Steady-state Core Research Articles

Coupling of CTF and RELAP5-3D Within an Enhanced-Fidelity Nuclear Power Plant Simulator

A robust and accurate multiphysics engineering simulator is being developed to model the core behavior and system response of pressurized water reactors. This simulator relies on the NESTLE and CTF computer codes to model the neutronics and thermal hydraulics (TH), respectively, inside the core on a nodal scale and on the Reactor Excursion and Leak Analysis Program—Three Dimensional (RELAP5-3D) to model the entire nuclear steam supply system. The RELAP5-3D model includes highly detailed nodalization and multidimensional flow modeling throughout the vessel. Previously, pin-resolved data generated via the Virtual Environment for Reactor Analysis core simulator were used to improve the accuracy of the NESTLE core predictions. The engineering simulator being developed as part of this work uses the 3KEYMASTER platform to couple the enhanced NESTLE model to a nodal-fidelity CTF model to balance run time with accuracy; NESTLE provides node-dependent powers to CTF, and CTF provides node-dependent coolant densities and fuel temperatures to NESTLE. An overlapping domain approach is used for the core TH in which RELAP5-3D provides core boundary conditions based on the system response and CTF provides a node-dependent coolant heating rate to the RELAP5-3D core solution. In the preliminary TH demonstration discussed in this paper, CTF and RELAP5-3D provided similar steady-state core predictions, indicating the hydraulic compatibility between the codes, as well as reasonable and expected behavior under hypothetical transient conditions. This provides an initial step in ongoing efforts toward a robust, multiscale TH/neutronics engineering simulator capability.

An Efficient High-to-Low Iterative Method for Light Water Reactor Analysis Based on NEAMS Tools

The so-called two-step method involving the consecutive lattice physics and core simulation has been successfully and widely used in large-scale nuclear reactor calculations thanks to its superior computational efficiency and a satisfactory level of accuracy. However, its performance is challenged by the ever-increasing level of heterogeneity in core designs due to the use of infinite lattice approximation in the lattice calculation and its inability to update cross-section sets according to the core environment change. This paper introduces an alternative approach for light water reactor steady-state core analysis. During the core calculation process, iterations between the local lattice transport calculation and the global core nodal calculation are conducted. These iterations continuously update the boundary condition applied to the lattice model and generate updated cross-section sets. This is done through the iteration between the local lattice transport calculation and the global core nodal simulation. The neutronics high-to-low (Hi2Lo) scheme was formulated using Nuclear Energy Advanced Modeling and Simulation or NEAMS codes, in particular, with the modified PROTEUS-MOC and PROTEUS-NODAL serving as the transport lattice solver and full-core nodal solver, respectively. The verification of the implemented Hi2Lo iterative scheme on the two-dimensional C5G7-TD benchmark problem shows that the Hi2Lo scheme outperforms the two-step approach in terms of prediction accuracy for the key responses of interest (e.g., the system eigenvalue and power distribution) at a computational cost lower than that of the direct full-core transport calculation. To further improve its efficiency, an acceleration method has been developed and implemented for the Hi2Lo approach, and the results indicate that the acceleration method can significantly reduce the run time of a full-core transport solution by a factor of 14 while generating solutions with comparable accuracy.

Modeling of sodium-cooled fast reactor core – A sensitivity study on number of core channels and axial meshes

System thermal–hydraulic codes are used in the dynamics analysis of sodium-cooled fast reactors (SFRs). In such simulations, the accuracy of the reactor core model is very important to establish plant safety. Coupled physical phenomena, namely, heat transfer, fluid dynamics, and neutronics, need to be simulated in the core. Due to the complexity, a simplified core modeling approach has been traditionally followed in system dynamics codes. Firstly, steady-state core neutronics calculations are carried out using a 2D axisymmetric model to obtain perturbation worth and power distributions. Later, this data is mapped to several core channels (usually 10 to 20), each representing a set of fuel subassemblies (SAs) for thermal-hydraulics calculations.In this paper, a more sophisticated core modeling approach has been followed to understand the importance of such an exercise. A 3D neutronics model was used to obtain subassembly-wise perturbation worths and power distributions. For thermal-hydraulics calculations, the number of core channels was increased with the idea of one subassembly per channel. The number of axial meshes in each channel was also increased. The traditional and new modeling approaches were used to simulate the benchmark problem of FFTF LOFWOS test#13 for comparison. Steady-state parameters from both the new and the old models were comparable. Notably, there was a peaking of the axial power profile in the new model compared to the old model, which resulted in slightly higher peak fuel temperatures in the new model. However, the difference in the evolution of transient parameters between the two models was insignificant, and both models compared reasonably well with the benchmark data.

