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At least 415 records · Page 23

Assessment of the Griffin Reactor Multiphysics Application Using the Empire Micro Reactor Design Concept

In late 2019, INL and ANL agreed to jointly develop the reactor physics code named Griffin based on the integration of the two code suites, MAMMOTH/Rattlesnake (INL) and MC2 - 3/PROTEUS (ANL). Griffin is being developed based on the MOOSE framework and MOOSE quality assurance procedures. This decision was made to be able to allow DOE-NE to efficiently invest funding to this area and to provide effective and timely support for existing and potential users; the latter includes industry and government organizations who are developing various types of advanced reactors in the near and long term. Since MAMMOTH/Rattlesnake has been developed based on the MOOSE framework, the INL/ANL Griffin development team agreed to build Griffin beginning with a merger of MAMMOTH and Rattlesnake into a single code and moving forward by implementing capabilities from the PROTEUS suite into Griffin. Moving forward, both ANL and INL efforts are equally invested in the Griffin project, with management support, to provide an advanced reactor multiphysics tool to assist in reactor design, optimization, and safety analysis. Much work remains in moving Griffin forward to migrate PROTEUS capabilities and to optimize performance to meet user needs. The main objective of this work is to assess the current status of Griffin capabilities in terms of performance and accuracy, to determine priorities for PROTEUS migration, and to identify capabilities and features to improve for supporting the code integration effort. For this assessment, the Empire micro reactor problem that was developed in the ARPA-E MEITNER program was selected as an advance reactor concept of interest to the technical community. The Empire reactor problem was expanded from its original incomplete specification to be a small heat-pipe-cooled micro reactor core with ~113 cm radius and 70 cm in height, composed of 18 fuel assemblies, 12 control drums, and beryllium radial and axial reflectors. In the current model, using 5 cm axial reflectors specified in the original Empire assembly model, more than 10% of neutrons leak axially and through the empty center safety hole, as well as through heat pipe channels in fuel assembly elements that extend through the top reflector region. Several calculation models of the core were defined for systematic assessment, including 2-D and 3-D fuel assemblies and whole cores with cylindrical boundaries. Cross sections were generated using Serpent 2, and meshes were produced using the Argonne mesh tool or the INL neutronics meshing tools combined with CUBIT. Cross sections and meshes were converted to the ISOXML and Exodus formats, respectively, so that Griffin and PROTEUS could use consistent data for solving the reactor problems. With the prepared cross sections and meshes, PROTEUS was run first to ensure that all input data were correctly generated and input options in terms of angle, mesh, and energy group were accurately determined. Comparisons against Serpent 2 solutions were made in terms of eigenvalue and pin power. The same calculations and comparisons were then conducted using Griffin. For the fuel assembly and whole core problems, the PROTEUS eigenvalues agreed well with reference Serpent 2 solutions within 100 and 30 pcm, respectively, and pin power differences relative to Serpent 2 were overall less than 2.2% and RMS 0.8% for the whole core models. This indicated that all input data were properly prepared. Using the same data, Griffin was run selecting the SAAF-CFEM SN solver with Legendre-Gaussian quadrature and NDA and DSA for acceleration. It was found that the SAAF-CFEM solver of Griffin required finer meshes to achieve eigenvalue and pin power solutions in good agreement with Serpent 2, consequently requiring more memory requirement and longer computation time. On the other hand, the SPH-Diffusion 2-D core calculations performed using Griffin were able to recover the exact eigenvalue from the reference Serpent 2 solutions, resulting in a pin-power distribution with an RMS of 0.6% and maximum absolute difference of less than 1.4%. The runtimes for SPH-Diffusion for the 2-D core were less than 3 minutes on 40 cores. During this evolution of this evaluation, many updates were made in Griffin by the Griffin development team of INL (focusing on software updates) and ANL (reviewing and supporting software updates) to complete this assessment. Observations from the code assessment are presented in the conclusion section of this report, followed by a discussion of recommendations for future work.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Pretest Simulations of a Supersonic Mixing and Combustion Validation Experiment to Assess Sensitivities

