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[en] The goal is for all modeling and simulation tools to be demonstrated accurate and reliable through a formal Verification and Validation (V and V) process, especially where such tools are to be used to establish safety margins and support regulatory compliance, or to design a system in a manner that reduces the role of expensive mockups and prototypes. Whereas the Verification part of the process does not rely on experiment, the Validation part, on the contrary, necessitates as many relevant and precise experimental data as possible to make sure the models reproduce reality as closely as possible. Hence, this report presents a limited selection of experimental data that could be used to validate the codes devoted mainly to Fast Neutron Reactor calculations in the US. Emphasis has been put on existing data for thermal-hydraulics, fuel and reactor physics. The principles of a new 'smart' experiment that could be used to improve our knowledge of neutron cross-sections are presented as well. In short, it consists in irradiating a few milligrams of actinides and analyzing the results with Accelerator Mass Spectroscopy to infer the neutron cross-sections. Finally, the wealth of experimental data relevant to Fast Neutron Reactors in the US should not be taken for granted and efforts should be put on saving these 30-40 years old data and on making sure they are validation-worthy, i.e. that the experimental conditions and uncertainties are well documented.
[en] One of the most reliable experimental methods for measuring the kinetic parameters of a subcritical assembly is the Sjoestrand method applied to the reaction rate generated from a pulsed neutron source. This study developed a new analytical methodology for characterizing the kinetic parameters of a subcritical assembly using the Sjoestrand method, which allows comparing the analytical and experimental time dependent reaction rates and the reactivity measurements. In this methodology, the reaction rate, detector response, is calculated due to a single neutron pulse using MCNP/MCNPX computer code or any other neutron transport code that explicitly simulates the fission delayed neutrons. The calculation simulates a single neutron pulse over a long time period until the delayed neutron contribution to the reaction is vanished. The obtained reaction rate is superimposed to itself, with respect to the time, to simulate the repeated pulse operation until the asymptotic level of the reaction rate, set by the delayed neutrons, is achieved. The superimposition of the pulse to itself was calculated by a simple C computer program. A parallel version of the C program is used due to the large amount of data being processed, e.g. by the Message Passing Interface (MPI). The new calculation methodology has shown an excellent agreement with the experimental results available from the YALINA-Booster facility of Belarus. The facility has been driven by a Deuterium-Deuterium or Deuterium-Tritium pulsed neutron source and the (n,p) reaction rate has been experimentally measured by a 3He detector. The MCNP calculation has utilized the weight window and delayed neutron biasing variance reduction techniques since the detector volume is small compared to the assembly volume. Finally, this methodology was used to calculate the IAEA benchmark of the YALINA-Booster experiment
[en] As part of the Global Nuclear Energy Partnership (GNEP), a fast reactor simulation program was launched in April 2007 to develop a suite of modern simulation tools specifically for the analysis and design of sodium cooled fast reactors. The general goal of the new suite of codes is to reduce the uncertainties and biases in the various areas of reactor design activities by enhanced prediction capabilities. Under this fast reactor simulation program, a high-fidelity deterministic neutron transport code named UNIC is being developed. The final objective is to produce an integrated, advanced neutronics code that allows the high fidelity description of a nuclear reactor and simplifies the multi-step design process by direct coupling with thermal-hydraulics and structural mechanics calculations. Currently there are three solvers for the neutron transport code incorporated in UNIC: PN2ND, SN2ND, and MOCFE. PN2ND is based on a second-order even-parity spherical harmonics discretization of the transport equation and its primary target area of use is the existing homogenization approaches that are prevalent in reactor physics. MOCFE is based upon the method of characteristics applied to an unstructured finite element mesh and its primary target area of use is the fine grained nature of the explicit geometrical problems which is the long term goal of this project. SN2ND is based on a second-order, even-parity discrete ordinates discretization of the transport equation and its primary target area is the modeling transition region between the PN2ND and MOCFE solvers. The major development goal in fiscal year 2008 for the MOCFE solver was to include a two-dimensional capability that is scalable to hundreds of processors. The short term goal of this solver is to solve two-dimensional representations of reactor systems such that the energy and spatial self-shielding are accounted for and reliable cross sections can be generated for the homogeneous calculations. In this report we present good results for an OECD benchmark obtained using the new two-dimensional capability of the MOCFE solver. Additional work on the MOCFE solver is focused on studying the current parallelization algorithms that can be applied to both the two- and three-dimensional implementations such that they are scalable to thousands of processors. The initial research into this topic indicates that, as expected, the current parallelization scheme is not sufficiently scalable for the detailed reactor geometry that it is intended for. As a consequence, we are starting the investigative research to determine the alternatives that are applicable for massively parallel machines. The major development goal in fiscal year 2008 for the PN2ND and SN2ND solvers was to introduce parallelism by angle and energy. The motivation for this is two-fold: (1) reduce the memory burden by picking a simpler preconditioner with reduced matrix storage and (2) improve parallel performance by solving the angular subsystems of the within group equation simultaneously. The solver development in FY2007 focused on using PETSc to solve the within group equation where only spatial parallelization was utilized. Because most homogeneous problems required relatively few spatial degrees of freedom (tens of thousands) the only way to improve the parallelism was to spread the angular moment subsystems across the parallel system. While the coding has been put into place for parallelization by space, angle, and group, we have not optimized any of the solvers and therefore do not give an assessment of the achievement of this work in this report. The immediate task to be completed is to implement and validate Tchebychev acceleration of the fission source iteration algorithm (inverse power method in this work) and optimize both the PN2ND and SN2ND solvers. We further intend to extend the applicability of the UNIC code by adding a first-order discrete ordinates solver termed SN1ST. Upon completion of this work, all memory usage problems are to be identified and studied in the solvers with the intent of making the new version of an exportable production code in either FY2008 or FY2009. This report covers the status of these tasks and discusses the work yet to be completed
