Fueled Core Research Articles

Uncertainty Quantification of 237Np, 241Am, and 243Am Reaction Rates in Highly Enriched Uranium Fuel Cores at Kyoto University Critical Assembly

Experimental analyses of 237Np, 241Am, and 243Am fission, as well as 237Np capture reaction rates, are conducted with the Serpent 2 code together with ENDF/B-VIII.0 and JENDL-5 using experimental data for the neutron spectra of thermal and intermediate regions obtained in the solid-moderated and solid-reflected cores with highly enriched uranium fuel at the Kyoto University Critical Assembly. Also, uncertainty quantification of the fission and capture reaction rate ratios of the test samples of 237Np, 241Am, and 243Am with reference samples of 235U and 197Au are evaluated by the MARBLE code system. In terms of the fission reaction rate ratios of 237Np/235U, 241Am/235U, and 243Am/235U, a comparison between experiments and Serpent 2 calculations shows an accuracy of about 5%, 15%, and 10%, respectively, together with ENDF/B-VIII.0 and JENDL-5. For the capture reaction rate ratios of 237Np/197Au, Serpent 2 calculations reveal a fairly good accuracy at the thermal neutron spectrum. The total uncertainties of the 237Np/235U, 241Am/235U, and 243Am/235U fission reaction rate ratios by MARBLE with the covariance data of ENDF/B-VIII.0 and JENDL-5 are found to be about 4% at most in all cores, except for about 8% for 243Am/235U with ENDF/B-VIII.0 at the intermediate neutron spectrum.

Admixed pellets for fast and efficient delivery of plasma enhancement gases: Investigations at AUG exploring the option for EU-DEMO

Gas and pellet injection are envisaged for particle fuelling in EU-DEMO. The gas system will provide edge and divertor fuelling and any further gas species required for operation. Pellets, mm-sized bodies formed from solid hydrogen fuel, are designed for efficient and fast core fuelling. However, they can also be employed for a more efficient delivery of plasma enhancement gases, by admixing them with the fuelling pellets. To check this option for EU-DEMO, explorative investigations have been performed at ASDEX Upgrade (AUG).The AUG system produces ice in a batch process sufficient for about 100 pellets, initially designed for operation with pure H2 or D2. On a trial basis, pellet formation was tested using an H2/D2 mixture and admixtures containing small amounts (up to 2 mol%) of N2, Ar, Kr or Xe in the D2 host. A homogeneous and reproducible ice composition was found for the H2/D2 = 1:1 case. For all the admixed gases, a depletion of the admixture in the ice with increasing atomic number is observed. Nevertheless, the fast and efficient delivery of admixed pellets was clearly demonstrated in dedicated plasma experiments at AUG. Detailed investigations showed that the Ar supplied via admixed pellets has a higher radiation efficiency and a faster radiation rise than an Ar/D2 gas puff. Furthermore, Ar density measurements in a discharge with admixed pellet injection show reasonable agreement with findings of a fading admixed species’ concentration along the ice rod and assumptions on the pellet ablation location in the plasma. Investigations performed at the Oak Ridge National Laboratory with a large batch extruder using up to 2 mol% Ne in D2 confirmed that production of much larger ice quantities can be achieved.These initial explorative investigations clearly reveal the great potential of admixed pellets, although they also demonstrate that further technology efforts are required before their benefits can be utilized.

Nodal method for handling irregularly deformed geometries in hexagonal lattice cores

The hexagonal nodal code RENUS has been enhanced to handle irregularly deformed hexagonal assemblies. The underlying RENUS methods involving triangle-based polynomial expansion nodal (T-PEN) and corner point balance (CPB) were extended in a way to use line and surface integrals of polynomials in a deformed hexagonal geometry. The nodal calculation is accelerated by the coarse mesh finite difference (CMFD) formulation extended to unstructured geometry. The accuracy of the unstructured nodal solution was evaluated for a group of 2D SFR core problems in which the assembly corner points are arbitrarily displaced. The RENUS results for the change in nuclear characteristics resulting from fuel deformation were compared with those of the reference McCARD Monte Carlo code. It turned out that the two solutions agree within 18 pcm in reactivity change and 0.46% in assembly power distribution change. These results demonstrate that the proposed unstructured nodal method can accurately model heterogeneous thermal expansion in hexagonal fueled cores.