The reliance on CFD simulations to develop, design, and optimize scramjet systems (or components) has become commonplace. This reliance inevitably hinges on the ability of the computational analyst to quantify the level of confidence in their computational results. Unfortunately, nearly all the measured data available for this assessment comes from antiquated experimental datasets, or from tests that focused on the extraction of scramjet system (or component) performance. The objective of a CFD validation experiment is to quantify the predictive accuracy of one or more of the CFD physics submodels, implying that other uncertainties related to replicating the facility flow environment (e.g., knowledge of boundary conditions) must be minimized to the extent possible. This inevitably places stringent requirements on the quality and quantity of measurements taken to accurately specify inflow, outflow, and surface conditions for the CFD simulations; in addition to the measurements taken for the validation of physics submodels. This places additional demands on the experimental process above and beyond those for test article performance assessment. A recent high speed code credibility workshop series sponsored by AFRL identified a gap in existing validation data for fundamental assessments of turbulent mixing and combustion CFD closure models at scramjet engine relevant conditions. To address this gap, engineers at AFRL have designed a coaxial jet flame configuration that will be tested at two facilities (Research Cell 19 at the Air Force Research Lab, and at Purdue University). The effort described here documents pretest simulations of this validation experiment with the goal of fleshing out the extent of the facility flowpath that must be included to adequately reproduce the facility test section flow environment. The findings indicate that the flow around the support structure for the fuel injection centerbody upstream of the facility nozzle generates disturbances that persist throughout the nozzle expansion process; corrupting the azimuthal symmetry that was desired in the fuel/air mixing region of the test section. Simulations without this support structure maintained a high degree of azimuthal symmetry up until the fuel injection plane. However, even in this scenario the azimuthal symmetry was not maintained once the centerbody boundary layer transitioned to a wake flow downstream of the fuel injection plane.

CFD↗

Pretest Simulations of a Supersonic Mixing and Combustion Validation Experiment to Assess Sensitivities

The reliance on CFD simulations to develop, design, and optimize scramjet systems (or components) has become commonplace. This reliance inevitably hinges on the ability of the computational analyst to quantify the level of confidence in their computational results. Unfortunately, nearly all the measured data available for this assessment comes from antiquated experimental datasets, or from tests that focused on the extraction of scramjet system (or component) performance. The objective of a CFD validation experiment is to quantify the predictive accuracy of one or more of the CFD physics submodels, implying that other uncertainties related to replicating the facility flow environment (e.g., knowledge of boundary conditions) must be minimized to the extent possible. This inevitably places stringent requirements on the quality and quantity of measurements taken to accurately specify inflow, outflow, and surface conditions for the CFD simulations; in addition to the measurements taken for the validation of physics submodels. This places additional demands on the experimental process above and beyond those for test article performance assessment. A recent high speed code credibility workshop series sponsored by AFRL identified a gap in existing validation data for fundamental assessments of turbulent mixing and combustion CFD closure models at scramjet engine relevant conditions. To address this gap, engineers at AFRL have designed a coaxial jet flame configuration that will be tested at two facilities (Research Cell 19 at the Air Force Research Lab, and at Purdue University). The effort described here documents pretest simulations of this validation experiment with the goal of fleshing out the extent of the facility flowpath that must be included to adequately reproduce the facility test section flow environment. The findings indicate that the flow around the support structure for the fuel injection centerbody upstream of the facility nozzle generates disturbances that persist throughout the nozzle expansion process; corrupting the azimuthal symmetry that was desired in the fuel/air mixing region of the test section. Simulations without this support structure maintained a high degree of azimuthal symmetry up until the fuel injection plane. However, even in this scenario the azimuthal symmetry was not maintained once the centerbody boundary layer transitioned to a wake flow downstream of the fuel injection plane.

