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At least 253 records · Page 14

A Gibbs Energy Minimization Approach for Modeling of Chemical Reactions in a Basic Oxygen Furnace

In modern steelmaking, the decarburization of hot metal is converted into steel primarily in converter processes, such as the basic oxygen furnace. The objective of this work was to develop a new mathematical model for top blown steel converter, which accounts for the complex reaction equilibria in the impact zone, also known as the hot spot, as well as the associated mass and heat transport. An in-house computer code of the model has been developed in Matlab. The main assumption of the model is that all reactions take place in a specified reaction zone. The mass transfer between the reaction volume, bulk slag, and metal determine the reaction rates for the species. The thermodynamic equilibrium is calculated using the partitioning of Gibbs energy (PGE) method. The activity model for the liquid metal is the unified interaction parameter model and for the liquid slag the modified quasichemical model (MQM). The MQM was validated by calculating iso-activity lines for the liquid slag components. The PGE method together with the MQM was validated by calculating liquidus lines for solid components. The results were compared with measurements from literature. The full chemical reaction model was validated by comparing the metal and slag compositions to measurements from industrial scale converter. The predictions were found to be in good agreement with the measured values. Furthermore, the accuracy of the model was found to compare favorably with the models proposed in the literature. The real-time capability of the proposed model was confirmed in test calculations.

36 MATERIALS SCIENCE↗

Core polarization effects: Oscillator strengths, transition probabilities and radiative lifetimes of levels in Bi IV

In this work a systematic analysis of Bi IV atomic properties is presented. Ab initio relativistic Hartree–Fock calculations in an intermediate coupling (IC) formalism and taking into account Core Polarization Effects (CPE) are used. We use for the IC calculations the standard method of least square fitting of experimental energy levels by means of computer codes from Cowan which was modified in order to include CPE. Our calculations have been made from 17 configurations of Bi IV: 5d{sup 10}6s{sup 2}, 5d{sup 10}6p{sup 2}, 5d{sup 10}6sns (n= 7–10) and 5d{sup 10}6snd (n= 6–9) for even parity and 5d{sup 10}6snp (n=6,7), 5d{sup 10}6snf (n=5,6), 5d{sup 9}6s{sup 2}6p and 5d{sup 9}6s{sup 2}nf (n=5,6) for odd parity. Transition probabilities and oscillator strengths for 100 lines and some radiative lifetimes of Bi IV have been calculated. Besides a comparison of our results with the scarce experimental values available in the bibliography are presented. The presence of spectral lines of bismuth in stellar atmospheres has been reported in different stars. This is the motivation by this work can be very interest in astrophysical area. Recently the Bi IV is taking importance because it is considered as the main dopant in the prospective used for the design and synthesis of new scintillators.

74 ATOMIC AND MOLECULAR PHYSICS↗

An algorithm to calculate a collapsed arc dose matrix in volumetric modulated arc therapy

Purpose: The delivery of volumetric modulated arc therapy (VMAT) is more complex than other conformal radiotherapy techniques. In this work, the authors present the feasibility of performing routine verification of VMAT delivery using a dose matrix measured by a gantry mounted 2D ion chamber array and corresponding dose matrix calculated by an inhouse developed algorithm.Methods: Pinnacle, v9.0, treatment planning system (TPS) was used in this study to generate VMAT plans for a 6 MV photon beam from an Elekta-Synergy linear accelerator. An algorithm was developed and implemented with inhouse computer code to calculate the dose matrix resulting from a VMAT arc in a plane perpendicular to the beam at isocenter. The algorithm was validated using measurement of standard patterns and clinical VMAT plans with a 2D ion chamber array. The clinical VMAT plans were also validated using ArcCHECK measurements. The measured and calculated dose matrices were compared using gamma ({gamma}) analysis with 3%/3 mm criteria and {gamma} tolerance of 1.Results: The dose matrix comparison of standard patterns has shown excellent agreement with the mean {gamma} pass rate 97.7 ({sigma}= 0.4)%. The validation of clinical VMAT plans using the dose matrix predicted by the algorithm and the corresponding measured dose matrices also showed good agreement with the mean {gamma} pass rate of 97.6 ({sigma}= 1.6)%. The validation of clinical VMAT plans using ArcCHECK measurements showed a mean pass rate of 95.6 ({sigma}= 1.8)%.Conclusions: The developed algorithm was shown to accurately predict the dose matrix, in a plane perpendicular to the beam, by considering all possible leaf trajectories in a VMAT delivery. This enables the verification of VMAT delivery using a 2D array detector mounted on a treatment head.

