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DOE OSTI2023

Simulation and Optimization of Volatile Fatty Acid Upgrading Strategies for Sustainable Transportation Fuel Production

The multicomponent nature of and variability of biomass makes chemical upgrading via a single process stream to a single end use infeasible for most feedstocks. A more promising approach is to identify upgrading strategies (encompassing bioprocessing, catalysis, and separations) which valorize varied biomass fractions to distinct products to which each is best suited. A robust procedure to carry out this goal should incorporate comparison of various upgrading procedures as well as suitability of products to a slate of identified end uses. We applied this strategy to one biomass-derived feedstock, volatile fatty acids (VFAs) derived from wet waste arrested anaerobic digestion, by developing a computer program, VFA Upgrading to Liquid Transportation fUels Refinery Estimation (VULTURE) which evaluates VFA catalytic upgrading to liquid transportation fuels. VULTURE considers multiple separations, catalysis (ketonization, hydrogenation), and fuel application options, generating hundreds of candidate scenarios for a given VFA stream, then selects several promising strategies that optimize bio-content of products with properties best suited for target fuel types. We find that VFAs are upgraded most efficiently when separate light alcohol (C3-6) and heavy hydrocarbon or alcohol (C7-13) fractions are targeted to create gasoline and heavy-duty (diesel or jet) fuels or fuel blends. Surrogate property testing of VULTURE-recommended fuels reveals that most predictive models employed are robust, while rigorous process simulation shows that the simple unit operation assumptions used in VULTURE are largely accurate, especially for heavy-duty fuel synthesis. Techno-economic and life-cycle analyses show that VFA-derived fuels are profitable and have dramatically (=57%) lower carbon intensities than fossil analogs.

BIOMASS FUELS↗

DOE OSTI2020

On the Formalization of Development and Assessment Process for Digital Twins in the Nearly Autonomous Management and Control System

In recent years, the autonomous control system has been encouraged in advanced reactors for restoring economic viability, simplifying the operation and maintenance, and enabling remote-site power generations [1]. Since the reactor is expected to be operated for a long period of time with a limited number of individuals onsite, it is recommended that the autonomous control system should have access to very realistic models of the state of processes in the whole lifecycle, together with these process behaviors in interaction with their environment in the real world. As a result, digital twin (DT) technology is suggested in autonomous control systems. DT is defined as a digital representation of a physical object or system, which contains a record for the histories of loads, operation and maintenance status, predictions for the near-term transient of important state variables, and decision-making process [2]. Since machine learning (ML) can recognize patterns within a complex system in real-time applications, it has been used to build DTs in the autonomous control systems for advanced reactors. Meanwhile, due to the rareness of operation data in accident scenarios, the development and assessment of DTs is expected to be mainly driven by simulations. Although the capability and feasibility of ML-based DTs are recognized in improving the safety and efficiency of reactor control, a major concern from the regulatory commission and the nuclear industry is whether the information from a DT is developed and assessed in accordance with expectation and requirements by the target decision. Such concerns not only affect the acceptance criteria for DTs, but also values that can be extracted from DTs and autonomous control system during operations. Inspired by the success of formal methods in improving the reliability and robustness of computer programming and software development, it is suggested that the development and assessment process (DAP) for both separate DTs and integral control system should be formalized in a transparent, consistent, and improvable manner. In this study, a digital-twin development and assessment process (DT-DAP) is proposed by adapting the evaluation model development and assessment process (EMDAP) [3] to requirements by the autonomous control system, ML algorithms, and DT technology. To demonstrate the framework, a baseline nearly autonomous management and control (NAMAC) system with ML-based DTs for diagnosis and prognosis is developed and assessed based on the framework. It is found that with selected testing methods and techniques, the DT-DAP can help identify errors in DTs and NAMAC which would otherwise be left unverified. Meanwhile, it is found that the DT-DAP can improve the DTs and NAMAC by continuously learning and iterating through different elements.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2022

Method and apparatus for real time, in situ sensing and characterization of roughness, geometrical shapes, geometrical structures, composition, defects, and temperature in three-dimensional manufacturing systems

Methods and apparatuses for manufacturing are disclosed, including (a) providing an apparatus having: a laser; scanner; powder injection system; powder spreading system; dichroic filter; imager-and-processor; and computer; (b) programming the computer with specifications of a sample; (c) using the computer to set initial parameters based on the sample specifications; (d) adjusting a stage to position the sample; (e) focusing and scanning electromagnetic radiation onto the sample while powder is concurrently injected onto the sample in order to deposit a layer; (f) capturing two-dimensional images of the sample and probing the sample to determine whether the deposited layer was manufactured per the specifications; (g) use the computer to adjust the three-dimensional manufacturing parameters based on the determination made in step (f) prior to additively manufacturing a subsequent layer or making repairs; and (h) repeating steps (d), (e), (f), and (g) until the manufacture is complete. Other embodiments are described and claimed.

