutions faster. These findings motivate incorporation of more physics to further aid in the quest of surpassing human intelligence. Future work will focus on scaling this method and addressing fuel performance and control rod limits to demonstrate that these new designs can be considered for a real reactor.

METAL: Methodology for liquid metal fast reactor core economic design and fuel loading pattern optimization

The design of Liquid Metal-cooled Fast Reactor (LMFR) cores encompass two levels of design. The first is the physical core design, which is concerned with the design of the fuel assembly geometry in the context of a full core. The second is the fuel loading design, which is an in-core fuel management problem concerned with the design of fuel enrichment and core loading pattern. In this paper, an optimized fuel design search is developed to improve core loading patterns. As part of the implementation, the multiphysics simulation suite LUPINE has been enhanced to allow for fuel shuffles and an in-core fuel management optimization methodology has been added with the objective of reducing the fuel cost. The design methodology is referred to as Methodology for Economical opTimization of Applied LMFR (METAL), and this methodology is demonstrated using the popular Super Power Reactor Innovative Small Modular (SPRISM) sodium-cooled fast reactor (SFR) as a reference core. One challenge currently facing the deployment of LMFRs is the availability of TRansUranium (TRU) and high assay low enriched uranium (HALEU) driver fuel. The two forms of fuel are expensive and not widely available, which poses a challenge on the startup of LMFR cores. To address the availability of driver fuel, and to lower fuel costs, low enriched uranium (LEU) blankets are investigated as replacements to the natural uranium or depleted uranium blankets typically used. The advantage of this solution is the wide availability of LEU and its ability to lower the amount of TRU or HALEU driver fuel needed. The design objectives, constraints, and optimization algorithm for a SFR are identified. Then, METAL is applied to design a SFR core fueled by TRU driver fuel assemblies and LEU blanket assemblies. To demonstrate the advantage of introducing LEU blankets, METAL is also used to design another SFR core following the same objectives and constraints, but using naturally enriched blankets. By comparing the two cores, it was found that the introduction of LEU blankets results in a ∼28% reduction in the driver fuel mass requirements. Upon development of a levelized fuel cycle cost (LFCC) model, the 28% reduction in driver fuel mass corresponds to a 10% decrease in the LFCC. Additional advantages of using LEU blankets include reducing the core conversion ratio (CR), and improving the assembly radial power peaking factor (RPPF), which enhances the safety and non-proliferation performances of the SFR core.

Fuel cycle cost comparison between lead and sodium cooled fast reactors

Sodium and lead coolants are the most recognized coolants for Liquid Metal-cooled Fast Reactor (LMFR). The two coolants have significantly different physical, thermal and chemical properties, which results in major core design differences between lead- and sodium-cooled systems. This work attempts to quantify the impact of the coolant type on the fuel cycle cost of a LMFR by comparing the Levelized Fuel Cycle Cost (LFCC).The paper is composed of two studies with the goal of comparing the advantage of the superior neutronic performance of lead versus the advantage of the tight lattice that can be achieved with sodium coolant. In the first study, the influence of the neutronic differences between lead and sodium on the fuel cycle cost are isolated and quantified. This is achieved by comparing the fuel cycle performance of a Lead-cooled Fast Reactor (LFR) core model to that of a Non-optimized SFR (NSFR) model. The two models have a similar geometry, fuel loading, fuel type, and structural materials. The only major difference between the two cores was the coolant (i.e. lead vs. sodium). Based on this comparison, it is shown that the better neutronic performance of lead coolant results in longer cycle length for the LFR core. It is also shown that this difference in cycle length can result in up to 30.8% better LFCC for a lead-cooled system compared to a sodium-cooled one. In the second study, a Sodium-cooled Fast Reactor (SFR) is designed that is optimized for the sodium coolant. The design process for the Long-life core Sodium Fast Reactor (LSFR) realizes the compatibility of sodium coolant with structural materials and allows for a more compact design. The design parameters and constraints of the design are presented, and the objective of the design is to reduce the LFCC. In this study, the design with the sodium coolant offers a 2% reduction in the LFCC, along with a smaller core size.The overall results show that if you compare sodium and lead using the same reactor geometry, the neutronic benefits of lead coolant can lead to a 30% reduction in the LFCC. However, if you optimize the geometry to take advantage of the higher power density allowed with sodium coolant, the sodium coolant can offer a 2% LFCC reduction or a smaller core size.

