Search

Most high-temperature reactors use graphite as a moderator and structural material. This includes high-temperature gas-cooled reactors with helium cooling and TRi-structural ISOtropic (TRISO) fuel particles embedded in graphite, as well as fluoride salt–cooled high-temperature reactors with clean salt coolant and TRISO fuel particles embedded in graphite and thermal spectrum molten salt reactors with a graphite moderator and fuel dissolved in the salt. The largest volume radioactive waste stream from these reactors is the irradiated graphite. We describe herein a roadmap for management of these graphite wastes that contain radioactive 14C, tritium, and other radionuclides. There may be some graphite wastes with sufficiently low radioactivity levels that can be treated as nonradioactive waste and managed like other graphite waste. Management options for the graphite include (1) direct disposal, (2) recycled back to the reactor or other nuclear applications, and (3) oxidizing the graphite with release as an effluent or underground sequestration of the carbon dioxide. Cosequestration of this carbon dioxide with carbon dioxide from industrial, biological, and cement production processes can isotopically dilute the 14C before sequestration to eliminate the possibility of exceeding individual radiation exposure limits. We also describe options for processing graphite-matrix TRISO fuel, including separating the bulk graphite to reduce the volumes of used fuel for disposal or processing to recover fissile materials. The inventories of radioactive isotopes in different carbon wastes vary by many orders of magnitude; thus, there is no single economic option for the management of all graphite waste. [ABSTRACT FROM AUTHOR]

AbstractBecause of their high impact on calculated temperatures, the offset gas gaps between the graphite holder and capsule wall are evaluated for the thermal model of the Advanced Gas Reactor 5/6/7 (AGR-5/6/7) irradiation experiment. Fuel compact temperatures are a crucial factor in assessing the irradiation performance of tristructural isotropic (TRISO)-coated-particle fuel for use in high-temperature gas-cooled reactors. TRISO fuel irradiation performance is a core tenet of the U.S. Department of Energy’s Advanced Gas Reactor Fuel Development and Qualification Program, conducted in the Advanced Test Reactor at Idaho National Laboratory. Specifically, the irradiation experiments were intended to provide irradiation performance data to support fuel process development, qualify fuel for normal operating conditions, support the development of fuel performance models and codes, and provide irradiated fuel and materials for postirradiation examination and safety testing.The thermal model used to predict the fuel compact temperatures was a three-dimensional finite element model that was subject to uncertainty. The most dominant factor in the uncertainty pertaining to calculated fuel temperatures is the gas gap uncertainty, which is caused by the graphite holder nub-to-capsule wall clearance. In AGR-5/6/7 capsules, the graphite holders can be shifted away from being perfectly centered inside the stainless steel capsule walls. This is due to irradiation-induced graphite shrinkage or to insufficient nubs on the graphite holder leading to azimuthally varying gas gaps during irradiation. A method was developed to find the best offset magnitude and direction at the top and bottom of the graphite holders in Capsules 1 and 2, which experienced higher temperatures than originally predicted among the five AGR-5/6/7 axially dispersed capsules irradiated between 2018 and 2020. The differences between the calculated and the measured temperatures of thermocouples (TCs) placed in the graphite holder near the fuel compacts were used as the basis for the holder offset determination. The best-fit offset assumes that the root-mean-square error between the measured and calculated TC temperatures is a minimum value. Typical offsets in the range of 0.002 to 0.004 in. increase the fuel compact temperatures in the range of 100°C to 200°C when compared to the zero offset for Capsules 1 and 2. Additionally, because of irradiation-induced changes in the capsule geometries, the holder offsets change throughout the irradiation period. [ABSTRACT FROM AUTHOR]

AbstractHeat-pipe microreactors (HPMRs) are very small-scale nuclear reactors that employ heat pipes (HPs) for heat removal. HPMRs can be easily integrated with other forms of renewable energies, can be used for emergency responses to disaster relief zones, can be deployed in remote locations not connected to the grid, and can be removed from sites and replaced by new ones. HPMRs can also be used for space missions as HPs do not rely on gravity for heat transfer. Conventional fuel materials, such as uranium oxide (UO2) and uranium oxycarbide (UCO), are currently considered in most existing HPMR designs, but ceramic uranium nitride (UN) fuel that has high uranium density, high thermal conductivity, and high melting point may become a better fuel candidate. Through neutronics calculations, this paper assesses the impact of using UN fuel in HPMRs with two different neutron spectra (fast and thermal) and two different fuel forms [traditional solid fuel pellets and TRi-structural-ISOtropic (TRISO) fuel compacts]. It was concluded that retrofitting HPMRs with UN fuel has the potential to reduce the initial 235U enrichment requirement by ~3 wt% (to keep the same cycle length) or increase the cycle length (by keeping the same initial 235U enrichment), which enables more compact and transportable HPMR core designs. However, using UN fuel decreases the control element worth [by up to 20% for the Special Purpose Reactor (SPR) and 5% for HP-MR] and is up to 80% more costly. Increasing 15N enrichment can further decrease the initial 235U enrichment requirement and increase the control element worth but is more costly. Compared to fast-spectrum HPMRs fueled with solid pellet fuels, retrofitting UN fuel is more suitable for thermal-spectrum HPMRs fueled with TRISO fuel compacts, where the neutron spectrum hardening caused by using UN is less significant. [ABSTRACT FROM AUTHOR]

The purpose of this work is to characterize the morphological, structural, and strength properties of model prototypes of new-generation TRi-structural ISOtropic particle fuel (TRISO) designed for Generation IV high-temperature gas reactors (HTGR-type). The choice of model structures consisting of inner pyrolytic carbon (I-PyC), silicon carbide (SiC), and outer pyrolytic carbon (O-PyC) as objects of research is motivated by their potential use in creating a new generation of fuel for high-temperature nuclear reactors. To fully assess their full functional value, it is necessary to understand the mechanisms of resistance to external influences, including mechanical, as in the process of operation there may be external factors associated with deformation and leading to the destruction of the surface of fuel structures, which will critically affect the service life. The objective of these studies is to obtain new data on the fuel properties, as well as their resistance to external influences arising from mechanical friction. Such studies are necessary for further tests of this fuel on corrosion and irradiation resistance, as closely as possible to real conditions in the reactor. The research revealed that the study samples have a high degree of resistance to external mechanical influences, due to the high strength of the upper layer consisting of pyrolytic carbon. The presented results of the radiation resistance of TRISO fuel testify to the high resistance of the near-surface layer to high-dose irradiation. [ABSTRACT FROM AUTHOR]

