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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.

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 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.

As Feiras de Ciências se constituem como uma possibilidade de rompimento com a abordagem tradicional na educação científica e tecnológica, podendo favorecer o estabelecimento de relações entre Ciência, Tecnologia e Sociedade e o desenvolvimento do Letramento Científico dos estudantes envolvidos no planejamento, execução e avaliação desse tipo de atividade pedagógica. Nesse sentido, este estudo tem por objetivo principal investigar, a partir da visão de professores da rede pública da cidade de Morro da Fumaça/ Santa Catarina (Brasil), quais as contribuições da participação em Feiras de Ciências para os estudantes da educação básica. Para isso, as percepções de nove docentes da área de Ciências da Natureza (das áreas de Química, Física e Biologia) sobre a contribuição das Feiras de Ciências para a formação dos estudantes foram investigadas qualitativamente, por meio de um estudo de caso cujo os dados foram produzidos a partir de um questionário respondido virtualmente. As respostas ao instrumento de produção de dados foram analisadas com inspiração na análise de discurso de linha franco-brasileira e em categorias de Letramento Científico publicadas na literatura. Foi identificado que as Feiras de Ciências, na visão dos professores participantes, possibilitam o desenvolvimento das dimensões conceitual, procedimental e afetiva do Letramento Científico, contribuindo para a formação científica dos estudantes. Nesse sentido, recomendamos que as Feiras de Ciências ocupem um lugar de maior destaque no planejamento das atividades curriculares da educação científica e tecnológica, incorporando a luta contra as injustiças sociais no escopo dessa atividade.

The Odeo Cornaro is a remarkable example of Renaissance architecture featuring an octagonal umbrella-vaulted hall surrounded by four adjacent barrel-vaulted spaces. According to the principles outlined by Vitruvius, the central octagonal hall was prized for its acoustical qualities, emphasizing sound propagation and vocal resonance. Due to its remarkably well-preserved condition, the structure continues to serve as a prestigious venue for musical and cultural events. This study investigates the acoustic of this remarkably preserved musical space, employing measurements and numerical models. Finite element analysis and geometrical acoustics models were employed to gain comprehensive insights into sound pressure level distribution across a broad frequency spectrum. The Odeo stands as an outstanding historical music venue due to its capacity to enhance vocal projection while maintaining an intimate ambience during musical performances.

The Cl35(n,p) and Cl35(n,α) cross sections at incident neutron energies between 2.42 and 2.74 MeV were measured using the Berkeley High Flux Neutron Generator. The cross sections for Cl35(n,p) were more than a factor of 3 to 5 less than all of the values in the neutron absorption data libraries, while the Cl35(n,α) cross sections are in reasonable agreement with the data libraries. The measured energy-differential cross section is consistent with a single resonance with a width of 293(46) keV. This result suggests that, despite the high incident neutron energy, any attempt to model (n,x) cross sections in the vicinity of the N=Z=20 shell gap requires a resolved resonance approach rather than a Hauser-Feshbach approach.

The Cl35(n,p) and Cl35(n,α) cross sections at incident neutron energies between 2.42 and 2.74 MeV were measured using the Berkeley High Flux Neutron Generator. The cross sections for Cl35(n,p) were more than a factor of 3 to 5 less than all of the values in the neutron absorption data libraries, while the Cl35(n,α) cross sections are in reasonable agreement with the data libraries. The measured energy-differential cross section is consistent with a single resonance with a width of 293(46) keV. This result suggests that, despite the high incident neutron energy, any attempt to model (n,x) cross sections in the vicinity of the N=Z=20 shell gap requires a resolved resonance approach rather than a Hauser-Feshbach approach.

The redox effect of activation reactions in molten salts has been previously qualitatively defined but never quantified. Here, we quantify the corrosion effects of neutron activation of FLiBe. Rates of change of activation products are computed for FHR and ARC. The impact of activation reactions on salt chemistry is quantitatively compared to other environmental effects of similar consequences on the salt. The results are applied to a redox control case-study and an irradiation-corrosion experiment. The results are presented in a manner useful to the reactor chemist and designer concerned with corrosion and chemistry control and tritium management in irradiated FLiBe, and design of salt-irradiation experiments.

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).

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.

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.

We discuss the challenge of selecting materials for nuclear applications and outline the need for comprehensive databases to assist scientists and engineers in choosing materials that meet interdependent physical, chemical, and nuclear criteria. In conventional engineering, chemical and physical properties and the electronic structure of materials are typically the primary considerations; nuclear applications must also consider the nuclear physics characteristics of a material. Development of databases that correlate physical, chemical, and nuclear properties would accelerate and facilitate innovations in nuclear design.

