Paul P. H. Wilson

Adjoint-Weighted Tallies for k -Eigenvalue Calculations with Continuous-Energy Monte Carlo

Abstract A Monte Carlo method is developed that performs adjoint-weighted tallies in continuous-energy k-eigenvalue calculations. Each contribution to a tally score is weighted by an estimate of the relative magnitude of the fundamental adjoint mode, by way of the iterated fission probability, at the phase-space location of the contribution. The method is designed around the power iteration method such that no additional random walks are necessary, resulting in a minimal increase in computational time. The method is implemented in the Monte Carlo N-Particle (MCNP) code. These adjoint-weighted tallies are used to calculate adjoint-weighted fluxes, point reactor kinetics parameters, and reactivity changes from first-order perturbation theory. The results are benchmarked against discrete ordinates calculations, experimental measurements, and direct Monte Carlo calculations.

ITER Nuclear Analysis Strategy and Requirements

Abstract The principle needs of ITER with regard to nuclear analysis can be divided into the broad categories of safety and licensing, plant operation, and decommissioning although there is much overlap and interdependence within these categories. This paper reviews the status of the methods applied to date and recommends the future strategy which ITER should adopt to address the continuing requirements and responsibilities. This is done by consideration of the application of radiation transport codes, the creation of an ITER reference neutronics model, the application of a neutronics results database, and the management tools which will be required. Areas in which new codes and techniques need to be developed will be identified.

THE ARIES-CS COMPACT STELLARATOR FUSION POWER PLANT

Abstract An integrated study of compact stellarator power plants, ARIES-CS, has been conducted to explore attractive compact stellarator configurations and to define key research and development (R&D) areas. The large size and mass predicted by earlier stellarator power plant studies had led to cost projections much higher than those of the advanced tokamak power plant. As such, the first major goal of the ARIES-CS research was to investigate if stellarator power plants can be made to be comparable in size to advanced tokamak variants while maintaining desirable stellarator properties. As stellarator fusion core components would have complex shapes and geometry, the second major goal of the ARIES-CS study was to understand and quantify, as much as possible, the impact of the complex shape and geometry of fusion core components. This paper focuses on the directions we pursued to optimize the compact stellarator as a fusion power plant, summarizes the major findings from the study, highlights the key design aspects and constraints associated with a compact stellarator, and identifies the major issues to help guide future R&D.

DESIGNING ARIES-CS COMPACT RADIAL BUILD AND NUCLEAR SYSTEM : NEUTRONICS, SHIELDING, AND ACTIVATION

Abstract Within the ARIES-CS project, design activities have focused on developing the first compact device that enhances the attractiveness of the stellarator as a power plant. The objectives of this paper are to review the nuclear elements that received considerable attention during the design process and provide a perspective on their successful integration into the final design. Among these elements are the radial build definition, the well-optimized in-vessel components that satisfy the ARIES top-level requirements, the carefully selected nuclear and engineering parameters to produce an economic optimum, the modeling - for the first time ever - of the highly complex stellarator geometry for the three-dimensional nuclear assessment, and the overarching safety and environmental constraints to deliver an attractive, reliable, and truly compact stellarator power plant.

THE BACK END OF THE FUSION MATERIALS CYCLE

Abstract Within the framework of the International Energy Agency, an international collaborative study on fusion radioactive waste has been initiated to examine the back end of the materials cycle as an important stage in maximizing the environmental benefits of fusion as an energy provider. The study addresses the management procedures for radioactive materials following the changeout of replaceable components and decommissioning of fusion facilities. We define this as “the back end” of the fusion materials cycle. It includes all the procedures necessary to manage spent radioactive materials from fusion facilities, from the removal of the components from the device to the reuse of these components through recycling/clearance, or to the disposal of the waste in geological repositories. Fusion devices have certain characteristics that make them environmentally friendly devices; minimization of long-lived waste that could be a burden for future generations is one of these characteristics. Recycling and clearance procedures and regulations have been recently revised, and the effects of these revisions on back-end fusion materials are examined in the paper. Finally, an integrated approach to the management of back-end fusion materials is proposed, and its application to three fusion reactor designs is discussed.

Evolution of Clearance Standards and Implications for Radwaste Management of Fusion Power Plants

Abstract The issue of radioactive waste management presents a top challenge for the nuclear industry. As an alternative to recycling or disposal in repositories, many countries are proceeding successfully with the process of developing clearance guidelines that allow solids and building rubble containing traces of radioisotopes to be cleared from regulatory control and unconditionally released to the commercial market after a specific storage period. With the emergence of new clearance standards, we took the initiative to compare U.S. to European and other international limits. This exercise is proving valuable in understanding the differences between the clearance standards and their implications for the radwaste management of fusion power plants. While clearance standards now exist for most radionuclides that are mainly important to the fission industry, no such standards are in place for many radionuclides of interest to fusion facilities. Before fusion penetrates the energy market, fusion-specific standards should be developed to address the safe release of fusion materials with trace levels of radioactive contamination.

