Search

This paper presents the development of a lattice physics–to–core simulator two-step procedure for the rapid analysis of the Advanced High Temperature Reactor (AHTR). Lattice physics, reflector, and...

Molten salt reactors (MSRs) are a class of next-generation nuclear reactors that have received recent industrial and research interest. A generalized species transport solver was implemented in the Virtual Environment for Reactor Applications (VERA) computing suite to extend this tool to analyze liquid-fueled MSRs. This core simulator has been extended to model the transport of fission product gases into a collection of circulating gas bubbles with the purpose of removing the gases. This paper presents the governing species transport equation, along with various nuclear source terms. Development of the source term for phase migration is discussed, along with a simplified interfacial area tracking method. Finally, a case study on a simplified MSR loop is presented in which modeling parameters were varied to assess their impact on gas removal. The steady state results show that parameters such as bubble diameter, gas injection rate and mass transfer coefficient have a low to moderate effect on the fraction of xenon in the core region. Removal efficiency has the greatest effect on the fraction in the core region. After the pump bowl, bubble diameter has a minor effect on the fraction of xenon in the gas void. These results point out that increasing parameters such as mass transfer coefficient, gas injection rate, and removal efficiency drives the xenon into the circulating gas void, while decreasing bubble diameter also drives xenon into the gas void by increasing interfacial area.

The development and demonstration of fusion power plant facilities require advancement of multiphysics modeling capabilities. These capabilities provide an integrated platform for design and analysis of a fusion engineering demonstration power plant. We present a state-of-the-art literature review of the status and needs of multiphysics modeling, including neutronics analysis, and thermal hydraulic/mechanical simulation tools in support of fusion powerplant system design and integration. In this paper we use the Fusion Nuclear Science Facility (FNSF) concept as an example system to assess future needs for advancement of multiphysics modeling capabilities. Specifically, we discuss the latest progress on R&D missions in the literature which are specific to the neutronics and thermal hydraulics design of the breeding blanket concept chosen for the FNSF, the Dual Coolant Lead Lithium (DCLL) blanket. We identify future R&D needs to be addressed through the development of multiphysics modeling capabilities that tie fusion neutronics to tritium breeding ratio (TBR), neutron wall loading (NWL), material damage and transmutation, shutdown dose, and safety assessments. The capabilities will support the design and licensing of a fusion engineering demonstration facility as well as a future engineering demonstration plant. We conclude that more integrated and prototypic tritium experiments must be developed to establish the database needed for the verification and validation of tritium transport simulations. Robust sensitivity and uncertainty (S/U) analysis must be integrated with the facility design to identify the significance of influencing parameters involved in tritium transport, along with quantification of their uncertainties towards output parameters. This enables design optimization of the DCLL blanket during the R&D phase for a fusion engineering demonstration facility. Additional separate- and integral-effects experiments for the validation of the multiphysics analysis is needed for particular components of interest for the DCLL blanket, such as the flow channel inserts (FCIs), along with multi-fidelity simulations which incorporate the magnetohydrodynamic (MHD) effects, and coupled analysis techniques that can capture the interactions between the thermal hydraulic effects and the tritium transport behavior. Simultaneously, the systematic S/U analysis for the multiphysics modeling framework must prove the reliability of the model to help define the technical gaps that need to be addressed by subsequent R&D activities for a fusion engineering demonstration facility. Lastly, S/U analysis of the uncertainties in design parameters for the evaluation of steady and transient safety analysis would be the final requirement that will elucidate the importance of these design parameters, as well as the quantification of their uncertainties on several identified key Figures of Merit (FoM) applicable to the safety analysis to provide guidance for the final design optimization of the DCLL.

A method of fuel injection rate shaping of the Diesel engine common rail fuel system with common rail injectors and solenoid control is proposed. The method envisages the impact on control current of impulses applied to the control solenoid valve of the common rail injectors for variation of the injection rate shape. At that, the fuel is supplied via two groups of injection holes. The entering edges of the first group with the coefficient of flow, ??B, were located in the sack volume and the entering edges of the second group (coefficient of flow, ??H) - on the locking taper surface of the nozzle body. The coefficients of flow, ??B, and ??H differ considerably and depend on the valve needle position. This enables to adjust the injection quantity by injection holes taking into account operating conditions of the Diesel engine and hence - by the combustion chamber zones. Using the constant fuel flow set-up, characteristic of the effective cross-section of the common rail fuel system injector holes was investigated. The diameter of injector holes was 0.12 ? 0.135 mm. The excessive pressure at the entering edges varied from 30 to 150 MPa and more and the excessive pressure in the volume behind the output edge - from 0 to 16 MPa.