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.

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.

The resource-renewable boiling water reactor (RBWR) is an innovative boiling water reactor that has the capability to breed or to burn transuranium elements (TRUs). Core characteristics of the RBWR of the TRU burner type were evaluated by two different core analysis methods. The RBWR core features an axially heterogeneous configuration, which consists of an internal blanket region between two seed regions, to achieve the TRU multi-recycling capability while maintaining a negative void reactivity coefficient. Axial power distribution of the TRU burner core tends to be more heterogeneous because the isotopic composition ratio of fertile TRUs to fissile TRUs becomes larger in the TRU burner-type core than in the breeder-Type core and the seed regions need to be axially shorter than that of the breeder-Type core. Thus core analysis of the TRU burner-type core is more challenging. A conventional diffusion calculation using nuclear constants prepared by two-dimensional lattice calculations was performed by Hitachi, while the calculation using nuclear constants prepared by three-dimensional calculations and axial discontinuity factors was performed by the University of Michigan to provide a more sophisticated treatment of the axial heterogeneity. Both calculations predicted similar axial power distributions except in the region near the boundary between fuel and plenum. Both calculations also predicted negative void reactivity coefficients throughout the operating cycle. Safety analysis was performed by Massachusetts Institute of Technology for the all-pump trip accident, which was identified as the limiting accident for the RBWR design. The analysis showed the peak cladding temperature remains below the safety limit. Detailed fuel cycle analysis by University of California, Berkeley, showed that per electrical power generated, the RBWR is capable of incinerating TRUs at about twice the rate at which they are produced in typical pressurized water reactors.

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.

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.

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.

The safety assessment of a geological repository for used nuclear fuel must ensure that future generations are shielded from radiation from fission products, in particular those released by re-criticality events. An investigation is required to understand whether or not criticality can actually be achieved. In fulfilling this end, this study assesses the uncertainty in the composition and total mass of precipitates forming in the far-field due to variation in transport parameters. The Latin Hypercube Sampling technique is employed to generate an accurate, random distribution of variables employed in the transport model and to assess the uncertainty of attaining a critical mass. The average characteristics of the damaged fuel from the Fukushima Daiichi reactor cores is used as the reference waste form. Results are compared to the minimum critical masses of previous studies to assess the criticality safety margin.

The RBWR-TR is a thorium-based reduced moderation BWR (RBWR) with a high transuranic (TRU) consumption rate. It is charged with LWR TRU and thorium, and it recycles all actinides an unlimited number of times while discharging only fission products and trace amounts of actinides through reprocessing losses. This design is a variant of the Hitachi RBWR-TB2, which arranges its fuel in a hexagonal lattice, axially segregates seed and blanket regions, and fits within an existing ABWR pressure vessel. The RBWR-TR eliminates the internal axial blanket, eliminates absorbers from the upper reflector, and uses thorium rather than depleted uranium as the fertile makeup fuel. This design has been previously shown to perform comparably to the RBWR-TB2 in terms of TRU consumption rate and burnup, while providing significantly larger margin against critical heat flux. This study examines the uncertainty in key neutronics parameters due to nuclear data uncertainty. As most of the fissions are induced by epithermal neutrons and since the reactor uses higher actinides as well as thorium and 233U, the cross sections have significantly more uncertainty than in typical LWRs. The sensitivity of the multiplication factor (keff) to the cross sections of many actinides is quantified using a modified version of Serpent 2.1.19 [1]. Serpent [2] is a Monte Carlo code which uses delta tracking to speed up the simulation of reactors; in this modified version, cross sections are artificially inflated to sample more collision, and collisions are rejected to preserve a "fair game." The impact of these rejected collisions is then propagated to the multiplication factor using generalized perturbation theory [3]. Covariance matrices are retrieved for the ENDF/B-VII.1 library [4], and used to collapse the sensitivity vectors to an uncertainty on the multiplication factor. The simulation is repeated for several reactor configurations (for example, with a reduced flow rate, and with control rods inserted), and the difference in keff sensitivity is used to assess the uncertainty associated with the change (the uncertainty in the void feedback and the control rod worth). The uncertainty in the RBWR-TR is found to be dominated by the epithermal fission cross section for 233U in reference conditions, although when the spectrum hardens, the uncertainty in fast capture cross sections of 232Th becomes dominant.

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.

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.