A Design and Optimization Methodology for Liquid Metal Fast Reactors

A liquid metal fast reactor (LMFR) design and optimization methodology (DOM) has been developed. The methodology effectively explores a search space by initially sampling the search space, excluding invalid design samples prior to performing expensive multiphysics analysis, and then performing local searches of the design space. The design samples are evaluated using the multiphysics capabilities of the LUPINE LMFR simulation suite. Two studies have been performed to demonstrate DOM. First, the Westinghouse long-life core lead fast reactor (WLFR) is optimized. This reactor is 950 MW th and fueled with uranium nitride (UN) fuel which has a natural nitrogen isotopic abundance. The objective of the optimization is the reduction of the levelized fuel cycle cost (LFCC) while complying with the design constraints. Considering the challenges associated with using natural nitrogen in nitride fuel, a second study was performed to design a competitive 15N-enriched UN-fueled long-life core LFR. Based on this design, the cost of the 15N enrichment process necessary to achieve a competitive LFCC was calculated.

Ab Initio Evaluation of Uranium CarbideS(α,β) and Thermal Neutron Cross Sections

Uranium Carbide (UC) is a nuclear fuel material which offers better neutron economy and lower fuel-cycle costs compared to conventional mixed-oxide fuels. UC’s lattice binding and dynamical properties impact thermal neutron scattering and low temperature epithermal resonance absorption. The Thermal Scattering Law (TSL) describes the scattering system available energy and momentum transfer states. There is no TSL evaluation for UC in the ENDF/B-VIII.0 database; herein,ab-initiolattice dynamics (AILD) techniques are invoked to calculate the phonon spectrum for UC using spin-orbit-coupling density functional theory (DFT). The TSLs, inelastic and elastic thermal scattering cross sections for Uranium238 and Carbon12 in UC, respectively, are calculated inFLASSHfor use in higher fidelity reactor design calculations.

I2S-LWR Top-Down Differential Economics Evaluation Approach

The Integral Inherently Safe Light Water Reactor (I2S–LWR) is a design concept of a large (~1000 MWe) light water reactor with integral primary circuit configuration. One of the key design features promoting inherent safety is implementation of an integral primary circuit configuration, which in turn requires a compact design of the core and primary circuit components. Assessments of the cost of I2S-LWR is an important aspect of the overall evaluation of the new reactor concept. There are several approaches to cost estimation and economics evaluation of the new nuclear power technologies. Frequently used guidelines rely on the Code of Accounts, originally developed in the U.S. Department of Energy (DOE) Energy Economics Data Base (EEDB) Program Code of Accounts, proposed as evaluation tool by C.R. Hudson, and further popularized in the guidelines for economic evaluation of bids, by The International Atomic Energy Agency (IAEA). The code of accounts allows to break down main costs (Total Capital Investment Cost, Fuel Cycle Cost, Operation and Maintenance) to individual systems and items. This work aims to implement and apply a top-down differential economics evaluation approach to the Code of Accounts based guidelines, to assess the costs of the I2S -LWR relative to a representative “standard” PWR. In this methodology, a representative PWR design was taken as a reference and the differential cost was estimated for each individual account based on the design difference (or similarity). Cost estimating techniques were applied to the accounts representing systems that differ from the ones of the reference PWR. In this manner, the cost of the common components cancels out, and the uncertainty in the estimate is reduced. While this preliminary evaluation yet needs to be completed, the indications so far are that the I2S-LWR LCOE will be economically competitive with a standard PWR.

Pre-conceptual high temperature gas-cooled microreactor design utilizing two-phase composite moderators. Part I: Microreactor design and reactor performance

In this work, an approach is demonstrated for designing a microreactor based on a High Temperature Gas-Cooled Reactor (HTGR) platform intended for use in remote regions or for emergency response. The pre-conceptual microreactor design generated in this study is based on the Idaho National Laboratory's High Temperature Gas-Cooled Test Reactor (HTGTR) point design. Major design changes to the HTGTR were made to transform the design to a microreactor based on constraints inspired by open literature information from the Office of Secretary of Defense (OSD)/Strategic Capabilities Office (SCO) Pele Program. The primary purpose of this study is to compare reactor performance by evaluating key figures of merit for systems employing two-phase composite moderators (magnesia matrix with either an entrained beryllium or hydride containing moderating phase) relative to a reference graphite moderated case. Design parameters were selected to maximize reactor cycle performance for a given microreactor design with additional performance metrics such as natural resource requirement and fuel cycle costs also calculated. Our results demonstrate that a microreactor design based on the HTGR can be achieved by implementing a high fuel density. By using two-phase composite beryllium- or hydride-based moderators instead of graphite, substantial improvements can be realized in the reactor cycle performance, natural resource utilization, and fuel cycle cost.

