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DOE OSTI2011

Going Beyond Code: A Guide for Creating Effective Green Building Programs for Energy Efficient and Sustainable Communities

This guide is designed to help state and local governments design and implement successful "beyond code" programs for new commercial and residential buildings. The goal is to help states and localities establish voluntary or mandatory programs that go well beyond traditional minimum code requirements for new buildings. The guide addresses keys to successful adoption and implementation and discusses the primary areas that are typically included in beyond code or green building programs, including energy-efficiency materials and resource conservation, water efficiency, indoor environmental quality, and site development and land use. Detailed descriptions and analysis of actual programs are discussed, including lessons learned and best practices. States and localities can use the information on local programs, model codes and standards, and the model energy-efficiency criteria for residential and commercial buildings to find the best approach for their jurisdiction to develop and implement an effective beyond code program.

Code adoption and implementation, sustainability, ↗

DOE OSTI2012

Computer code for space-time diagnostics of nuclear safety parameters

The computer code ECRAN 3D (Experimental and Calculation Reactor Analysis) is designed for continuous monitoring and diagnostics of reactor cores and databases for RBMK-1000 on the basis of analytical methods for the interrelation parameters of nuclear safety. The code algorithms are based on the analysis of deviations between the physically obtained figures and the results of neutron-physical and thermal-hydraulic calculations. Discrepancies between the measured and calculated signals are equivalent to obtaining inadequacy between performance of the physical device and its simulator. The diagnostics system can solve the following problems: identification of facts and time for inconsistent results, localization of failures, identification and quantification of the causes for inconsistencies. These problems can be effectively solved only when the computer code is working in a real-time mode. This leads to increasing requirements for a higher code performance. As false operations can lead to significant economic losses, the diagnostics system must be based on the certified software tools. POLARIS, version 4.2.1 is used for the neutron-physical calculation in the computer code ECRAN 3D. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2012

Dynamical analysis of innovative core designs facing unprotected transients with the MAT5 DYN code

Since 2007, advanced Sodium-cooled Fast Reactors (SFR) are investigated by CEA, AREVA and EDF in the framework of a joint French collaboration. A prototype called ASTRID, sets out to demonstrate progress made in SFR technology, is due to operate in the years 2020's. The modeling of unprotected transients by computer codes is one of the key safety issues in the design approach to such SFR systems. For that purpose, the activity on CATHARE, which is the reference code for the transient analysis of ASTRID, has been strengthened during last years by CEA. In the meantime, EDF has developed a simplified and multi-channel code, named MAT5 DYN, to analyze and validate innovative core designs facing protected and unprotected transients. First, the paper consists in a description of MAT5 DYN: a code based on the existing code MAT4 DYN including major improvements on geometry description and physical modeling. Second, two core designs based on the CFV core design developed at CEA are presented. Then, the dynamic response of those heterogeneous cores is analyzed during unprotected loss of flow (ULOF) transient and unprotected transient of power (UTOP). The results highlight the importance of the low void core effect specific to the CFV design. Such an effect, when combined with a sufficient primary pump halving time and an optimized cooling group scheme, allows to delay (or, possibly, avoid) the sodium boiling onset during ULOF accidents. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2012

Analysis of Phenix end-of-life natural convection test with the MARS-LMR code

The end-of-life test of Phenix reactor performed by the CEA provided an opportunity to have reliable and valuable test data for the validation and verification of a SFR system analysis code. KAERI joined this international program for the analysis of Phenix end-of-life natural circulation test coordinated by the IAEA from 2008. The main objectives of this study were to evaluate the capability of existing SFR system analysis code MARS-LMR and to identify any limitation of the code. The analysis was performed in three stages: pre-test analysis, blind posttest analysis, and final post-test analysis. In the pre-test analysis, the design conditions provided by the CEA were used to obtain a prediction of the test. The blind post-test analysis was based on the test conditions measured during the tests but the test results were not provided from the CEA. The final post-test analysis was performed to predict the test results as accurate as possible by improving the previous modeling of the test. Based on the pre-test analysis and blind test analysis, the modeling for heat structures in the hot pool and cold pool, steel structures in the core, heat loss from roof and vessel, and the flow path at core outlet were reinforced in the final analysis. The results of the final post-test analysis could be characterized into three different phases. In the early phase, the MARS-LMR simulated the heat-up process correctly due to the enhanced heat structure modeling. In the mid phase before the opening of SG casing, the code reproduced the decrease of core outlet temperature successfully. Finally, in the later phase the increase of heat removal by the opening of the SG opening was well predicted with the MARS-LMR code. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2012

