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Hyper-fidelity (HxF) depletion of pebble bed reactors (PBRs) is the capability to model depletion for every pebble while accounting for motion through the core. Previous HxF work demonstrated feasibility to deplete hundreds of thousands of stationary pebbles concurrently within reasonable timeframes. This work illustrates the second step towards HxF, coupling depletion with a discrete motion scheme. The model assumes an ordered bed with pebbles occupying fixed positions. Motion is simplified as discrete since pebbles move in straight lines from one set position to another. The methodology was implemented in Serpent 2, combined with its transport and depletion capabilities. Ad-hoc routines were developed ensuring compatibility with domain decomposition and pebble recirculation after each pass based on discharge criteria and fresh pebble insertion. Capabilities of HxF with discrete motion are demonstrated using a full-scale high-temperature gas-cooled reactor model. Specifically, an approach to equilibrium is performed, and example results are shown for in-core and discarded pebbles. The data illustrates how HxF provides unique insights into PBR fuel, producing information on statistical distributions rather than average values only, as obtained by traditional methods that rely on spectral zoning for depletion. Knowledge of these distributions can greatly improve analysis and assessment of PBRs.

The accuracy of neutrons modeling and simulation tools strongly depends on the quality of the nuclear data. Data libraries are generated by evaluators combining physics-based model codes and experimental data. There are many instances where experimental data are not available, are not reported rigorously or are discordant. In such cases, the evaluators need to make an expert judgment exposing the generated data to human bias and large uncertainties. This work proposes to support the evaluators’ complex tasks by leveraging Machine Learning (ML) and Artificial Intelligence (AI). Two proof-of-concept ML models, a Decision Tree and K-Nearest-Neighbor, were developed to fit nuclear data from the EXFOR database in order to infer neutron induce reaction cross sections. Both models were used to predict nuclear data for 233U, a well-characterized isotope in literature, and 35Cl, a less studied but important nuclide for some advanced nuclear reactors. The predicted values for 233U were validated using the 233U Jezebel benchmark in Serpent2 model. The predicted values for 35Cl(n,p) cross section were compared against recent new measurement not available in EXFOR. The predicted ML/AI values matched more accurately the new measurements than any of the evaluated data libraries, which overestimate experimental results by up to a factor of five. In turn, the proof-of-concept models explored in this work, reliant on learning underlying patterns of cross section data from other radionuclides, demonstrate evidence that ML models can aid traditional physics-guided models and have a role to play in nuclear data evaluations. Furthermore, incorporating ML models in the nuclear data pipeline can allow evaluators to make faster bias-free decisions in areas of uncertainty as well as better inform future data measurement campaigns on areas of greatest sensitivity in EXFOR.

The hybrid resolution (HR) method is a recently developed calculational algorithm tailored for efficient nuclear reactor analysis by leveraging multigroup (MG) Monte Carlo (MC) coupled with continuous energy MC calculations. While the HR method has demonstrated promising results in various scenarios, an anticipated challenge arises from inconsistencies of MG MC results caused by spatial homogenization when relatively coarse mesh is adopted, particularly in heterogeneous problems characterized by a strong flux gradient. To address this issue, the HR method incorporates MG MC calculations on the same explicit geometry used in continuous energy MC calculations. In other words, this approach preserves accuracy in geometry handling while solely simplifying energy treatment. In our investigation, a 3-dimensional BWR-type fuel assembly was focused on for a comparative study regarding the different geometry treatments in MG MC calculations. Various reactor parameters such as k-effective, power, flux, and atomic densities were estimated, and numerical performance was also compared in terms of the computing time and figure-of-merits. Our evaluation revealed that the explicit geometry-based HR method yielded more consistent and reliable parameters compared to the simplified coarse mesh-based scheme, reducing errors in power and flux distributions by about 10%.