Core Materials for Sodium-Cooled Fast Reactors: Past to Present and Future Prospects

Abstract The Sodium-Cooled Fast Reactor (SFR) is the most promising and technologically evolved among the six nuclear reactor concepts selected by the Generation IV International Forum. Although it is compatible with a closed fuel cycle for sustainable energy production, its competitiveness depends on being able to achieve high fuel burn-up up to 200 GWd/t. This would enable efficient fuel utilization to minimize fuel cycle costs. Hence, development of optimized core materials, especially for fuel cladding tubes that are subjected to extreme conditions of intense fast-spectrum neutron irradiation, high temperatures, and mechanical and chemical fuel–cladding interactions, continues to be a priority worldwide. Resistance to void swelling, irradiation creep, and embrittlement are required to be enhanced to minimize dimensional changes and loss of ductility in core components, which limit achievable burn-up. Current SFR core materials comprise austenitic stainless steels (SS) and ferritic-martensitic (FM) steels. Cold-worked austenitic SS grades 304, 316, 316Ti, niobium-stabilized grades FV548, EI-847, EP-172, and Ti-modified 15Cr-15Ni SS, such as D9, 1.4970, 15/15Ti SS, and ChS-68, have been used for fuel cladding and ducts in SFRs built to date, and they withstand neutron irradiation damage up to 80–100 dpa. Subsequent improvements have been made by optimizing minor elements (titanium, silicon, and phosphorous) and multi-stabilization to develop advanced grades such as Si-modified 15-15Ti, Indian Fast Reactor Advanced Clad (IFAC-1), PNC-316, EK-164, and HT-UPS, which could support burn-up of 150 GWd/t. In the case of FM steels, several commercial grades are found suitable in view of their inherent void swelling resistance to 200 dpa, and are being developed for improved creep strength by oxide dispersion strengthening to realize burn-up of 200–250 GWd/t. This article presents the development of structural materials for SFR fuel pin cladding and ducts and associated enhancement of permissible fuel burn-up. Indian and international experience resulting from extensive in-pile and out-of-pile mechanical testing and fuel–cladding interactions has been covered.

Techno-Economic Assessment of Fuel Cycle Facility of System Integrated Modular Advanced Reactor (SMART)

The economic assessment of advanced nuclear power reactors is very important, specifically during the early stages of concept design. Therefore, a study was performed to calculate the total cost estimation of fuel cycle supply for a system modular advanced reactor (SMART) by using the Generation-IV economic program called G4-ECONS (Generation 4 Excel-based Calculation of Nuclear Systems). In this study, the detailed description of each model and methodology are presented including facility, operations, construction matrix, post-production model, and fuel cycle cost estimation model. Based on these models, six Generation-III+ and Generation-IV nuclear reactors were simulated, namely System 80+ with benchmark data, System 80+ with uranium oxide (UOx) and mixed oxide (MOx) fuel assemblies, fast reactor, PBMR (Pebble Bed Modular Reactor), and PWR (Pressurized Water Reactor), with partially closed and benchmarked cases. The total levelized costs of these reactors were obtained, and it was observed that PBMR showed the lowest cost. The research was extended to work on the SMART reactor to calculate the total levelized fuel cycle cost, capital cost, capital component cost, fraction of capital spent, and sine curve spent pattern. To date, no work is being reported to calculate these parameters for the SMART reactor. It was observed that SMART is the most cost-effective reactor system among other proven and advanced pressurized water-based reactor systems. The main objective of the research is to verify and validate the G4-ECONS model to be used for other innovative nuclear reactors.