Full 3D visualization tool-kit for Monte Carlo and deterministic transport codes

We propose a package of tools capable of translating the geometric inputs and outputs of many Monte Carlo and deterministic radiation transport codes into open source file formats. These tools are aimed at bridging the gap between trusted, widely-used radiation analysis codes and very powerful, more recent and commonly used visualization software, thus supporting the design process and helping with shielding optimization. Three main lines of development were followed: mesh-based analysis of Monte Carlo codes, mesh-based analysis of deterministic codes and Monte Carlo surface meshing. The developed kit is considered a powerful and cost-effective tool in the computer-aided design for radiation transport code users of the nuclear world, and in particular in the fields of core design and radiation analysis. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2012

Calculation of the Phenix end-of-life test in natural circulation with the CATHARE code

The Inst. of Radiological Protection and Nuclear Safety (IRSN) acts as technical support to French public authorities. As such, IRSN is in charge of safety assessment of operating and under construction reactors, as well as future projects. In this framework, one current objective of IRSN is to evaluate the ability and accuracy of numerical tools to foresee consequences of accidents. One of the advantages pointed up for fast reactors cooled by heavy liquid metal is the possibility of decay heat removal based on natural convection. The promotion of this passive cooling mode in future safety demonstrations will involve the use of adapted and validated numerical codes. After the final shutdown of the Phenix sodium cooled fast reactor in 2009, a set of tests covering different areas was conducted for code validation, including a natural circulation test in the primary circuit. Experimental data were issued by CEA to organize a benchmark exercise in the frame of an IAEA Coordinated Research Project (CRP), with the objective to assess the system-codes capability in simulating the thermal-hydraulics behavior of sodium cooled fast reactors in such accidental conditions. IRSN participated to this benchmark with the CATHARE code. This code, co-developed by CEA, EDF, AREVA and IRSN and widely used for PWR safety studies, was recently extended for sodium applications. This paper presents the CATHARE modeling of the Phenix primary circuit and the results obtained. A relatively good agreement was found with available measurements considering the experimental uncertainties. This work stressed the local aspects of phenomena occurring during the natural convection establishment and the limits of a 0D/1D approach. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2012

SPACE code simulation of cold leg small break LOCA in the ATLAS integral test

SPACE code is a system analysis code for pressurized water reactors. This code uses a two-fluid and three-field model. For a few years, intensive validations have been performed to secure the prediction accuracy of models and correlations for two-phase flow and heat transfer. Recently, the code version 1.0 was released. This study is to see how well SPACE code predicts thermal hydraulic phenomena of an integral effect test. The target experiment is a cold leg small break LOCA in the ATLAS facility, which has the same two-loop features as APR1400. Predicted parameters were compared with experimental observations. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2012

CATHARE thermal-hydraulic system code for HLM preliminary validation in natural convection tests

The innovative nuclear systems cooled by Heavy Liquid Metal (HLM) are the subject of an ongoing interest both in Europe and outside, evidenced by a number of projects in progress. In the frame of the European Framework Programmes have been evidenced the need to adopt a thermalhydraulic system code capable to treat lead and Lead-Bismuth Eutectic (LBE) systems, with a particular interest in developing a 'European' code. Considering this scenario, within a specific collaboration between ENEA and CEA, the CATHARE French system code has been modified extending its capabilities to simulate HLM systems. In the present paper, the state of the validation process of CATHARE-HLM is discussed. The activity aims to assess the capabilities and limitations of the code to simulate the behavior of integral facilities, in particular in natural circulation conditions. The experimental data come from NACIE LBE-cooled facility sited at the ENEA Brasimone laboratories. The results obtained show a good capability in reproducing the systems behavior, despite some uncertainties on the experimental measurements. Future improvements on the code are going to be planned within the collaboration ENEA/CEA. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2012