Previous studies assessed the feasibility of designing Sodium-cooled Fast Reactors (SFR) in a novel Seed-and-Blanket (S&B) configuration in which a significant fraction of the core power is generated by a radial metallic thorium-fueled blanket that operates in the Breed-and-Burn (B&B) mode. The radiation damage on the cladding material in both seed and blanket does not exceed the presently acceptable constraint of 200 Displacements per Atom (DPA). This paper investigates a number of blanket fuel options other than metallic thorium fuel, including thorium dioxide fuel, thorium hydride fuel, thorium nitride Fully Ceramic Encapsulated (FCM) fuel, and Pressurized Water Reactor (PWR) Used Nuclear Fuel (UNF) after limited reprocessing. Since the neutron spectrum of the oxide fueled blanket is softer than that of the reference metallic thorium fueled blanket, it can discharge its fuel at an average burnup of around 110 MWd/kg versus 65 MWd/kg of the metallic thorium. The two thorium hydride fueled blankets feature an even softer neutron spectrum than the oxide fueled blanket and, therefore, a higher average discharge burnup — 192 MWd/kg and 245 MWd/kg when the H-to-Th ratio is, respectively, 0.5 and 2.0. The FCM fuel is composed of SiC matrix and cladding that is expected to self-anneal the neutrons induced damage. With the assumption that the FCM fuel will indeed be proven not limited by radiation damage, its discharge burnup could be up to 482 MWd/kg. As a result, the corresponding thorium utilization is the highest of all options examined – close to 80 times the utilization of natural uranium in present Light Water Reactors (LWRs). The amount of Trans-Th isotopes accumulated in the thorium blanket per unit of generated electricity decreases as the blanket fuel is discharged at a higher burnup. Therefore, a higher discharge blanket burnup results in lower long-term radioactivity and radiotoxicity. It is also found that the 232U/233U ratio in the discharged thorium fuel is over 3 times higher in the thorium hydride than in the other blankets thus improving the thorium-hydride blanket's proliferation resistance. Softening of the blanket spectrum leads to softening of the seed spectrum region interfacing with the blanket and this enables to discharge the seed fuel at a higher average burnup. The higher discharge burnup of the seed fuel reduces the required reprocessing and fuel fabrication capacities per unit of generated electricity. The cores having softer neutron spectrum blankets also feature less positive feedback to sodium voiding and much more negative Doppler coefficient. When PWR UNF is used after limited reprocessing to fuel the blanket, the subcritical blanket driven by the excess neutrons from the seed can discharge this fuel at an average burnup of ∼120 MWd/kg. Combined with the burnup in the first stage PWRs, this two-stage energy system can achieve an accumulated uranium burnup of ∼170 MWd/kg. This increases the fuel utilization of natural uranium by a factor of three relative to once-through PWRs while at the same time reducing the amount of high-level waste that needs to be disposed of per unit of generated electricity. Another scenario examined in this paper is a 3-stage closed energy system: TRU extracted from Stage-1 PWRs used fuel is charged to the seed of Stage-2 S&B cores while the Trans-Th (mainly 233U) extracted from Stage-2 S&B blanket discharged fuel is used as the fissile feed for Stage-3 PWRs. It is estimated that 1GWe of S&B SFRs in such a 3-Stage energy system can support 3.2 GWe of PWRs versus ∼1.7 GWe of PWRs that can be supported by 1 GWe of CR = 0.5 Advanced Burner Reactor (ABR).

