John C. Wagner

FW-CADIS Method for Global and Regional Variance Reduction of Monte Carlo Radiation Transport Calculations

Abstract This paper presents a hybrid (Monte Carlo/deterministic) method for increasing the efficiency of Monte Carlo calculations of distributions, such as flux or dose rate distributions (e.g., mesh tallies), as well as responses at multiple localized detectors and spectra. This method, referred to as Forward-Weighted CADIS (FW-CADIS), is an extension of the Consistent Adjoint Driven Importance Sampling (CADIS) method, which has been used for more than a decade to very effectively improve the efficiency of Monte Carlo calculations of localized quantities (e.g., flux, dose, or reaction rate at a specific location). The basis of this method is the development of an importance function that represents the importance of particles to the objective of uniform Monte Carlo particle density in the desired tally regions. Implementation of this method utilizes the results from a forward deterministic calculation to develop a forward-weighted source for a deterministic adjoint calculation. The resulting adjoint function is then used to generate consistent space-and energy-dependent source biasing parameters and weight windows that are used in a forward Monte Carlo calculation to obtain more uniform statistical uncertainties in the desired tally regions. The FW-CADIS method has been implemented and demonstrated within the MAVRIC (Monaco with Automated Variance Reduction using Importance Calculations) sequence of SCALE and the ADVANTG (Automated Deterministic Variance Reduction Generator)/MCNP framework. Application of the method to representative real-world problems, including calculation of dose rate and energy-dependent flux throughout the problem space, dose rates in specific areas, and energy spectra at multiple detectors, is presented and discussed. Results of the FW-CADIS method and other recently developed global variance-reduction approaches are also compared, and the FW-CADIS method outperformed the other methods in all cases considered.

Fusion Nuclear Science Facility (FNSF) Before Upgrade to Component Test Facility (CTF)

Abstract The compact (R0~1.2-1.3m) Fusion Nuclear Science Facility (FNSF) is aimed at providing a fully integrated, continuously driven fusion nuclear environment of copious fusion neutrons. This facility would be used to test, discover, and understand the complex challenges of fusion plasma material interactions, nuclear material interactions, tritium fuel management, and power extraction. Such a facility properly designed would provide, initially at the JET-level plasma pressure (~30%T2) and conditions (e.g., Hot-Ion H-Mode, Q<1)), an outboard fusion neutron flux of 0.25 MW/m2 while requiring a fusion power of ~19 MW. If and when this research is successful, its performance can be extended to 1 MW/m2 and ~76 MW by reaching for twice the JET plasma pressure and Q. High-safety factor q and moderate-plasmas are used to minimize or eliminate plasma-induced disruptions, to deliver reliably a neutron fluence of 1 MW-yr/m2 and a duty factor of 10% presently anticipated for the FNS research. Success of this research will depend on achieving time-efficient installation and replacement of all internal components using remote handling (RH). This in turn requires modular designs for the internal components, including the single-turn toroidal field coil center-post. These device goals would further dictate placement of support structures and vacuum weld seals behind the internal and shielding components. If these goals could be achieved, the FNSF would further provide a ready upgrade path to the Component Test Facility (CTF), which would aim to test, for ≤6 MW-yr/m2 and 30% duty cycle, the demanding fusion nuclear engineering and technologies for DEMO. This FNSF-CTF would thereby complement the ITER Program, and support and help mitigate the risks of an aggressive world fusion DEMO R&D Program. The key physics and technology research needed in the next decade to manage the potential risks of this FNSF are identified.

