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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 775 records · Page 43
All sourcesNASA NTRSDOE OSTI
DOE OSTI2011

Neutronics qualification of the Jules Horowitz reactor fuel by interpretation of the VALMONT experimental program - Transposition of the uncertainties on the reactivity of JHR with JEF2.2 and JEFF3.1.1

The new European material testing Jules Horowitz Reactor (JHR), currently under construction in Cadarache center (CEA France), will use LEU (20% enrichment in {sup 235}U) fuels (U{sub 3}Si{sub 2} for the start up and UMoAl in the future) which are quite different from the industrial oxide fuel, for which an extensive neutronics qualification database has been established. The HORUS3D/N neutronics calculation scheme, used for the design and safety studies of the JHR, is being developed within the framework of a rigorous verification-validation-qualification methodology. In this framework, the experimental VALMONT (Validation of Aluminium Molybdenum uranium fuel for Neutronics) program has been performed in the MINERVE facility of CEA Cadarache (France), in order to qualify the capability of HORUS3D/N to accurately calculate the reactivity of the JHR reactor. The MINERVE facility using the oscillation technique provides accurate measurements of reactivity effect of samples. The VALMONT program includes oscillations of samples of UAl{sub x}/Al and UMo/Al with enrichments ranging from 0.2% to 20% and Uranium densities from 2.2 to 8 g/cm{sup 3}. The geometry of the samples and the pitch of the experimental lattice ensure maximum representativeness with the neutron spectrum expected for JHR. By comparing the effect of the sample with the one of a known fuel specimen, the reactivity effect can be measured in absolute terms and be compared to computational results. Special attention was paid to the rigorous determination and reduction of the experimental uncertainties. The calculational analysis of the VALMONT results was performed with the French deterministic code APOLLO2. A comparison of the impact of the different calculation methods, data libraries and energy meshes that were tested is presented. The interpretation of the VALMONT experimental program allowed the qualification of JHR fuel UMoAl8 (with an enrichment of 19.75% {sup 235}U) by the Minerve-dedicated interpretation tool: PIMS. The effect of energy meshes and evaluations put forward the JEFF3.1.1/SHEM scheme that leads to a better calculation of the reactivity effect of VALMONT samples. Then, in order to quantify the impact of the uncertainties linked to the basic nuclear data, their propagation from the cross section measurement to the final computational result was analysed in a rigorous way by using a nuclear data re-estimation method based on Gauss-Newton iterations. This study concludes that the prior uncertainties due to nuclear data (uranium, aluminium, beryllium and water) on the reactivity of the Begin Of Cycle (BOC) for the JHR core reach 1217 pcm at 2{sigma}. Now, the uppermost uncertainty on the JHR reactivity is due to aluminium. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

A Station-hybrid HVDC System Structure and Control Strategies for Cross-seam Power Transmission

Cross-seam high voltage DC (HVDC) transmission in the United States can provide an efficient and flexible bulk power delivery highway between the interconnections. Recent studies have shown significant economic benefits of cross-seam HVDC transmission systems by taking the advantages of load diversity, frequency response, renewable energy diversity, and energy arbitrage in the different interconnections. This paper introduces a station-hybrid system scheme for cross-seam interconnections. Here, the proposed station-hybrid system combines the advantages of line commutated converter (LCC) and voltage source converter (VSC) technologies, thus enabling reliable and flexible bidirectional power flows across the interconnections. Control strategies of the station-hybrid system are investigated with a focus on power flow reversal control under normal operation conditions and grid support control under emergency conditions. The feasibility and effectiveness of the proposed station-hybrid system and corresponding control strategies are verified by simulations in PSCAD/EMTDC.