61 RADIATION PROTECTION AND DOSIMETRY↗

Identification and Quantification of Beta-Emitters in Waste Packages by Segmented Gamma-Scanning - 20114

A commonly applied method for the non-destructive characterization of the radioactive inventory of waste packages (e.g. 200-l drums) is segmented gamma-scanning. The gamma-emitting inventory is identified via their characteristic lines in the measured spectra. The corresponding net peak areas allow for quantification applying appropriate models. The identification and quantification of beta-emitters present in the waste package require further evaluation. While some of them emit gamma-rays, too, which can be used for identification and quantification, pure beta-emitters contribute to the gamma-spectra only by their Bremsstrahlung-spectra. {sup 32}P, {sup 90}Sr and {sup 90}Y are typical representatives of these isotopes. A method is developed for identification and quantification of these pure beta-emitters in radioactive waste packages. Origin of the procedure is the evaluation of the measured gamma-spectrum. After the identification of all lines present in the spectrum (characteristic gamma-lines, single- and double-escape-lines, X-ray-lines, annihilation peak etc.) this information is used together with calibration data of the detection system and further a-priori information like matrix composition, cross-section data from databases etc. to simulate the spectra without the Bremsstrahlung-components by an appropriate method. This may be Monte Carlo simulations or, like in our case, a computer code taking into account mathematical descriptions of the physical effects like photo effect, (multiple) Compton scattering, backscattering etc. and the attenuation effects of the waste matrix. In any case, the resulting spectrum must simulate the real spectrum without its Bremsstrahlung-component as good as possible. Subtracting the simulated from the measured spectrum then results in the Bremsstrahlung-components caused by the beta-emitters. Prior to evaluation, a set of Bremsstrahlung-spectra is calculated and stored for different beta-emitters in different matrices using the Bethe-Heitler equation. This data set can be extended for new conditions, if necessary, and is allocated in a data base for future application. Applying a non-linear optimization using the generalized reduced gradient method for comparing the calculated with the derived Bremsstrahlung-spectra results in the identification of the beta-emitter and a factor, which is proportional to its activity. The proportionality is determined by previous calibration measurements. The application of the method is demonstrated by measurements on a 200-l mock-up drum and a real waste package. (authors)

07 ISOTOPE AND RADIATION SOURCES↗

Beam dynamics in an ultra-low energy storage rings (review of existing facilities and feasibility studies for future experiments)

Storage rings operating at ultra-low energies and in particular electrostatic storage rings have proven to be invaluable tools for atomic and molecular physics. Due to the mass independence of the electrostatic rigidity, these machines are able to store a wide range of different particles, from light ions to heavy singly charged bio-molecules. However, earlier measurements showed strong limitations on beam intensity, fast decay of ion current, reduced life time etc. The nature of these effects was not fully understood. Also a large variety of experiments in future generation ultra-low energy storage and decelerator facilities including in-ring collision studies with a reaction microscope require a comprehensive investigation of the physical processes involved into the operation of such rings. In this paper, we present review of non-linear and long term beam dynamics studies on example of the ELISA, AD Recycler, TSR and USR rings using the computer codes BETACOOL, OPERA-3D and MAD-X. The results from simulations were benchmarked against experimental data of beam losses in the ELISA storage ring. We showed that decay of beam intensity in ultra-low energy rings is mainly caused by ion losses on ring aperture due to multiple scattering on residual gas. Beam is lost on ring aperture due to small ring acceptance. Rate of beam losses increases at high intensities because of the intra-beam scattering effect adds to vacuum losses. Detailed investigations into the ion kinetics under consideration of the effects from electron cooling and multiple scattering of the beam on a supersonic gas jet target have been carried out as well. The life time, equilibrium momentum spread and equilibrium lateral spread during collisions with this internal gas jet target were estimated. In addition, the results from experiments at the TSR ring, where low intensity beam of CF{sup +} ions at 93 keV/u has been shrunk to extremely small dimensions have been reproduced. Based on these simulations, conditions for stable ring operation with extremely low emittance beam are presented. Finally, results from studies into the interaction of ions with a gas jet target at 3–30 keV energy range are summarized.