Validation of theory-based integrated modeling and new insights for a high-performance steady-state scenario with only RF heating on EAST

Theory-based integrated modeling is validated against high-performance steady-state core plasmas on EAST in the high poloidal beta (β p) regime with only RF heating. Reasonably good agreement between the modeling results and experimental measurements is obtained not only for the temperature profiles but also for the 11-chord line-integrated densities and Faraday angles for the first time. This validation effort demonstrates that the safety factor profiles can be non-reversed in high β p experiments on EAST. The inaccessibility for LH waves observed in conventional ray-tracing simulations for some high β p experiments is effectively mitigated by including the modeling of wave propagation in the scrape-off layer. The observed confinement improvement with density increasing (Gong et al 2019 Nucl. Fusion 59 086030) can be attributed to the reduction of turbulent transport by the collisional stabilization in trapped electron modes, rather than by the Shafranov shift stabilization effect which was proposed to be the major cause of confinement enhancement in previous literature. Based on the successful validation and newly gained physical insights, predictive modeling is performed for core plasma considering the future upgrade capacity of LH wave system and shows that the high-performance steady-state H-mode scenario on EAST can be extended to the regime with q 95 to be ITER relevant.

Physics design of new lower tungsten divertor for long-pulse high-power operations in EAST

A new lower tungsten divertor has been developed and installed in the EAST superconducting tokamak to replace the previous graphite divertor with power handling capability increasing from <2 MW m−2 to ∼10 MW m−2, aiming at achieving long-pulse H-mode operations in a full metal wall environment with the steady-state divertor heat flux of ∼10 MW m−2. A new divertor concept, ‘corner slot’ (CS) divertor, has been employed. By using the ‘corner effect’, a strongly dissipative divertor with the local buildup of high neutral pressure near the corner can be achieved, so that stable detachment can be maintained across the entire outer target plate with a relatively lower impurity seeding rate, at a separatrix density compatible with advanced steady-state core scenarios. These are essential for achieving efficient current drive with low-hybrid waves, a low core impurity concentration and thus a low loop voltage for fully non-inductive long-pulse operations. Compared with the highly closed small-angle-slot divertor in DIII-D, the new divertor in EAST exhibits the following merits: (1) a much simpler geometry with integral cassette body structure, combining vertical and horizontal target plates, which are more suitable for actively water-cooled W/Cu plasma facing components, facilitating installation precision control for minimizing surface misalignment, achieving high engineering reliability and lowering the capital cost as well; (2) it has much greater flexibility in magnetic configurations, allowing for the position of the outer strike point on either vertical or horizontal target plates to accommodate a relatively wide triangularity range, δ l = 0.4–0.6, thus enabling to explore various advanced scenarios. A water-cooled copper in-vessel coil has been installed under the dome. Five supersonic molecular beam injection systems have been mounted in the divertor to achieve faster and more precise feedback control of the gas injection rate. Furthermore, this new divertor allows for double null divertor operation and slowly sweeping the outer strike point across the horizontal and vertical target plates to spread the heat flux for long-pulse operations. Preliminary experimental results demonstrate the ‘corner effect’ and are in good agreement with simulations using SOLPS-ITER code including drifts. The EAST new divertor provides a test-bed for the closed divertor concept to achieve steady-state detachment operation at high power. Next step, a more closed divertor, ‘sharp-cornered slot’ divertor, building upon the current CS divertor concept, has been proposed as a candidate for the EAST upper divertor upgrade.