This paper presents results obtained by reproducing a VVER-1000 lattice physics benchmark using the Serpent2 code. The benchmark data are related to the initial cycles of the Khmelnitskiy nuclear power plant. This study compares results obtained with the Serpent2 code against those calculations published in the benchmark document. Detailed descriptions are provided for two fuel assemblies; one without gadolinium in the fuel pins, and the other with gadolinium. A sensitivity study is performed utilizing different nuclear data libraries, different depletion sequences (including pin by pin burnup) and different number of radial regions within the fuel pins containing gadolinium. The infinite multiplication factor behavior and isotopic concentrations of major isotopes are compared and discussed. It is concluded that Serpent2 provided reasonable results in comparison with the other codes used in the benchmark.

The proper calculation of core periphery discontinuity factors is important for accurate modeling when using an advanced nodal diffusion simulator. In cores with hexagonal assemblies, such as in the VVER-1000, most fuel assemblies share two faces with the radial reflector, and some even three faces. For this reason, use of a two-dimensional (2D) reflector model will more accurately capture the neutron physics near the core periphery. This article illustrates key points related to the use of 2D discontinuity factors in the reflector region, first by using an algorithm that applies the methodology proposed by Mittag et al. (2003) after correcting some minor typographical errors in the original publication, and then by employing the SCALE transport module NEWT to compute the appropriate quantities. Large and even negative discontinuity factors are an acceptable fact of this methodology when the diffusion approximation becomes invalid due to the problem’s localized features and the large flux gradients.

This work applies to recent initiatives at the Radiochemical Engineering Development Center at Oak Ridge National Laboratory to optimize the production of transcurium isotopes in the High F...

The Monte Carlo code KENO contains thermal scattering data for a wide variety of thermal moderators. These data are processed from Evaluated Nuclear Data Files (ENDF) by AMPX and stored as double differential probability distribution functions. The method examined in this paper uses S ( α , β ) probability distribution functions derived from the ENDF data files directly instead of being converted to double differential cross sections. This allows the size of the cross section data on the disk to be reduced substantially amount. KENO has also been updated to allow interpolation in temperature on these data so that problems can be run at any temperature. Results are shown for several simplified problems for a variety of moderators. In addition, benchmark models based on the KRITZ reactor in Sweden were run, and the results are compared with the previous versions of KENO without the direct S ( α , β ) method. Results from the direct S ( α , β ) method compare favorably with the original results obtained using the double differential cross sections. Sampling the data increases the run-time of the Monte Carlo calculation, but memory usage is decreased substantially.

Detailed three-dimensional (3D) spent nuclear fuel (SNF) isotopic composition analysis using a full-core 3D nodal simulator is not a widespread practice within production-level nuclear analysis fuel management codes. A 3D full-core pin-wise detailed isotopic analysis can be very time consuming, so more often than not, reduced geometries are used, either based on two-dimensional (2D) lattice-physics, single fuel pins, or point-reactor (0D) estimations. In fact, the typical macroscopic cross section based codes do not usually have the ability to evaluate the full detailed composition of SNF, but instead may track only a few isotopes in spatially homogenized regions. Accordingly, a microscopic cross section based depletion sequence was revisited within the otherwise macroscopic-based NESTLE code. To illustrate this capability, the plutonium content of SNF within a VVER-1000 benchmark was calculated and its non-proliferation risk was estimated. The total plutonium masses and 239Pu fractions were calculated on a 3D node-by-node basis using an established VVER-1000 NESTLE benchmark model. The total plutonium mass after the first cycle is estimated at 502.6 kg, and it increased to 523.4 kg after 90 days of decay. The mass of plutonium with a 239Pu content greater than 80% at the end of cycle is about 7.8 kg, while all remaining content shows to have a 239Pu fraction below 87%. The plutonium content with a 239Pu fraction greater than 80% after 20 days of the end of cycle is less than 7.9 kg. This article highlights a feature of the NESTLE code that can help make isotopic assessments radially and axially for fuel assemblies located in different regions of the core, thus accounting for the varying 3D conditions and flux distributions as a function of exposure.