43 PARTICLE ACCELERATORS↗

Performance Analyses of Passive Containment Cooling System for a Concrete Type Containment

The containment of nuclear power plants is the final barrier to keep the radioactive materials from uncontrolled release to the environment. To keep the containment integrity, pressure and temperature must be controlled by the spray system which relies on active components for PWRs with a concrete containment. To achieve a passive means to keep the containment integrity, therefore, cooling should be achieved by forming a natural circulation loop between internal cooling devices connected to an external heat sink. A system like this is often called Passive Containment Cooling System (PCCS). In the early phase of APR-1400 development, this concept was studied and VVER-1200 is also known to adopt the PCCS. The PCCS consists of a cooling device installed at the inside of the containment which is a tube-bank type heat exchanger where cooling occurs at the exterior of the tubes and a heat sink which is called Passive Containment Cooling Tank (PCCT) located at the outside of the containment on the top of the auxiliary building. In this paper, we introduce performance analysis results, utilizing the GOTHIC computer code. The PCCS performance analysis model was developed for APR-1400 which has a thermal power of 4000 MW and considers PCCS as an option to improve its safety characteristics. The PCCS which is considered in this paper is to cope with both Design Basis and Severe Accidents, considering SBO conditions. The pressure and temperature transients due to the Loss of Coolant Accident (LOCA) were analyzed with and without PCCS together with a normal case using the spray system. And, in particular, limitations of the PCCS in regard to mission time will be discussed. Through this study, it was found that the PCCS being considered as an option to improve the safety of APR-1400 can have almost same capability to control the containment pressure for LBLOCA as the containment spray system does. Although it would depends on the heat removal capacity of a given design such as the size of the heat transfer area, the PCCS could even replace the containment spray system. However, the PCCS may lose its capability after 24 hours as its temperature reaches that of the containment atmosphere. To extend the mission time of PCCS, either cooling or supplementary cold water needs to be provided to the external heat sink of PCCS. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Radiometric characterization process for locating radioactivity hold-up and measuring non-uniformly distributed sources - 15342

A novel approach to facility radiometric characterization for non-uniformly distributed radiation sources has been developed by ANTECH and is currently being employed at the Argonne National Laboratory. It is being applied to the measurement of a wide range of objects and the interiors and contents of laboratories and buildings. The integrated non destructive assay measurement process involves the use of a scanning spectroscopic gamma ray camera as well as high purity Germanium detectors with different collimators and hence differing detector fields of view to measure and quantify non-uniformly distributed sources of radioactivity in laboratory equipment, glove boxes, storage tanks, pipework and general laboratory objects and process components. The purpose of the radiometric characterization is to quantify radiation sources and radioactive contamination levels in facilities containing nuclear material and to ensure compliance with facility characterization requirements and for decontamination and decommissioning planning and execution. The integrated measurement process also involves automated data capture and processing software and a database with which both measurement instrument operators and subject matter expert analysts can interact. An appropriate measuring instrument (gamma camera, scintillation detector, high purity Germanium detector system, neutron detector or alpha-beta counter) records the measurement data and the measurement site is recorded in a photograph. All of the data relevant to the measurement (or set of measurements for a particular object) are recorded in the database including measured gamma ray spectra, equipment description and relevant calibration, measurement site photograph and measurement descriptive data. Once data analysis protocols are established for specific geometries and measurement situations they can be applied to the measured data in a partially automated manner, reducing analysis and transcription errors. In addition to measurements and data analysis the process also utilises radiometric modelling computer codes including MicroShield and Monte Carlo Neutrons Photons in order to perform model calculations where appropriate. In a typical measurement, the ANTECH RadSearch spectroscopic gamma camera is utilized to find areas of concentrated activity so that these areas can be measured and calculated separately from areas with diffuse contamination. The use of the gamma camera in scanning mode helps to ensure that the assay does not overestimate the detected radiation source, which is a frequent consequence of using standard far field geometry and attenuation analysis codes that assume uniformity of both density and radioactive source distribution. Once the hot spots are identified, high purity Germanium detector measurements are made using various collimators with wide or narrow fields of view as required. These measurements are then processed through a semi-automatic calculation process assuming an appropriate source configuration as identified by RadSearch. These source configurations include point sources, line sources or surface sources. The Generalized Geometry Hold-up algorithms are employed to determine an estimate of the activity of the various sources that are identified in a measurement. The ANTECH integrated measurement process provides an automated calculation procedure where the analyst can select the source type, geometry and appropriate input data, and then the software calculates the activity by radionuclide for all sources identified in the measurement. The measurement and analysis process is controlled under the ANTECH ISO9001: 2008 Quality Assurance System and managed by a site specific Quality Assurance Plan. Both Argonne National Laboratory and Department of Energy technical reviewers have vetted the Integrated Measurement and Analysis Process. The application of the measurement and analysis process is described and measurement examples are provided. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Preliminary Design of the TEX-MOX Criticality Experimental Program using Optimization Algorithms Implemented in the IRSN PROMETHEE Tool