A comparison of the influence of plasmoid-drift mechanisms on plasma fuelling by cryogenic pellets in ITER and Wendelstein 7-X

Pellet injection is the most promising technique to achieve efficient plasma core fuelling, key for attaining stationary scenarios in large magnetic confinement fusion devices. In this paper, the injection of pellets with different volumes and speeds into standard plasma scenarios in ITER (tokamak) and Wendelstein 7-X (stellarator) is studied by modeling the pellet ablation and particle deposition, focusing on the evaluation of the expected differences in pellet plasmoid drifts in tokamaks and stellarators. Since the efficiency of the damping-drift mechanisms is predicted to depend on the magnetic configuration, device-specific characteristics are expected for the temporal evolution of the plasmoid drift acceleration. For instance, plasmoid-internal Pfirsch–Schlüter currents dominate the drift damping process for stellarators, while plasmoid-external currents are more relevant for tokamaks. Also, relatively larger drifts are in principle expected for W7-X due to higher field gradients in relation to machine dimensions. However, shorter plasmoid-internal charge reconnection lengths result in the drift damping due to internal Pfirsch–Schlüter currents being more effective than in a tokamak. Therefore, the average relative drift displacement during the whole plasmoid homogenization may a priori be comparable in both magnetic configurations. Moreover, High Field Side (HFS) injection is expected to be highly advantageous to maximize pellet particle deposition in ITER, whereas it may only be beneficial in medium to high β environments in W7-X. Finally, there may be means for the optimization of pellet injection configurations in both ITER and W7-X for the considered plasma scenarios despite the sizeable differences in the relative importance of the mechanisms of plasmoid drift acceleration and deceleration in play.

Comparative Study on the Impact of endf/b-vii.1 Nuclear Library Uncertainty on the CERCER and CERMET Fueled Accelerator-Driven System Calculation Results Using Monte Carlo Method Codes

Abstract A new design of a small power accelerator-driven system (ADS) using inert matrix fuel—ceramic–ceramic matrix (CERCER) and ceramic-metallic matrix (CERMET) was proposed in our previous work for high transmutation rate, small reactivity loss due to burnup, and enhanced safety features. The fast neutron spectrum in the ADS showed a benefit for reducing minor actinides inventories in the fuel cycle. However, the reliability of the nuclear cross section data of evaluated libraries, especially at high energy, needs to be investigated and its impact on CERCER and CERMET fueled ADS calculation results needs to be assessed. In this study, the impact of endf/b-vii.1 nuclear library uncertainty on the CERCER and CERMET fueled ADS reactivity calculation are evaluated using the Monte Carlo method (scale6.2 and whisper-1.1). The sensitivity coefficients and uncertainty obtained by the two codes show a good agreement. The results reveal the uncertainty of the endf/b-vii.1 nuclear data caused significant uncertainty in the reactivity calculations (about 1000 pcm) and it is particularly higher in the CERMET fueled core due to a harder neutron spectrum. The paper identifies the major contributes—isotopes and related cross sections—to the core reactivity uncertainty by suggesting the enhancement of these nuclear data.

A design study on a metal fuel fast reactor core for high efficiency minor actinide transmutation by loading silicon carbide composite material

ABSTRACT A metal fuel fast reactor core for high efficiency minor actinide (MA) transmutation was designed by loading silicon carbide composite material (SiC/SiC) which can improve sodium-cooled fast reactor (SFR) core safety characteristics such as sodium void reactivity worth and Doppler coefficient due to neutron moderation. Based on a 750 MWe metal fueled SFR core concept designed in a prior work, the reactor core loading fuel subassemblies with SiC/SiC wrapper tubes and moderator subassemblies was designed. To improve the reactor core safety characteristics efficiently, three layers of SiC/SiC moderator subassemblies were loaded in the core by replacing 108 out of 393 fuel subassemblies with the moderators. The reactor core with approximately 20 wt% MA-containing metal fuel satisfied all safety design criteria and achieved the MA transmutation amount as high as 420 kg/GWe-y which is twice as high as that of the axially heterogeneous core with inner blanket and upper sodium plenum, and two-thirds of that of the accelerator-driven system.