Use of CAD Generated Geometry Data in Monte Carlo Transport Calculations for ITER

Abstract An extensive benchmark exercise has been conducted on ITER with the objective to test and validate different approaches for the use of CAD generated geometry data for Monte Carlo transport calculations with the MCNP code. The exercise encompassed the generation of a dedicated neutronics CATIA model based on available engineering CAD design data, the conversion into MCNP geometry, the verification of the converted models, and a number of calculations to compare the different approaches with regard to the performance and the validity of the results obtained. The paper briefly reviews the different approaches and provides a detailed description of the ITER benchmark effort, its results and conclusions showing that the approaches have reached the maturity level to allow their application to real ITER design analyses. This is considered an essential step forward for neutronics analysis tools to satisfy ITER quality assurance rules.

Multigrid in energy preconditioner for Krylov solvers

We have added a new multigrid in energy (MGE) preconditioner to the Denovo discrete-ordinates radiation transport code. This preconditioner takes advantage of a new multilevel parallel decomposition. A multigroup Krylov subspace iterative solver that is decomposed in energy as well as space-angle forms the backbone of the transport solves in Denovo. The space-angle-energy decomposition facilitates scaling to hundreds of thousands of cores. The multigrid in energy preconditioner scales well in the energy dimension and significantly reduces the number of Krylov iterations required for convergence. This preconditioner is well-suited for use with advanced eigenvalue solvers such as Rayleigh Quotient Iteration and Arnoldi.

Three Dimensional Neutronics Analysis of the ITER First Wall/Shield Module 13

Radiation shielding and energy removal for ITER are provided by an array of first wall/shield modules (FWS). Nuclear analysis of the shield modules is important for understanding their performance and lifetime in the system. While one-dimensional (1-D) analysis provides an adequate first approximation, three-dimensional (3-D) analysis is needed to validate the 1-D analysis and resolve fine geometric details that result from heterogeneities in the module. Using MCNPX-CGM, a coupling of traditional MCNPX with the Common Geometry Module (CGM), high-fidelity 3-D neutronics analysis is now possible. Particles are transported in the CAD geometry reducing analysis time, eliminating input error, and preserving geometric detail. A detailed 3-D CAD model of FWS module 13 is inserted into a 1-D radial approximation including homogenized representations of the inboard FWS, plasma, and vacuum vessel (VV). A 14.1 MeV uniform source between the inboard and outboard sides is used to simulate the ITER plasma. Reflecting boundary conditions approximate the full extent of ITER in the poloidal and toroidal directions. Heating, radiation damage, and helium production profiles through module 13 are determined using high-resolution mesh tallies. In the front manifold of the shield block, heating and helium production were found to be lower in steel than the homogenized 1-D model suggests. Peaking in nuclear heating and helium production in steel is observed at the interface with adjacent water zones.

ITER Neutronics Modeling Using Hybrid Monte Carlo/Deterministic and CAD-Based Monte Carlo Methods

Abstract The immense size and complex geometry of the ITER experimental fusion reactor require the development of special techniques that can accurately and efficiently perform neutronics simulations with minimal human effort. This paper shows the effect of the hybrid Monte Carlo (MC)/deterministic techniques—Consistent Adjoint Driven Importance Sampling (CADIS) and Forward-Weighted CADIS (FW-CADIS)—in enhancing the efficiency of the neutronics modeling of ITER and demonstrates the applicability of coupling these methods with computer-aided-design-based MC. Three quantities were calculated in this analysis: the total nuclear heating in the inboard leg of the toroidal field coils (TFCs), the prompt dose outside the biological shield, and the total neutron and gamma fluxes over a mesh tally covering the entire reactor. The use of FW-CADIS in estimating the nuclear heating in the inboard TFCs resulted in a factor of ˜275 increase in the MC figure of merit (FOM) compared with analog MC and a factor of ˜9 compared with the traditional methods of variance reduction. By providing a factor of ˜21 000 increase in the MC FOM, the radiation dose calculation showed how the CADIS method can be effectively used in the simulation of problems that are practically impossible using analog MC. The total flux calculation demonstrated the ability of FW-CADIS to simultaneously enhance the MC statistical precision throughout the entire ITER geometry. Collectively, these calculations demonstrate the ability of the hybrid techniques to accurately model very challenging shielding problems in reasonable execution times.