Chemical spectral shift control method for VVER-1000 LEU fuel assembly benchmark

The spectral shift control (SSC) is a promising method to decrease the fuel cycle cost and enhance the utilization of fuel resources (U, Pu) in nuclear reactors. In the chemical SSC method, reactivity control is accomplished by varying the ratio of heavy water to light water moderator (D2O/H2O). In the current work, we apply the chemical SSC method to analyze the VVER-1000 LEU fuel assembly from the OECD benchmark. Additionally, we compare the chemical SSC method to the conventional poison method presented in the benchmark. The results showed that the discharge burnup was improved by 60% and the conversion ratio value reached 0.83. The k∞ was suppressed at the beginning of cycle by −8% compared to the benchmark model. Furthermore, the dedication of 12 burnable absorber rods with 4.0 wt% Gd2O3 would save the D2O quantity during the adjustment process resulting in reducing the capital cost of the SSC method. However, it’s found that the physical changes associated with applying the SSC method including a high contribution of 239Pu and 241Pu to power generation, the low effect of 135Xe and 149Sm on the reactivity system, and reduction of control rod worth should be considered in the safety analysis of the SSC design.

Transition to thorium fuel cycle on a heavy water moderated molten salt reactor by using low enrichment uranium

Lacking fissile element (U-233) in nature is a key issue faced by Th-U fuel cycle, and existing available fuels should be taken as starting fuel to convert Th-232 to U-233 to realize transition to thorium fuel cycle. Low enrichment uranium (LEU, U-235 enrichment < 20%) is the most accessible fuel, and has been widely studied as starting fuel to realize transition to Th-U fuel cycle in graphite moderated molten salt reactors. Due to an inevitable introduction of large amount of fertile element U-238, however, this transition way has proven to be almost impossible. Heavy water moderated molten salt reactor (HWMSR) is an innovative molten salt reactor adopting heavy water as the moderator and heavy metal elements dissolved in fluoride salt as the fuel. A high neutron economy resulting from heavy water paves the way for the transition from LEU to thorium fuel cycle. The results obtained from this study indicated that transition can be achieved by using LEU with enrichment no lower than 5%. Raising enrichment would also enhance the breeding performance, but exert negative/positive influences on the attractiveness of uranium/plutonium contained in the fuel salt, decrease the heavy metal utilization efficiency and impose a small impact on the fuel cycle cost.

Comparison and analysis of different selection strategies of genetic algorithms for fuel reloading optimization of Thorium-based HTGRs

The nuclear fuel cycle cost can be effectively reduced through fuel reloading optimization. Genetic algorithm (GA) is a classic optimization algorithm that is widely applied in fuel reloading optimization. In the GA, selection is a key operator. However, few studies have compared and analyzed different selection strategies. In this study, 1/6 core of thorium-based block-type high temperature gas-cooled reactor was considered as an example, and ten different selection strategies were compared and analyzed. Five of these strategies were the roulette wheel and proportionate selection, tournament selection, uniform sorting, exponential sorting, and deterministic selection, whereas the other five were the aforementioned selection strategies combined with the truncation selection strategy. These ten different selection strategies were evaluated for single-objective and multi-objective problems. In single-objective optimization problems, the effective neutron multiplication factor was selected as the only optimization objective, whereas in multi-objective optimization problems, the effective neutron multiplication factor and power peak factor were considered as optimization objectives. The results indicated that exponential sorting was the best selection strategy for single-objective optimization problems, whereas hybrid truncation exponential sorting was the best selection strategy for multi-objective optimization problems.

Optimization under uncertainty for robust fuel cycle analyses

SummaryOptimization of nuclear fuel cycles is essential for experts and policy makers for studying and analyzing the future of the nuclear energy. In the case of advanced electronuclear transition scenarios, multiple parameters with a complex dependence have to be fine‐tuned in order to achieve a set of predefined objectives. However, in the presence of uncertainties the solutions obtained in this way may not be stable since small perturbations could break the delicate balance between different parts of the scenario. In this work, the optimization of an uncertainty European‐based sustainable transition scenario has been studied. This scenario, which has been analyzed with the TR_EVOL nuclear fuel cycle simulator system, is aimed at reducing the transuranic inventory masses while keeping the fuel cycle costs. To that end, an extension of the DEMO evolutionary multiobjective algorithm has been implemented within TR_EVOL for allowing the inclusion of constraints and uncertainties with a methodology that can be used by any fuel cycle simulator. Results show the importance of coupling optimization and uncertainty analyses due to the suboptimal and unstable solutions that can be obtained if not considered jointly. In addition, the uncertainties shrink the decision space. It was found that in their presence the transuranic mass can be reduced and stabilized by a factor ranging between 65% and 71% with an increase of the cost of 16% and 18.5% after 300 years of operation by using advanced systems when compared with an open fuel cycle strategy.