Liu, Jian↗

DOE OSTI2023

Apparatus and method for shaped waveform interrogation

Described are an apparatus, computer program product, and associated methods for shaped waveform acoustic interrogation of substances and materials to determine one or more properties of the materials or substances. In some embodiments, a shaped waveform is formed by summing two or more different waveforms and an acoustic wave is generated according to the shaped waveform. The acoustic wave is transmitted by one or more transmitting transducers through the substance or material and received by one or more receiving transducers. The shaped waveform acoustic wave can have a duration or a period that is less than about 20 μs and can comprise predetermined frequency content. Characteristics of the shaped waveform acoustic wave, as received at the receiving transducer(s), including characteristics such as amplitude, frequency, time of flight, etc., can be associated with said one or more properties of the substance or material to provide for real-time monitoring of these properties.

Sinha, Dipen N.↗

DOE OSTI2015

Response Time Analysis and Test of Protection System Instrument Channels for APR1400 and OPR1000

Safety limits are required to maintain the integrity of physical barriers designed to prevent the uncontrolled release of radioactive materials in nuclear power plants. The safety analysis establishes two critical constraints that include an analytical limit in terms of a measured or calculated variable, and a specific time after the analytical limit is reached to begin protective action. Keeping with the nuclear regulations and industry standards, satisfying these two requirements will ensure that the safety limit will not be exceeded during the design basis event, either an anticipated operational occurrence or a postulated accident. Various studies on the setpoint determination methodology for the safety-related instrumentation have been actively performed to ensure that the requirement of the analytical limit is satisfied. In particular, the protection setpoint methodology for the advanced power reactor 1400 (APP1400) and the optimized power reactor 1000 (OPR1000) has been recently developed to cover both the design basis event and the beyond design basis event. The developed setpoint methodology has also been quantitatively validated using specific computer programs and setpoint calculations. However, the safety of nuclear power plants cannot be fully guaranteed by satisfying the requirement of the analytical limit. In spite of the response time verification requirements of nuclear regulations and industry standards, it is hard to find the studies on the systematically integrated methodology regarding the response time evaluation. In cases of APR1400 and OPR1000, the response time analysis for the plant protection system is partially included in the setpoint calculation and the response time test is separately performed via the specific plant procedure. The test technique has a drawback which is the difficulty to demonstrate completeness of timing test. The analysis technique has also a demerit of resulting in extreme times that not actually possible. Thus, the establishment of the systematic response time evaluation methodology is needed to justify the conformance to the response time requirement used in the safety analysis. This paper proposes the response time evaluation methodology for APR1400 and OPR1000 using the combined analysis and test technique to confirm that the plant protection system can meet the analytical response time assumed in the safety analysis. In addition, the results of the quantitative evaluation performed for APR1400 and OPR1000 are presented in this paper. The proposed response time analysis technique consists of defining the response time requirement, determining the critical signal path for the trip parameter, allocating individual response time to each component on the signal path, and analyzing the total response time for the trip parameter, and demonstrates that the total analyzed response time does not exceed the response time requirement. The proposed response time test technique is composed of defining the response time requirement, determining the critical signal path for the trip parameter, determining the test method for each component on the signal path, performing the response time test, and demonstrates that the total test result does not exceed the response time requirement. The total response time should be tested in a single test that covers from the sensor to the final actuation device on the instrument channel. When the total channel is not tested in a single test, separate tests on groups of components or single components including the total instrument channel shall be combined to verify the total channel response. For APR1400 and OPR1000, the ramp test technique is used for the pressure and differential pressure transmitters and the step function testing technique is applied to the signal processing equipment and final actuation device. As a result, it can be demonstrated that the response time requirement is satisfied by the combined analysis and test technique. Therefore, the proposed methodology in this paper plays a crucial role in guaranteeing the safety of the nuclear power plants systematically satisfying one of two critical requirements from the safety analysis. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2017