The Seed-and-Blanket (S&B) Sodium-cooled Fast Reactor (SFR) core concept was proposed for generating a significant fraction of the core power from thorium fueled breed-and-burn (B&B) blankets without exceeding the presently verified radiation damage constraint of 200 Displacement per Atom (DPA). To make beneficial use of the excess neutrons from fast reactors, the S&B core is designed to have an elongated TRU transmuting (or “TRU burner”) seed from which over 20% of the fission neutrons leak into a subcritical thorium blanket that radially surrounds the seed. The seed fuel is recycled while the blanket operates in a once-through breed-and-burn (B&B) mode. The objective of this paper is to compare the fuel cycle performance of the S&B reactor against an Advanced Burner Reactor (ABR) and a conventional Pressurized Water Reactor (PWR). For the fast reactors (SFR: ABR and S&B) the fuel cycle performance is evaluated based on a 2-stage PWR-SFR energy system while the reference nuclear system is made of once-through PWRs. It was found that relative to the ABR, the S&B core has a lower fuel cycle cost, higher capacity factor, and comparable short-term radioactivity. The discharged seed fuel from the S&B core features lower fissile Pu-to-Pu ratio, higher 238Pu-to-Pu ratio, higher specific plutonium decay heat, higher spontaneous fission rate, and lower overall material attractiveness for weapon use. Due to the significant amount of 233U discharged from the breed-and-burn thorium fueled blankets, the S&B core has much higher long-term radioactivity and radiotoxicity. Since the thorium fueled blanket operates in the breed-and-burn mode and requires no fuel reprocessing, the discharged blanket fuel is unattractive for weapons application. Compared with a PWR, the S&B core has a lower fuel cycle cost, much lower short-term radioactivity and radiotoxicity but higher long-term values, and higher proliferation resistance for the discharged plutonium. The natural uranium utilization of the 2-stage PWR-S&B system is approximately 60% higher than that of present PWRs; it is few percent higher than that of the 2-stage PWR-ABR system. Approximately 7% of the thorium fed to the blanket is converted into energy, which makes the thorium fuel utilization approximately 12 times the utilization of natural uranium in PWRs. A comprehensive fuel cycle evaluation performed with the methodology developed by the recent U.S Department of Energy's Nuclear Fuel Cycle Evaluation and Screening campaign concludes that the PWR-S&B system has similar fuel cycle performance characteristics as the PWR-ABR system. The S&B concept may potentially feature improved economics and resource utilization relative to the ABR.

In this work, the capability of computing adjoint-weighted kinetic parameters, including effective delayed neutron fraction and neutron generation time, was implemented in the Reactor Monte Carlo (RMC) code based on the iterated fission probability (IFP) method. Three algorithms, namely, the Non-Overlapping Blocks (NOB) algorithm, the Multiple Overlapping Blocks (MOB) algorithm and the superhistory algorithm, were implemented in RMC to investigate their accuracy, computational efficiency and estimation of variance. The algorithms and capability of computing kinetic parameters in RMC were verified and validated by comparison with MCNP6 as well as experimental results through a set of multi-group problems and continuous-energy problems.

Metal Matrix Micro-encapsulated (M3) fuel consists of TRISO coated fuel particles directly dispersed in a matrix of zirconium metal to form a solid rod. In this integral fuel concept the cladding tube and the failure mechanisms associated with it have been eliminated; therefore, M3 fuel, compared to existing fuel designs, is expected to provide greatly improved operational performance. The main challenge to the deployment of M3 fuels is the low heavy metal load. This needs to be compensated with high density fuel (uranium nitride), larger rods and mainly higher enrichment. This study found that M3 fuel requires about 15.5% enriched UN TRISO particles to match the cycle length of standard fuel when loaded in a PWR. In order to achieve comparable reactivity feedbacks, in particular moderator temperature, the pitch to rod diameter ratio should be reduced to 1.20 compared to the typical 1.326. As M3 fuel provides a better path for heat transfer from the fuel, such smaller pitch is expected to be feasible, but further evaluations especially to examine thermal-hydraulic aspects are required. Analyses of possible load patterns for burnable poisons showed that M3 fuel maintains similar cycle length to that of standard fuel and pin power peaks are also comparable. Nevertheless, further optimization will be required to limit assembly power peaks.