CFD↗

On Calculation Methods and Results for Straight Cylindrical Roller Bearing Deflection, Stiffness, and Stress

The purpose of this study was to assess some calculation methods for quantifying the relationships of bearing geometry, material properties, load, deflection, stiffness, and stress. The scope of the work was limited to two-dimensional modeling of straight cylindrical roller bearings. Preparations for studies of dynamic response of bearings with damaged surfaces motivated this work. Studies were selected to exercise and build confidence in the numerical tools. Three calculation methods were used in this work. Two of the methods were numerical solutions of the Hertz contact approach. The third method used was a combined finite element surface integral method. Example calculations were done for a single roller loaded between an inner and outer raceway for code verification. Next, a bearing with 13 rollers and all-steel construction was used as an example to do additional code verification, including an assessment of the leading order of accuracy of the finite element and surface integral method. Results from that study show that the method is at least first-order accurate. Those results also show that the contact grid refinement has a more significant influence on precision as compared to the finite element grid refinement. To explore the influence of material properties, the 13-roller bearing was modeled as made from Nitinol 60, a material with very different properties from steel and showing some potential for bearing applications. The codes were exercised to compare contact areas and stress levels for steel and Nitinol 60 bearings operating at equivalent power density. As a step toward modeling the dynamic response of bearings having surface damage, static analyses were completed to simulate a bearing with a spall or similar damage.

Krantz, Timothy L.↗

Enhancements supporting IC usage of PEM libraries on next-gen platforms

This milestone reports on the culmination of several years of effort by multiple PEM support software development teams to provide capabilities for use in LLNL-developed integrated codes on next-gen ASC platforms, including GPU support. We will provide a survey of relevant Application Program Interfaces (API) that are required to support LLNL IC code capability on relevant architectures, with a focus on Sierra and El Capitan. We will identify and summarize all dependencies between PEM supported libraries and IC supported physics codes. We will provide an assessment of algorithmic improvements that have been deployed, as well as future developments that are required to complete the GPU porting efforts. This assessment will include a description of programming models adopted by each of the PEM projects, distinct algorithmic challenges for each of the capabilities, and information about sharing GPU memory between the APIs and host codes. We will develop targeted test problems to assess computational performance. Finally, this milestone will result in identification of gaps in our effort to assist the LLNL ASC program in prioritization of effort for porting software to El Capitan.

97 MATHEMATICS AND COMPUTING↗

Development of the FHR advanced natural circulation analysis code and application to FHR safety analysis

The University of California, Berkeley (UCB) is performing thermal hydraulics safety analysis to develop the technical basis for design and licensing of fluoride-salt-cooled, high-temperature reactors (FHRs). FHR designs investigated by UCB use natural circulation for emergency, passive decay heat removal when normal decay heat removal systems fail. The FHR advanced natural circulation analysis (FANCY) code has been developed for assessment of passive decay heat removal capability and safety analysis of these innovative system designs. The FANCY code uses a one-dimensional, semi-implicit scheme to solve for pressure-linked mass, momentum and energy conservation equations. Graph theory is used to automatically generate a staggered mesh for complicated pipe network systems. Heat structure models have been implemented for three types of boundary conditions (Dirichlet, Neumann and Robin boundary conditions). Heat structures can be composed of several layers of different materials, and are used for simulation of heat structure temperature distribution and heat transfer rate. Control models are used to simulate sequences of events or trips of safety systems. A proportional-integral controller is also used to automatically make thermal hydraulic systems reach desired steady state conditions. A point kinetics model is used to model reactor kinetics behavior with temperature reactivity feedback. The underlying large sparse linear systems in these models are efficiently solved by using direct and iterative solvers provided by the SuperLU code on high performance machines. Input interfaces are designed to increase the flexibility of simulation for complicated thermal hydraulic systems. In conclusion, this paper mainly focuses on the methodology used to develop the FANCY code, and safety analysis of the Mark 1 pebble-bed FHR under development at UCB is performed.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Extended Plate and Beam Wall System: Concept Investigation and Initial Evaluation

A new and innovative High-R wall design, referred to as the Extended Plate & Beam (EP&B), is under development. The EP&B system uniquely integrates foam sheathing insulation with wall framing such that wood structural panels are installed exterior of the foam sheathing, enabling the use of standard practices for installation of drainage plane, windows and doors, claddings, cavity insulation, and the standard exterior foam sheathing installation approach prone to damage of the foam during transportation of prefabricated wall panels. As part of the ongoing work, the EP&B wall system concept has undergone structural verification testing and has been positively vetted by a group of industry stakeholders. Having passed these initial milestone markers, the advanced wall system design has been analyzed to assess cost implications relative to other advanced wall systems, undergone design assessment to develop construction details, and has been evaluated to develop representative prescriptive requirements for the building code. This report summarizes the assessment steps conducted to-date and provides details of the concept development.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗

Description of Transport Codes for Space Radiation Shielding

This slide presentation describes transport codes and their use for studying and designing space radiation shielding. When combined with risk projection models radiation transport codes serve as the main tool for study radiation and designing shielding. There are three criteria for assessing the accuracy of transport codes: (1) Ground-based studies with defined beams and material layouts, (2) Inter-comparison of transport code results for matched boundary conditions and (3) Comparisons to flight measurements. These three criteria have a very high degree with NASA's HZETRN/QMSFRG.

Kim, Myung-Hee Y.↗

Space Shuttle Debris Impact Tool Assessment Using the Modern Design of Experiments

Complex computer codes are used to estimate thermal and structural reentry loads on the Shuttle Orbiter induced by ice and foam debris impact during ascent. Such debris can create cavities in the Shuttle Thermal Protection System. The sizes and shapes of these cavities are approximated to accommodate a code limitation that requires simple "shoebox" geometries to describe the cavities -- rectangular areas and planar walls that are at constant angles with respect to vertical. These approximations induce uncertainty in the code results. The Modern Design of Experiments (MDOE) has recently been applied to develop a series of resource-minimal computational experiments designed to generate low-order polynomial graduating functions to approximate the more complex underlying codes. These polynomial functions were then used to propagate cavity geometry errors to estimate the uncertainty they induce in the reentry load calculations performed by the underlying code. This paper describes a methodological study focused on evaluating the application of MDOE to future operational codes in a rapid and low-cost way to assess the effects of cavity geometry uncertainty.

DeLoach, Richard↗

ASME Code Activities

U.S. Nuclear Regulatory Commission (NRC) is currently assessing ASME Section III, Division 5 (2017 Edition) and accompanying Code Cases for endorsement Draft Regulatory Guide (DG-1380) and draft NUREG (2245) on the endorsement of Division 5 were issued and public comment phase completed DG-1380 places conditions in the areas of general requirements, mechanical design, metallic materials allowable stresses, and graphite materials/design of Division 5 DG-1380 is a revision to Regulatory Guide 1.87, Revision 1, 1975 (which endorsed the 159X series of ASME Code Cases, with conditions) NRC staff is currently reviewing the low-temperature and elevated temperature Alloy 617 code cases for endorsement Results of the review will be incorporated into DG-1380 and changes will be announced publicly The formal endorsement of Section III, Division 5 and accompanying code cases, with conditions, will be made through Regulatory Guide 1.87, Revision 2

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Energy Reporting: Device Demonstration, Communication Protocols, and Codes and Standards

Energy reporting is the principle that all energy-using devices in buildings should be able to track their own energy use and report this to the local network. Energy reporting can provide building owners with easy access to highly granular energy use data. This report makes the case that energy reporting should become a free basic feature of all devices, and reports on a project intended to move us towards that goal. The project collected a set of demonstration devices with energy reporting features, including products that were modified by the project team or the manufacturer, or are already available for sale. To show these devices operating live at meetings and conferences, the team created a management system that queries the energy reporting devices for their data, stores the data, and displays it in compelling visualizations. The devices covered a wide range, including heating, ventilation, and air conditioning (thermostat and air purifier); lighting (individual bulb, task light, and auto-dimming overhead light); a vehicle charger; a water heater; electronics (notebook personal computer and universal serial bus charger); and three external meters (one integral with a dimming light switch). The demonstration uses a variety of communication protocols. The report reviews existing communication protocols that support energy reporting and describes how to use them with a proposed reference data model for energy reporting. It also assesses ways that energy codes and standards processes can be leveraged to drive energy reporting technology into the market. Energy reporting could ultimately save California on the order of 2.5 terawatt-hours per year and about $0.8 billion per year. Energy reporting is a highly practical technology with minimal (sometimes no) cost to consumers and manufacturers. This report discusses creation of the energy reporting devices themselves, analysis and recommendations for data models and protocols for energy reporting, and energy codes and standards implications of energy reporting technology. While energy reporting does not directly save energy, it provides information for better decision-making to save energy in changing equipment operation, maintenance, and replacement.