Computational Safeguards Analysis of PWR Spent Nuclear Fuel in Dry Storage - 17435

The master plan for spent nuclear fuel (SNF) management in Korea recently adopted by Korean atomic energy promotion council. According to the master plan, the detailed practical and R and D strategies for SNF transferring, storage and disposal should be determined in the near future. Besides, it is also necessary to make plan how to secure SNF at each step in terms of nuclear safeguards, safety, security and nonproliferation. Several nondestructive analysis (NDA) techniques have been studied for assuring the safety of SNF in PWR methods. Obviously, gamma-ray is most dominant radiation emitted by SNF so several gamma-ray measurement methods have been developed and applied for the purpose of SNF monitoring. Those methods, however, suffer from a lack of nuclear data and interfering background radiation. To date, several neutron measurement methods have been developed as a solution to the problem of gamma-ray methods. Korea Institute of Nuclear Nonproliferation and Control (KINAC) currently develops a new novel technique of fast neutron detection for spent fuel monitoring at dry spent fuel storage facilities. As the first step, a computational program has been developed in order to evaluate neutron flux on the surface of PWR dry storage cask and study for some potential sabotage events. Typically, neutron flux on the surface of dry storage cask can be evaluated via two steps. The source term, based on burn-up rate of SNF, its cooling time and others related to operation condition at nuclear power plants, can be firstly defined by ORIGEN code. The surface flux can be then evaluated via MCNP code. These two separated steps are complicated and time consuming so KINAC developed a new algorithm and user-interface, coupled ORIGEN and MCNP. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2015

Tracking radioactivity content in the spent fuel pool in PWRS - 15150

Over the last decade, significant progress has been made in managing radioisotopes generated in the PWR primary systems. These include limiting the use of cobalt containing materials, along with improved fuel performance and zinc injection to passivate the corrosion layers to limit the accumulation of Co-58 and Co-60 on primary coolant surfaces. Attention here is directed to the spent fuel pool where accumulation of spent fuel and activated material represents a growing source of mobile radioactivity supplemented with periodic exchanges of water from the reactor cooling system. Characterization of this source is increasingly important to effective management of disposal options. This paper examines the radioactive resin source term as determined from routine sampling of the coolant and fuel pool liquids. Targeted resin stream sampling and radiochemical analyses conducted at Diablo Canyon serve to inform and provide context for this examination. This examination is supplemented with computer modelling using the DW James Consulting computer program, 3R-SCAN. At most operating plants daily reactor coolant samples are counted using gamma spectroscopy and recorded. These samples serve to track fuel performance and maintain water chemistry. A secondary use for this data is to maintain a day to day tracking of releases of radioisotopes important to waste disposal. Given the coolant concentrations, plant operating history, and corroborative mathematical models and sampling, accumulation of these radioisotopes in coolant cleanup system ion-exchangers can be predicted. Fuel pool sampling is less frequent, but at the same time, things change more slowly there. The fuel pool contains about ten times the volume of the reactor coolant but its cleanup system process rate is comparable to reactor coolant letdown, i.e. 5.047 - 8.833 liters per second [lps] (80-140 gallons per minute [gpm]). A spent fuel pool clean-up demineralizer flow rate is typically about 5.047 lps (80 gpm). Concentrations of longer lived radioisotopes in the spent fuel pool can be comparable to those in reactor coolant. Releases of corrosion products and fission products from the stored fuel occur at a low rate but given the overall inventory can be significant. Recognized scaling factor ratios, particularly those to Co-60, are skewed by the aging of the material in the pool. An additional consideration stems from refueling operations themselves. Prior to refueling, it is common practice to conduct a forced oxidation of the corrosion films within the reactor. This spikes the activity within the reactor coolant. Most of this activity is removed by the reactor coolant purification system prior to removing the reactor head. This paper examines the accumulation of activity in the fuel pool purification ion exchange resins and how it impacts the overall classification of those ion exchange resins. The following discussion focuses on the transport of nickel and cobalt isotopes particularly as it relates to determination of Ni-63 which has become increasingly prominent in disposal classification. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2016

Nuclide Release Pathways in and Around a Trench Repository for Radioactive Waste - 16014