This paper assesses the feasibility of designing seed-and-blanket (S&B) sodium-cooled fast reactor (SFR) cores to generate a significant fraction of the core power from radial thorium-fueled blankets that operate in the breed-and-burn (B&B) mode. The radiation damage on the cladding material in both seed and blanket does not exceed the presently acceptable constraint of 200 displacements per atom (dpa). The S&B core is designed to have an elongated seed (or driver) to maximize the fraction of neutrons that radially leak into the subcritical B&B blanket and reduce the neutron loss via axial leakage. A specific objective of this study is to maximize the fraction of core power generated by the B&B blanket that is proportional to the neutron leakage rate from the seed to the blanket. Since the blanket feed fuel is very inexpensive and requires no reprocessing and remote fuel fabrication, a larger fraction of power from the blanket will result in a lower fuel cycle cost per unit of electricity generated by the SFR core. It is found possible to design the seed of the S&B core to have a lower transuranics (TRU) conversion ratio (CR) than a conventional advanced burner reactor (ABR) core without deteriorating core safety. This is due to the unique synergism between a low CR seed and the B&B thorium blanket. The benefits of the synergism are maximized when using an annular seed surrounded by inner and outer thorium blankets. Two high-performance S&B cores are designed to benefit from the annular seed concept: (1) an ultra-long-cycle core having a CR = 0.5 seed and a cycle length of ∼7 effective full-power years (EFPYs) and (2) a high-transmutation core having a TRU CR of 0.0. The TRU transmutation rate of the latter core is comparable to that of the reference ABR with a CR of 0.5, and the thorium blanket can generate close to 60% of the core power. Because of the high blanket power fraction along with the high discharge burnup of the CR = 0 seed, the reprocessing capacity per unit of core power required by this S&B core is only approximately 1/6th that of the reference ABR core with a TRU CR of 0.5. Although the seed fuel CR is nearly zero, the burnup reactivity swing is low enough to enable a cycle length of more than 4 EFPYs. This is attributed to a combination of reactivity gain in the thorium blankets over the cycle and the relatively high heavy metal inventory. Moreover, despite the very low leakage, the S&B cores feature a less positive coolant reactivity coefficient and large enough negative Doppler coefficient even when using nonfertile fuel for the seed, because of the unique physics properties of the U and Th in the thorium blankets. With the long cycles, the S&B SFR is expected to have a higher capacity factor, and therefore a lower cost of electricity, than conventional ABRs. The discharge burnup of the thorium blanket fuel is typically 70 MWd/kg such that the thorium fuel utilization is approximately 12 times that of natural uranium in light water reactors. A sensitivity study is subsequently undertaken to quantify the trade-off between the core performances and several design variables: amount of zirconium in the inert matrix seed fuel, active core height, coolant pressure drop, and radiation damage constraint. The effect of the criterion used for quantifying acceptable radiation damage is evaluated as well. It is concluded that a viable S&B core can be designed without significant deviation from typical SFR core design practices.

Coupled Monte Carlo (MC) methods are becoming widely used in reactor physics analysis and design. Many research groups therefore, developed their own coupled MC depletion codes. Typically, in such coupled code systems, neutron fluxes and cross sections are provided to the depletion module by solving a static neutron transport problem. These fluxes and cross sections are representative only of a specific time-point. In reality however, both quantities would change through the depletion time interval. Recently, Generalized Perturbation Theory (GPT) equivalent method that relies on collision history approach was implemented in Serpent MC code. This method was used here to calculate the sensitivity of each nuclide and reaction cross section due to the change in concentration of every isotope in the system. The coupling method proposed in this study also uses the substep approach, which incorporates these sensitivity coefficients to account for temporal changes in cross sections. As a result, a notable improvement in time dependent cross section behavior was obtained. The method was implemented in a wrapper script that couples Serpent with an external depletion solver. The performance of this method was compared with other existing methods. The results indicate that the proposed method requires substantially less MC transport solutions to achieve the same accuracy.

Capabilities for nuclear data sensitivity and uncertainty analysis have been recently implemented in numerous continuous-energy Monte Carlo codes, including the Reactor Monte Carlo (RMC) code. Previous work developed the capability for RMC of computing sensitivity coefficients of the effective multiplication factor and related uncertainties, due to nuclear data. In this work, such capability was extended to generalized responses in the form of ratios of linear response functions of the forward flux based on the collision history-based approach as implemented in SERPENT2. The superhistory algorithm was also adopted in RMC to reduce memory consumption for generalized sensitivity calculations. These new capabilities of RMC were verified by comparing results of TSUNAMI-1D in SCALE6.1 code package, and SERPENT2 through Jezebel, Flattop and the UAM TMI PWR pin cell benchmark problems.