High-temperature gas-cooled reactors present significant technical competitiveness in various fields, such as low-carbon electricity, thermal energy, and hydrogen supply. Silicon Carbide (SiC) matrix dispersed with TRi-structural ISOtropic (TRISO) particle fuel is characterized by high-temperature tolerance, radiation resistance, and good neutron economy, which has emerged as a pivotal technological pathway for reactor core design. With the rapid advancement of SiC Additive Manufacturing (AM) technology, there is a substantial increase in core design flexibility, and an innovative gas-cooled reactor design was proposed by Monte Carlo (MC) method in this paper. The Non-uniform Rational B-spline (NURBS), Latin Hypercube Sampling (LHS), second-order polynomial response surface, as well as Genetic Algorithm (GA) were coupled to realize the optimization design for the coolant channel. Besides, a modular fuel design was proposed, and the AM SiC shell was fabricated using Binder Jetting (BJ) technology. Furthermore, the densification of the AM SiC shell and the matrix with SiC powder and SiC microspheres (TRISO simulator) was investigated by Chemical Vapor Infiltration (CVI). Finally, the microscopic morphology analysis and thermal-mechanical performance characterization of the samples were conducted, which demonstrated the technical feasibility and advanced capabilities of the SiC matrix dispersed with TRISO fuel based on AM technology. • A modular helium-cooled reactor core was designed with innovative modules for space utilization improvement and heat transfer enhancement. • NURBS and Genetic Algorithm were used for the flow channel optimization. • Binder Jetting was applied for SiC shell, and the densification of SiC shell, powder, and microspheres were achieved by Chemical Vapor Infiltration. [ABSTRACT FROM AUTHOR]

• Irradiated kernels mainly consisted of UO 2 and UC(O 0.0x) phases. • Zr, Nb, Mo, Ru, Tc concentrate in uranium carbide phase. • Fission gas bubbles stay in uranium carbide phase. • Excess Mo, Ru, Tc form U(Mo, Tc)C 2 or URuC 2 , and Nb forms oxide shell at pore surface. [ABSTRACT FROM AUTHOR]

Tri-structural isotropic (TRISO) fuel particles are a key component in several previous and current reactors as well as in a variety of novel nuclear reactor designs. Interest in TRISO fuel is on the rise, necessitating considerable computer modeling of TRISO fuel behavior in order to support related design and licensing activities. The Bison nuclear fuel performance code, which offers a full set of capabilities for modeling TRISO fuels, makes it easier to explore the various important aspects of TRISO fuel behavior. One key advantage of Bison is its ability to create meshes in 1D, 2D, and 3D. Users can customize these meshes for specific geometries, mesh densities, and use cases. This enables a wide variety of analyses, including thermal, structural, mass diffusion, homogenization, and statistical failure analyses. Furthermore, the meshing capability simplifies analysts' workflows. The inherent mesh generation capability eliminates the need for separate mesh-generating software and mesh file management. Also, the fact that the meshes are customizable makes it straightforward to automate an investigation over a range of geometric parameters or mesh densities. The present paper highlights the ease with which Bison may be used to create meshes for both simple and relatively complex TRISO fuel particles, and it explores the types of analyses enabled by these meshes. • The Bison code provides 1D, 2D, and 3D meshing capability for TRISO particle fuels. • Input commands are shared, and consistent mesh feature names are used. • 1D meshes form the basis of nearly all studies since they run efficiently. • 2D meshes are necessary for aspherical meshes. • 3D meshes are required for computing homogenized properties. [ABSTRACT FROM AUTHOR]

The SCALE code system was used to model, deplete, and compare several different tri-structural isotropic (TRISO)–fueled reactor designs: a helium-cooled prismatic reactor, a helium-cooled pebble-bed reactor (PBR), and a fluoride-lithium-beryllium (FLIBE) molten-salt-cooled PBR. The purpose of this comparison was to understand how differences in the reactor designs affect the radioactivity of the fuel after discharge and whether those differences are significant. First, the various reactor designs were built and depleted in the TRITON module for each design and fuel enrichment. Then, the TRITON outputs were used to create burnup-dependent reactor libraries. These libraries were then used by ORIGEN to determine the activities of discharged fuel for each reactor, which were compared to generic Westinghouse 17 × 17 fuel. Overall, the results showed that short-term activities are dominated by reactors with higher operating powers, and the reactor type, initial fuel enrichment, and maximum burnup are of only secondary importance. Although this analysis only focuses on activities in Becquerels, these dependencies are consistent with the expected behavior of decay heat. However, analysis of long-term time periods post irradiation shows that the reactor type and maximum burnup have strong impacts on the activities; initial fuel enrichment has a secondary impact while operating power is inconsequential. These results would be useful for analyses, such as dose assessment and modeling in postrelease scenarios, normal fuel handling operations, and spent fuel transport, storage, and disposal. Of particular interest, the results in this technical note show that analyses that focus on spent nuclear fuel of advanced reactors need to consider each parameter carefully. Unsurprisingly, if the correct operating power is not used in short-term analyses, the results will not be correct. Perhaps unexpectedly, however, if the correct reactor type is not used, then the long-term results will also be incorrect, especially for areas such as permanent disposal. Even though this technical note focuses on the total activity of nuclear fuel, it provides initial results on the effects of various input parameters and also provides a framework to extend the work into other analyses of spent fuel from advanced reactors, especially those employing TRISO fuel. [ABSTRACT FROM AUTHOR]

The random distribution of TRISO fuel particles in the fuel elements may form local hotspots, thereby increasing the thermal stress on TRISO particle layers that may cause particle failures. Optimization of the distribution of TRISO fuel particles can significantly decrease the peak temperature, reducing the stress, and thus improve the safety and economy of the reactor. In this study, two optimization methods, namely octahedral and icosahedral-based segmentation optimization methods, were introduced to decrease the maximum temperature of the pebble fuel with packing fractions from 5% to 40%. ANSYS 19.1 was used to calculate the temperature distribution of pebble fuel elements with optimized and random TRISO particle distributions. The results showed that the optimization for the distribution of TRISO particles dispersed in the pebble fuel element is essential and feasible. The necessity and effectivity of optimization increase as packing fraction increases. The peak temperatures of the fuel pebbles with optimized TRISO particle distributions were significantly decreased in comparison with the ones with randomly distributed TRISO particle. The icosahedral-based optimization method is recommended among the two kinds of optimization methods because the effectivity of the icosahedral-based optimization method is higher than the octahedral-based ones and the icosahedral-based optimization method can also get a higher target packing fraction.