MARS-KS code validation activity through the atlas domestic standard problem

The 2 nd Domestic Standard Problem (DSP-02) exercise using the ATLAS integral effect test data was executed to transfer the integral effect test data to domestic nuclear industries and to contribute to improving the safety analysis methodology for PWRs. A small break loss of coolant accident of a 6-inch break at the cold leg was determined as a target scenario by considering its technical importance and by incorporating interests from participants. Ten calculation results using MARS-KS code were collected, major prediction results were described qualitatively and code prediction accuracy was assessed quantitatively using the FFTBM. In addition, special code assessment activities were carried out to find out the area where the model improvement is required in the MARS-KS code. The lessons from this DSP-02 and recommendations to code developers are described in this paper. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2013

Development of GPU-based Monte Carlo code for fast CT imaging dose calculation on CUDA Fermi architecture

This paper describes the development of a Graphics Processing Unit (GPU) accelerated Monte Carlo photon transport code, ARCHER{sub GPU}, to perform CT imaging dose calculations with good accuracy and performance. The code simulates interactions of photons with heterogeneous materials. It contains a detailed CT scanner model and a family of patient phantoms. Several techniques are used to optimize the code for the GPU architecture. In the accuracy and performance test, a 142 kg adult male phantom was selected, and the CT scan protocol involved a whole-body axial scan, 20-mm x-ray beam collimation, 120 kVp and a pitch of 1. A total of 9 x 108 photons were simulated and the absorbed doses to 28 radiosensitive organs/tissues were calculated. The average percentage difference of the results obtained by the general-purpose production code MCNPX and ARCHER{sub GPU} was found to be less than 0.38%, indicating an excellent agreement. The total computation time was found to be 8,689, 139 and 56 minutes for MCNPX, ARCHER{sub CPU} (6-core) and ARCHER{sub GPU}, respectively, indicating a decent speedup. Under a recent grant funding from the NIH, the project aims at developing a Monte Carlo code with the capability of sub-minute CT organ dose calculations. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗

DOE OSTI2016

JASMINE: a fuel rod thermal-mechanical performance's code

JASMINE, which is developed by China General Nuclear Power Corporation (CGNPC), is a code for calculating thermal-mechanical response of PWR fuel rods. The code models many physical phenomena, including heat conduction through the fuel and cladding to the coolant, cladding elastic and plastic deformation, fuel elastic deformation, fuel swelling and densification,fuel-cladding mechanical interaction, fission gas release from the fuel, rod internal pressure, cladding oxidation and hydriding. Just as most fuel rod codes, JASMINE, which is a quasi-two dimensional code, includes a thermal module, a mechanical module, a fission gas release (FGR) module, and an internal pressure module. For thermal module, the time dependent thermal conduction equation of axisymmetric cylindrical geometry, Halden's fuel thermal conductivity, thermal conductivity integration method and multi-dimension Newton-Raphson procedure are used; for the mechanical module, generalized plane strain hypothesis, cladding-fuel axial lockup model, cladding thermal stress and transfer matrix method are applied; for fission gas release module, the time dependent diffusion equation for a spherical grain, Speight-Turnbull grain surface boundary condition, and Forsberg-Massih's method are used; for the internal pressure module, the ideal gas state equation is used. The code, which is on validation stage, will be used for fuel rod design of nuclear power plants in China in a near future. So far, the preliminary validation process for pellet centerline temperature and fission gas release, which is based on fuel rod test data in FUMEX-III project, has finished. According to the validation results, JASMINE has the ability to reasonably predicts fuel behavior and performance in-pile. JASMINE has good prediction in fuel temperature, and good prediction on FGR as a whole. However, for FGR in the range 1% to 15%, JASMINE over-predicts FGR; for other range, JASMINE underestimates FGR. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