The XGPT method extends the Generalized Perturbation Theory capabilities of Monte Carlo codes to continuous-energy sensitivity functions. In this work, this method is proposed as a new approach to nuclear data uncertainty propagation. XGPT overcomes some of the limitations of legacy perturbation-based approaches. In particular, it allows the nuclear data uncertainty propagation to be performed adopting continuous energy covariance matrices, instead of discretized (multi-group) data. The XGPT capabilities are demonstrated in three simple fast criticality benchmarks for 239Pu and 208Pb cross section uncertainties. The new method is also applied in selected cases to estimate higher moments of the keff distribution, starting from TENDL random evaluations. The XGPT estimates, when compared against reference Total Monte Carlo (TMC) results, show a good agreement and a significant reduction in computational requirements with respect to the TMC approach. Finally, the capabilities for uncertainty propagation involving adjoint-weighted response functions are demonstrated.

The University of California, Berkeley (UCB), has developed a preconceptual design for a commercial pebble-bed (PB), fluoride salt-cooled, high-temperature reactor (FHR) (PB-FHR). The baseline design for this Mark-I PB-FHR (Mk1) plant is a 236-MW(thermal) reactor. The Mk1 uses a fluoride salt coolant with solid, coated-particle pebble fuel. The Mk1 design differs from earlier FHR designs because it uses a nuclear air-Brayton combined cycle designed to produce 100 MW(electric) of base-load electricity using a modified General Electric 7FB gas turbine. For peak electricity generation, the Mk1 has the ability to boost power output up to 242 MW(electric) using natural gas co-firing. The Mk1 uses direct heating of the power conversion fluid (air) with the primary coolant salt rather than using an intermediate coolant loop. By combining results from computational neutronics, thermal hydraulics, and pebble dynamics, UCB has developed a detailed design of the annular core and other key functional features. Both an active normal shutdown cooling system and a passive, natural-circulation-driven emergency decay heat removal system are included. Computational models of the FHR-validated using experimental data from the literature and from scaled thermal-hydraulic facilities-have led to a set of design criteria and system requirements for the Mk1 to operate safely and reliably. Three-dimensional, computer-aided-design models derived from the Mk1 design criteria are presented.

An attractive feature of using liquid lithium as the breeder and coolant in fusion blankets is that it has very high tritium solubility and results in very low levels of tritium permeation throughout the facility infrastructure. However, lithium metal vigorously reacts with air and water and presents plant safety concerns. The Lawrence Livermore National Laboratory is carrying an effort to develop a lithium-based ternary alloy that maintains the beneficial properties of lithium (e.g. high tritium breeding and solubility) and at the same time reduces overall flammability concerns. This study evaluates the neutronics performance of lithium-based alloys in the blanket of an inertial fusion energy chamber in order to inform such development. 3-D Monte Carlo calculations were performed to evaluate two main neutronics performance parameters for the blanket: tritium breeding ratio (TBR), and the fusion energy multiplication factor (EMF). It was found that elements that exhibit low absorption cross sections and higher q-values such as Pb, Sn, and Sr, perform well with those that have high neutron multiplication such as Pb and Bi. These elements meet TBR constrains ranging from 1.02 to 1.1. However, most alloys do not reach EMFs greater than 1.15. Additionally, it was found that enriching lithium with 6Li significantly increases the TBR and decreases the minimum lithium concentration by more than 60%. The amount of enrichment depends on how much total lithium is in the alloy to begin with. Alloys that performed well in the TBR and EMF calculations were considered for activation analysis. Activation simulations were executed with 50 years of irradiation and 300 years of cooling. It was discovered that bismuth is a poor choice due to achieving the highest decay heat, contact dose rates, and accident doses. In addition, it does not meet the waste disposal ratings (WDR). Some of the activation results for alloys with Sn, Zn, and Ga were in the higher end and should be considered secondary to elements such as Sr and Ba that had overall better results. The results of this study along with other considerations such as thermodynamics, and chemical reactivity will help down select a preferred lithium ternary alloy.