Summary: The Fixed Bed Nuclear Reactor (FBNR) is a modular reactor for small energy needs (70‐120 MWel). FBNR is pressurized water reactor, but uses GEN‐IV reactors fuel. UO2 kernels of 500 μm diameter with 25 μm thick SiC cladding are embedded in graphite matrix form the spherical fuel elements with a diameter of 15 mm and covered by 300 μm thick Zircaloy‐4 cladding. FBNR has high level safety features and operates without On‐site Refueling. Approximately 1.62 million spherical fuel elements are suspended in a core height of 200 cm and inner and outer diameters of 32 cm and 224 cm by upwards flowing light water coolant with 280°C entry and 340°C exit temperatures. The unit lattice geometry is dodecahedron with fuel element in center and surrounded with light water coolant‐moderator.In the present work, 9% enriched UO2 fuel is used. Neutronic analysis has been performed with the MCNPX‐2.7.0 code. As, it is not practical to tackle millions of fuel elements separately, at first run unit fuel cell calculations are performed in spherical geometry for one single lattice consisting of separate fuel, cladding and volume‐equivalent moderator regions. In the second run, the fuel cell is homogenized. Infinite cell calculations have been executed for variable moderator/fuel (H/235U) ratios in order to determine under‐ and over‐moderated criticality values. The criticality values for the homogenized cell geometry are slightly lower and remains <5% range, that is, on the conservative side, which is acceptable for the purpose of this study. This allows us to homogenize the entire core for the full 3‐D reactor with core and reflector regions. Reactor criticality increases with increasing H/235U ratios in the, under moderated region, where H/235U = 18 is selected for further investigations. Temporal variation of the reactor criticality is pursued for full reactor power of 210 MWth (~70 MWel net) with keff = 1.36 at the beginning of life down to keff = 1.02 over 1260 days with a plant factor of 1.0. The final fuel burn up was found as 57.3 GWd/ton. [ABSTRACT FROM AUTHOR]

The purpose of this work is to characterize the morphological, structural, and strength properties of model prototypes of new-generation TRi-structural ISOtropic particle fuel (TRISO) designed for Generation IV high-temperature gas reactors (HTGR-type). The choice of model structures consisting of inner pyrolytic carbon (I-PyC), silicon carbide (SiC), and outer pyrolytic carbon (O-PyC) as objects of research is motivated by their potential use in creating a new generation of fuel for high-temperature nuclear reactors. To fully assess their full functional value, it is necessary to understand the mechanisms of resistance to external influences, including mechanical, as in the process of operation there may be external factors associated with deformation and leading to the destruction of the surface of fuel structures, which will critically affect the service life. The objective of these studies is to obtain new data on the fuel properties, as well as their resistance to external influences arising from mechanical friction. Such studies are necessary for further tests of this fuel on corrosion and irradiation resistance, as closely as possible to real conditions in the reactor. The research revealed that the study samples have a high degree of resistance to external mechanical influences, due to the high strength of the upper layer consisting of pyrolytic carbon. The presented results of the radiation resistance of TRISO fuel testify to the high resistance of the near-surface layer to high-dose irradiation.

This paper presents a deterministic performance assessment for spent fuel from deep-burn modular hightemperature reactors (DBMHRs) in the proposed Yucca Mountain repository. Typical DBMHR designs utilize fuel elements manufactured from graphite. The fuel itself is made of TRISO particles containing the fissile material. The performance of the DBMHR spent fuel (DBSF) was evaluated in terms of the annual dose to the reasonably maximally exposed individual (RMEI) under various hydrogeological conditions. Part of this evaluation was an analysis of the graphite waste matrix and of the TRISO particles under repository conditions, the result of which indicates that the lifetime of the graphite matrix greatly exceeds that of the TRISO particles and that it is the graphite, not the TRISO particles, that serves to sequester the radionuclides within the fuel matrix. Under all 14 cases considered, DBSF is seen to comply with the annual dose standards set in Part 197 of Title 40 of the Code of Federal Regulations, for exposure via groundwater contamination under current climatic conditions. Parametric studies for the effect of waste matrix lifetime on annual dose received by the RMEI indicate that repository performance is sensitively linked to waste matrix durability because most radionuclides including actinides are likely to be released congruently with the graphite matrix.

The double heterogeneity of the tristructural isotropic (TRISO) fuel in the prismatic-core advanced high-temperature reactor should be accurately and correctly modeled and analyzed, especially for a large-scale loaded with the double-heterogeneity effect. The reactivity-equivalent physical transformation method was developed and employed to enable homogenizing TRISO fuel in a high temperature reactor considering the double heterogeneity and taking into account the large problem involved in performing the whole-core burnup calculation using Monte Carlo transport codes with double-heterogeneity problems. In this work, the heterogeneous effects of a collision of probability calculation method were used to represent the effects of scattering anisotropy on the leakage rates and the isotropic streaming effects due to low optical density in the model. The WIMS and DRAGON codes have been used to perform the calculations of double heterogeneity for the TRISO fuel, fuel compact, and fuel element and the results are compared with the SERPENT Monte Carlo code. [ABSTRACT FROM AUTHOR]