Group Structure Optimization Using the PyGroup Code

Group constants are used as coefficients in the energy domain discretized Linear Boltzmann Equation (LBE), in which the Continuous-Energy (CE) fluxes are also integrated over energy bins and transformed into group fluxes. This procedure, based on the Multi-Group (MG) theory, is widely applied in radiation transport and diffusion simulations Careful preparation of MG cross section data, or group constants, is thus a prerequisite for success in acquiring high-fidelity LBE solutions. Generation of MG cross section data itself is a complicated multi-step task, since group constants are indispensably coupled with the solution of the LBE, i.e. the flux spectrum in the problem to be solved. Theoretically, a MG cross section library is only valid for the problem that it is built upon. Nevertheless, in practice MG cross section libraries are prepared for similar type of problems, such as thermal reactors, fast reactors, etc., to facilitate the neutronic simulations. Generally, CE point-wise cross section data set, generated by both theoretical models and experimental data, is first converted to Fine-Group (FG) libraries using a pre-defined or precalculated flux spectra, then, Broad-Group (BG) libraries are collapsed from the fine-group libraries using problem-dependent flux spectra. Proper treatments of energy and spatial self-shielding and resonance cross section are required in the CE-to-MG converting process Previously, we have developed a computer code, YGROUP, for fine-to-broad group cross section collapsing. To maintain high accuracy of the resulting BG libraries, YGROUP provides users several options on collapsing strategies and weighting schemes for group binning and evaluation of BG cross section data in users' specific problems.. An approach based on Particle Swarm Optimization was applied to automatically determine the optimal group structure. We also developed a hybrid weighting scheme to further improve BG library performance for Special Nuclear Materials (SNM) detection applications. In this work, a new computer code, PyGroup, has been developed to implement the PSO approach with an improved simulation-driven fitness function design that includes both response and forward/adjoint flux errors. Together with YGROUP, and deterministic transport code TITAN/PENMSHXP, the PyGroup code package is tested on a Dual Range Coincident Counter (DRCC) model. In this paper, we demonstrated the capability of the PyGroup code for group structure optimization using the DRCC detector model. In the future, an iteration scheme can be implemented to determine the number of broad groups based on a user-defined tolerance. we will also apply PyGroup to more complicated reactor models, and investigate physics-based fitness function designs to further reduce computational cost. (authors)

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2017

Examples of Use of SINBAD Database for Nuclear Data and Code Validation

Benchmark experiments can serve several purposes and the availability of suitable experimental data is fundamental to ensure that modelling tools meet the requirements of nuclear industry and licensing needs. Essentially, they are used for: - validation of nuclear cross-section data, - validation of nuclear codes and models used, - qualification and training of users through benchmarking and code comparison exercises. The Shielding Integral Benchmark Archive and Database (SINBAD) project was started in the early 1990's at the OECD Nuclear Energy Agency Data Bank (NEA DB) and the Radiation Safety Information Computational Center (RSICC) as a collection of quality experiments for validating and benchmarking computer codes and data for radiation transport and shielding applications. Currently, the SINBAD database comprises over 100 benchmark compilations and evaluations of relevance to: - reactor shielding, pressure vessel dosimetry (48); - fusion blanket neutronics (31); - accelerator shielding (23). In order to facilitate their use, the data are organized in the standardized format structure. The index and the master document (called abstract) use the Hyper Text Markup Language (HTML) format with links to text files, figures, Portable Document Format (pdf) documents and computer code inputs. The format allows easy access and forward/backward navigation among the data and has proved to be suitable, general enough and above all easy (and low-cost) to maintain. In addition to the characterization of the radiation source, description of the shielding set-up, the instrumentation and the relevant detector measurements, most sets in SINBAD contain also the deterministic or Monte Carlo radiation transport computer model used for the interpretation of the experiment and, where available, results from uncertainty analysis. The set of primary documents used for the benchmark compilation and evaluation are provided in computer readable form. The database is intended for different types of users, including nuclear data evaluators, computer code developers, experiment designers and university students. SINBAD is available from the NEA Data Bank and RSICC (see https://www.oecdnea.org/science/wprs/shielding/sinbad/). Up to now several hundreds of SINBAD packages were distributed from the NEA DB and RSICC. Since 2011 the maintenance and further development of the database is coordinated in the scope of the OECD NEA Nuclear Science Committee (NSC) Working Party on Scientific Issues of Reactor Systems (WPRS) Expert Group on Radiation Transport and Shielding (EGRTS). A key objective of the group is to identify, evaluate and preserve experimental data on shielding benchmarks. The Expert Group maintains close links with the International Reactor Physics Experiment Evaluation (IRPhE) and International Criticality Safety Benchmark Evaluation (ICSBEP) projects, as well as in co-ordination with the JEFF and EFF NEA Data Bank projects and with the SATIF community of accelerator shielding specialists. In the last years, much effort was devoted to the quality review of the available benchmark information.

