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AbstractMicroreactors are expected to be deployed in locations or situations where standard macrogrid electricity is not readily or reliably available. Locations such as remote mining communities, military installations, and other similar communities may seek to decarbonize their electricity with a microreactor. Traditional remote communities typically pay much higher energy prices than nonremote communities, so there may be a higher degree of support for microreactors despite higher expected (per unit capacity) capital costs. There is also interest in deploying microreactors on sites that utilize their own microgrids, for example, federal installations or universities, in different locations across the United States. To model the deployment prospects of microreactors on microgrids, HOMER Pro was utilized to generate optimized microgrid configurations under an assumption of net zero, i.e. with no fossil fuels permitted. The economic performance of microreactors on microgrids was analyzed in a variety of locations throughout the United States. It was found that the minimal capital cost required for entry of a microreactor into the system was significantly higher in locations with both low solar resource (less than ~4.5kW∙h/m2∙day−1) and (in particular) low wind (<5 m/s average wind velocity at 80-m hub height), roughly corresponding to locations somewhat inland in both the eastern and the western United States. Under the financial assumptions made in this study, depending on location, the capital cost at which microreactors could enter the system varied from around $24 000/kW(electric) in areas of high wind and/or solar resource to over $90 000/kW(electric) in Alaska, with a range of intermediate results. It was found that the minimum allowed power level of the microreactor did not have a substantial impact on results. A utility presence was found to have a major impact on the financial performance of a microreactor, especially in locations that have strong renewable resources. [ABSTRACT FROM AUTHOR]

Increasing the number of pins within a pressurized water reactor (PWR) assembly reduces pin temperature for a given assembly power. In conjunction with a core retrofit, this presents a potential route to PWR uprate, which is of growing interest given recent increases in electricity prices. However, most PWRs utilize regular lattice designs with fixed guide tube positions, such as the very common 17 × 17 lattice design with 25 guide/instrumentation tubes. These tubes are aligned with penetrations in the reactor pressure vessel, which presents a prohibitive obstacle to retrofit, and more widely, may "lock" many PWRs to this particular fuel configuration. In this paper, an irregular PWR fuel assembly is proposed. It is shown that a backward-compatible lattice with 324 fuel pins per assembly (BL324), uniform enrichment, and the same hydrogen-to–heavy metal ratio as a reference 17 × 17 assembly with 264 fuel pins can achieve within-assembly power peaking within 2% of the reference assembly under equivalent conditions while fixing the guide tube positions. Power peaking can be further reduced to reach that of the existing fuel assembly by reducing the enrichment of 36 of the pins by 0.2 wt%. The fuel assembly could potentially either support a significant uprate of up to ~20% in conjunction with low-enriched uranium plus (LEU+) fuel or a more aggressive cycle design, and hence, improved discharge burnup at the same power and batch strategy. A subchannel analysis shows that the coolant heat-up distribution is comparable to the reference assembly. However, the pressure drop is estimated to be 4% higher, which would challenge the performance of transition cores containing both 17 × 17s and BL324s. Further incremental changes to BL324 may be attractive, either to improve manufacturability or to slightly improve performance through formal optimization. [ABSTRACT FROM AUTHOR]

Variable renewable energy availability has increased the volatility in energy prices in most markets. Nuclear power plants, with a large ratio of capital to variable costs, have historically operated as base load energy suppliers but the need for more flexible operation is increasing. We simulate the techno-economic performance of a 950 MW t nuclear power plant, based on the Westinghouse lead-cooled fast reactor, coupled with molten salt thermal storage as a method for flexible energy dispatch. We use the System Advisor Model to model the nuclear reactor thermal power input and power cycle operating modes. We combine this robust engineering model with a mixed-integer linear program model for optimized dispatch scheduling. We then simulate the coupled nuclear and thermal storage system under different market scenarios with varying price volatility. We find that the coupled plant outperforms the base plant under markets where energy price peaks fluctuate by a factor of two or more about the mean price. We show that a calculated power purchase agreement price for the plant improves by up to 10% when operating under California energy market conditions. Sensitivity analysis on the thermal storage cost shows that the optimal design remains unchanged even when doubling costs. [ABSTRACT FROM AUTHOR]