The assessment of economic viability of a new reactor concept is crucial particularly during the early stages of its concept development. The G4-ECONS methodology provides a standardized top-down estimate of electricity cost and parametric sensitivities, not specifically targeted toward an accurate prediction of the final cost when deployed, but rather seeking an approximation of cost variations relative to other systems. This study presents an analysis of the Integral Molten Salt Reactor (IMSR) concept in comparison with a consistent analysis of an advanced PWR reactor (represented by AP1000). Estimation of levelized unit electricity costs, as well as sensitivity analyses to the discount rate and uranium or SWU prices, are presented using this methodology.

In this paper, we introduce a quick method for improving the accuracy of Monte Carlo simulations by generating one- and two-dimensional cross sections at a user-defined temperature before performing transport calculations. A finite difference method is used to Doppler-broaden cross sections to the desired temperature, and unit-base interpolation is done to generate the probability distributions for double differential two-dimensional thermal moderator cross sections at any arbitrarily user-defined temperature. The accuracy of these methods is tested using a variety of contrived problems. In addition, various benchmarks at elevated temperatures are modeled, and results are compared with benchmark results. Lastly, the problem-dependent cross sections are observed to produce eigenvalue estimates that are closer to the benchmark results than those without the problem-dependent cross sections.

Three-dimensional (3D) full-core calculations are an integral part of fuel reload design for today’s light water reactors (LWRs). The current approaches are typically based on the nodal diffusion codes that calculate criticality state points and power distributions as a function of burnup. The CASL VERA (Consortium for Advanced Simulation of Light Water Reactors, Virtual Environment for Reactor Applications) code represents one of the latest advancements in 3D full-core calculation analysis based on transport theory methods employing the Method of Characteristics (MOC) and coupled multi-physics. In this article, a publicly released version of the NuScale reactor core is analysed with the VERA code (version 3.9 and 4.1) and contrasted against some static Serpent and Polaris based simulations. The analysis shows an excellent agreement for the lattice-level calculations as well as with some of the 3D full-core models. However, larger deviations were found in cases with heavy reflector models, whereby the reflector composition was found to impact the differences between the VERA and Serpent results. In this analysis, it was determined that greater than 90% stainless steel content in the heavy reflector leads to higher deviations between the VERA results and the Serpent results. The burnup calculations showed that the presence of the heavy reflector extends the cycle length and also leads to a flatter power distribution in the core, which can generally be interpreted as more efficient.

Safety analyses at the high flux isotope reactor (HFIR) are required to qualify experiment targets for the production of plutonium-238 (238Pu) from neptunium dioxide/aluminum cermet (NpO2/Al) pellets. High heat generation rates (HGRs) due to fissile material and low melting temperatures require a sophisticated set of steady-state thermal simulations in order to ensure sufficient safety margins. These simulations are achieved in a fully coupled thermo-mechanical analysis using comsolmultiphysics for four different preliminary target designs using an evolving set of pre- and postirradiation data inputs, and subsequently evolving solution scopes, from the unique pellet and target designs. A new comprehensive presentation of these preliminary analyses is given and revisited analyses of the first prototypical target designs are presented to reveal the effectiveness of evolving methods and input data.

Uncertainty quantification is an indispensable analysis for nuclear reactor simulation as it provides a rigorous approach by which the credibility of the predictions can be assessed. Focusing on propagation of multi-group cross-sections, the major challenge lies in the enormous size of the uncertainty space. Earlier work has explored the use of the physics-guided coverage mapping (PCM) methodology to assess the quality of the assumptions typically employed to reduce the size of the uncertainty space. A reduced order modeling (ROM) approach has been further developed to identify the active degrees of freedom (DOFs) of the uncertainty space, comprising all the cross-section few-group parameters required in core-wide simulation. In the current work, a sensitivity study, based on the PCM and ROM results, is applied to identify a suitable compressed representation of the uncertainty space to render feasible the quantification and prioritization of the various sources of uncertainties. While the proposed developments are general to any reactor physics computational sequence, the proposed approach is customized to the TRITON-NESTLE computational sequence, simulating the BWR lattice model and the core model, which will serve as a demonstrative tool for the implementation of the algorithms.

A new computer code for performing BWR in-core and out-of-core fuel cycle optimization for multiple cycles simultaneously has been developed. Parallel simulated annealing (PSA) is used to optimize the new fuel inventory and placement of new and reload fuel for each cycle considered. Several algorithm improvements were implemented and evaluated. The most significant of these are variable sampling probabilities and sampling new fuel types from an ordered array. A heuristic control rod pattern (CRP) search algorithm was also implemented, which is useful for single CRP determinations, however, this feature requires significant computational resources and is currently not practical for use in a full multi-cycle optimization. The PSA algorithm was demonstrated to be capable of significant objective function reduction and finding candidate loading patterns without constraint violations. The use of variable sampling probabilities was shown to reduce runtime while producing better results compared to using constant sampling probabilities. Sampling new fuel types from an ordered array was shown to have a mixed effect compared to random new fuel type sampling, whereby using both random and ordered sampling produced better results but required longer runtimes.