Tristructural isotropic (TRISO) particle fuel is one of the most promising fuel concepts enabling high temperature and high burnup reactor operation. One dominant source of radioactivity released from the TRISO particles is silver (Ag), which is subject to a high release fraction and long decay life compared to other fission products. Previous modeling efforts using the fuel performance code BISON indicated nonnegligible uncertainties in modeling the diffusion process of fission products in TRISO compared to the Advanced Gas Reactor experiments. The overall uncertainties observed when modeling the fission product diffusion can result from uncertainties in model parameters, noisy experimental measurements, and deficiencies in the developed models. The three types of underlying uncertainties have not yet been properly quantified in open literature. This paper presents the Bayesian uncertainty quantification (UQ) using massively parallelizable Markov chain Monte Carlo samplers. The uncertainties due to model parameters, model inadequacy, and experimental measurement noise are quantified, with the σ term used to represent the sum of the model inadequacy and measurement noise uncertainties. It is worth noting that this is the first time the σ term is inferred for nuclear fuel experiments, as compared to using prescribed values for uncertainty quantification in previous work. The parallelizable Markov chain Monte Carlo samplers efficiently infer the model parameters and the σ term, giving insight into physical parameters like diffusion coefficients and the combined model discrepancy and measurement noise. A subsequent forward uncertainty quantification (UQ) is also performed based on the calibration results to generate more accurate predictions of the Ag release. The model inadequacy plus experimental noise is the most dominant source of uncertainty compared to the parametric uncertainty. All the UQ analyses presented in this work are based on the second series of the irradiation experiments in the Advanced Gas Reactor program. • Inverse UQ of TRISO diffusion model parameters using parallel Bayesian inference. • Sigma term (model inadequacy plus experimental noise uncertainty) is also inferred. • Forward UQ of model parameters and sigma term for predictive uncertainty in Ag. • Model inadequacy plus experimental noise most dominant source of uncertainty. • Bayesian median Ag release predictions match experimental data better. [ABSTRACT FROM AUTHOR]

Highlights • FCM fuel is expected to have similar behavior in HTGR design basis accidents as conventional fuel. • Fuel temperatures were predicted to be higher with FCM fuel. • The fuel cycle cost of FCM fuel in HTGRs may be higher. Abstract 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. [ABSTRACT FROM AUTHOR]

Highlights • FCM fuel is expected to have similar behavior in mHTGR design basis accidents as conventional fuel. • Power distribution is an important factor that drives differences in maximum fuel temperature during both normal operation and accident scenarios. • The impact of SiC annealing on the progression of mHTGR design basis accidents is expected to be marginal. Abstract We conducted reactor performance calculations to assess the potential design basis accident performance of HTGR fuel designs. Three Fully Ceramic Microencapsulated (FCM) fueled HTGR designs were developed in a previous work (Lu et al., 2018). The maximum fuel temperature in the cores fueled by these three FCM fuels was predicted to be higher than that in the reference 350-MWt mHTGR core in both normal operating conditions and during representative design basis accidents (Lu and Brown, 2019). To better understand the potential safety margins in mHTGR design basis accidents, we performed thermal-hydraulics sensitivity studies to investigate how maximum fuel temperature varies considering various parameters, e.g. thermal properties, within the ranges corresponding to the differences between the FCM-fueled prismatic mHTGR cores and the reference core with conventional fuel compacts. We found that the difference in the steady-state axial power distribution contributed the most to the difference in the maximum fuel temperature, in both normal operation and design basis accidents. Experimental data suggested that the annealing process of irradiation defects in SiC would be rapid at mHTGR relevant fuel temperatures. The bounding potential impact of the SiC annealing on the maximum fuel temperature was analyzed considering both the thermal conductivity recovery and the Wigner energy release due to the annealing of SiC. We found that the defect annealing process in SiC would at most increase the peak maximum fuel temperature of an FCM-fueled core by 40 K in loss of forced cooling accidents and by 10 K in a control rod withdrawal accident. Additional experiments on the SiC defect annealing kinetics and Wigner energy release in more relevant conditions are needed. [ABSTRACT FROM AUTHOR]

Highlights • High-temperature fluoride salt cooled reactor is evaluated. • A comprehensive method for the analysis of the on-line refueling has been developed. • On-line refueling eliminates cycle length issue and improves fuel utilization. • On-line refueling provides a reduction of the fuel cycle cost. • On-line refueling maintains adequate safety with respect to multi-batch refueling. Abstract Several academic and commercial organizations around the world are developing the Fluoride salt-cooled High-temperature Reactor (FHR) technology, due to its safety features and potential to generate high temperature energy for electricity and process heat applications. The Advanced High Temperature Reactor (AHTR) being considered in this study is a FHR design developed at Oak Ridge National Laboratory (ORNL). It is based on the use of graphite as moderator, FLiBe as coolant, and hexagonal fuel elements with fuel plates (or "planks") composed of TRISO particles embedded in a carbonaceous matrix. The AHTR reference design is based on traditional batch refueling approach, which requires to shut down the reactor and replace/reshuffle a certain amount of fuel assemblies in the core at a specific frequency. Several options have been evaluated in the design process, in order to maximize the cycle length and optimize the use of fuel. However, the relatively short cycle and poor fuel utilization are intrinsic features of this family of reactors, due to the low heavy metal loading in the core and insufficient moderation, which are competing aspects in terms of core volume fraction. Since the fuel is expected to be more expensive than the fuel of light water reactors (LWR), this issue might challenge the economic viability of the AHTR. In order to eliminate or ameliorate this issue, a novel approach to refueling has been developed and proposed, consisting of continuous on-power refueling, or on-line refueling, in which the refueling procedure is performed at full power or partially reduced power (the reactor is not shut down) and a single assembly is removed for each refueling operation. A systematic neutronic and thermal-hydraulic analysis approach has been developed and performed to assess the viability and safety of the refueling operations, followed by the evaluation of the core design requirements and a quantification of the economic advantages resulting from the implementation of this procedure. [ABSTRACT FROM AUTHOR]