Considerable attention is currently directed towards space nuclear reactor systems with significantly higher power outputs, catering to applications like nuclear electric propulsion and fission surface power, in contrast to past historic designs, which were predominantly ≤ 0.1 MWe. In addition, many previous designs have used Highly Enriched Uranium (HEU), although there is a recent drive to consider only High Assay Low Enriched Uranium (HALEU) for non-proliferation reasons. This requirement in particular motivates a contemporary review. In this paper, we highlight the drivers for high-power outputs (≥ 1 MWe), provide a comprehensive overview of previously-proposed space fission power system technologies and outline the attractiveness of various technologies. Unlike previous reviews that focused on nuclear thermal propulsion or lower power nuclear electric systems with HEU fuel, this study focuses on higher power nuclear electric systems operating with HALEU and key design considerations. While the reactor core technologies (i.e. coolants, fuels, and core materials) are the focus of this paper, it is important to recognise the interplay between the reactor core, heat rejection, and power conversion systems. Therefore, a systematic and cross-cutting review of the underpinning technologies associated with such space reactor systems is performed, and the technical maturity of relevant technologies is assessed. The most promising technologies for high-power HALEU designs are outlined and important areas for research are also identified. [ABSTRACT FROM AUTHOR]

As energy systems worldwide change to mitigate against climate change, the sustainability of electricity production methods is becoming increasingly relevant. Nuclear energy is constrained by the need for the efficient conversion of uranium to electricity to define itself as sustainable. The thermodynamic measure of available energy of work – exergy – is used to analyse nuclear fuel cycle processes by accounting for material and energy flows over whole conversion chains. This study proposes that a more comprehensive definition of nuclear exergy is needed to account for the additional exergy released in nuclear decay. This allows the extension of nuclear exergy analysis to the back end of the nuclear fuel cycle and novel decay heat applications. Furthermore, nuclear decay exergy is shown to add approximately 10% to previous definitions of the exergy of fissile isotopes which accounts for the additional energy released by fission products in decay. To aid future research an evaluation of this nuclear decay exergy is performed for 3000+ nuclear isotopes. The new definition of nuclear exergy is employed in investigating the sustainability of advanced nuclear fuel cycles. This analysis uses quantitative exergy efficiency metrics and Sankey diagrams to represent certain sustainability trade-offs numerically and visually. • Nuclear exergy can be used to analyse nuclear fuel cycle sustainability. • Trade-offs in advanced nuclear cycle selection can be shown with Sankey diagrams. • Nuclear decay exergy increases the previous value of nuclear exergy. • A computation of nuclear decay exergy is performed for 3000 isotopes. [ABSTRACT FROM AUTHOR]

The article offers information on the main attributes and evolutionary developments of Generation III and III+ reactor designs that are under consideration to improve safety, increase plant life and optimise performance in Great Britain in 2016.

• A concept for a telescopic control rod containing ∼5 concentric annuli is proposed. • This could significantly reduce silo depth and provide substantial cost savings. • The telescopic control rod can achieve greater than 90% of the worth of a typical control rod. • A minor redesign of the control rod or core can compensate for any deficit in worth. • The rod can be constructed using materials already qualified for HTR environments. High temperature gas-cooled reactors (HTRs) are intentionally tall and thin to enable passive decay heat removal through conduction, with a core height of 8 m being typical. Units are typically embedded in the ground, which has numerous safety advantages but requires excavating a deep silo, the cost of which grows super-linearly with depth. HTRs contain rigid control rods that add additional space above the core and make the silo deeper. A concept for a telescopic control rod containing ∼ 5 concentric annuli is proposed. Depending on the design details of the HTR, this has the potential to significantly reduce the silo depth and hence provide substantial reactor capital cost savings. Neutronic analysis demonstrates that a telescopic control rod consisting of 5 concentric annuli can achieve greater than 90% of the control rod worth of a typical articulated HTR control rod without further compensatory measures. The addition of a more control rods or a minor redesign of the rod can recover the deficit in worth if necessary to ensure shutdown margin requirements are met. The telescopic control rod will deplete faster than the reference control rod, which may necessitate more frequent replacement, but this is not anticipated to be a serious drawback as HTR control rods are not inserted extensively at power (especially in pebble bed reactors). The telescopic control rod can be constructed using materials already qualified for HTR environments. Further work is required to demonstrate comprehensively mechanical design, reactor physics and materials feasibility, and the necessary steps are identified in the paper. For mechanical design, a failure modes and effects analysis is required to identify possible failure mechanisms and their consequences, so that they can be designed out or mitigated. The friction/wear resistance of mechancial components must also be evalauted through experimentation on a prototype control rod in a high temperature helium environment. The arrangement of the upper head must also be considered. For reactor physics, the differential control rod worth must be investigated. Finally, a cost benefit analysis should be performed. [ABSTRACT FROM AUTHOR]