Station-hybrid system↗
DOE OSTI2017

Numerical Simulation of Measurements during the Reactor Physical Startup at Unit 3 of Rostov NPP

The results of numerical calculations and measurements of some reactor parameters during the physical startup tests at unit 3 of Rostov NPP are presented. The following parameters are considered: the critical boron acid concentration and the currents from ionization chambers (IC) during the scram system efficiency evaluation. The scram system efficiency was determined using the inverse point kinetics equation with the measured and simulated IC currents. The results of steady-state calculations of relative power distribution and efficiency of the scram system and separate groups of control rods of the control and protection system are also presented. The calculations are performed using several codes, including precision ones.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Reduced Model for Direct Induction Startup Scenario Development on MAST-U and NSTX-U

A reduced semi-empirical model using time-dependent axisymmetric vacuum field calculations is used to develop the prefill and feed-forward coil current targets required for reliable direct induction (DI) startup on the new MA-class spherical tokamaks, MAST-U and NSTX-U. The calculations are constrained by operational limits unique to each device, such as the geometry of the conductive elements and active coils, power supply specifications and coil heating and stress limits. The calculations are also constrained by semi-empirical models for sufficient breakdown, current drive, equilibrium and stability of the plasma developed from a shared database. A large database of DI startup on NSTX and NSTX-U is leveraged to quantify the requirements for achieving a reliable breakdown (Ip ~ 20 kA). It is observed that without pre-ionization, STs access the large E/P regime at modest loop voltage (Vloop) where the electrons in the weakly ionized plasma are continually accelerating along the open field lines. This ensures a rapid (order millisecond) breakdown of the neutral gas, even without pre-ionization or high-quality field nulls. The timescale of the initial increase in Ip on NSTX is reproduced in the reduced model provided a mechanism for impeding the applied electric field is included. Most discharges that fail in the startup phase are due to an inconsistency in the evolution of the plasma current (Ip) and equilibrium field or loss of vertical stability during the burn-through phase. The requirements for the self-consistent evolution of the fields in the weakly and full-ionized plasma states are derived from demonstrated DI startup on NSTX, NSTX-U and MAST. The predictive calculations completed for MAST-U and NSTX-U illustrate that the maximum Ip ramp rate (dIp/dt) in the early startup phase is limited by the voltage limits on the poloidal field coils on MAST-U and passive vertical stability on NSTX-U.

Direct induction↗
DOE OSTI2011

Photovoltaics (Fact Sheet)

The fact sheet summarizes the goals and activities of the DOE Solar Energy Technologies Program efforts within its photovoltaics subprogram.

14 SOLAR ENERGY↗
DOE OSTI2015

Preliminary Analysis of the Transient Reactor Test Facility (TREAT) with PROTEUS

The neutron transport code PROTEUS has been used to perform preliminary simulations of the Transient Reactor Test Facility (TREAT). TREAT is an experimental reactor designed for the testing of nuclear fuels and other materials under transient conditions. It operated from 1959 to 1994, when it was placed on non-operational standby. The restart of TREAT to support the U.S. Department of Energy’s resumption of transient testing is currently underway. Both single assembly and assembly-homogenized full core models have been evaluated. Simulations were performed using a historic set of WIMS-ANL-generated cross-sections as well as a new set of Serpent-generated cross-sections. To support this work, further analyses were also performed using additional codes in order to investigate particular aspects of TREAT modeling. DIF3D and the Monte-Carlo codes MCNP and Serpent were utilized in these studies. MCNP and Serpent were used to evaluate the effect of geometry homogenization on the simulation results and to support code-to-code comparisons. New meshes for the PROTEUS simulations were created using the CUBIT toolkit, with additional meshes generated via conversion of selected DIF3D models to support code-to-code verifications. All current analyses have focused on code-to-code verifications, with additional verification and validation studies planned. The analysis of TREAT with PROTEUS-SN is an ongoing project. This report documents the studies that have been performed thus far, and highlights key challenges to address in future work.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