Neutronic analysis of the NuScale core using accident tolerant fuels with different coating materials

The design flexibility of small modular nuclear reactors in synergy with accident-tolerant fuels provides a longer time to react to the loss of active cooling under severe accidents and offers comparable fuel performance during normal operation. The NuScale reactor has an integrated design with a fuel enrichment below 5%, a thermal power of 160 MW and 24-month operating cycles, also guaranteeing safe shutdown by redundant passive systems. In this work, the core of the NuScale reactor is simulated with the Serpent Monte Carlo code, with the objective of analyzing important steady-state core parameters and making a comparison using U3Si2 fuel with different coating materials. The results predict the behavior of the core through the neutron multiplication factor and ppm of boron vs fuel burnup, the Doppler, the moderator density and the temperature reactivity coefficients, the radial and axial power distributions and the neutron spectrum of each of the studied fuels and coating materials. The results show that the operation of the reactor with these materials is safe and feasibly according to the analyzed neutron parameters.

HEXAGONAL PWR-CORE MODELING AND SIMULATION WITH APPLICATION OF NECP-BAMBOO

In this paper, the modeling and simulation of the PWRs loaded with hexagonal fuel assemblies has been implemented with the NECP-Bamboo code. NECP-Bamboo, consisting of a 2D lattice code named Bamboo-Lattice and a 3D steady-state core code named Bamboo-Core, was primitively designed for the PWRs loaded with the rectangular fuel assemblies. As the capability extension for PWRs with hexagonal fuel assemblies, four aspects of improvement have been implemented in NECP-Bamboo. Firstly, the Constructive Solid Geometry (CSG) has been implemented in Bamboo-Lattice for the lattice modeling. Secondly, the explicit modeling of the reflector assembly has been applied to provide more reliable few-group constants, compared with the conventional 1D model for the reflector assembly. Thirdly, the assembly-homogenization capability has been extended to the hexagonal assembly. Fourthly, the diffusion solver in Bamboo-Core based on the Variational Nodal Method (VNM) has been extended to handle hexagonal geometry. With application of the capability-extended NECP-Bamboo, the modeling and simulations for the VVER-1000 benchmark loaded with MOX fuel has been implemented. It can be observed that the numerical results provided by NECP-Bamboo can agree well with corresponding results by the Monte-Carlo code.

The impacts of heterogeneity on CO2 capillary trapping within the Captain Sandstone - a core to field scale study.

To ensure storage security, it is vital we understand, and can effectively model, the physical and chemical trapping mechanisms for CO2 storage. A key trapping mechanism underpinning storage security and immobilising a significant proportion of the CO2 plume is capillary trapping. Capillary trapping is considered to be responsible for 90% of the storage capacity in saline aquifers in the US, the largest potential CO2 storage resource. The ability to model and predict capillary trapping over large spatial scales in complex geological systems is essential to minimise risks and evaluate capacity. This paper investigates the effect of natural rock heterogeneities on capillary trapping across spatial scales in the Captain D Sandstone, from the Goldeneye formation, North Sea. That is, how heterogeneities affect CO2 saturation and distribution within rock core samples, and how those effects manifest, as a pseudo residual trapping, when upscaled to the field. A comprehensive dataset of 48 core plugs over a 65m interval have been studied. The location has industrial application as a target injection site for the discontinued Peterhead CCS project, with the aim to store 10 Mt CO2 over 10 years. These results are also applicable to other storage sites of similar geology. We have carried out steady-state core flooding experiments using medical x-ray CT, providing a detailed characterisation of continuum multiphase flow properties, including residual trapping characteristics, over cm scales. The results show that individual core plugs exhibit a large range in apparent residual trapping at the centimetre scale which averages out over scales of 10s of cm’s or greater. The average Land trapping value at the deca-centimetre scale is 1.7, however, at the cm scale, it varies between 0.8 and 2.8, representing a very large variation in locally trapped CO2. Consequently, at the plug scale for large initial saturations, approaching 1, the apparent residual saturation may vary between 0.2 and 0.5, given the variations in Land trapping parameter. At lower initial saturations, the variations are not as great in absolute terms, but the relative variation is just as large. Continuum numerical models are constructed from observations of residual trapping in heterogeneous rock cores using characterisation methods developed by Jackson et al. (2018, 2020). The trapping stage of a CO2 injection and migration project is simulated over a 65m region, while retaining the impact of heterogeneities at the cm-scale. Results from these models indicate that great care should be taken when upscaling the Land trapping parameter for use in field simulations, to estimate the long-term trapping of CO2 by capillary heterogeneities. Through quantifying the proportion of trapping resulting from capillary heterogeneities, the implications of modelling the system with the average and maximum experimental Land trapping parameter are evaluated. Modelling the system with the maximum experimental core plug Land trapping parameter results in the proportion of trapping caused by capillary heterogeneities to approximately double compared to that obtained using the average experimental core plug Land trapping parameter.