An effective model for a safety assessment of a conceptual trench repository system, in which low-and intermediate level radioactive waste (LILW) is disposed of has been developed. The repository system is assumed to be located in the surface area of a mountain hill near the seashore. An easy-to-use computer program based on this model, developed as a GoldSim template, has been extended based on the previous one in consideration of the rather possible complex pathways in and around the trench for its performance assessment, which is able to evaluate a nuclide release from the repository and farther transport into the geosphere and biosphere under various normal, disruptive natural, and manmade events, and scenarios that can occur after a failure of waste drums with associated uncertainty. To demonstrate the nuclide release and transport behavior through these various possible pathways in the near- and far-fields of the repository system under certain scenarios, some illustrative evaluations and a simple sensitivity investigation with C-14 for the concrete barriers of the trench were made. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2014

Determination of attenuation coefficients of heavyweight concretes containing colemanite by using {sup 133}Ba radioactive isotope source - Paper 140

As it is well known, colemanite is a boron ore. Boron is one of the most important underground richness of Turkey. Boron is employed as a constituent material of neutron shielding because of its high absorption of thermal neutrons. Barite ore includes BaSO{sub 4}, which is a well photon radiation absorbent. The literature about radiation attenuation in heavy concretes (including boron and barite ores) is fairly abundant. Abdo and Abdo et al. theoretically calculated and determined cross section by using XCOM computer program for photon and neutron radiations shielding. Akkurt et al. measured radiation transmission of heavy concretes including normal and barite aggregates for different γ-ray energies and calculated linear attenuation coefficients. Bashter et al. studied heavy concretes including hematite-serpentine, ilmenite-limonite as control absorber in nuclear reactors γ-rays and neutron particles shielding. Bashter and Bashter et al. studied heavy concretes including hematite-serpentine, ilmenite-limonite, basalt-magnetite, ilmenite, basalt, steel and magnetite for only photon radiation shielding and calculated linear and mass attenuation coefficients from 10 keV to 1 GeV. Dealmeid et al. studied radiation transmission and angular distribution for 25 MV X-rays by using LINAC. Demir and Keles measured radiation transmission of concrete including boron waste for 59.54 and 80.99 keV γ-rays. Kaplan reported about many heavy concretes studies in his book. Kitis et al. studied heavy concretes including barite aggregates for cyclotron shielding. Mollah et al. determined concretes including ilmenite and magnetite aggregates for neutron shielding by using {sup 252}Cf radioactive source and BF{sub 3} neutron detector. Yarar and Bayuelken studied concretes including colemanite aggregates for neutron shielding. In this paper, we have measured radiation transmission of thirteen different heavy concretes produced by using barite, colemanite and normal aggregates for 276, 303, 356 and 383 keV γ-ray energies of {sup 133}Ba radioactive isotope by using CdTe semiconductor detector and calculated linear and mass attenuation coefficients. Results were compared to X-ray mass attenuation coefficient values of NIST. The attenuation of γ-ray photons is governed primarily by the atomic number and density of the concrete as a shielding material. Photon attenuation depends on pair production, photoelectric effect and Compton scattering. In this study, linear attenuation coefficients (μ) decrease with the increasing the photon energy for these concretes. Heavy elements there is a fairly sharp minimum in the absorption cross section at about this energy region. It is indicated that attenuation coefficient is useful parameter for radiation shielding. As known, the linear attenuation coefficients of the concretes used in this study at these γ-ray energies are not available in the literature by using CdTe semiconductor detectors. The data shows the effects of including boron-rich colemanite or barium-rich barite in addition to normal aggregate in the concrete mixtures. As would be expected, the barite is desirable for photon attenuation, while the colemanite is desirable for neutron attenuation. The concretes including barite and colemanite would be preferred as materials for building construction of radiation oncology departments, radiology departments, nuclear reactors and radiation research laboratories against photon and neutron particle radiations. But other aggregate properties of colemanite such as abrasion resistance and freeze-thaw durability are also found to be important for concrete but they cannot be quantified at this study. (authors)

36 MATERIALS SCIENCE↗

DOE OSTI2024

Energy storage system

A system, a method, and a computer program product for providing heterogeneous unifying battery storage. A state-of-health value of a battery is determined. The state-of-health value of the battery is less than an original capacity value of the battery. The battery is connected to an electrical power source for re-conditioning. A target state-of-health value for the battery and a number of cycles required to achieve the target state-of-health value of the battery are determined. Each cycle in the number of cycles includes at least one of: a charging the battery and a discharging the battery. The battery is re-conditioned by cycling the battery using the determined number of cycles. Cycling includes drawing electrical power from the electrical power source.