In the frame of its missions, IRSN is widely involved in design, evaluation and analyses of critical experiments in order to contribute to computer codes and nuclear data validation. The work described in this paper was carried out in the framework of the PRINCESS IRSN research project and is part of the IRSN/LLNL collaboration on the TEX program in the frame of the NCSP. The need for epithermal and intermediate energy range critical benchmarks is an established international criticality safety data need, underpinned by the subsequent use of data assimilation methods like GLLSM to reduce Nuclear Data uncertainties. The goal of the TEX program is to address this need by executing experiments with NCSP fissile assets, which can be used to create critical assemblies that span a wide range of fission energy spectra, from the thermal energy range (below 0.625 eV), through the intermediate energy range (0.625 eV to 100 keV), to the fast one (above 100 keV). The TEX program involves different materials of interest, such as TEX-Ta experiments using plutonium ZPPR plates proposed by LLNL. This paper presents an overview of the studies performed at IRSN using numerical optimization algorithms for the design of critical experiments involving Mixed U and Pu fuel plates, referred to as TEX-MOX program. This program ultimately aims to perform critical experiments to validate the MOX fuel configurations encountered in the MOX fuel fabrication process. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

TIME-DEPENDENT PHOTOIONIZATION OF GASEOUS NEBULAE: THE PURE HYDROGEN CASE

We study the problem of time-dependent photoionization of low density gaseous nebulae subjected to sudden changes in the intensity of ionizing radiation. To this end, we write a computer code that solves the full time-dependent energy balance, ionization balance, and radiation transfer equations in a self-consistent fashion for a simplified pure hydrogen case. It is shown that changes in the ionizing radiation yield ionization/thermal fronts that propagate through the cloud, but the propagation times and response times to such fronts vary widely and nonlinearly from the illuminated face of the cloud to the ionization front (IF). IF/thermal fronts are often supersonic, and in slabs initially in pressure equilibrium such fronts yield large pressure imbalances that are likely to produce important dynamical effects in the cloud. Further, we studied the case of periodic variations in the ionizing flux. It is found that the physical conditions of the plasma have complex behaviors that differ from any steady-state solution. Moreover, even the time average of ionization and temperature is different from any steady-state case. This time average is characterized by overionization and a broader IF with respect to the steady-state solution for a mean value of the radiation flux. Around the time average of physical conditions there is a large dispersion in instantaneous conditions, particularly across the IF, which increases with the period of radiation flux variations. Moreover, the variations in physical conditions are asynchronous along the slab due to the combination of nonlinear propagation times for thermal fronts/IFs and equilibration times.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗

Obtain3D_open

This computer code uses the concept of the parallax to compute the x, y and z coordinates of points found using transmission electron microscopy (TEM), or any transmission imaging technique, using two images, each taken at a different perspective of the region containing the points. Points correspond, but are not limited, to the center of cavities or precipitates, positions of irradiation black dot damage, positions along a dislocation line, or positions along where an interface meets a free surface. Features can include dislocations, interfaces, cavities, precipitates, inclusions etc. The x, y and z coordinates of the points are output in a text file. The code also allows the user to visualize the features containing the points in three dimensions.