HSI Prototypes for Human Systems Simulation Laboratory

This report describes in detail the design and features of three Human System Interface (HSI) prototypes developed by the Institutt for Energiteknikk (IFE) in support of the U.S. Department of Energy’s Light Water Reactor Sustainability Program under Contract 128420 through Idaho National Laboratory (INL). The prototypes are implemented for the Generic Pressurized Water Reactor simulator and installed in the Human Systems Simulation Laboratory at INL. The three prototypes are: 1) Power Ramp display 2) RCS Heat-up and Cool-down display 3) Estimated time to limit display The power ramp display and the RCS heat-up/cool-down display are designed to provide good visual indications to the operators on how well they are performing their task compared to their target ramp/heat-up/cool-down rate. The estimated time to limit display is designed to help operators restore levels or pressures before automatic or required manual actions are activated.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Initial Comparison of Direct and Legacy Modeling Approaches for Radial Core Expansion Analysis

Radial core expansion in sodium-cooled fast reactors provides an important reactivity feedback effect. As the reactor power increases due to normal start up conditions or accident scenarios, the core and surrounding materials heat up, causing both grid plate expansion and bowing of the assembly ducts. When the core restraint system is designed correctly, the resulting structural deformations introduce negative reactivity which decreases the reactor power. Historically, an indirect procedure has been used to estimate the reactivity feedback due to structural deformation which relies upon perturbation theory and coupling legacy physics codes with limited geometry capabilities. With advancements in modeling and simulation, radial core expansion phenomena can now be modeled directly, providing an assessment of the accuracy of the reactivity feedback coefficients generated by indirect legacy methods. Recently a new capability was added to the PROTEUS-SN unstructured geometry neutron transport solver to analyze deformed meshes quickly and directly. By supplying the deformed mesh in addition to the base configuration input files, PROTEUS-SN automatically processes material adjustments including calculation of region densities to conserve mass, calculation of isotopic densities according to material models (for example, sodium density as a function of temperature), and subsequent re-homogenization of materials. To verify the new capability of directly simulating deformed meshes, PROTEUS-SN was used to compute reactivity feedback for a series of contrived yet representative deformed configurations for the Advanced Burner Test Reactor design. The indirect legacy procedure was also performed to generate reactivity feedback coefficients for the same deformed configurations. Interestingly, the legacy procedure consistently overestimated reactivity feedbacks by 35% compared to direct simulations by PROTEUS-SN. This overestimation indicates that the legacy procedures are in fact not conservative and could be overestimating reactivity feedback effects that are closely tied to reactor safety. We conclude that there is indeed value in performing direct simulation of deformed meshes despite the increased computational expense. PROTEUS-SN is already part of the SHARP multi-physics toolkit where both thermal hydraulics and structural mechanical feedback modeling can be applied but this is the first comparison of direct simulation to legacy techniques for radial core expansion.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Operational Characteristics of an Accelerator Driven Fissile Solution System

Operational characteristics represent the set of responses that a nuclear system exhibits during normal operation. Operators rely on this behavior to assess the status of the system and to predict the consequences of off-normal events. These characteristics largely refer to the relationship between power and system operating conditions. The static and dynamic behavior of a chain-reacting system, operating at sufficient power, is primarily governed by reactivity effects. The science of reactor physics has identified and evaluated a number of such effects, including Doppler broadening and shifts in the thermal neutron spectrum. Often these reactivity effects are quantified in the form of feedback coefficients that serve as coupling coefficients relating the neutron population and the physical mechanisms that drive reactivity effects, such as fissile material temperature and density changes. The operational characteristics of such nuclear systems usually manifest themselves when perturbations between system power (neutron population) and system operating conditions arise. Successful operation of such systems requires the establishment of steady equilibrium conditions. However, prior to obtaining the desired equilibrium (steady-state) conditions, an approach from zero-power (startup) must occur. This operational regime may possess certain limiting system conditions that must be maintained to achieve effective startup. Once steady-state is achieved, a key characteristic of this operational regime is the level of stability that the system possesses. Finally, a third operational regime, shutdown, may also possess limiting conditions of operation that must be maintained. This report documents the operational characteristics of a “generic” Accelerator Driven Fissile Solution (ADFS) system during the various operational regimes of startup, steady-state operation, and shutdown. Typical time-dependent behavior for each operational regime will be illustrated, and key system parameters, such as response times, will be quantified. A generalized linear systems analysis of steady-state operations will be performed to evaluate the level of stability of ADFS systems. This information should provide a basic understanding of typical ADFS system operational behavior, and facilitate the development of monitoring procedures and operator aids.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2013