A description of the Atomic, Molecular and Optical Sciences (AMO) instrument at the Linac Coherent Light Source is presented. Recent scientific highlights illustrate the imaging, time-resolved spectroscopy and high-power density capabilities of the AMO instrument., The Atomic, Molecular and Optical Science (AMO) instrument at the Linac Coherent Light Source (LCLS) provides a tight soft X-ray focus into one of three experimental endstations. The flexible instrument design is optimized for studying a wide variety of phenomena requiring peak intensity. There is a suite of spectrometers and two photon area detectors available. An optional mirror-based split-and-delay unit can be used for X-ray pump–probe experiments. Recent scientific highlights illustrate the imaging, time-resolved spectroscopy and high-power density capabilities of the AMO instrument.

CONSTRUCCIÓN DE REDES DE ALCANTARILLADO SANITARIO EN EL CASCO URBANO DEL MUNICIPIO DE VILLAGARZON PUTUMAYO QUE BENEFICIA A LOS BARRIOS GALILEA Y LOS PASTOS, LEVANTAMIENTOS A PREDIOS DE PROPIEDAD PRIVADA DE LOS MUNICIPIOS VILLAGARZON MOCOA Y MIRAFLORES DE LOS DEPARTAMENTOS PUTUMAYO Y CAUCA., Tabla de Contenido Agradecimientos ........................................................................................................................ 3 Tabla de Ilustraciones ................................................................................................................ 6 Introducción ............................................................................................................................... 9 2. Objetivos .............................................................................................................................. 10 2.1 General ........................................................................................................................... 10 2.2 Específicos ..................................................................................................................... 10 3. Glosario ................................................................................................................................ 11 4. Capítulo I .............................................................................................................................. 12 4.1 Descripción de la Empresa ............................................................................................. 12 4.2 Razón social ............................................................................................................... 12 4.3 Actividad a la que se dedica ....................................................................................... 12 4.4 Misión ........................................................................................................................ 12 4.5 Visión ......................................................................................................................... 12 4.6 Organigrama ............................................................................................................... 12 5. Capítulo II ............................................................................................................................ 13 5.1 Lugar Donde se Realizó la Pasantía Laboral ................................................................. 13 6. Capítulo III ........................................................................................................................... 14 6.1 Descripción del Lugar de Pasantías ............................................................................... 14 7. Capítulo IV ........................................................................................................................... 15 5 7.1 Actividades Que se Realizaron en la Pasantía ............................................................... 15 7.2 Plan de trabajo ............................................................................................................ 15 8. Conclusiones ........................................................................................................................ 23, Pregrado, Tecnólogo en Topografía

This study evaluated the neutronics and some of the fuel cycle characteristics of using uranium-based fully ceramic microencapsulated (FCM) fuel in a pressurized water reactor (PWR). Specific PWR l...

This article presents a benchmark study of a VVER 1000 core simulation of hot zero power (HZP) tests. Measured experimental data from the Temelin Unit 2 nuclear power plant’s operation test during its first reactor startup were compared using MCNP as well as the combination of Serpent + NESTLE lattice-to-diffusion nodal simulation. Results from full-core 3D models included multiplication factor and moderator temperature coefficients (MTC) that were compared to experimental data, which revealed a very good agreement with criticality states during HZP tests, and a reasonable agreement for MTC. Therefore, it is concluded that the Serpent-NESTLE combination and MCNP, both, provided reasonable results for this benchmark. The Serpent-NESTLE code combination was used for the first time in this study and has shown to be capable of providing high quality simulations in an efficient fashion. To encourage future benchmarks by other researchers, this article makes a special effort to document the core design information and details that would facilitate such efforts.