Investigation of the density shoulder formation by using self-consistent simulations of plasma turbulence and neutral kinetic dynamics

Simulations of plasma turbulence in a lower single-null magnetic configurations are presented. The plasma dynamics is modelled by the drift-reduced two-fluid Braginskii equations that are coupled with a kinetic model for a single neutral species that considers ionization, charge–exchange, recombination and elastic collisions. The effect of increased core fuelling on the plasma scrape-off layer (SOL) density and temperature profile is investigated. The increase in core fuelling leads to an increase of the e-folding length in the near SOL and, in the far SOL, to an increase of the plasma density. These results are in agreement with experimental measurements, and in particular with the observations of the formation of a density shoulder in high-fuelling scenarios. The physical mechanisms underlying the increase of the far SOL density are analysed comparing parallel and perpendicular fluxes in the SOL and considering also simulations with similar parameters but without neutrals. Despite the increase of the blob size, the increase of the far SOL density observed at high-fuelling rates is found to be mainly caused by the strong decrease of parallel transport, due to the cooling of electrons resulting from ionization events.

Performance Improvement for Multi-recycling HTGR Incorporated with MA burning

Multi-recycling HTGR has been investigated by JAEA in order to reduce the environmental burden and non-proliferation of Pu. In the previous study, it is found that all actinoids except neptunium, which is not problematic from the viewpoint of toxicity and proliferation, can be recycled. However, the cycle length is slightly decreased compared with uranium fueled core due to the cumulated fertile TRU in the fuel composition. In the present study, Pu recycling HTGR is designed to incorporate with MA burning. As a result, the cycle length is increased by approximately 15% compared with TRU multi-recycling core, and if the MA burner can be achieved by IFR, the cost is decreased by 0.14 yen/kWh.

Neutron Cross Section Generation for PWR MOX Fuel Assemblies with SCALE and Serpent

In this study, the SCALE/TRITON code (based on deterministic method) and the Serpent 2 code (based on Monte Carlo method) were utilized to prepare the group constants of the pressurized water reactor (PWR) mixed-oxide (MOX) fuel assemblies for transient analyses of PWR MOX fueled cores in normal operation and control rod ejection accident condition with 3D reactor kinetics codes. The PWR MOX fuel assemblies were modeled with TRITON and Serpent, and their infinite neutron multiplication factors (k-inf) versus burnup and respective two-group neutron cross sections were calculated and compared against the available benchmark data obtained with the HELIOS code. The comparative results generally show a good agreement between TRITON and Serpent with HELIOS within 643 pcm for the k-inf values and within 5% for the two-group neutron cross sections. Therefore, it indicates that the TRITON and Serpent models developed herein for the PWR MOX fuel assemblies can be applied to group constant generation to be further used in transient analyses of PWR MOX fueled cores.

Обоснование требований к объему пускового комплекса технических средств радиационного контроля при физическом пуске энергоблоков Нововоронежской АЭС-2

The physical start-up is a stage of the power unit commissioning, which includes reactor core fuelling, achieving the first criticality and completion of all required tests and measurements at the power level. At the stages of physical start-up the priority task is to determine the scope of radiation monitoring equipment, put into operation and sufficient for absolute safety assurance and unambiguous assessment of the power unit radiation exposure of personnel, population and environment.

Study on Control Rod Reactivity of Small Pebble Bed Reactor with Wallpaper Fuel Design

The control rod reactivity is a crucial reactor parameter designed to shut down the reactor and to control the safety of reactor core under any operation conditions. This paper discusses the control rod reactivity of small pebble bed reactor with wallpaper fuel design. The geometric configuration of RDE was selected for the reactor core model. A number of simulations with Monte Carlo n-particle transport code MCNP6 based on evaluated nuclear data library ENDF/B-VII were employed to investigate the total and individual control rod worth and reactivity in stuck rod condition. In the simulation model, the total amount of uranium per pebble was kept constant at 5 g by varying the hexagonal lattice pitch calculated for different radius of the central graphite zone. The simulation results show that the total control rod achieves the best effectiveness on the central graphite radius of 1.0 cm. The highest worth of individual rods in the wallpaper fueled core are almost greater than that of standard fueled core. The stuck rod criterion in all simulation results confirm that, the failure of one rod with highest worth will not prevent the control system from shutting down the reactor. These results conclude that selecting a central graphite radius of wallpaper fuel is one of the strategies to be carefully considered and a radius of 1.0 cm is the best one in controlling reactor safety and ensuring the control system could shut down the reactor safely.