Eftink, Benjamin↗

Applying fast calorimetry analysis on spent fuel calorimeter - 15620

Swedish Nuclear Fuel and Waste Management Company (SKB) is responsible for the back-end of the Swedish nuclear industry. This includes transport of spent fuel by boat from the power plants to the Central Interim Storage Facility for Spent Fuel (Clab), and in the future, encapsulation of the spent fuel and further transportation to a geological repository. In order to ensure safety for the planned geological repository, the decay heat has to be well known. For this purpose the SKB has performed calorimetric measurements on over 200 pressurized water reactor (PWR) and boiling water reactor (BWR) assemblies since 2003. The results from earlier analyses have been used to improve computer codes predicting heat as well as to acquire deeper knowledge about the fuel inventory. The work complements research being done as part of the Next Generation of Safeguards Initiative - Spent Fuel project, whose purpose is to develop and test nondestructive assay (NDA) technologies to improve safeguards measurements of spent nuclear fuel. The project is working to accomplish multiple goals, one of which is to determine the heat content of spent fuel assemblies. To shorten the measurement time, the temperature increase method was used instead of the equilibrium temperature method. In the temperature increase method, the temperature inside of the calorimeter is lowered below the surrounding water temperature before the measurement, and the temperature is allowed to increase due to the higher surrounding water temperature during the measurement. The rate of the temperature increase when the temperature is the same as the surrounding temperature is then compared to a calibration curve where an electric heater has been used in the same way. The price for the higher throughput is less accuracy and dependency of fuel geometry since the heat flow will be different for different fuel types, especially between PWR and BWR assemblies. Recently at Los Alamos National Laboratory, more sophisticated analysis methods have been considered for the use of calorimetry for treaty verification. These methods aim to predict the equilibrium temperature based on early data and therefore be able to shorten the measurement time while maintaining good accuracy and eliminate the dependency on fuel geometry. Even though the techniques show promise for future calorimetry of spent fuel, certain design features of the current system have been identified to interfere with the accuracy of equilibrium measurements in general such as changing pool temperatures and heat leakage during calibration. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Rich-Text based PSA Model Transformation and Verification in RiskA

Nuclear Energy is one of the most possible alternative to fossil energy in large-scale. Nuclear safety is the lifeline of nuclear energy development, and has been an important part of national security, especially after the accident of Fukushima. On the basis of computer codes, reliability engineering and Probabilistic Risk Assessment (PSA) is effective methods to improve the safety of nuclear power plants. After many years of development, the methodologies, applications and relevant soft programs have been quite sufficient in the area of nuclear safety. However, it is still difficult for the current PSA technologies and software to meet the needs of dealing with the problems of quantitatively public risk assessment in the digital society. Especially, there are so many software that adopted different modeling and analyzing methods, which brings difficulties to the comparison, evaluation or verification for different software and models. By using its own academic advantages, the advanced nuclear energy research team, i.e., FDS Team has invested a few hundreds person-years of independent research and development of a reliability and probabilistic safety assessment software system RiskA. The system consists of three main functions: device reliability optimization management, probabilistic safety assessment and instantaneous risk monitoring, and the comprehensive public risk assessment in digital society. RiskA establishes an integrated risk evaluation system for the device reliability, nuclear power plant (NPP) risk monitoring and public risk assessment. Currently, RiskA has successfully been applied in more than 600 organizations and institutes from a variety of fields in 39 countries, involving nuclear, aerospace, defense and so on. For example, RiskA has been used in the safety analysis of International Thermonuclear Experimental Reactor (ITER), FDS series fusion reactor, and the Strategic Priority Research Program of Chinese Academy of Sciences 'the Future Advanced Nuclear Fission Energy - Accelerator Driven Sub-critical (ADS) transmutation system'. Also, RiskA has been used in public risk communication and risk assessment. Among them, Third Qinshan nuclear power plant Risk Monitor (TQRM), which is the China's first domestic nuclear safety analysis software used in the actual online operation, has been highly apprised by renowned experts in the field. Meanwhile, it has been identified as the first domestic nuclear power plant risk monitoring system with completely independent intellectual property rights, and has reached the international advanced level. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Assessment of defense-related uranium mines in the United States and other US DOE office of legacy management domestic and international uranium initiatives - 15335