Исследование влияния характеристик стационарного состояния активной зоны ВВЭР-1000 (1200) на устойчивость реактора к ксеноновым колебаниям

Исследование влияния характеристик стационарного состояния активной зоны ВВЭР-1000 (1200) на устойчивость реактора к ксеноновым колебаниям

Cause analysis and solutions of water blocking damage in cracked/non-cracked tight sandstone gas reservoirs

After hydraulic fracturing treatment, a reduction in permeability caused by the invasion of fracturing fluids is an inevitable problem, which is called water blocking damage. Therefore, it is important to mitigate and eliminate water blocking damage to improve the flow capacities of formation fluids and flowback rates of the fracturing fluid. However, the steady-state core flow method cannot quickly and accurately evaluate the effects of chemical agents in enhancing the fluid flow capacities in tight reservoirs. This paper introduces a time-saving and accurate method, pressure transmission test (PTT), which can quickly and quantitatively evaluate the liquid flow capacities and gas-drive flowback rates of a new nanoemulsion. Furthermore, scanning electron microscopy (SEM) was used to analyze the damage mechanism of different fluids and the adsorption of chemical agents on the rock surface. Parallel core flow experiments were used to evaluate the effects of the nanoemulsion on enhancing flowback rates in heterogeneous tight reservoirs. Experimental results show that the water blocking damage mechanisms differ in matrices and fractures. The main channels for gas channeling are fractures in cracked cores and pores in non-cracked cores. Cracked cores suffer less damage from water blocking than non-cracked cores, but have a lower potential to reduce water saturation. The PTT and SEM results show that the permeability reduction in tight sandstones caused by invasion of external fluids can be list as guar gum fracturing fluid > slickwater > brine. Parallel core flow experiments show that for low-permeability heterogenous sandstone reservoirs with a certain permeability ratio, the nanoemulsion can not only reduce reverse gas channeling degree, but also increase the flowback rate of the fracturing fluid. The nanoemulsion system provides a new solution to mitigate and eliminate water blocking damage caused by fracturing fluids in tight sandstone gas reservoirs.

Application of Cross Sections Uncertainty Propagation Framework to Light and Heavy Water Reactor Systems

Abstract Uncertainty quantification has been recognized by the community as an essential component of best-estimate reactor analysis simulation because it provides a measure by which the credibility of the simulation can be assessed. In a companion paper, a framework for the propagation of nuclear data uncertainties from the multigroup level through lattice physics and core calculations and ultimately to core responses of interest has been developed. The overarching goal of this framework is to automate the propagation, prioritization, mapping, and reduction of uncertainties for reactor analysis core simulation. This paper employs both heavy and light water reactor systems to exemplify the application of this framework. Specifically, the paper is limited to the propagation of the nuclear data starting with the multigroup cross section covariance matrix and down to core responses, e.g., eigenvalue and power distribution, in steady-state core wide calculations. The goal is to demonstrate how the framework employs reduction techniques to compress the uncertainty space into a very small number of active degrees-of-freedom (DOFs), which renders the overall process computationally feasible for day-to-day engineering evaluations.

Quantifying the effects of gaseous pore pressure and net confining stress on low-permeability cores using the “RaSSCAL” steady-state permeameter