压水堆核电站实心燃料堆芯向环形燃料堆芯过渡方案研究

压水堆核电站实心燃料堆芯向环形燃料堆芯过渡方案研究

CALCULATION OF 2-DIMENSIONAL PWR MOX/UO2 CORE BENCHMARK OECD NEA 6048 WITH SRAC CODE

The mixed uranium-plutonium oxide fuel (MOX/UO2) is an interesting fuel for future power reactors. This is due to the large amount of plutonium that can be processed from spent fuel of nuclear plants or from plutonium weapons. MOX/UO2 fuel is very flexible to be applied in thermal reactors such as PWR and it is more economical than UO2 fuel. However, due to the different nature of neutron interactions of MOX in PWR, it will change the reactor core design parameters and also its safety characteristic. The purpose of this study is to determine the accuracy of SRAC2006 code system in generation of cross-sections and calculation of reactor core design parameters such as criticality, reactivity of control rods and radial power distribution. In this study, PWR MOX/UO2 Core Transient Benchmark is used to verify the code that models a MOX/UO2 fueled core. SRAC-CITATION result is different from DeCART by 0.339% from. SRAC-CITATION result of single rod worth in All Rods Out (ARO) conditions are quite good with a maximum difference of 6.34% compared to BARS code and 4.74% compared to PARCS code. In All Rods In (ARI) condition, SRAC-CITATION results compared to the PARCS code is quite good where the maximum difference is 9.72%, but compared to BARS code, it spikes up to 33.24% at maximum difference. In the other case, overall radial power density results are quite good compared to the reference. Its maximum deviation from DeCART code is 5.325% in ARO condition and 6.234% in ARI condition. Based on the results of these calculations, SRAC code system can be used to generate cross-section and to calculate some neutronic parameters. Hence, it can be used to evaluate the neutronic parameters of the MOX/UO2 PWR core design.Keywords: MOX/UO2 fuel, Criticality, Power peaking factor, SRAC2006

Assessment of CASL VERA for BWR analysis and application to accident tolerant SiC/SiC channel box

Application of the Virtual Environment for Reactor Applications (VERA) to BWR analysis is assessed in this paper by comparing results to those calculated using other widely-used modeling tools, namely the U.S. Nuclear Regulatory Commission’s PARCS/PATHS and the Serpent Monte Carlo particle transport code. Additionally, VERA is used to calculate 3-D temperature and fast neutron flux distributions in silicon carbide (SiC) fiber-reinforced, SiC matrix composite (SiC/SiC) BWR channel boxes, which are being studied as an Accident Tolerant Fuel core structural material concept. The code-to-code comparisons were favorable, and the SiC/SiC channel box evaluation demonstrates the many advanced modeling features VERA provides while also highlighting the non-uniformity in fast neutron flux distributions that can play a role in potential SiC/SiC channel box deformation. Traditional BWR analysis tools do not have the calculation fidelity necessary for coupled assessment of flux and temperature gradients in a SiC/SiC channel box. VERA is a state-of-the-art modeling environment that was developed to increase the safety and economic competitiveness of nuclear power through improved modeling accuracy. While VERA has already been deployed in the nuclear industry for PWR applications, the current study is a vital initial step in the extensive development, validation, and verification that VERA must go through to be useful for BWR applications.