[en] Recently a discretization (in time) scheme for the transient even-parity neutron transport equation was successfully developed and implemented in the VARIANT/KIN3D code. Starting from this result we have first more thoroughly investigated the order of approximation, with respect to the time variable, arising from such a scheme especially having in mind the use of a variable time step. Based on the results of this analysis we have then introduced a time discretization scheme capable of keeping the mathematical structure of the equation also in the framework of a variable time step. Starting from this new scheme we have also introduced an automatic time step control option based on the estimation of the third order time derivative by comparing the backward and centered Euler schemes. In the paper, the developed theoretical background is presented and results for a small size fast reactor are shown. We have investigated the time step control option and find its behavior being in agreement with the physical phenomena. In particular the increase of the delayed neutron fraction changes the behavior in the expected manner. The time step control was, in fact, able to detect the presence of additional time scales introduced by two families of delayed neutrons in comparison with the case with almost no delayed neutrons. (authors)
[en] In the last year INL has internally pursued the development of a new reactor analysis tool: PHISICS. The software is built in a modular approach to simplify the independent development of modules by different teams and future maintenance. Most of the modules at the time of this summary are still under development (time dependent transport driver, depletion, cross section I/O and interpolation, generalized perturbation theory), while the transport solver INSTANT (Intelligent Nodal and Semi-structured Treatment for Advanced Neutron Transport) has already been widely used1, 2, 3, 4. For this reason we will focus mainly on the presentation of the transport solver INSTANT
[en] The PHISICS (Parallel and Highly Innovative Simulation for INL Code System) software is under an intensive development at INL. In the last months new features have been added and improvements of the previously existing one performed. The modular approach has created a friendly development environment that allows a quick expansion of the capabilities. In the last months a little amount of work has been dedicated to the improvement of the spherical harmonics based nodal transport solver while the implementation of a solver based on the self adjoint formulation of the discrete ordinate is in the test phase on structured mesh. PHISICS now include a depletion solver with the option to use two different algorithms for the solution of the Bateman equation: the Taylor development of the exponential matrix and the Chebyshev Rational Approximation Method. The coupling with RELAP5 is also available at least in the steady state search mode. The coupling between RELAP5 and PHISICS can also take advantage of the new cross section interpolation module so that the coupling could be performed using an arbitrary number of energy groups.
[en] A code package called UNIC is currently under development at Argonne (Argonne National Laboratory). This new code package is focused on high fidelity solutions for nuclear reactor plant operations. The focus of this paper is the development of a spherical harmonics method used to solve the neutron transport equation. This new code, called PNFE, is just one of the components of UNIC. PNFE is targeted for use on massively parallel platforms using the PETSc library developed at ANL. In this paper, an overview of the theory behind the spherical harmonics method is given along with some results obtained with a parallel implementation of the code.
[en] RAVEN is a software tool under development at the Idaho National Laboratory (INL) that acts as the control logic driver and post-processing tool for the newly developed Thermal-Hydraulic code RELAP-7. The scope of this paper is to show the software structure of RAVEN and its utilization in connection with RELAP-7. A short overview of the mathematical framework behind the code is presented along with its main capabilities such as on-line controlling/ monitoring and Monte-Carlo sampling. A demo of a Station Black Out PRA analysis of a simplified Pressurized Water Reactor (PWR) model is shown in order to demonstrate the Monte-Carlo and clustering capabilities. (authors)
[en] The time-dependent second-order PN equation is discretized by a fully implicit finite-difference scheme; the obtained pseudo-steady-state PN equations are then used in the framework of the Variational Nodal Method (VNM). This scheme is implemented in the VARIANT/KIN3D code. A new code option is presented that takes in account neutron wave phenomena The code has been validated for a very simple case against an analytic solution; other simple tests are performed in order to investigate differences compared to the original version of KIN3D/VARIANT. (authors)
[en] This report presents the software engineering effort under way at ANL towards a comprehensive integrated computational framework (SHARP) for high fidelity simulations of sodium cooled fast reactors. The primary objective of this framework is to provide accurate and flexible analysis tools to nuclear reactor designers by simulating multiphysics phenomena happening in complex reactor geometries. Ideally, the coupling among different physics modules (such as neutronics, thermal-hydraulics, and structural mechanics) needs to be tight to preserve the accuracy achieved in each module. However, fast reactor cores in steady state mode represent a special case where weak coupling between neutronics and thermal-hydraulics is usually adequate. Our framework design allows for both options. Another requirement for SHARP framework has been to implement various coupling algorithms that are parallel and scalable to large scale since nuclear reactor core simulations are among the most memory and computationally intensive, requiring the use of leadership-class petascale platforms. This report details our progress toward achieving these goals. Specifically, we demonstrate coupling independently developed parallel codes in a manner that does not compromise performance or portability, while minimizing the impact on individual developers. This year, our focus has been on developing a lightweight and loosely coupled framework targeted at UNIC (our neutronics code) and Nek (our thermal hydraulics code). However, the framework design is not limited to just using these two codes