A case study in manual and automated Monte Carlo variance reduction with a deep penetration reactor shielding problem

Abstract Certain reactor transients cause a reduction in moderator temperature and, hence, increased attenuation of neutrons and decreased response of excore detectors. This decreased detector response is of concern because of the credit assumed for detector-initiated reactor trip to terminate the transient. Explicit modeling of this phenomenon presents the analyst with a difficult problem because of the dense and optically thick neutron absorption media, given the constraint that precise response characteristics must be known in order to account for this phenomenon. The solution in this study was judged to be the use of Monte Carlo techniques coupled with robust variance reduction to accelerate problem convergence. A fresh discussion on the motivation for variance reduction is included, followed by separate accounts of manual and automated applications of variance reduction techniques. Finally, the results of both manual and automated variance reduction techniques are presented and compared.

SIMULTANEOUS OPTIMIZATION OF TALLIES IN DIFFICULT SHIELDING PROBLEMS

Abstract Monte Carlo is quite useful for calculating specific quantities in complex transport problems. Many variance reduction strategies have been developed that accelerate Monte Carlo calculations for specific tallies. However, when trying to calculate multiple tallies or a mesh tally, users have had to accept different levels of relative uncertainty among the tallies or run separate calculations optimized for each individual tally. To address this limitation, an extension of the Consistent Adjoint Driven Importance Sampling (CADIS) method, which is used for difficult source/detector problems, has been developed to optimize several tallies or the cells of a mesh tally simultaneously. The basis for this method is the development of an importance function that represents the importance of particles to the objective of uniform Monte Carlo particle density in the desired tally regions. This method utilizes the results of a forward discrete ordinates solution, which may be based on a quick coarse-mesh calculation, to develop a forward-weighted source for the adjoint calculation. The importance map and the biased source computed from the adjoint flux are then used in the forward Monte Carlo calculation to obtain approximately uniform relative uncertainties for the desired tallies. This extension is called forward-weighted CADIS, or FW-CADIS.

STUDY OF FUKUSHIMA DAIICHI NUCLEAR POWER STATION UNIT 4 SPENT-FUEL POOL

A study on the Fukushima Daiichi nuclear power station spent-fuel pool (SFP) at Unit 4 (SFP4) is presented in this paper. We discuss the design characteristics of SFP4 and its decay heat load in detail and provide a model that we developed to estimate the SFP evaporation rate based on the SFP temperature. The SFP level of SFP4 following the March 11, 2011, accident is predicted based on the fundamental conservation laws of mass and energy. Our predicted SFP level and temperatures are in good agreement with measured data and are consistent with Tokyo Electric Power Company evaluation results.

REMOTE HANDLING AND PLASMA CONDITIONS TO ENABLE FUSION NUCLEAR SCIENCE R&D USING A COMPONENT TESTING FACILITY

Abstract The use of a fusion component testing facility to study and establish, during the ITER era, the remaining scientific and technical knowledge needed by fusion Demo is considered and described in this paper. This use aims to test components in an integrated fusion nuclear environment, for the first time, to discover and understand the underpinning physical properties, and to develop improved components for further testing, in a time-efficient manner. It requires a design with extensive modularization and remote handling of activated components, and flexible hot-cell laboratories. It further requires reliable plasma conditions to avoid disruptions and minimize their impact, and designs to reduce the divertor heat flux to the level of ITER design. As the plasma duration is extended through the planned ITER level (∼103 s) and beyond, physical properties with increasing time constants, progressively for ∼104 s, ∼105s, and ∼106 s, would become accessible for testing and R&D. The longest time constants of these are likely to be of the order of a week (∼106 s). Progressive stages of research operation are envisioned in deuterium, deuterium-tritium for the ITER duration, and deuterium-tritium with increasingly longer plasma durations. The fusion neutron fluence and operational duty factor anticipated for this “scientific exploration” phase of a component test facility are estimated to be up to 1 MW-yr/m2 and up to 10%, respectively.