US DOE's Office of Legacy Management (LM) has taken several initiatives to improve domestic uranium mining and milling operations in the United States and to contribute to the management of legacy uranium mining and milling sites around the world. Initiatives include completing the 2014 Report to Congress on Defense-Related Uranium Mines, managing the Title X Uranium and Thorium Program with the Office of Environmental Management, and assisting IAEA with uranium issues in member state countries. The 2014 Report to Congress identified abandoned uranium mines (AUMs) in the United States that provided uranium ore to the US Atomic Energy Commission (AEC) between 1947 and 1970. The report identified AUM locations, reclamation/remediation status, radiological risks and other hazards, costs for reclamation and remediation, and potential prioritization schemes for mine cleanup. To develop reclamation and remediation cleanup cost estimates and assess risks, AUMs were grouped into six categories by tons of uranium ore produced, ranging from Small (less than 91 metric tons [100 tons]) to Very Large (greater than 453,900 metric tons [500,000 tons]). Radiological risk was calculated using the RESRAD computer code for five exposure scenarios: offsite resident, onsite resident, occasional visitor, recreational visitor, and mine reclamation worker. Results indicate that 69 percent of the AUMs are in Colorado and Utah and 23 percent are in Arizona, Wyoming, and New Mexico; and that the majority of the AUMs (68 percent) were considered Small or Small/Medium production mines, producing less than 908 metric tons (1,000 tons) of uranium ore. Although New Mexico has fewer AUMs (247) than the other states in the Colorado Plateau region, AUMs in that state (primarily those in the Grants Mineral Belt) produced 45 percent of the 79.5 million tons of AEC-purchased uranium ore. About 50 percent of the AUMs are on public land managed by the US Bureau of Land Management; 435 AUMs are on the Navajo Nation. Risk estimates for the onsite resident scenario (plausible on tribal and non-federal land) could result in an incremental cancer risk greater than 10{sup -4}. Radon inhalation was the dominant contributor to radiological risk for the five exposure scenarios evaluated. Different state and federal agencies are conducting cleanup of some AUMs under various remedial and legal authorities. Activities constituting mine 'reclamation' may reduce radiological risks to humans to acceptable levels for many mines on federal public lands if the occasional visitor and recreational visitor exposure scenarios are assumed. Addressing physical hazards (e.g., open shafts) at mines where conditions could cause serious injuries is typically a priority of public land management agencies. At locations where mine-related groundwater contamination occurs, it is a significant contributor to cleanup costs. However, most Small and Small/Medium AUMs were likely developed above the water table, and some wet mines are located in areas where groundwater has high, naturally occurring levels of the same constituents (including radioactive elements) that are typically associated with uranium mines. LM is also administering, with the US DOE Office of Environmental Management, the Title X program, which was developed by Congress to reimburse private companies (i.e., those that held active operating licenses prior to 1978) for a portion of their reclamation costs for mill sites that processed ore sold to AEC. LM recently updated the program guidance document and established internal protocols for annually auditing companies. LM is assisting IAEA on two different initiatives: assisting developing countries with training on the basic steps to take to perform reclamation of uranium mines, and developing technical guidance and case studies related to safety and environmental assessments and post-closure management of uranium legacy sites as part of the IAEA Regulatory Supervision of Legacy Sites initiative. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

User Guide for the R5EXEC Coupling Interface in the RELAP5-3D Code

This report describes the R5EXEC coupling interface in the RELAP5-3D computer code from the users perspective. The information in the report is intended for users who want to couple RELAP5-3D to other thermal-hydraulic, neutron kinetics, or control system simulation codes.

97 MATHEMATICS AND COMPUTING↗

Monte Carlo Second- and Third-Order Many-Body Green’s Function Methods with Frequency-Dependent, Nondiagonal Self-Energy

We fully develop the Monte Carlo many-body Green’s function (MC-GF) method with the following enhancements: (1) The truncation order of the perturbation expansion of the Dyson self-energy is raised from the second order (MC-GF2) to the third order (MC-GF3) with the aid of a computerized procedure to enumerate and transform all 84 third-order Goldstone diagrams into Monte Carlo integrable expressions and then into central processing unit (CPU)/graphical processing unit (GPU)-parallel computer codes. (2) An efficient algorithm is proposed that computes all off-diagonal and diagonal elements of the MC-GF2 and MC-GF3 self-energy matrices by common subexpression elimination. (3) The frequency-independent approximation is lifted by introducing a method that computes frequency derivatives of the MC-GF2 and MC-GF3 self-energies up to any arbitrarily high order at nearly no additional computational cost. (4) The imaginary-time integration in the Laplace-transformed expressions of the self-energy is carried out stochastically (instead of using a quadrature in the previous implementations), resulting in a 50- to 200-fold speedup. (5) The efficiency of the redundant-walker convergence acceleration scheme is analyzed numerically, and the guidelines are established to select an optimal number of walkers for maximal efficiency. When such an optimal number is used, the cost per sample is constant of molecular size on either many CPUs or many GPUs. (6) The computational cost to obtain a binding energy within a given statistical uncertainty is observed to increase as (tentatively) O(n 4 ) and O(n 5 ) of molecular size (n is the number of electrons) with and without the diagonal approximation, respectively, as opposed to O(n 5 ) and O(n 6 ) of the corresponding deterministic algorithms. With this method applied to the electron binding energies of C 60 , we show that the third-order corrections to the self-energies are much greater in electron binding energies than in ground-state energies. As a result, they display a sign of oscillatory convergence toward experimental results, not necessarily improving the agreement with increasing perturbation order, justifying MC-GF3 and motivating even higher-order methods.