Subsurface engineering systems undergo large changes in pore pressure, which can have drastic effects on their ability to transport and store fluids. For rock samples retrieved from conventional reservoirs, standard core analysis procedures provide suitable estimates of petrophysical properties, such as porosity and permeability. However, performing similar investigations on low-permeability systems like shales and tight gas sands remains a challenge. Steady-state gas permeametry is considered the gold standard for analyzing low-permeability cores, but researchers often forego it in favor of faster but less reliable methods. Attempting a conventional permeability test on a tight plug sample typically results in miniscule flow rates that only expensive, specialized equipment can measure directly. Furthermore, these time-consuming tests are best run multiple times under varying ranges of pore pressure and confining stress. Doing so allows one to anticipate suitably the changes in permeability that will occur because of the expected changes in reservoir conditions.This study describes the methodology of the Randolph Steady-State Core Analysis Laboratory (RaSSCAL), a new experimental design and data analysis algorithm that accounts for changes in permeability over the range of pore pressures and confining stresses that occur during processes like unconventional hydrocarbon recovery and geologic carbon sequestration. Routine steady-state core analysis requires relatively small pressure drops across the sample so a uniform apparent permeability can be assumed. The approach outlined here enables an effective interpretation of core-flood data in the presence of arbitrarily large pressure gradients. The experiment maintains a near-constant pressure difference of arbitrary magnitude across a core-plug sample and indirectly measures the flow rate induced by that gradient using a differential pressure transmitter. Tests were performed on a Marcellus Shale sample from Bedford, Pennsylvania and an Oriskany Sandstone sample from Berkeley Springs, West Virginia over a wide array of pore pressures and net confining stresses using helium and methane gas. Having a quantifiable understanding of how the permeability of the system may change during the various stages of subsurface activity should significantly enhance the ability to optimize design outcomes. This will ultimately lead to improved oil and gas recovery and better predictions of injectivity and storage capacity during carbon-storage operations.

Progress on in-vessel poloidal field coils optimization design for alternative divertor configuration studies on the EAST tokamak

An upgrade to the lower divertor is currently being planned for EAST superconducting tokamak, aiming at reaching over 400 s long-pulse H-mode operations with a full metal wall and a divertor heat load of ˜10 MW/m2. A new divertor concept for EAST, “Tightly Baffled Divertor”, suited to water- cooled W/Cu plasma face components (PFC) with minimized divertor volume, has been proposed to achieve Te,target <5 eV across entire outer target at lower separatrix plasma density and optimized pumping by a simple closed divertor structure combining horizontal target with inclined baffle, dome and duct. This divertor should allow access to high-triangularity small Edge Localized Mode (ELM) H-mode regimes and also allow achieving advanced magnetic divertor configurations with the assistance of two water-cooled in-vessel divertor coils (Divertor coils). Preliminary engineering design of in-vessel Divertor coils indicates a maximum current of 8 kA for long-pulse discharges, and 20 kA for the shortest ones. However, flexibility on Divertor coils position optimization is limited to the water cooling system. Initial plasma equilibrium studies by EFIT code, used in combination with CREATE-NL and FIXFREE tools, show that the distance of the two nearby divertor poloidal field nulls, can be decreased up to ˜ 0.95 m with a plasma current IP ˜ 400 kA, leading to a configuration with the secondary X-point located close to the target, with a significant increase of magnetic poloidal flux expansion and connection length. This may provide a promising divertor solution compatible with advanced steady-state core scenarios.

Evaluation of permeability damage caused by drilling and fracturing fluids in tight low permeability sandstone reservoirs

Tight sandstone reservoirs with low porosity and permeability are always vulnerable to damage due to drilling and fracturing fluids, which significantly impedes hydrocarbon production. Therefore, exploring damage evaluation methods has been an undeniable demand of formation protection and efficient development. However, the common steady-state core flow method is time-consuming and inaccurate for evaluating permeability damage in tight low permeability reservoirs. Aiming at this problem, this paper introduced pressure transmission test (PTT), a time-saving and accurate method, to quantitatively evaluate the core permeability damage degree from macroscopic perspective. Furthermore, nuclear magnetic resonance (NMR) was also used to analyze the damage mechanisms of fracturing fluids from microscopic perspective. In the fracturing fluids damage experiments, the results of PTT show that the ultimate damage degree of fracturing fluids is approximately 40%, of which water-sensitivity, water-block, and solid-phase are approximately 10%, 12% and 16%, respectively. With the technique of NMR, the damage mechanisms due to fracturing fluids are also investigated. In the drilling and fracturing fluids damage experiments, the results of PTT show that the drilling fluids cause the major damage. In the experiments of acidizing treatment, the chelating acid shows excellent properties in reliving core permeability damage caused by drilling and fracturing fluids. This study provides a new idea for evaluating permeability damage in low permeability reservoirs, such as shale, tight sandstone reservoirs, bedrock reservoirs, and so on.