devices↗

Developing and Implementing the Data Mining Algorithms in RAVEN

The RAVEN code is becoming a comprehensive tool to perform probabilistic risk assessment, uncertainty quantification, and verification and validation. The RAVEN code is being developed to support many programs and to provide a set of methodologies and algorithms for advanced analysis. Scientific computer codes can generate enormous amounts of data. To post-process and analyze such data might, in some cases, take longer than the initial software runtime. Data mining algorithms/methods help in recognizing and understanding patterns in the data, and thus discover knowledge in databases. The methodologies used in the dynamic probabilistic risk assessment or in uncertainty and error quantification analysis couple system/physics codes with simulation controller codes, such as RAVEN. RAVEN introduces both deterministic and stochastic elements into the simulation while the system/physics code model the dynamics deterministically. A typical analysis is performed by sampling values of a set of parameter values. A major challenge in using dynamic probabilistic risk assessment or uncertainty and error quantification analysis for a complex system is to analyze the large number of scenarios generated. Data mining techniques are typically used to better organize and understand data, i.e. recognizing patterns in the data. This report focuses on development and implementation of Application Programming Interfaces (APIs) for different data mining algorithms, and the application of these algorithms to different databases.

97 MATHEMATICS AND COMPUTING↗

The 2014 Sandia Nonlinear Mechanics and Dynamics Summer Research Institute

A collaborative research institute was organized and held at Sandia Albuquerque for a period of six weeks. This research institute brought together researchers from around the world to work collaboratively on a set of research projects. These research projects included: developing experimental guidelines for studying variability and repeatability of nonlinear structures; decoupling aleatoric and epistemic uncertainty in measurements to improve dynamic predictions; a numerical round robin to assess the performance of five different numerical codes for modeling systems with strong nonlinearities; and an assessment of experimentally derived and numerically derived reduced order models. In addition to the technical collaborations, the institute also included a series of seminars given by both Sandians and external experts, as well as a series of tours and field trips to local places of scientific and engineering importance. This report details both the technical research and the programmatic organization of the 2014 Sandia Nonlinear Mechanics and Dynamics Summer Research Institute.

36 MATERIALS SCIENCE↗

Modeling and Analysis of FCM UN TRISO Fuel Using the PARFUME Code

The PARFUME (PARticle FUel ModEl) modeling code was used to assess the overall fuel performance of uranium nitride (UN) tri-structural isotropic (TRISO) ceramic fuel in the frame of the design and development of Fully Ceramic Matrix (FCM) fuel. A specific modeling of a TRISO particle with UN kernel was developed with PARFUME, and its behavior was assessed in irradiation conditions typical of a Light Water Reactor (LWR). The calculations were used to access the dimensional changes of the fuel particle layers and kernel, including the formation of an internal gap. The survivability of the UN TRISO particle was estimated depending on the strain behavior of the constituent materials at high fast fluence and burn-up. For nominal cases, internal gas pressure and representative thermal profiles across the kernel and layers were determined along with stress levels in the pyrolytic carbon (PyC) and silicon carbide (SiC) layers. These parameters were then used to evaluate fuel particle failure probabilities. Results of the study show that the survivability of UN TRISO fuel under LWR irradiation conditions might only be guaranteed if the kernel and PyC swelling rates are limited at high fast fluence and burn-up. These material properties are unknown at the irradiation levels expected to be reached by UN TRISO fuel in LWRs. Therefore, more effort is needed to determine them and positively conclude on the applicability of FCM fuel to LWRs.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

RELAP5-3D Developmental Assessment. Comparison of Version 4.3.4i on Linux and Windows

Figures have been generated comparing the parameters used in the developmental assessment of the RELAP5-3D code, version 4.3i, compiled on Linux and Windows platforms. The figures, which are the same as those used in Volume III of the RELAP5-3D code manual, compare calculations using the semi-implicit solution scheme with available experiment data. These figures provide a quick, visual indication of how the code predictions differ between the Linux and Windows versions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Improvements to a method for the geometrically nonlinear analysis of compressively loaded stiffened composite panels