Reactor performance and safety characteristics of two-phase composite moderator concepts for modular high temperature gas cooled reactors

Graphite moderators have an extensive historical performance record, but also feature inherent challenges for modular High Temperature Gas-Cooled Reactors (mHTGRs). Challenges with graphite include non-uniform expansion and contraction under irradiation and build-up of potential energy during the bombardment of high energy neutrons that results in a large energy release under annealing. These challenges have led to the investigation and development of alternative moderators to be utilized in mHTGRs, including beryllium- and hydride-based concepts with compositions selected for favorable moderating power and the potential for improved in-service lifetime as compared to graphite. The proposed moderators are fabricated as two-phase composites with magnesium oxide, MgO, as the radiation-stable host matrix and beryllium metal, Be, beryllium oxide, BeO, or zirconium hydride, ZrHx=1 (to account for hydrogen loss from the hydride phase during processing), as the entrained moderating phase. Here, we evaluate the reactor performance and safety characteristics of these moderator concepts relative to a graphite reference using a Ft. Saint Vrain-style fuel block. We assessed the cycle length, discharge burnup, natural resource utilization, neutron flux spectra, moderating power, moderating ratio, critical size, moderator and fuel temperature feedback, fuel cycle cost, spent nuclear fuel and high level waste radioactivity per unit energy generated, and environmental impact per unit energy generated. The results demonstrate that the advanced moderators have the potential for comparable or enhanced cycle performance to that of the graphite reference case with significantly improved performance for an optimized moderator-to-fuel ratio design. These advanced moderators are also assessed from a reactor safety standpoint for Design Basis Accidents (DBAs) including Pressurized Loss of Forced Cooling and Depressurized Loss of Forced Cooling accidents for a 350 megawatt thermal prismatic-type mHTGR. The full core thermohydraulic analysis of DBAs show that the high volumetric heat capacity of the beryllium-based moderator grants them a greater margin to fuel failure in these analyses than a conventional graphite moderated system, but the lower thermal conductivity of the beryllium-based moderators leads to longer times at elevated temperatures.

Simulating IRIS assembly containing erbium oxide as a burnable poison

In traditional pressurized water reactors fuelled with uranium oxide (UOX), extended burnup can be achieved through the use of a high uranium enrichment and high percentage of gadolinium as burnable poison. Such a high enrichment of fuel and gadolinium would lead to increasing fuel cycle costs. Consequently, searching for another burnable poison that can achieve a long core cycle and at the same time may be mixed with nuclear fuel with lower concentration was one of the main interests of nuclear power plants in recent years. Erbium is one of the burnable poisons that can satisfy these goals. The potential interest of this poison (Er-167) is that it presents a relatively low absorption cross section in the thermal range and a non-negligible resonance integral. This will lead to very long cycles owing to its slow consumption in the reactor core. The main objective of this work is to investigate the effect of mixing erbium oxide with uranium oxide on some neutronic safety parameters of IRIS (International Reactor, Innovative and Secure) assembly. Among these parameters are the effective multiplication factor, the thermal flux, the total flux, the residual reactivity penalty, the radial power and the axial power distributions, the fuel temperature coefficient, the fuel temperature coefficient and peaking factor.Three assembly models are simulated using MCNPX 2.7 code; the first model uses uranium oxide fuel with no Erbium (4.95 % U-235), while the other two models use Erbium with different concentrations (1% Er2O3 and 2.2 % Er2O3). The obtained results for k-effective showed that the suggested loadings of the Erbium bearing pins can sustain very long burnup cycles up to 60 GW d/t. It is also noted that the assemblies containing erbium have higher amounts of U-235 at 60 GW d/t (the end of life). Furthermore, erbium models present a higher spectral index than that of the free Erbium model. Additionally, it is observed that there is an increase in the pin power peaking factors for erbium assemblies. Finally, FTCs and MTCs are more negative for assemblies bearing erbium especially for the model containing 2.2 % Er. This guarantees the safe operation of the reactor.