AUTOMATED VARIANCE REDUCTION APPLIED TO NUCLEAR WELL-LOGGING PROBLEMS

Abstract Simulating nuclear well-logging devices with Monte Carlo methods is computationally challenging and requires significant variance reduction to compute detector responses with low statistical uncertainties in reasonable lengths of time. The consistent adjoint-driven importance sampling (CADIS) method, which provides consistent source and transport biasing parameters based on a deterministic adjoint (importance) function, has been demonstrated to be very effective for well-logging simulations and other deep-penetration problems. A recent extension to the CADIS method, FW-CADIS (forward-weighted CADIS), is designed to optimize the calculation of several tallies at once by using an adjoint function based on an adjoint source weighted by the inverse of the forward flux. These advanced variance reduction methods have been incorporated and automated into the MAVRIC sequence of SCALE, making them very easy to use. The CADIS and FW-CADIS methods are demonstrated and compared on simple benchmark models of both neutron- and photon-based well-logging devices. Both advanced variance reduction methods offer a substantial reduction in computing time, compared to analog simulation, for these applications.

Monte Carlo transport calculations and analysis for reactor pressure vessel neutron fluence

The application of Monte Carlo methods for reactor pressure vessel (RPV) neutron fluence calculations is examined. As many commercial nuclear light water reactors approach the end of their design lifetime, it is of great consequence that reactor operators and regulators be able to characterize the structural integrity of the RPV accurately for financial reasons, as well as safety reasons, due to the possibility of plant life extensions. The Monte Carlo method, which offers explicit three-dimensional geometric representation and continuous energy and angular simulation, is well suited for this task. A model of the Three Mile Island unit 1 reactor is presented for determination of RPV fluence; Monte Carlo (MCNP) and deterministic (DORT) results are compared for this application; and numerous issues related to performing these calculations are examined. Synthesized three-dimensional deterministic models are observed to produce results that are comparable to those of Monte Carlo methods, provided the two methods utilize the same cross-section libraries. Continuous energy Monte Carlo methods are shown to predict more (15 to 20%) high-energy neutrons in the RPV than deterministic methods.

Sensitivity and Uncertainty Analysis of Commercial Reactor Criticals for Burnup Credit

Abstract This paper provides insights into the neutronic similarities between a representative high-capacity rail-transport cask containing typical pressurized water reactor (PWR) spent nuclear fuel assemblies and critical reactor state-points, referred to as commercial reactor critical (CRC) state-points. Forty CRC state-points from five PWRs were analyzed, and the characteristics of CRC state-points that may be applicable for validation of burnup-credit criticality safety calculations for spent fuel transport/storage/disposal systems were identified. The study employed cross-section sensitivity and uncertainty analysis methods developed at Oak Ridge National Laboratory and the TSUNAMI set of tools in the SCALE code system as a means to investigate neutronic similarity on an integral and nuclide-reaction-specific level. The results indicate that except for the fresh-fuel-core configuration, all analyzed CRC state-points are either highly similar, similar, or marginally similar to the representative high-capacity cask containing spent nuclear fuel assemblies with burnups ranging from 10 to 60 GWd/tU in terms of their shared uncertainty in keff due to cross-section uncertainties. On a nuclide-reaction-specific level, the CRC state-points provide significant coverage, in terms of neutronic similarity, for most of the actinides and fission products relevant to burnup credit. Hence, in principle, the evaluated CRC state-points could serve as part of a set of benchmark experiments for determining a bias and bias uncertainty to be applied to the calculated keff of a spent fuel transport/storage/disposal system to correct for approximations in computational methods and errors and uncertainties in nuclear data. Note, however, that an evaluation to quantify the uncertainties associated with various CRC modeling parameters (e.g., fuel isotopic compositions, physical characteristics of reactor core components, and reactor operating history information), which has relevance to the use of these critical configurations for bias determination, was not performed as part of this study.

Review and Prioritization of Technical Issues Related to Burnup Credit for LWR Fuel

The present invention relates to a polymer composition, to a layer element, preferably to at least one layer element of a photovoltaic module, comprising the polymer composition and to an article which is preferably said at least one layer of a layer element, preferably of a layer element of a photovoltaic module.