• Estimation of the impact of geometrical and material property uncertainties on failure fraction of TRISO particles. • Uncertainty quantification in TRISO fuel performance analysis. • Optimization potential of TRISO particles from mechanical perspectives. Tri-structural isotropic (TRISO) fuel particles are a fuel form being considered for potential use in next-generation nuclear reactors (i.e., high-temperature gas-cooled reactors). Though the TRISO fuel manufacturing process has continually advanced in recent years, particle comparisons still reveal statistical variations and uncertainties in terms of geometric configurations and material properties. Given that the physical processes ongoing in TRISO fuel particles during reactor operation are highly correlated with each other, a small degree of uncertainty in one model may lead to significant uncertainty in another. This makes appropriate uncertainty quantification of TRISO fuel particles essential. However, one may wonder about the extent to which the current version of TRISO particles has been optimized, and whether any room remains for further improvements. This paper quantifies TRISO fuel performance model uncertainties that stem from geometric and material data. For this analysis, the BISON code was used, and the Advanced Gas Reactor (AGR)-2 experiment served as a reference case. A total of 105 calculations was performed for the uncertainty and optimization analysis, altering the geometric and material data within their uncertainty range. Lastly, the optimization potential of TRISO particles is evaluated from a fuel performance perspective. [ABSTRACT FROM AUTHOR]

Nuclear spent fuel management is one of the top major subjects in the utilization of nuclear energy. Hence, solutions to this problem have been increasingly researched in recent years. The basic aim of this work is to examine the fissile breeding and transuranic fuel transmutation potentials of a gas cooled accelerator-driven system. In line with this purpose, firstly, the conceptually designed system is optimized by using several target materials and fuel mixtures, from the point of neutronic. Secondly, three different material compositions, namely, pure lead bismuth eutectic (LBE), LBE+natural UO2, and LBE+15 % enrichment UO2, are considered as target material. The target zone is separated to two sub-zones but as one within the other. The outer sub-zone is pure LBE target, and the inner sub-zone is either UO2 or pure LBE target. The UO2 target sub-zone is cooled with helium gas. Finally, the thorium dioxide mixed with transuranic dioxides, discharged from PWR-MOX spent fuel, in pebbles composed of graphite and TRISO-coated spherical fuel particles, is used for breeding fissile fuel and transmuting transuranic fuels. Three different thorium-transuranic mixtures, (Th, Pu)O2, (Th, Cm)O2, (Th, Pu, MA)O2, are examined with various mixture fractions. The packing fractions of the fuel pebbles in the transmutation core and the tristructural-isotropic coated fuel particles in a pebble are assumed as 60 % and 29 %, respectively. The transmutation core is also cooled with a high-temperature helium coolant. In order to produce high-flux neutrons that penetrate through the transmutation core, the target is exposed to the continuous beams of 1 GeV protons. The computations have been carried out with the high-energy Monte Carlo code MCNPX using the LA150 library. The numerical outcomes show that the examined accelerator-driven system has rather high neutronic data in terms of the energy production and fissile fuel breeding.

Ceramography was performed on cross sections from four tristructural isotropic (TRISO) coated particle fuel compacts selected from the AGR-2 experiment, which was irradiated between June 2010 and October 2013 in the Advanced Test Reactor (ATR). The individual fuel specimens examined in this study contained coated particles with either uranium oxide (UO 2 ) kernels or uranium oxide/uranium carbide (UCO) kernels that were irradiated to final burnup values between 9.0 and 11.5% FIMA. These examinations were intended to explore kernel and coating morphology evolution during irradiation. This included observations of kernel porosity, kernel swelling, and irradiation-induced TRISO coating layer fracture and separation. Variations in behavior within a specific cross section, which could be related to temperature or burnup gradients within the fuel compact, were also explored. Results were compared with similar investigations performed as part of the earlier AGR-1 irradiation experiment. The criteria for categorizing AGR-1 particle post-irradiation morphologies during ceramographic exams were applied to the AGR-2 compact particles examined. This paper presents the results of the AGR-2 examinations and discusses the key implications for fuel irradiation performance. [ABSTRACT FROM AUTHOR]

Gamma spectrometry has been used to evaluate the burnup and fission product inventory of different components from the US Advanced Gas Reactor Fuel Development and Qualification Program’s second TRISO-coated particle fuel irradiation test (AGR-2). TRISO fuel in this irradiation included both uranium carbide/uranium oxide (UCO) kernels and uranium oxide (UO 2 ) kernels. Four of the 6 capsules contained fuel from the US Advanced Gas Reactor program, and only those capsules will be discussed in this work. The inventories of gamma-emitting fission products from the fuel compacts, graphite compact holders, graphite spacers and test capsule shell were evaluated. The measured activities were used to calculate the fractional release of fission products such as Cs-137, Cs-134, Eu-154, Ce-144, and Ag-110m from the compacts. The fraction of Ag-110m retained in the compacts ranged from 1.8% to full retention. Additionally, the activities of the radioactive cesium isotopes (Cs-134 and Cs-137) have been used to evaluate the burnup of all US TRISO fuel compacts in the irradiation. The experimental burnup evaluations compare favorably with burnups predicted from physics simulations. Predicted burnups for UCO compacts range from 7.26 to 13.15% fission per initial metal atom (FIMA) and 9.01–10.69% FIMA for UO 2 compacts. Measured burnup ranged from 7.3 to 13.1% FIMA for UCO compacts and 8.5–10.6% FIMA for UO 2 compacts. Results from gamma emission computed tomography performed on compacts and graphite holders that reveal the distribution of different fission products in a component will also be discussed. Gamma tomography of graphite holders was also used to locate the position of TRISO fuel particles suspected of having silicon carbide layer failures that lead to in-pile cesium release. [ABSTRACT FROM AUTHOR]