The NLPAN computer code uses a finite-strip approach to the analysis of thin-walled prismatic composite structures such as stiffened panels. The code can model in-plane axial loading, transverse pressure loading, and constant through-the-thickness thermal loading, and can account for shape imperfections. The NLPAN code represents an attempt to extend the buckling analysis of the VIPASA computer code into the geometrically nonlinear regime. Buckling mode shapes generated using VIPASA are used in NLPAN as global functions for representing displacements in the nonlinear regime. While the NLPAN analysis is approximate in nature, it is computationally economical in comparison with finite-element analysis, and is thus suitable for use in preliminary design and design optimization. A comprehensive description of the theoretical approach of NLPAN is provided. A discussion of some operational considerations for the NLPAN code is included. NLPAN is applied to several test problems in order to demonstrate new program capabilities, and to assess the accuracy of the code in modeling various types of loading and response. User instructions for the NLPAN computer program are provided, including a detailed description of the input requirements and example input files for two stiffened-panel configurations.

Stoll, Frederick↗

The Phenix ultimate natural convection test

The French sodium cooled fast reactor Phenix was shut down in 2009 after 35 years of operation. Before decommissioning, a final set of tests were performed by the CEA during 9 months. Several topics were involved such as thermal hydraulics, core physics and fuel behaviour. Among these ultimate experiments, two thermal hydraulic tests were performed: an asymmetrical test consisting in a trip of one secondary pump and a natural convection test in the primary circuit. Recognizing the unique opportunity offered by these Phenix ultimate tests, IAEA decided in 2007 to launch a Coordinated Research Project (CRP) devoted to benchmarking analyses with system codes on the Phenix natural convection test. One objective of the natural convection test in Phenix reactor is the assessment of the CATHARE system code for safety studies on future and advanced sodium cooled fast reactors. The aim of this paper is to describe this test, which was performed on June 22-23, 2009, and the associated benchmark specifications for the CRP work. The paper reminds briefly the Phenix reactor with the main physical parameters and the instrumentation used during the natural convection test. After that, the test scenario is described: - initial state at a power of 120 MWth, - test beginning resulting from a manual dry out of the two steam generators, - manual scram, - manual trip on the three primary pumps without back-up by pony motors, - setting and development of natural convection in the primary circuit, in a first phase without significant heat sink in the secondary circuits and in a second phase with significant heat sink in the secondary circuits, by opening the casing of steam generators to create an efficient heat sink, by air natural circulation in the steam generators casing. The benchmark case ends after this second phase, which corresponds to the experimental test duration of nearly 7 hours. The paper presents also the benchmark specifications data supplied by the CEA to all participants of this CRP in order to perform calculations (core, primary circuit, primary pumps, IHX, shutdown system, operating parameters, test scenario and real test conditions). Finally, main test results and analyses are presented including the evolution of the core and of the heat exchangers inlet and outlet temperatures, and some local temperature measurements. The natural convection has been easily set up in the pool type reactor Phenix with different boundary conditions at the secondary side, with or without heat sink. The data obtained during this unique test represent some very useful and precious results for the development of SFR in a wide range of thematic such as numerical methods dedicated to thermal-hydraulics safety analyses (system codes, CFD codes and coupling system and CFD codes) and instrumentation. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Energy levels, radiative rates, and electron impact excitation rates for transitions in Li-like ions with 21≤Z≤28

We report calculations of energy levels, radiative rates, and electron impact excitation rates for transitions in Li-like ions with 21≤Z≤28. The General-Purpose Relativistic Atomic Structure Package is adopted for calculating energy levels and radiative rates, while for determining the collision strengths and subsequently the excitation rates, the Dirac Atomic R-matrix Code is used. Oscillator strengths, radiative rates, and line strengths are listed for all E1, E2, M1, and M2 transitions among the lowest 24 levels of the Li-like ions considered. Collision strengths have been averaged over a Maxwellian velocity distribution, and the effective collision strengths obtained are given over a wide temperature range up to 10{sup 7.8} K. Additionally, lifetimes are listed for all calculated levels of these ions. Finally, extensive comparisons are made with results available in the literature, as well as with our analogous calculations for all parameters with the Flexible Atomic Code, in order to assess the accuracy of the results.

74 ATOMIC AND MOLECULAR PHYSICS↗