Integrated energy systems can improve flexibility on future energy grids with one option being Nuclear-Solar hybrid systems. Integrating solar generated heat from parabolic troughs into the feedwater line allows the plant to alleviate turbine bled steam and transiently power boost 15% above nominal power for a nominal small modular PWR cycle. This article presents a parametric study of the design of such a system and provides a full system dynamic model written in the Modelica language to analyse the dispatch in transient load following operation. The control of the system is presented, and the transient analysis is shown to help inform sizing of the concrete storage and parabolic trough arrays. The trade off in design between improved degrees of power boosting and system nominal efficiency is investigated with the work suggesting that higher steam generator entry temperatures offer improved opportunities of up to 42% for power boosting flexibility. • An open-source model of the full proposed energy system is written in Modelica. • The model allows a high time fidelity simulation of all key system variables. • Solar feedwater heating allows power boosting without changing reactor conditions. • Transient analysis informs sizing the concrete thermal storage and solar array. [ABSTRACT FROM AUTHOR]

Nuclear power is typically deployed as a baseload generator. Increased penetration of variable renewables motivates combining nuclear and renewable technologies into Integrated Energy Systems (IES) to improve dispatchability, component synergies and, through cogeneration, address multiple markets. However, combining multiple energy resources heavily depends on the proper selection of each system's location and design limitations. In this paper, co-siting options for IES that couple nuclear and concentrating solar power (CSP) with thermal desalination are investigated. A comprehensive siting analysis is performed that utilizes global information survey data to determine possible co-siting options for nuclear and solar thermal generation in the United States. Viable co-siting options are distributed across the Southwestern U.S., with the greatest concentration of siting options in the southern Great Plains, although siting with higher solar direct normal irradiance is possible in other states such as Arizona and New Mexico. Brackish water desalination is also attractive across the southwest U.S. due to high water stress, but for brackish water desalination reverse osmosis (an electricity driven process) is most cost- and energy-efficient, which does not require co-siting with the thermal generator. The most attractive state for nuclear and thermal desalination (which is more attractive when using seawater) is Texas, although other areas may become attractive as water stress increases over the coming decades. Co-siting of all CSP and thermal desalination is challenging as attractive CSP sites are not coastal. [ABSTRACT FROM AUTHOR]

The NuScale small modular reactor (SMR) has been modeled using the Virtual Environment for Reactor Applications multiphysics environment and the results compared with the publicly reported data in the Design Certification Application. The results show an excellent agreement for the compared axial and radial power distributions, temperature coefficients of reactivity, boron and control rod worths, and fast neutron flux. This NuScale model is then used to investigate the effect of different operational modes on reactor components to determine how the flexible load-following operation may affect control rod and reactor pressure vessel (RPV) lifetimes. The control rod degradation is confirmed to primarily affect the silver-indium-cadmium rod tip. The degradation rate is observed to follow a nonlinear function of core power level where the increase in degradation decreases with insertion depth. For the variation in core power levels expected with current load-following schemes, the total control rod degradation is found to be mild, at 5% to 10% of usable life per cycle for a reactor operating at <80% power. Nonetheless, this enables load-following strategies to be confirmed and/or modified to ensure that control rods do not need to be replaced during the 60+ year life of the reactor. The RPV degradation was found to be almost directly proportional to the core power level and was not overly sensitive to flux shape perturbations. Future work is planned using these damage functions to optimize operation over multiple NuScale SMR units and to develop strategies for prognostics and health management. [ABSTRACT FROM AUTHOR]

The relevance of nuclear power plant flexible power operation (FPO) is rising due to increased penetration of variable renewables. Small Modular Reactors (SMRs) such as NuScale are envisaged to have multiple units on one site. This provides opportunities for coordinating FPO between units, but also introduces challenges as common services (e.g., refueling equipment) must be shared between units and fuel loadings may be standardized. It is therefore important to quantify whether FPO with SMRs is economically beneficial. This paper quantifies the economics of FPO on different timescales to analyze refueling outages of such multi-unit SMRs. A 24-h price-taker profit maximization of SMR operation is first solved considering revenues from wholesale power and ancillary service (AS) markets. The results are then used to find the most profitable operation strategy over several years for multiple units at the site accounting for the physics and materials limits on NPP operation, including refueling outages. Results show a small but appreciable participation of nuclear into AS, contributing 4% of revenues. Due to the longer fuel cycle, over a decade of operation the units' refueling outages drifted by 3 months, ultimately leading to refueling during summer. • We develop a model for multi-unit small modular nuclear reactor operation. • Mixed integer linear programming is used to optimize for revenue. • Short and long timescale models are used in combination. • It is found that flexible power operation is economically beneficial. • Outage scheduling drifts over time due to longer fuel cycle. [ABSTRACT FROM AUTHOR]