Development of SNTA Code System for SCWR Core Steady-State Analysis

A coupled neutronics/thermal-hydraulics (N/T) three-dimensional code system SNTA is developed for supercritical water-cooled reactor (SCWR) core steady-state analysis by modular coupling the improved neutronics nodal methodological code and SCWR thermal-hydraulic subchannel code. The appropriate outer iteration coupling method and self-adaptive relaxation factor are proposed for enhancing convergence, stability, and efficiency of coupled N/T calculation. The steady-state analysis for the CSR1000 core is applied to verify SNTA. The results calculated by SNTA agreed well with those by CASIR and SRAC. SNTA is more efficient than CASIR and SRAC, where the neutronics modules are based on the finite-difference method. The numeric results show that SNTA can be applied to SCWR core steady-state analysis and core concept design.

Uncertainty analysis for the assembly and core simulation of BEAVRS at the HZP conditions

Based on the “two-step” scheme for the reactor-physics calculations, the capability of uncertainty analysis for the core simulations has been implemented in the UNICORN code, an in-house code for the sensitivity and uncertainty analysis of the reactor-physics calculations. Applying the statistical sampling method, the nuclear-data uncertainties can be propagated to the important predictions of the core simulations. The uncertainties of the few-group constants introduced by the uncertainties of the multigroup microscopic cross sections are quantified first for the lattice calculations; the uncertainties of the few-group constants are then propagated to the core multiplication factor and core power distributions for the core simulations. Up to now, our in-house lattice code NECP-CACTI and the neutron-diffusion solver NECP-VIOLET have been implemented in UNICORN for the steady-state core simulations based on the “two-step” scheme. With NECP-CACTI and NECP-VIOLET, the modeling and simulation of the steady-state BEAVRS benchmark problem at the HZP conditions was performed, and the results were compared with those obtained by CASMO-4E. Based on the modeling and simulation, the UNICORN code has been applied to perform the uncertainty analysis for BAEVRS at HZP. The uncertainty results of the eigenvalues and two-group constants for the lattice calculations and the multiplication factor and the power distributions for the steady-state core simulations are obtained and analyzed in detail.

Contactless, non-intrusive core temperature measurement of a solid body in steady-state

Accurate measurement of temperature is critical for understanding thermal behavior and monitoring safety and performance of engineering systems involving heating and cooling. While a number of methods are available for measurement of temperature on the outside surface of solid bodies, there is a lack of contactless, non-invasive methods for determining temperature inside solid bodies. Development of such methods is likely to impact a wide range of engineering systems. This paper describes and validates a method for measurement of internal core temperature of a solid body in steady-state based on measurement of the temperature distribution on its outside surface based on a theoretical thermal conduction model. This method is validated by determining the steady-state core temperature of a thermal test cell using infrared temperature measurement on the surface, and comparing with measurements from an embedded thermocouple. The two measurements are found to agree well with each other in a variety of heat generation and cooling conditions. While this validation is presented for a cylindrical body, the method lends itself easily to bodies of other shapes. This work contributes towards fundamental thermal metrology, with possible applications in a wide variety of engineering systems.

Validation and verification of the coupled neutron kinetic/thermal hydraulic system code DYN3D/ATHLET