Conceptual design and uncertainty analysis of a typical 50 kW aqueous homogeneous reactor aimed for medical isotope production

Production of medical isotopes such as 99Mo, 90Sr, 131I and other fission products using Aqueous Homogeneous Reactors (AHRs) offers several advantages in comparison to other methods. In this method, after about one week operation, required isotopes can be directly extracted from the solution and the remaining fuel solution will be returned to the core for the next campaign. Moreover, AHRs are inherently safe and reliable due to their large negative reactivity feedbacks. This paper demonstrates technical features and conceptual scheme of the Iranian homogeneous reactor named RAHA, which is mainly used for the production of 99Mo/99mTc. Neutronic and thermal-hydraulic characteristics of the reactor were analyzed with MCNPX-2.6 and RELAP5/Mod3.3 codes, respectively. It was assumed that the reactor is fueled by Uranyl Nitrate, and is able to remove generated heat using natural circulation. The coolant circulation velocity was obtained as 6.5 cm/s. Four different types of neutron absorbers were considered as control rod material, and BORA resin with a diameter of 1.4 cm and Stainless-Steel cladding was chosen. As a new characteristic which may improve safety of the system, the possibility of reactor startup without any external neutron source for the case of fresh fueled core was assessed. Finally, effects of the possible uncertainties on some physical and material properties of the core were investigated and sensitivity of the effective multiplication factor relative to the uncertain parameters was also extracted using Spearman's formula.

Neutronic investigation of the thorium-based mixed-oxide as an alternative fuel in the TRIGA Mark-II research reactor – Part I: A beginning of life calculations

This article presents a neutronic investigation of the effect of replacing some uranium zirconium hydride (U-ZrH1.6) fuels with thorium-based mixed-oxide ((Th,U)O2) fuels in the TRIGA Mark-II research reactor without changing the core geometry, while the only variable is the fuel material.The possibility of utilizing thorium mixed oxide fuel in the TRIGA Mark-II reactor is assessed using the famous seed-blanket configuration or also known as Radkowsky Thorium Fuel (RTF) concept. Using the RTF concept, a long core lifetime can be achieved by enhancing the internal conversion rate of fertile to fissile materials, thanks to the thermal capture cross-section of 232Th that is around three times that of 238U. Additionally, on the fuel cycle aspect of using the thorium-based fuel is that the thorium cycle produces much less plutonium and other transuranic radioactive elements than uranium fuel cycle.Actually, neutronic and kinetic calculations, at the beginning of life conditions, for nine different seed-blanket configurations have been investigated using the MCNPX2.6 code. These calculations for all proposed configurations in comparison with the conventional U-ZrH1.6 fueled core were performed to pick out an optimal core(s) design ensuring the most efficient use of (Th,U)O2 fuel in it and with minimum changes in the main safety parameters. The parameters consist of effective multiplication factor keff, delayed neutron fraction βeff, neutron flux distribution, power distribution and radial hot rod power peaking factor (FHR), and shutdown margin (SDM).The calculation results exhibit that the highest initial criticality achieved is 1.07716. The results of the other neutronic investigations of thorium-based mixed-oxide fuel elements and their positions, including their numbers in the core are summarized and discussed in the following paper.

On benchmarking of simulations of particle transport in ITER

We report results of benchmarking of core particle transport simulations by a collection of codes widely used in transport modelling of tokamak plasmas. Our analysis includes formulation of transport equations, difference between electron and ion solvers, comparison of modules of the pellet and edge gas fuelling on the ITER baseline scenario. During the first phase of benchmarking we address the particle transport effects in the stationary phase. Firstly, simulations are performed with identical sources, sinks, transport coefficients, and boundary conditions prescribed in the flattop H-mode phase. The transformation of ion particle transport equations is introduced so to directly compare their results to electron transport solvers. Secondly, the pellet fuelling models are benchmarked in various conditions to evaluate the dependency of the pellet deposition on the pellet volume, injection side, pedestal, and separatrix parameters. Thirdly, edge gas fuelling is benchmarked to assess sensitivities of source profile predictions to uncertainties in plasma conditions and detailed model assumptions. At the second phase, we address particle transport effects in the time-evolving plasma including the current ramp-up to the ramp-down phase. The ion and the electron solvers are benchmarked together. Differences between the simulation results of the solvers are investigated in terms of equilibrium, grid resolution, radial coordinate, radial grid distribution, and plasma volume evolution term. We found that the selection of the radial coordinate can yield prominent differences between the solvers mainly due to differences in the edge grid distribution. The simulations reveal that electron and ion solvers predict noticeably different density peaking for the same diffusion and pinch velocity while with the peaked profile of helium, expected in fusion reactors. The fuelling benchmarking shows that gas puffing is not efficient for core fuelling in H-modes and density control should be done by the high field side pellet injection in contrast to present machines.