Tong, Shijie↗

DOE OSTI2017

GAUSS XI

Gauss XI is an interpretive computer program for the analysis of the gamma-ray spectra.

Egger, AnnE.↗

DOE OSTI2017

Development of a utility computer-based training program on setpoint methodology for nuclear power plants - 129

The purpose of this paper is to provide an overview of the development of the Computer- Based Training (CBT) Program for utility engineers on Setpoint Methodology for nuclear power plants. This computer based training module was prepared by the Advanced Engineering Training initiative. The Advanced Engineering Training initiative is a nuclear industry effort to develop technical computer-based training modules for targeted areas of engineering expertise. The target audience is engineers at commercial nuclear power facilities in the U. S. These modules are being developed using the systematic approach to training and are being designed to maintain industry engineering expertise and to facilitate engineering knowledge transfer. The objective of this training effort is to support the development of the next generation of site technical experts for specific topic areas. The training is one part of three key components that would facilitate knowledge transfer to the future site Subject Matter Expert (SME). The following are the major focus areas covered in the Setpoint Methodology CBT: Chapter 1 provides an overview of the role instrument setpoints perform in the defense-in-depth safety analysis for the plant as well introducing key definitions related to setpoints. Chapter 2 provides the history and current regulatory requirements related to setpoint control including the laws and guidance provided by the U.S. NRC. Chapter 3 provides describes the setpoint methodology including terms, formulae, methods of combination, graded approach and methods for drift analysis. This is a major part of the training module in covering the incremental uncertainty terms that need to be considered and how they are combined. Chapter 4 provides and overview of the role of setpoints in Technical Specifications and associated documents. Chapter 5 describes how the regulatory process, related to setpoint control, has been implemented in NRC inspections and how the major elements of the 10 CFR 50.59 reviews apply to the performance of a new or revised setpoint calculation. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2024

Transforming Energy Through Computational Excellence: NREL HPC Resources for High Performance Computing for Energy Innovation (HPC4EI) Program

NREL hosts computing facilities for the U.S. Department of Energy's Office of Energy Efficiency and Renewable Energy (EERE). In 2024, NREL introduced Kestrel, the 3rd generation, EERE-sponsored supercomputer dedicated to renewable energy and energy efficiency research. Kestrel has already been used for hundreds of research projects by NREL, other national laboratories, and university partners. This includes HPC4EI-sponsored industrial partnerships.

high-performance computing↗

DOE OSTI2021

The Early Years and Evolution of the DOE Computational Science Graduate Fellowship Program

The U.S. Department of Energy Computational Graduate Fellowship Program, celebrating 30 years of existence in 2021, is one of the most successful graduate fellowships in the world as well as one of the longest running programs in the U.S. Department of Energy. Furthermore, this article discusses the conception, early years and evolution of the fellowship over the past thirty years.

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2014

High Performance Computing - Power Application Programming Interface Specification (V.1.0)

Measuring and controlling the power and energy consumption of high performance computing systems by various components in the software stack is an active research area [13, 3, 5, 10, 4, 21, 19, 16, 7, 17, 20, 18, 11, 1, 6, 14, 12]. Implementations in lower level software layers are beginning to emerge in some production systems, which is very welcome. To be most effective, a portable interface to measurement and control features would significantly facilitate participation by all levels of the software stack. We present a proposal for a standard power Application Programming Interface (API) that endeavors to cover the entire software space, from generic hardware interfaces to the input from the computer facility manager.

Laros III, James H.↗

DOE OSTI2016

High Performance Computing: Power Application Programming Interface Specification (V.1.3)

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2017

High Performance Computing - Power Application Programming Interface Specification Version 2.0.

Measuring and controlling the power and energy consumption of high performance computing systems by various components in the software stack is an active research area. Implementations in lower level software layers are beginning to emerge in some production systems, which is very welcome. To be most effective, a portable interface to measurement and control features would significantly facilitate participation by all levels of the software stack. We present a proposal for a standard power Application Programming Interface (API) that endeavors to cover the entire software space, from generic hardware interfaces to the input from the computer facility manager.

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2016

High Performance Computing - Power Application Programming Interface Specification Version 1.4

97 MATHEMATICS AND COMPUTING↗