Externally driven subcritical systems are closely associated with thorium, partially because thorium has no naturally occurring fissile isotopes. Both accelerator-driven systems (ADSs) and fusion-driven systems have been proposed. This paper highlights key literature related to the use of thorium in externally driven systems (EDSs) and builds upon this foundation to identify potential roles for EDSs in thorium fuel cycles. In fuel cycles with natural thorium feed and no enrichment, the potential roles are (1) a once-through breed-and-burn fuel cycle and (2) a fissile breeder (mainly 233U) to support a fleet of critical reactors. If enriched uranium is used in the fuel cycle in addition to thorium, EDSs may be used to burn transuranic material. These fuel cycles were evaluated in the recently completed U.S. Department of Energy Evaluation and Screening of nuclear fuel cycle options relative to the current once-through commercial nuclear fuel cycle in the United States. The evaluation was performed with respect to nine specified high-level criteria, such as waste management and resource utilization. Each of these fuel cycles presents significant potential benefits per unit energy generation compared to the present once-through uranium fuel cycle. A parametric study indicates that fusion-fission-hybrid systems perform better than ADSs in some missions due to a higher neutron source relative to the energy required to produce it. However, both potential externally driven technology choices face significant development and deployment challenges. In addition, there are significant challenges associated with the use of thorium fuel and with the transition from a uranium-based fuel cycle to a thorium-based fuel cycle.

In severe accident conditions with loss of active cooling in the core, zirconium alloys, used as fuel cladding materials for current light water reactors (LWR), undergo a rapid oxidation by high temperature steam with consequent hydrogen generation. Novel fuel technologies, named accident tolerant fuels (ATF), seek to improve the endurance of severe accident conditions in LWRs by eliminating or at least mitigating such detrimental steam-cladding interaction. Most ATF concepts are expected to work within the design framework of current and future light water reactors, and for that reason they must match or exceed the performance of conventional fuel in normal conditions. This study analyzed the neutronic performance of ATF when employed in both pressurized and boiling water reactors. Two concepts were evaluated: (1) coating the exterior of zirconium-alloy cladding with thin ceramics to limit the zirconium available for reaction with higherature steam; (2) replacing zirconium alloys with alternative materials possessing slower oxidation kinetics and reduced hydrogen production. Findings show that ceramic coatings should remain 10-30 μm thick to limit the neutronic penalty. Alternative cladding materials, with the exception of SiC, enhance neutron loss compared to zirconium-alloys. An extensive parametric analysis concluded that reference performance metrics can be met by employing 300-μm or less thick cladding or increasing fuel enrichment by up to 1.74% depending on material and geometry.

Fully ceramic microencapsulated (FCM) fuel consists of TRISO (tristructural-isotropic) fuel particles embedded in a ceramic matrix (SiC) to form fuel pellets and rods and offers improved fission product retention and lower operating temperature with expected superior performance in normal and off-normal conditions compared to conventional fuel. When coupled with SiC cladding, FCM fuel eliminates zirconium altogether and is expected to drastically reduce hydrogen generation during a beyond-design- basis accident. In order to be deployed in current or future pressurized water reactors (PWRs), FCM fuel must meet or exceed the neutronic performance of conventional fuel. Limited by low heavy metal loading, an FCM fuel assembly requires increased enrichment and large fuel rods to match the cycle length of a conventional fuel assembly. This study investigated the core design, neutronics, and thermal hydraulics of a PWR loaded with FCM fuel and sought to optimize the assembly design to minimize the enrichment required to reach fuel performance similar to that of conventional fuel. It was found that the implementation of FCMfuel in a 17 X 17 assembly requires close to 20% enrichment and large fuel rods. Such design performs comparably to conventional fuel (4.5% enrichment) in terms of cycle length, reactivity coefficients, intra-assembly power peaking factor, burnable poison penalty, and control rod worth but requires an increase of pumping power. A parametric analysis spanned a large design space varying fuel outer diameter and pitch-to- diameter ratio (P/D) and downselected two alternate assembly designs: 11 X 11 (1.65-cm outer diameter and 1.18 P/D) and 9 X 9 (2.12-cm outer diameter and 1.12 P/D). These designs meet the cycle length requirement with 18.6% and 16.2% enrichments, respectively, but feature a smaller minimum departure from nucleate boiling ratio (MDNBR) compared to a reference assembly. It was estimated that a slight increase in rod outer diameter increases MDNBR to the desired level and implies a pressure drop increase of 10%.