Fully ceramic microencapsulated fuel in prismatic high-temperature gas-cooled reactors: Design basis accidents and fuel cycle cost

Fully ceramic microencapsulated (FCM) fuel may enhance the already strong inherent reactor safety characteristics of a conventional High-Temperature Gas-Cooled Reactor (HTGR). FCM fuel uses an SiC matrix that exhibits higher stability under irradiation with limited swelling relative to a conventional graphite matrix. Additionally, the SiC matrix is expected to exhibit improved mechanical performance relative to graphite. The feasibility of FCM-fueled HTGRs has been explored in a previous study (Lu et al., 2018), and three FCM fuel concepts were identified to be able to maintain the cycle length of the reference conventional General Atomics 350 MWt prismatic modular HTGR.In this work, the previous investigation of the reactor core safety characteristics of HTGRs fueled by these three FCM fuels was significantly extended to detailed thermal hydraulics and neutronics analysis of normal operation and design basis accident scenarios. We found that these FCM-fueled cores would have a maximum fuel temperature ∼20–40 K higher than that of the reference HTGR core at the beginning of cycle (BOC) and ∼30–60 K higher at the end of cycle (EOC) during normal operating conditions. Pressurized (P-) and depressurized (D-) loss of forced cooling (LOFC) accidents as well as the control rod withdrawal accident were explored. The FCM-fueled cores would at most have a maximum fuel temperature ∼30–50 K higher than that of the reference core in a P-LOFC accident, ∼10–30 K in a D-LOFC accident and ∼30–50 K in a rod withdrawal accident. The conclusion of the study is that FCM fuel has a small impact on the reactor performance and safety characteristics during normal operation and design basis accident conditions within the existing and generic 350 MWt prismatic modular HTGR design we considered.Additionally, a fuel cycle cost comparison was also performed in the current study. We found that the FCM fuels may be up to 40%–74% more expensive than the conventional fuel. Further analysis of the economics of FCM fuel are needed, because the uncertainties in the assumptions for this study are high.

Regional Cooperation for Nuclear Spent Fuel Management in East Asia: Costs, Benefits and Challenges – Part II

ABSTRACT East Asia is home to some of the most dynamic economies on earth, but also a locus of current and historical national and international conflicts. Some of the largest economies lack domestic energy resources, and nuclear power has been adopted as a perceived key to energy security. Lacking, however, is a concerted strategy for managing nuclear spent fuel at the regional or even, for the most part, national levels. Regional coopera tion on spent fuel management offers many benefits, as well as challenges. This two-part paper explores four different “scenarios” of regional spent-fuel management, ranging from a national “go-it-alone” option to regional cooperation on uranium supply, enrichment and “back-end” spent fuel management. Results for physical flows of nuclear materials and other inputs and outputs, and for the relative costs of each scenario, are presented. Results suggest that the costs of scenarios that include reprocessing are higher than those without reprocessing, and costs for increased dry-cask storage (reducing the amount of fuel stored in high-density spent fuel pools) are likely to be a tiny part of overall nuclear fuel cycle costs. As a result, radiological risk and attendant political, social and legal concerns should drive decisions regarding spent fuel management, not costs.

Development of methodology for efficient fuel design evaluation of the Advanced High Temperature Reactor (AHTR)

This paper presents a methodology for developing a neutronics model to be used for a fluoride-salt high temperature reactor (FHR) fuel design optimization study. The reactor design is based on the Advanced High-Temperature Reactor (AHTR) with hexagonal fuel element with fuel “planks” (plates) developed at the Oak Ridge National Laboratory. The overall objective of this research was to develop a methodology accurate enough, yet practical and fast enough for a fuel plank design optimum search that minimizes the fuel cycle cost (FCC) of FHR. Resonance self-shielding calculations for double heterogeneous fuel planks of AHTR design is currently unavailable in most codes. Therefore, automated modeling of the double heterogeneity was implemented to allow accurate use of multi-group simulations. The methodology generates MCDancoff Factors using SCALE6.2; its adequate accuracy was verified. A MCDancoff Factor surrogate model was created so on-the-fly MCDancoff Factors could be provided for depletion calculations over the design space for parametric studies. A simple non-linear reactivity model with assembly-to-whole-core reactivity correction was applied to analyze performance of multi-batch refueling. Results indicate that cycle length is maximized at 40% packing fraction with a carbon-to-heavy metal ratio (CHM) of approximately 250, which minimizes the cost of outages. In contrast, discharge burnup and fuel utilization is maximized at around 20% packing fraction and a CHM of approximately 600, which minimizes fuel cost, but not necessarily the FCC cost, which also accounts for outage penalty. Void and power reactivity coefficients were determined over the design phase space to establish the envelope for stable design. The minimum FCC is the optimum trade-off between the two, within the limits of the stability envelope. In combination with the cost model the objective function is non-linear and requires global optimization methods. This paper presents the methodology needed for global optimization, while the optimization itself will be presented in a follow up companion paper.