Advanced nuclear reactor technologies have the potential to expand the missions of nuclear energy while reducing carbon emissions. This paper presents scoping reactor physics and thermal hydraulics analysis of a high temperature gas-cooled reactor (HTGR) using the fully ceramic microencapsulated (FCM) fuel form, and demonstrates the feasibility of FCM fueled HTGRs. FCM fuel consists of tr istructural iso tropic (TRISO) coated fuel particles embedded in a matrix of silicon carbide (SiC). The potential advantages of FCM fuel, which uses a monolithic SiC matrix, over conventional HTGR fuels with a carbon-based matrix include: a long refueling interval; high stability of the SiC matrix under irradiation with limited swelling; high fission product retention of the fuel form, with the SiC matrix acting as an additional barrier to fission product release; and enhanced oxidation resistance during normal operation and air ingress accidents. In addition, the literature shows that the effective thermal conductivity of SiC fuel compacts and conventional HTGR compacts are expected to be similar. The key finding of this study is that FCM fuel, within the form factor of a typical General Atomics prismatic graphite block, exhibits similar fuel cycle performance to conventional HTGR fuel. The reactor cycle length, discharge burnup, and natural resource utilization are similar. However, the reduced moderation in the FCM designs considered here does marginally reduce the discharge burnup, and therefore natural resource utilization, versus the reference HTGR design. The hardened neutron flux spectrum resulting from the SiC matrix, which displaces carbon from the core, requires a slightly higher packing fraction of conventional uranium oxy-carbide (UCO) fuel kernels or the use of higher density uranium mononitride (UN)-based fuel kernels. These options will marginally increase the decay power, because they harden the neutron flux energy spectrum and increase the density of 238 U in the fuel. In one case considered, this will increase the absorption of neutrons in 238 U, and the resultant impact of 239 Np isotope on the decay power. The Doppler coefficients normalized per total fuel heat capacity are weaker in the FCM-fueled designs than in the reference HTGR design. This impacts the energy deposition in a control rod ejection accident, and hence the design of potential transient tests of these fuel forms. In addition, analyses of loss-of-forced cooling accidents indicate that the fuel temperature during these design basis accidents are up to ∼30 °C higher with FCM fuel than with conventional HTGR fuels due to the increased decay power. [ABSTRACT FROM AUTHOR]

SUMMARY This article focuses on calculation of reactivity and kinetic parameters in different states (different core heights and boron concentrations in the coolant water) of the fixed bed nuclear reactor (FBNR). The numerical calculations are performed with the SRAC code. Uranium dioxide (UO2) with 235U enrichment grade of 5% is used as spherical fuel pellets in the TRISO fuel type. The core height is changed by a core height level limiter. The main results of this study can be summarized as follows: The effective neutron multiplication coefficient is varied with the core height and boron concentration in the coolant water. When there is no-boron in the coolant water and the core height is over 120 cm, the reactor control can be carried out by the movement of a core height level limiter, without the fine control rods in the core center; but when there is a boron concentration, the reactor may be controlled by the level limiter at the lower core heights. The kinetic parameters (the prompt-neutron lifetime and effective delayed neutron fraction) of the reactor also are changed to the reactor core height and to the boron concentration in the coolant water. Copyright © 2016 John Wiley & Sons, Ltd. [ABSTRACT FROM AUTHOR]

TRistructural ISOtropic (TRISO) fuels are commonly used in High-Temperature Gas cooled Reactors due to their resistance to high temperatures, irradiation, and oxidation. Reactor physics modelling of a fuel form consisting of millions of layered particles is quite challenging, however. Homogenisation techniques are often used to simplify the fuel form, representing it as a solid material to reduce computational costs. The impact of these approximations on reactor physics parameters was evaluated using OpenMC. Geometries at various scales (i.e. fuel pin, fuel block, full core) were developed based on the Modular High-Temperature Gas-cooled Reactor. It was found that homogenisation results in a significant reduction in simulation time. The Reactivity equivalent Physical Transformation (RPT) method was found to be superior to simple Volume-Weighted Homogenization (VWH). The neutron energy spectra and flux distributions calculated by RPT had insignificant differences to the explicit model. Reactivity coefficients were found to be similar, but diverged at very high temperatures. • Explicit neutronic modelling of TRISO particles is computationally expensive. • Fuel homogenisation was studied using OpenMC models of the MHTGR benchmark. • Reactor physics parameters of fuel pin, fuel block, and full core were compared. • The RPT method was found to be an adequate approximation, greatly reducing run time. [ABSTRACT FROM AUTHOR]

In many cases, a X̅ control chart based on a performance variable is used in industrial fields. Typically, the control chart monitors the measurements of a performance variable itself. However, if the performance variable is too costly or impossible to measure, and a less expensive surrogate variable is available, the process may be more efficiently controlled using surrogate variables. In this paper, we present a model for the economic statistical design of a VSI (Variable Sampling Interval) X̅ control chart using a surrogate variable that is linearly correlated with the performance variable. We derive the total average profit model from an economic viewpoint and apply the model to a Very High Temperature Reactor (VHTR) nuclear fuel measurement system and derive the optimal result using genetic algorithms. Compared with the control chart based on a performance variable, the proposed model gives a larger expected net income per unit of time in the long-run if the correlation between the performance variable and the surrogate variable is relatively high. The proposed model was confined to the sample mean control chart under the assumption that a single assignable cause occurs according to the Poisson process. However, the model may also be extended to other types of control charts using a single or multiple assignable cause assumptions such as VSS (Variable Sample Size) X̅ control chart, EWMA, CUSUM charts and so on. [ABSTRACT FROM AUTHOR]

This article is about calculation of fuel burn-up and some physics parameters of the Fixed Bed Nuclear Reactor (FBNR), using spherical fuel elements similar to TRISO fuel, which is investigated by SRAC code. The fuel that was used is a uranium dioxide sphere (UO 2 ) with U 235 enrichment of 5%. The main results of this study can be summarized as follows: (1) Variation of nuclear isotope densities and effective neutron multiplication coefficient in process of fuel burn-up is determined and the reactor can operate during the time of 2.73 years, without on-site refueling. The U 235 burn-up reaches value 71.84% of its initial density at the end of the reactor fuel cycle. (2) When the core height is over 120 cm, the reactor control can be carried out by the movement of a core height level limiter, without the fine control rods in the core center. (3) The kinetic parameters of the reactor (the prompt-neutron lifetime and effective delayed neutron fraction) change in process of reactor operation. [ABSTRACT FROM AUTHOR]