A feasibility study of a subcritical fission-fusion hybrid reactor using lead-lithium eutectic as a coolant and minor actinides (MAs) as fuel is presented. Such a reactor could support the fission community by transmuting MAs and the fusion community by breeding tritium. The feasibility of such a reactor for the burnup of MAs is assessed in terms of burnup performance, tritium breeding, and safety characteristics. Tandem mirrors are a promising neutron source technology, and a deuterium-tritium tandem mirror is considered here for the neutron source with power Psource = 1.13 MW assumed for scoping purposes. Subcritical reactivities from keff = 0.9800 to keff = 0.9950 were considered, representing the initial reference for subcritical reactivity and the assessed upper limit, respectively. Stability analyses indicated the reactivity would be stable under perturbations of fuel, coolant, and inlet temperatures, with a positive reactivity insertion expected during reactor shutdown. This k e f f range corresponded to nuclear heating values of 150 to 650 MW and mass burn rates of 53 to 216 kg/year. The upper mass burn rate limit would require 1110 reactor years with a capacity factor of 0.9 to fission the global supply of MAs and could offset the annual U.S. MA production with eight reactors. Tritium breeding was assessed for keff = 0.9800 and 3.795% 6Li enrichment in the coolant, and a tritium breeding ratio of 1.602 ± 0.017 was tallied, suggesting the reactor could, without elevated 6Li enrichment, produce tritium to both sustain operation and supply tritium for other fusion devices. Time-series modeling of fuel burnup was conducted for a four-batch loading scheme and three different fuel residence times at keff = 0.9800, which showed system performance would drop with burnup, and that the rate of this drop was lower for longer fuel residence times, motivating a means of reactivity control. Last, changes in fuel composition with burnup were assessed for relative concentrations of MAs, transmutation products, and fission products. The breeding of plutonium in the blanket was calculated and found to be of minimal proliferation concern. [ABSTRACT FROM AUTHOR]

• Material model for FeCrAl implemented in MELCOR-2.2 with increased detail. • Recovered PWR LOCA scenario analyzed for high burnup ATF and reference zircaloy case. • For scenario analyzed, additional coping time from FeCrAl is important. • Increased burnup increases fission product inventory. • Lower fission product release with FeCrAl compared to zircaloy maintained at high burnup. It has been established that incremental improvements in beyond design basis accident performance can be achieved through accident tolerant fuel (ATF). However, they have the potential to recover margin with respect to conventional fuel and therefore enhance plant economics through uprate or cycle length increase. To realize this potential, it is necessary to quantify the reduction in source term due to use of ATF, and correspondingly how this is affected by increasing the cycle length and/or burnup. This requires development of a risk informed analysis methodology for ATF under high burnup conditions, which is being developed within LWRS. To this end, a MELCOR simulation of FeCrAl ATF in a recovered Large Break LOCA (LBLOCA) scenario has been developed, using a model based on the Zion Pressurized Water Reactor (PWR). The new user defined material capability, along with the inclusion of detailed neutronics- and depletion-derived parameters such as core power distribution, decay heat behavior, and fission product inventories, allows a more detailed simulation of FeCrAl clad material properties and behavior than has previously been possible using MELCOR. These detailed FeCrAl results were compared with a zircaloy clad model to investigate the differences between the two clad types and quantify the benefits of FeCrAl ATF with an 18-month cycle. Next, the fuel cycle with ATF was extended to 24 months, to determine whether any additional safety margin provided by FeCrAl ATF could be leveraged to implement high burnup FeCrAl-clad fuel while retaining the same operating and safety limits as current zircaloy-clad fuel. For the particular scenario analyzed, the delay to fuel failure from using ATF was of the same order as the LOCA recovery time, and hence significant in reducing fission product release, with the 24-month FeCrAl core performing better than the 18-month zircaloy core. It is noted that for other transients, the reduction in release due to using FeCrAl may be less significant. The material model developed here can be used in such further studies in support of determining the overall source term. [ABSTRACT FROM AUTHOR]

This work provides fully reproducible benchmark models of the Systems for Nuclear Auxiliary Power (SNAP) 8 Experimental Reactor. Validation benchmarks of the criticality configuration experiments under dry conditions with no coolant are presented. In addition, relevant reactivity worth experiments, and measurement of power distributions are validated and presented here. Discrepancies between modeled and experimental results are at most 110 pcm for fuel and poison worths, and critical configurations that do not manipulate control elements are within 50 pcm. Larger discrepancies found in reflector element worth experiments are due to experimental methodology which was noted as being simplified due to limited calculational capabilities at the time, while discrepancy in the control element manipulated critical configuration is due the absence of structural components for modeling simplification. Power distributions closely follow what was observed in experiment with expected peaking in reflecting elements. All models are well documented with cited references describing and justifying assumptions, simplifications, and adjustments to reproduce the models with any nuclear safety codes; model inputs and outputs are stored in the snapReactors GitHub repository. [ABSTRACT FROM AUTHOR]