42 ENGINEERING↗

DOE OSTI2016

Potential Energy Cost Savings from Increased Commercial Energy Code Compliance

An important question for commercial energy code compliance is: “How much energy cost savings can better compliance achieve?” This question is in sharp contrast to prior efforts that used a checklist of code requirements, each of which was graded pass or fail. Percent compliance for any given building was simply the percent of individual requirements that passed. A field investigation method is being developed that goes beyond the binary approach to determine how much energy cost savings is not realized. Prototype building simulations were used to estimate the energy cost impact of varying levels of non-compliance for newly constructed office buildings in climate zone 4C. Field data collected from actual buildings on specific conditions relative to code requirements was then applied to the simulation results to find the potential lost energy savings for a single building or for a sample of buildings. This new methodology was tested on nine office buildings in climate zone 4C. The amount of additional energy cost savings they could have achieved had they complied fully with the 2012 International Energy Conservation Code is determined. This paper will present the results of the test, lessons learned, describe follow-on research that is needed to verify that the methodology is both accurate and practical, and discuss the benefits that might accrue if the method were widely adopted.

code compliance↗

DOE OSTI2016

Potential Energy Cost Savings from Increased Commercial Energy Code Compliance

code compliance, commercial energy code, energy co↗

DOE OSTI2024

Single-Shot Decoding of Good Quantum LDPC Codes

Abstract Quantum Tanner codes constitute a family of quantum low-density parity-check codes with good parameters, i.e., constant encoding rate and relative distance. In this article, we prove that quantum Tanner codes also facilitate single-shot quantum error correction (QEC) of adversarial noise, where one measurement round (consisting of constant-weight parity checks) suffices to perform reliable QEC even in the presence of measurement errors. We establish this result for both the sequential and parallel decoding algorithms introduced by Leverrier and Zémor. Furthermore, we show that in order to suppress errors over multiple repeated rounds of QEC, it suffices to run the parallel decoding algorithm for constant time in each round. Combined with good code parameters, the resulting constant-time overhead of QEC and robustness to (possibly time-correlated) adversarial noise make quantum Tanner codes alluring from the perspective of quantum fault-tolerant protocols.

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2024

PDMATLAB2D: A Peridynamics MATLAB Two-dimensional Code

PDMATLAB2D is a meshfree peridynamics implementation in MATLAB suitable for simulation of two-dimensional fracture problems. The purpose of this code is twofold. First, it provides an entry-level peridynamics computational tool for educational and training purposes. Second, it serves as an accessible and easily modifiable computational tool for peridynamics researchers who would like to adapt the code for a multitude of peridynamics simulation scenarios. The current version of the code implements a bond-based brittle elastic peridynamic model and a critical stretch criterion for bond breaking. However, the code is designed to be extendable for other peridynamic models and computational features. In this paper, we provide an overview of the code structure and functions with illustrative examples. In conclusion, due to the integrated computation and postprocessing MATLAB capabilities, PDMATLAB2D can serve as an effective testbed for testing new constitutive models and advanced numerical features for peridynamics computations.

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2024

The g-code file

The slicing process for additive manufacturing (AM) involves many steps, including slicing the object to be printed and fitting toolpaths to each resulting layer. The toolpaths define the trajectories that the machine must follow during printing to construct the object. Once the slicing is complete and all toolpaths have been found and properly planned, this data must be transferred to the 3D printer in a way the printer can use to construct the object. This is achieved by exporting a file, usually a g-code file, from the slicing software. The g-code file is a text-based file that contains instructions for the machine, controlling the system and defining the motions for each toolpath. The file, which often contains thousands of lines of code, is mostly composed of numbers that command positions for axis motions. This chapter will outline the contents of a g-code file, discuss various commands and their implementation, then provide example codes and files for object construction.

Roschli, Alex↗