Tristructural isotropic (TRISO)-coated particle fuel is a robust nuclear fuel and determining its reliability is critical for the success of advanced nuclear technologies. However, TRISO failure probabilities are small and the associated computational models are expensive. We used coupled active learning, multifidelity modeling, and subset simulation to estimate the failure probabilities of TRISO fuels using several 1D and 2D models. With multifidelity modeling, we replaced expensive high-fidelity (HF) model evaluations with information fusion from two low-fidelity (LF) models. For the 1D TRISO models, we considered three multifidelity modeling strategies: only Kriging, Kriging LF prediction plus Kriging correction, and deep neural network (DNN) LF prediction plus Kriging correction. While the results across these multifidelity modeling strategies compared satisfactorily, strategies employing information fusion from two LF models called the HF model least often. Next, for the 2D TRISO model, we considered two multifidelity modeling strategies: DNN LF prediction plus Kriging correction (data-driven) and 1D TRISO LF prediction plus Kriging correction (physics-based). The physics-based strategy, as expected, consistently required the fewest calls to the HF model. However, the data-driven strategy had a lower overall simulation time since the DNN predictions are instantaneous, and the 1D TRISO model requires a non-negligible simulation time. • TRISO, a robust nuclear fuel, is associated with small failure probabilities. • Active learning with multifidelity modeling to efficiently estimate TRISO failure. • Multifidelity information fusion from two low-fidelity models most efficient. • Physics-based multifidelity strategy reduces calls to high-fidelity model. • Data-driven multifidelity strategy has less overall simulation time. [ABSTRACT FROM AUTHOR]

A new daily as-run thermal analysis was performed at the Idaho National Laboratory for the advanced gas cooled reactor (AGR) test experiment number two (AGR-2) in the Advanced Test Reactor (ATR). This thermal analysis incorporates gas gaps changing with time during the irradiation experiment due to graphite shrinkage resulting from neutron damage. The purpose of this analysis was to calculate the daily average temperatures of each TRISO (tristructural isotropic)-particle fuel compact. A steady-state thermal analysis was performed daily for each capsule with the commercial finite element heat transfer code ABAQUS. These new thermal predictions show the compact fuel temperature dependence on the variable gas gap method. Comparison between measured and calculated temperatures is discussed. [ABSTRACT FROM AUTHOR]

Low-enriched uranium oxycarbide (LEUCO) and surrogate tristructural isotropic (TRISO)- coated-particle compacts with particle volumetric packing fractions of 25%, 40%, and 48% were imaged utilizing X-ray-computed tomography. Subsequent 3D image analysis identified and further quantified kernel size, sphericity, and observed porosity. In addition, the spatial distribution, coordination number, and kernel-nearest neighbors were analyzed and compared for the different packing fractions. Metrics such as observed porosity and sphericity enabled quantification and screening for abnormal kernels within TRISO compacts. Assessment of TRISO particle location confirmed and further quantified a non-uniform distribution of TRISO particles with the spatial distribution in the radial direction being roughly described as a dampened sinusoidal function. The amplitude and frequency of this non-uniform distribution increased with increasing packing fraction. Measured kernel-nearest neighbor distances indicated two regions along the radial surfaces of compacts where TRISO particles are more likely to be in intimate contact with one another. These regions were found: (1) at upper and lower faces of compacts (i.e., corners); (2) offset ∼10% of a compact's length from the axial center near the exterior surface. Within these regions, small quantities of defective TRISO surrogate particles (48% packing fraction) and defective LEUCO TRISO particles (40% packing fraction) were found. No defective particles were found within 25% packing fraction LEUCO TRISO compacts. [ABSTRACT FROM AUTHOR]

• A mesoscale effective Ag diffusion coefficient in SiC is derived from atomistic calculations of bulk and GB diffusivity. • The effective Ag diffusion coefficient in SiC accounts for temperature and microstructure effects. • Microstructures with grains that are smaller in the direction perpendicular to diffusion promote faster Ag diffusion in SiC. • The multiscale effective Ag diffusivity is within an order of magnitude of the empirical diffusion coefficient currently used in Bison. • Accounting for the microstructure effects on Ag diffusion increases the prediction accuracy of Ag release from AGR-1 TRISO fuel. The silicon carbide (SiC) layer in tristructural isotropic (TRISO) fuel particles serves as a barrier to prevent the escape of fission products produced and not retained in the fuel kernel. The release of silver (Ag) is a concern due to the long half-life of the 110 m Ag isotope. However, accurately determining the fission gas release rate requires knowing the diffusion coefficient through the SiC layer. In this study, we leverage atomistic calculations of Ag diffusivity in SiC bulk and grain boundaries (GBs) to develop a mesoscale effective Ag diffusion coefficient (D e f f) in SiC. Since GBs serve as pathways for Ag diffusion, D e f f is defined as a function of temperature and microstructure variables. In particular, the size of SiC grains in the direction perpendicular to diffusion is shown to significantly affect Ag diffusion. The prediction of the mechanistic, mesoscale approach falls within one order of magnitude of empirical values. The temperature and microstructure-dependent effective Ag diffusivity in SiC is implemented in the fuel performance code Bison with a correction factor to predict Ag release from AGR-1 TRISO fuel particles. We hereby quantify the impact of SiC grain size on Ag release and improve Bison's predictions. [Display omitted] [ABSTRACT FROM AUTHOR]

The behavior of oxygen in uranium oxycarbide (UCO) has been investigated using density functional theory. To assess the role of carbon on the stability of oxygen point defects, we first determined the formation energies of oxygen vacancy and interstitial for different carbon configurations. Subsequently, we evaluated the barriers for migration of various oxygen defects in the UCO matrix. Our results show that carbon atoms create strong attraction centers for oxygen atoms, as a consequence the spontaneous formation of oxygen Frenkel pairs in the fluorite structure is promoted. Moreover, the strong affinity of oxygen atoms for sites in the first coordination shell of a carbon center leads to the formation of CO, and thus reducing the mobility of oxygen atoms in the UCO fuel. Upon formation, the CO remains fixed at the carbon site due to significant hybridization with uranium atoms, therefore behaving as trap sites for oxygen atoms. Our findings indicate that the formation of CO molecules increases the retention of oxygen atoms inside the UCO matrix. Consequently, carbon atoms in the UCO matrix contribute to mitigating the deleterious effects that oxygen release from the fuel kernel has on the structural integrity of the silicon carbide layer in tristructural isotropic (TRISO) particles. [ABSTRACT FROM AUTHOR]