Algorithms↗

Advanced Potential Energy Surfaces for Molecular Simulation

Advanced potential energy surfaces are defined as theoretical models that explicitly include many-body effects that transcend the standard fixed-charge, pairwise-additive paradigm typically used in molecular simulation. However, several factors relating to their software implementation have precluded their widespread use in condensed-phase simulations: the computational cost of the theoretical models, a paucity of approximate models and algorithmic improvements that can ameliorate their cost, underdeveloped interfaces and limited dissemination in computational code bases that are widely used in the computational chemistry community, and software implementations that have not kept pace with modern high-performance computing (HPC) architectures, such as multicore CPUs and modern graphics processing units (GPUs). Here in this Feature Article we review recent progress made in these areas, including well-defined polarization approximations and new multipole electrostatic formulations, novel methods for solving the mutual polarization equations and increasing the MD time step, combining linear-scaling electronic structure methods with new QM/MM methods that account for mutual polarization between the two regions, and the greatly improved software deployment of these models and methods onto GPU and CPU hardware platforms. Finally, we have now approached an era where multipole-based polarizable force fields can be routinely used to obtain computational results comparable to state-of-the-art density functional theory while reaching sampling statistics that are acceptable when compared to that obtained from simpler fixed partial charge force fields.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

SNL/JAEA Collaborations on Sodium Fire Benchmarking.

Two sodium spray fire experiments performed by Sandia National Laboratories (SNL) were used for a code - to - code comparison between CONTAIN - LMR and SPHINCS. Both computer codes are used for modeling sodium accidents in sodium fast reactors. The comparison between the two codes provides insights into the ability of both codes to model sodium spray fires. The SNL T3 and T4 experiments are 20 kg sodium spray fires with sodium spray temperature s of 200 deg C and 500 deg C, respe ctively. Given the relatively low sodium temperature in the SNL T3 experiment, the sodium spray experienced a period of non - combustion. The vessel in the SNL T4 experiment experienced a rapid pressurization that caused of the instrumentation ports to fail during the sodium spray. Despite these unforeseen difficulties, both codes were shown in good agreement with the experiment s . The subsequent pool fire that develops from the unburned sodium spray is a significant characteristic of the T3 experiment. SPHIN CS showed better long - term agreement with the SNL T3 experiment than CONTAIN - LMR. The unexpected port failure during the SNL T4 experiment presented modelling challenges. The time at which the port failure occurred is unknown, but is believed to have occur red at about 11 seconds into the sodium spray fire. The sensitivity analysis for the SNL T4 experiment shows that with a port failure, the sodium spray fire can still maintain elevated pressures during the spray.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

COMPUTATION FLUID DYNAMICS ANALYSIS FOR GENERIC SMALL MODULAR REACTOR CONTAINMENT SEPARATE EFFECTS TEST

It is desirable for fourth-generation Small Modular Reactors to be passively cooled in standard and accident operations. Passive Containment Cooling Systems can reject heat from the containment structure, without using pumps or blowers. The targeted design containment structure is a large, domed, stainless steel, cylindrical vessel. In a postulated Design Basis Accident, steam will flash inside containment. Steam condensation occurs on the inner containment wall and transfers heat through the steel containment into a large body of water known as the annular reservoir (AR) surrounding the vessel serving as the ultimate heat sink. Natural circulation drives the flow in the AR and heat will be released to the environment by evaporation of water. Unique containment geometry requires a separate effects test (SET) facility for the verification and validation of the computer code and evaluation model development and assessment for reactor licensing efforts. In this study, STAR-CCM+, a computational fluid dynamics (CFD) code was used to inform the decision-making process on the design of the SET. The CFD simulation modeled, a two-phase turbulent flow with fluid film development and heat transfer for different containment geometries. The Reactor Excursion and Leak Analysis Program will also be used in a code-to-code verification against the CFD results.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