Nuclear fuel cycle advancements will result in new types of fissile material, including nuclear wastes, that require security and safeguards. Nuclear wastes may be more vulnerable for diversion by non-state actors, and chemical processing to recover fissile material is not an insurmountable challenge. Previous work has applied a figure of merit (FOM) to assess material attractiveness and security risks. This analysis applies the material attractiveness FOM to wastes produced by fuel cycles from the Fuel Cycle Evaluation and Screening (FCES) study. Two aspects of security risk are studied: (1) the time before the fissile material in the waste becomes attractive and (2) the number of waste packages required to obtain a critical mass of fissile material. Two fuel cycles are presented to highlight detailed results: (1) once-through use of low-enriched U in light water reactors (LWRs) and (2) continuous recycle of Pu in sodium fast reactors (SFRs). Increasing LWR used nuclear fuel (UNF) package loading increases the time to attractiveness, but the larger packages contain enough Pu for multiple critical masses. The high-level waste (HLW) from processing the SFR fuels has similar FOM behavior but longer time to attractiveness due to the concentration of fission products. More HLW packages are required to obtain a critical mass; that number can be further increased by increasing the separation efficiency. Extended to all FCES fuel cycles, the minimum time before attractiveness is generally lower for UNF than for HLW because radioactivity is concentrated in HLW. For nearly all fuel cycles that produce UNF, only one package is required to recover enough fissile material for a critical mass. Notably, some advanced fuel cycles produce HLW, of which only two packages need to be recovered to obtain a critical mass, even when the target fissile material is recycled. Going forward, an assessment of the security risks posed by fissile material in nuclear wastes will need to quantify the challenge posed by separations. Ultimately, the assessment could inform security and response measures; whether any of the observations might affect these measures could be an area for future work. Finally, future analysis could study whether different fuel cycle wastes are more attractive for use in radiological dispersal devices or radiological exposure devices. [ABSTRACT FROM AUTHOR]

This study establishes a procedure for constructing explicit and adjoint-based implicit sensitivities with MCNP5. Using these methods, an instantaneous sensitivity-based uncertainty analysis is performed on the depleted uranium hybrid LIFE (Laser Inertial Fusion Energy) blanket. Explicit sensitivities and uncertainties are calculated for (n, 2n), tritium production, fission, and radiative capture reaction rates during the fuel lifecycle. Nuclear data uncertainties and Monte Carlo counting precision are compared in a convergence study and the compounding of the two is quantified to gauge the validity of the analysis. A multi-group cross-section library is generated for adjoint calculations and selected adjoint distributions are shown and discussed.

Molten Salt Reactors (MSRs) have recently gained resurged research and development interest in the advanced reactor community. Several computational tools are being developed to capture the strong neutronics/thermal-hydraulics coupling effect in this special reactor configuration. This paper presents a consistent one-dimensional (1D) multigroup neutron diffusion model for MSR analysis, with the primary aim for fast and accurate calculations for long transients, as well as sensitivity and uncertainty analysis of the reactor. A fictitious radial leakage cross section is introduced in the model to properly account for the radial leakage effects of the reactor. The leakage cross section and other consistent neutronics parameters are generated with the Monte Carlo code Serpent using high-fidelity three-dimensional (3D) models. The accuracy of the 1D consistent model is verified by the reference solution from the Monte Carlo model on the Molten Salt Reactor Experiment (MSRE) configuration. The 1D consistent model successfully reproduced the integrated flux from the 3D model and the reactor multiplication factor keff with the error in the range of 95 to 397 pcm (per cent mille), depending on discretized energy group structures. The developed model is also extended to estimate the reactivity loss due to fuel circulation in MSRE. The estimate of reactivity loss in dynamics analysis is in great agreement with the experimental data. This model functions as the first step in the development of a 1D fully neutronics/thermal-hydraulics coupled model for short- and long-term MSRE transient analysis. [ABSTRACT FROM AUTHOR]