Progress on Plant-Level Modeling and Simulation of Used Nuclear Fuel Dissolution: Startup Condition with Quasi-Developed Flow and Varying Acid Concentration

An update to a previous report is presented here describing the startup condition of a continuous dissolver. The need for a careful implementation of a startup condition is twofold. First, by varing the used fuel composition and various other parameters of the model, it becomes difficult to solve the governing equations accurately or to obtain a converged solution for an arbitrary initial condition. This is evident when coupling the dissolver module with a third-party user simulator. The second impetus for improving the modeling of the startup process is that an improved realism is obtained and the model is more complete and useful for accounting material balances rigorously. The aforementioned previous report on this dissolver module is a required reading for the understanding of this update.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Comparison of spatial dynamics and point kinetics approaches in multiphysics modeling of the molten salt reactor experiment

In this work, we present validation test results of fully coupled neutronics and thermal-hydraulics models of the Molten Salt Reactor Experiment (MSRE) against experimental data of the zero power pump transients and the natural circulation tests at low power. To capture the strong coupling between neutronics and thermal-hydraulics due to fuel circulation, and to account for the delayed neutron precursor (DNP) distribution, the porous media thermal-hydraulics solver Pronghorn was fully coupled to the spatial neutron dynamics code Griffin, which solves the neutron diffusion equation, and to the 0-D point kinetics solver Squirrel, using a 2-D homogenized representation of the MSRE. The validation test results show very good agreement with experimental data for both point kinetics and spatial dynamics simulations, capturing the strong feedback effect and DNP losses in the MSRE. The 0-D code Squirrel accurately predicted the time-dependent behavior in the MSRE given the steady-state spatial dynamics solution of Griffin.

42 - ENGINEERING↗
NASA NTRS2012

In-Flight Performance of the TES Loop Heat Pipe Rejection System: Seven Years in Space

The Tropospheric Emission Spectrometer (TES) is an infrared, high spectral resolution Fourier transform spectrometer with a 3.3 to 15.4 micron wavelength coverage. TES is a scanning instrument intended for determining the chemical state of the Earth's lower atmosphere (troposphere) from the surface to 30+ km. TES produces vertical profiles of important pollutant and greenhouse gases such as carbon monoxide, ozone, methane, and water vapor on a global scale every other day. TES was launched into orbit onboard NASA's earth Observing System Aura spacecraft on July 15, 2004 from Vandenberg Air Force Base, California.

tropospheric emission spectrometer↗
DOE OSTI2012

Addressing Control of Hazardous Energy (COHE) Requirements in a Laser Safety Program

OSHA regulation 29CFR1910.147 specifies control of hazardous energy requirements for 'the servicing and maintenance of machines and equipment in which the unexpected energization or start up of the machines or equipment, or release of stored energy could cause injury to employees.' Class 3B and Class 4 laser beams must be considered hazardous energy sources because of the potential for serious eye injury; careful consideration is therefore needed to safely de-energize these lasers. This paper discusses and evaluates control of hazardous energy principles in this OSHA regulation, in ANSI Z136.1 ''Safe Use of Lasers,'' and in ANSI Z244.1 ''Control of Hazardous Energy, Lockout/Tagout and Alternative Methods.'' Recommendations are made for updating and improving CoHE (control of hazardous energy) requirements in these standards for their applicability to safe laser operations.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2012