One of the most intensively developing areas in the LWR multi-physics is a coupling of different best estimate 3-D neutron kinetic (BIPR, DYN3D, KIKO3D, NEM, PARCS, etc.) and thermal hydraulic (ATHLET, CATHARE, RELAP5, etc.) codes. Resulting coupled code systems have advanced capabilities of modeling both steady-state spatial distributions of the core power and their evolutions during different kinds of reactor transients. They are also highly useful in the analyses of possible reactor instabilities. Initial steady-state core power distributions can be disturbed by changes in the reactor loop mass flow rates and/or temperatures, by relocations of the low-temperature/diluted-boron water slugs within the primary system or by movements of control rods.The coupled code used for LWR simulations in HZDR is DYN3D/ATHLET, which includes the 3-D core neutron kinetic and thermal hydraulic model of own development – DYN3D. The paper reports major capabilities of DYN3D as well as different ways of its coupling with the thermal hydraulic code ATHLET (external, internal and parallel), but mainly focuses on the validation and verification of the coupled code DYN3D/ATHLET. In the course of DYN3D/ATHLET validation/verification nearly 20 real plant transients and dynamic benchmarks have been simulated and analysed. The LWR types covered by these tasks are: VVER-440 (Bohunice-3, Greifswald-5 and Loviisa-1 units), VVER-1000 (Balakovo-1, Balakovo-4, Kalinin-3, Kozloduy-6, Saporozhye-6 and Temelin-2 units), B&W PWR (TMI-1) and BWR/4 MK-1 (Peach Bottom-2).For each reactor unit a computational model was developed according with the benchmark specifications. The simulated tasks describe different scenarios of increasing complexity, including the transients initiated by the main steam header or main steam line breaks, switching off/on of main circulation pumps, turbine trip and generator load drop. Some of the transients are characterized by a strongly asymmetric behavior of the primary system (e.g. caused by a steam line break), and the processes of coolant mixing in the lower and upper reactor plenums as well as in the downcomer are important for these cases. The coupled code DYN3D/ATHLET models the primary coolant mixing in two ways – whether by an appropriate nodalization of the mixing area or by using two specific models for mixing in the lower plenum. The first of the lower plenum mixing models is a part of the coupling interface at the core inlet plane, while the second one is the analytical coolant mixing model developed for the downcomer and the lower plenum regions of VVER-440.The results of DYN3D/ATHLET simulations were assessed both against available measured data and calculations performed with similar multi-physics codes. The paper includes an overview of the simulated problems and the most representative results of DYN3D/ATHLET validation/verification for all coupling modes. The obtained experience of code validation provides a better understanding of reactor transients with a strong interaction between neutron kinetics and thermal hydraulics, helping to improve computation models. This experience is also useful for the present and further activities in coupling of DYN3D with other best estimate codes, like CFX, TRANSURANUS, etc.

LAVENDER: A steady-state core analysis code for design studies of accelerator driven subcritical reactors

Accelerator driven subcritical reactors (ADSRs) have been proposed and widely investigated for the transmutation of transuranics (TRUs). ADSRs have several special characteristics, such as the subcritical core driven by spallation neutrons, anisotropic neutron flux distribution and complex geometry etc. These bring up requirements for development or extension of analysis codes to perform design studies. A code system named LAVENDER has been developed in this paper. It couples the modules for spallation target simulation and subcritical core analysis. The neutron transport-depletion calculation scheme is used based on the homogenized cross section from assembly calculations. A three-dimensional SN nodal transport code based on triangular-z meshes is employed and a multi-channel thermal-hydraulics analysis model is integrated. In the depletion calculation, the evolution of isotopic composition in the core is evaluated using the transmutation trajectory analysis algorithm (TTA) and fine depletion chains. The new code is verified by several benchmarks and code-to-code comparisons. Numerical results indicate that LAVENDER is reliable and efficient to be applied for the steady-state analysis and reactor core design of ADSRs.

Transition to CASMO-5M and SIMULATE-3K for Stability Analyses of the Swiss Boiling Water Reactors

For stability analyses of the Swiss operating boiling water reactors, the methodology employed and validated so far at the Paul Scherrer Institute (PSI) was based on the RAMONA-3 code with a hybrid upstream static lattice/core analysis approach using CASMO-4 and PRESTO-2. More recently, steps were undertaken toward a new methodology based on the SIMULATE-3K (S3K) code for the dynamical analyses combined with the CMSYS system, which relies on the CASMO/SIMULATE-3 suite of codes and was established at PSI to serve as framework for the development and validation of reference core models of all the Swiss reactors and operated cycles.This paper presents a first validation of the new methodology on the basis of a benchmark recently organized by a Swiss utility and including the participation of several international organizations with various codes/methods. Now in parallel, a transition from CASMO-4E (C4E) to CASMO-5M (C5M) as basis for the CMSYS core models was also recently initiated at PSI. Consequently, it was considered adequate to address the impact of this transition both for the steady-state core analyses as well as for the stability calculations and to achieve thereby an integral approach for the validation of the new S3K methodology. Therefore, a comparative assessment of C4 versus C5M is also presented in this paper, with particular emphasis on the void coefficients and their impact on the downstream stability analysis results.