The effective delayed neutron fraction is an important reactor kinetics parameter. In flowing liquid-fuel reactors, this differs from the delayed neutron fraction because of the emission of delayed neutrons with a lower energy spectrum than prompt and the delayed neutron precursor (DNP) drift due to the fuel movement. In general, neglecting delayed neutron precursor drift leads to an over-estimation of the effective delayed neutron fraction. Nevertheless, the capability to simulate this peculiar phenomenon is not available in most reactor physics tools. In this project, a multi-physics approach to modeling DNP drift is developed using the GeN-Foam toolkit, and it benchmarked against available experimental data from the Molten Salt Reactor Experiment (MSRE). GeN-Foam couples a neutron diffusion solver with a thermal-hydraulics solver. Additionally, a new function was added for solving adjoint multi-group diffusion eigenvalue problems and calculating effective delayed neutron fraction. For benchmarking, an R-Z model of the MSRE was developed in GeN-Foam. The porous media model was applied, and cross sections were generated using the Monte Carlo code Serpent-2 with ENDF/B-VII.1 nuclear data library. In order to evaluate the impact of DNP drift, two steady-state conditions (stationary and flowing salt at 1200 gpm) were simulated. A reactivity change of -241 pcm was calculated using GeN-Foam for the MSRE between static and flowing fuel, which is in a good agreement with the experimental value of -212 pcm. The total effective delayed neutron fraction change was calculated to be -230 pcm vs. -304 pcm reported for the MSRE and analytical calculated during the experimental campaign. Three transient accidents were also analyzed.

The Autonomous Reactivity Control (ARC) system is a passive safety system aiming to provide an additional negative reactivity feedback during reactor transient scenarios. This paper shows how the performance of the ARC system can be enhanced by introducing a hydraulic diode that allows for different engagement and disengagement speeds of system. The benefits of the proposed system is assessed in a reference soidum-cooled fast reactor (SFR) during multiple postulated transient scenarios. The reactor and plant dynamic response are evaluated using SAS4A/SASSYS-1, whereas for the internal ARC system fluid dynamics, the SAM code is adopted. The two codes are externally coupled using a custom driver script which coordinates data exchange and restart calculations at each time step, with Picard iterations used to converge each time step. All the transients analyzed in this work show that the enhanced ARC system is effective in reducing peak temperatures and in reducing the oscillatory behavior encountered in some cases with the standard ARC system.

A set of benchmarks based on the experimental data from the Molten Salt Reactor Experiment (MSRE) is being compiled as part of International Reactor Physics Experiments Evaluation Reactor Physics Experiments Evaluation Project (IRPhEP). The initial benchmark that will be available in the 2019 edition of the IRPhEP handbook covers the first zero-power criticality experiment. Follow up benchmarks are under development based on the series of control rod calibration experiments performed at the MSRE, which consisted in progressive addition of a small amount (85g) of 235U in the salt followed by the insertion of the control rods acts to compensate for the excess reactivity insertion. Multiple reactivity effects and coefficients measurements are included in the benchmark: differential worth of a control rod, reactivity equivalent of 235U addition, control rod bank worth, reactivity effect of fuel circulation, isothermal temperature coefficient and fuel temperature coefficient. An uncertainty of 2% is attributed to the reported reactivity measurements from experimenters and it was believed that the uncertainty of reactor period measurement contributed the most of the experimental uncertainty. An additional 2% uncertainty was added to all reactivity measurements to represent the uncertainty for the correction factor applied to pull all the measurements on the same uranium concentration and this uncertainty was reasonably inferred by evaluating this factor on the MSRE benchmark model. The calculated reactivity equivalent of 235U additions (0.2228±0.0014, represented as the change of reactivity over the relative change of 235U mass in loop) matches well with the experiment value (0.223±0.006). Most of other calculations, including the control rod bank worth, reactivity effects of fuel circulation and isothermal and fuel temperature coefficients fall within one standard deviation from the experimental values as well.