Core design of long life-cycle fast reactors operating without reactivity margin

In this paper we consider a possibility of designing a fast reactor core that operates without reactivity margin for a long time. This study is based on the physical principle of fast reactor operating in a self-adjustable neutron-nuclear regime (SANNR-1) introduced by L.P. Feoktistov (1988-1993) and improved by V. Ya. Gol'din SANNR-2 (1995). The mathematical modeling of active zones of fast reactors in SANNR modes is held by authors since 1992. The numerical simulation is based on solving the neutron transport equation coupled with quasi-diffusion equations. The calculations have been performed using standard 26 energy groups. We use a hierarchy of spatial models of 1D, 1.5D, 2D, and 3D geometries. The spatial models of higher dimensionality are used for verification of results. The calculations showed that operation of the reactor in this mode increases its efficiency, safety and simplifies management. It is possible to achieve continuous work of the reactor in SANNR-2 during 7-10 years without fuel overloads by means of further optimization of the mode. Small reactivity margin is used only for the reactor start up. After first 10-15 days the reactor in SANNR-2 operates without reactivity margin. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Testing and Starting up of the Plasma Melting Facility for Industrial Treatment of Radioactive Waste - 17362

The operation and maintenance of nuclear power plants, the nuclear fuel cycle in general, research laboratories and pharmaceutical, medical and industrial facilities generate low-level radioactive waste which, along with the historical radioactive waste from past nuclear activities, needs to be treated and stored, awaiting final disposal. Plasma technology offers a very effective way of treating this waste with a high volume reduction factor (VRF), free from organics, liquids and moisture, and meets without a doubt the acceptance criteria for safe storage and disposal. By means of a plasma beam of approximately 5000 deg. C, the inorganic materials are melted into a glassy slag, containing most of the radioactive isotopes while the organic material is gasified and afterwards oxidized in an afterburner and purified in an off-gas cleaning system. This paper describes the principles of plasma, the different waste feed systems, off-gas treatment and operational experience and future plasma plants. More specifically, it presents the new full-scale plasma facility for the treatment of radioactive waste at the Kozloduy Nuclear Power Plant in Bulgaria for which construction and assembly was finalized end of August 2016. It also describes in more detail the Site Acceptance Tests (SAT) of this facility. This facility is designed and built as an EPC contract by the Joint Venture Iberdrola Ingenieria y Construccion (Spain) and Belgoprocess (Belgium). (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Transition to a Thorium-based Fuel Cycle - Choices, Challenges, and Options

The Charter for the Evaluation and Screening (E and S) of Nuclear Fuel Cycle Options requested by the US Department of Energy calls for the identification of promising fuel cycle options to help guide the future directions of research and development in this area. The final report for this study, which was produced by various national laboratory and industry experts, collected the expansive set of fuel cycle options into 40 evaluation groups (EGs). In support of this study, detailed steady-state physics and mass flow data for each of these groups were compiled into a set of Fuel Cycle Data Packages (FCDPs). The data in the FCDPs were then used to quantify the nine criteria detailed in the study to identify the most promising EGs. In addition to supporting the E and S study, data from these FCDPs can be used to benchmark fuel cycle simulation tools for equilibrium calculations. Once benchmarked, their dynamic modeling capabilities can support the continuing effort to determine the most effective approaches for transitioning to proposed advanced future fuel cycles. While steady-state conditions can be calculated directly in some cases, fuel cycle simulation tools allow for the transition scenarios to be simultaneously optimized on multiple parameters such as the fissile material availability, the amount of desirable material in storage, and the time required for transition. In addition to fuel cycle simulations where material availability is considered and is often a constraint to achieving transition, determining the technology needs to enable a transition to take place is also a key consideration (e.g., what technology is needed, when it is needed, and whether it is needed only during the transition and not for the equilibrium fuel cycle). This paper evaluates the material and technology issues and challenges associated with a transition from a uranium-based fuel cycle (such as that operating in the US today, represented by EG01) to a thorium-based fuel cycle, recognizing that thorium requires fissile material to start up. Based on the findings of the E and S report, the two most promising thorium fuel cycle options are considered in this report: (1) EG26, a thermal spectrum molten salt reactor (MSR), and (2) EG28, represented by a sodium fast reactor (SFR, fast spectrum) and a resource-renewable boiling water reactor (RBWR, intermediate spectrum). (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