The Radiochemical Engineering Development Center at Oak Ridge National Laboratory is the world’s leader in production of 252 Cf. This and other heavy actinides are produced by irradiation of mixed curium/americium targets in the High Flux Isotope Reactor. Due to the strong dependence of isotopic cross-sections upon incoming neutron energy, the efficiency with which an isotope is transmuted is highly dependent upon the energy spectrum and intensity of the neutron flux. There are certain energy ranges in which the rate of fission absorptions in feedstock materials is reduced relative to the rate of (n, γ) captures. Using a variety of computational models it is shown that by perturbing the flux spectrum, it is possible to alter the net consumption of curium feedstock, as well as the yields of key isotopes for the heavy element research program, such as 249 Bk and 252 Cf. This flux spectrum perturbation is accomplished by means of focused resonance shielding through the use of filter materials. It is further shown that these perturbations can alter the target yields in a significant way, increasing the amount of 252 Cf produced per unit curium consumption by over 40%.

Prostatitis is a very common urogenital disease of the male with prevalence ranging from 2.2 to 9.7% worldwide. Interestingly, some recent evidences have showed a significant association between chronic prostatitis (CP) and male infertility including a detrimental effect on sperm parameters, reduction of zinc concentration on semen sperm and production of anti-semen antibodies (ASAs). The aim of the current meta-analysis was to evaluate the association between CP and alteration of semen parameters. This analysis was conducted according to the Preferred Reporting Items for Systematic Reviews and Meta-analysis guidelines and we included in the final analysis 27 studies, with a total of 3241 participants, including 381 (11.75%) with chronic bacterial prostatitis (CBP), 1670 (51.53%) with chronic prostatitis/chronic pelvic pain syndrome (CP/CPPS) and 1190 (36.72%) controls. CBP was associated with reduction of sperm concentration, sperm vitality, sperm total and progressive motility, while CP/CPPS was related to the reduction of semen volume, sperm concentration, sperm progressive motility and sperm normal morphology. We found that CP was significantly associated with reduced zinc concentration on seminal plasma (SMD: −20.73; p = 0.005). Finally, CP statistically increased the risk of developing ASA on seminal plasma (OR 3.26; p

This article describes the light water reactor optimizer (LWROpt), a fuel cycle optimization code originally developed for BWRs, which has been adapted to perform core fuel reload and/or operational control rod management for pressurized water reactors (PWRs) and small modular reactors (SMRs), as well. Additionally, the eighth-core symmetric shuffle option is introduced to help expedite large-scale optimizations. These new features of the optimizer are tested by performing optimizations starting from a base case of an SMR core model that was developed manually and unrodded. The new fuel inventory (NFI) and loading pattern (LP) search in LWROpt was able to eliminate all of the constraint violations present in the initial base solution. However, independent control rod pattern (CRP) searches for the best several LPs found were not successful in generating CRPs without any constraint violations. This indicates that fully decoupling the fuel loading from the CRP optimization can increase the computational tractability of these calculations but at the expense of effectiveness. To improve on the individual search results, a coupled fuel loading (NFI and LP) and CRP search was performed, which produced a better overall result but still with some small constraint violations, emphasizing the fact that optimizing the fuel loading arrangement in a small high-leakage unborated core while concurrently determining its operational rod patterns for a 4-year operational cycle is no easy feat even to an experienced core designer; thus, this process can be greatly aided by employing automated combinatorial optimization tools.

The High Flux Isotope Reactor located at the Oak Ridge National Laboratory utilizes a large cylindrical beryllium reflector that is subdivided into three concentric regions and encompasses the compact reactor core. Nuclear transmutations caused by neutron activation occur in the beryllium reflector regions, which leads to unwanted neutron absorbing and radiation emitting isotopes. During the past year, two topics related to the HFIR beryllium reflector were reviewed. The first topic included studying the neutron poison (helium-3 and lithium-6) buildup in the reflector regions and its affect on beginning-of-cycle reactivity. A new methodology was developed to predict the reactivity impact and estimated symmetrical critical control element positions as a function of outage time between cycles due to helium-3 buildup and was shown to be in better agreement with actual symmetrical critical control element position data than the current methodology. The second topic included studying the composition of the beryllium reflector regions at discharge as well as during decay to assess the viability of transporting, storing, and ultimately disposing the reflector regions currently stored in the spent fuel pool. The post-irradiation curie inventories were used to determine whether the reflector regions are discharged as transuranic waste or become transuranic waste during themore » decay period for disposal purposes and to determine the nuclear hazard category, which may affect the controls invoked for transportation and temporary storage. Two of the reflector regions were determined to be transuranic waste at discharge and the other region was determined to become transuranic waste in less than 2 years after being discharged due to the initial uranium content (0.0044 weight percent uranium). It was also concluded that all three of the reflector regions could be classified as nuclear hazard category 3 (potential for localized consequences only).« less