Highlights: [•] We developed MATLAB and RELAP5models of the single channel of the AHTR. [•] Single channel analysis indicates design envelope for effective heat removal. [•] The reactivity feedback evaluated by SCALE supports safe operation of the reactor. [•] We developed RELAP5 models of the fuel assembly and full core. [•] The fully passive DRACS protects the reactor during a LOFC accident. [ABSTRACT FROM AUTHOR]

Highlights: [•] AFPR-100 reactor considered as a small nuclear reactor without on-site refueling originally based on TRISO micro-fuel element. [•] The AFPR-100 reactor was re-designed using the new Spherical Cermet fuel element. [•] The adoption of the Cermet fuel instead of TRISO fuel reduces the core lifetime operation by 3.1 equivalent full power years. [•] We discussed the new micro-fuel element candidate for small and medium sized reactors. [ABSTRACT FROM AUTHOR]

Abstract: The early version of the conceptual modified design of the Laser Inertial Confinement Fusion Fission Energy (LIFE) engine consists of a spherical fusion chamber of 5m diameter, surrounded by a multi-layered blanket. The first wall is made of 2cm thick ODS and followed by a Li17Pb83 zone (2cm), acting as neutron multiplier, tritium breeding and front coolant zone. It is separated by an ODS layer (2cm) from the FLIBE molten salt zone (50cm), containing fissionable fuel. A 3rd ODS layer (2cm) separates the molten salt zone on the right side from the graphite reflector (30cm). Calculations have been conducted for a constant fusion driver power of 500MWth in S8-P3 approximation using 238-neutron groups. Reactor grade (RG) plutonium carbide fuel in form of TRISO particles with volume fractions of 2%, 3%, 4%, 5% and 6% have been dispersed homogenously in the FLIBE coolant. Tritium breeding ratio (TBR) values per incident fusion neutron for the above cited cases start with TBR=1.35, 1.52, 1.73, 2.02 and 2.47, respectively. With the depletion of fissionable RG-Pu isotopes, TBR decreases gradually. At startup, higher fissionable fuel content in the molten salt leads to higher blanket energy multiplication, namely M 0 =3.8, 5.5, 7.7, 10.8 and 15.4 with 2%, 3%, 4%, 5% and 6% TRISO volume fraction, respectively. Calculations have led to very high burn up values (>400,000 MD.D/MT). TRISO particles can withstand such high burn ups. Such high burn ups would lead to drastic reduction of final nuclear waste per unit energy production. [Copyright &y& Elsevier]

Abstract: High temperature gas reactors (HTGRs) are being considered for near term deployment in the United States under the GNEP program and farther term deployment under the Gen IV reactor design (). A common factor among current HTGR (prismatic or pebble) designs is the use of TRISO coated particle fuel. TRISO refers to the three types of coating layers (pyrolytic carbon, porous carbon, and silicon carbide) around the fuel kernel, which is both protected and contained by the layers. While there have been a number of reactors operated with coated particle fuel, and extensive amount of research has gone into designing new HTGRs, little work has been done on modeling and analysing the degradation rates of spent TRISO fuel for permanent geological disposal. An integral part of developing a spent fuel degradation modeling was to analyze the waste form without taking any consideration for engineering barriers. A basic model was developed to simulate the time to failure of spent TRISO fuel in a repository environment. Preliminary verification of the model was performed with comparison to output from a proprietary model called GARGOYLE that was also used to model degradation rates of TRISO fuel. A sensitivity study was performed to determine which fuel and repository parameters had the most significant effect on the predicted time to fuel particle failure. Results of the analysis indicate corrosion rates and thicknesses of the outer pyrolytic carbon and silicon carbide layers, along with the time dependent temperature of the spent fuel in the repository environment, have a significant effect on the time to particle failure. The thicknesses of the kernel, buffer, and IPyC layers along with the strength of the SiC layer and the pressure in the TRISO particle did not significantly alter the results from the model. It can be concluded that a better understanding of the corrosion rates of the OPyC and SiC layers, along with increasing the quality control of the OPyC and SiC layer thicknesses, can significantly reduce uncertainty in estimates of the time to failure of spent TRISO fuel in a repository environment. [Copyright &y& Elsevier]

Abstract: Reactivity feedback coefficients have been calculated for a compact sized PWR core that utilizes carbon coated micro fuel particles instead of standard cylindrical fuel pellets with an inventive composition. A small amount of Pu-240 with 5 w/o has also been added in tristructural-isotropic (TRISO) fuel in place of U-238 for the reduction of excess reactivity. The values of fuel, moderator and void reactivity coefficients have been calculated at the middle of fuel cycle. All the reactivity coefficients were found negative which meet the design safety criteria. It was also observed that all reactivity feedback coefficients are interlinked and their effects are pronounced when coupled together. [ABSTRACT FROM AUTHOR]

HTRs use a high performance particulate TRISO fuel with ceramic multi-layer coatings due to the high burn up capability and very neutronic performance. TRISO fuel because of capable of high burn up and very neutronic performance is conducted in a D-T fusion driven hybrid reactor. In this study, TRISO fuels particles are imbedded body-centered cubic (BCC) in a graphite matrix with a volume fraction of 68%. The neutronic effect of TRISO coated LWR spent fuel in the fuel rod used hybrid reactor on the fuel performance has been investigated for Flibe, Flinabe and LiSn coolants. The reactor operation time with the different first neutron wall loads is 24 months. Neutron transport calculations are evaluated by using XSDRNPM/SCALE 5 codes with 238 group cross section library. The effect of TRISO coated LWR spent fuel in the fuel rod used hybrid reactor on tritium breeding (TBR), energy multiplication (M), fissile fuel breeding, average burn up values are comparatively investigated. It is shown that the high burn up can be achieved with TRISO fuel in the hybrid reactor. [ABSTRACT FROM AUTHOR]