Addressing the feasibility of inboard direct-line injection of high-speed pellets, for core fueling of DEMO

Pellet injection represents, to date, the most promising option for core fuelling of the EU-DEMO tokamak. Simulations with the HPI2 pellet ablation/deposition code indicate, however, that sufficiently deep fuel deposition requires injection from the High Field Side (HFS) at velocities ≳1 km/s. Two complementary inboard injection schemes are being explored: one makes use of guide tubes with curvature radii ≥6 m in the attempt of preserving pellet integrity at speeds of ˜1 km/s, the other is investigating the feasibility of injecting high-speed (˜3 km/s) pellets along “direct line of sight” (DLS) trajectories, from either the HFS or a vertical port. Options using quasi-vertical DLS paths routed across the upper vertical port have been explored first, as they can be more easily integrated, Unfortunately, the radial position of the available vertical access (≳9 m from the machine axis) turns out to be unfavorable; further simulations with the HPI2 code predict indeed that vertical injection may be effective only if pellets trajectories are well inboard the magnetic axis. High-speed injection through oblique inboard “DLS” paths, not interfering with the Central Solenoid (CS), are instead predicted to yield good performance, provided that the injection location is ≲2.5 m from the equatorial mid-plane. The angular spread of high-speed free-flight pellets, recently measured using an existing facility, turns out to be enclosed within ˜ 0.7°. This scatter cone may require significant cut off volume of the Breeding Blanket (BB). Moreover, DLS in-vessel conical penetrations may increase the neutron flux outside of the bio-shield, and also result in a significant heat load in the cryogenic pellet source. These issues are being investigated, to identify suitable shielding strategies; preliminary results are reported. The suitability of straight guide tubes to reduce the scatter cone, and hence the corresponding open cross section on BB penetration and the neutron streaming, will be explored as a further step.

Study on MOX Core Characteristics of Experimental Power Reactor using MCNP6 Code

In experimental power reactor (reaktor daya eksperimental, RDE), a full MOX fueled core can be accommodated in its operation without significant modifications on the reactor core design, but the MOX core tends to produce less favorable safety features and deliver transient behavior into unwanted fatal accidents. This paper aimed to investigate the MOX core characteristics of RDE through a series of calculations with MCNP6 code and ENDF/B-VII library. The calculation results show that MOX core with 235U enrichment of above 5 % can reach the criticality condition, maintain and run the reactor during the operation cycle. Utilizing MOX fuel in RDE with lower 235U enrichment will have a more negative impact on temperature coefficient of reactivity. The lower βeff makes the MOX core is more difficult to control, especially with low 235U enrichment. These results conclude that the selection of 235U enrichment in the MOX core must be carefully considered because it is one of the strategies to ensure the reactor safety criteria are fulfilled.

Uncertainty quantification of sodium-cooled fast reactor based on the UAM-SFR benchmarks: From pin-cell to full core

To quantify the uncertainties in the neutronics calculations of sodium-cooled fast reactor (SFR), a series of sensitivity and uncertainty (S&U) analysis were performed on two typical SFRs, which were a large-size MOX-fueled core and a medium-size metallic fueled core. The analysis was done from the pin-cell level to full core level to investigate the uncertainty propagation in the two-step neutronics calculation scheme. The responses analyzed contained the kinf, homogenized cross-sections, sodium void reactivity (SVR), Doppler constant (KD) and control rod worth (CRW) at the sub-assembly level, and the keff and power distribution at the full core level. The direct numerical perturbation method was used to get the relative sensitivity coefficients and the ‘Sandwich Rule’ was used to get the relative uncertainties. The results showed that the relative uncertainties propagated linearly from pin-cell to full core for the kinf and keff. Relative uncertainties of reactivity like SVR were much bigger than the ones of kinf and keff. However, the relative uncertainty of power distribution was smaller than other responses. Differences were observed in the S&U analysis between the MOX fueled and metallic fueled SFRs. Softer spectrum of the large-size MOX fueled core caused bigger uncertainty compared to the metallic fueled core.