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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 91 records · Page 5
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Nuclear data uncertainty propagation applied to the versatile test reactor conceptual design

We report the Versatile Test Reactor (VTR) currently under development is a 300 MWth sodium-cooled fast reactor (SFR) fueled with ternary metal alloy fuel, which aims to accelerate the testing of advanced nuclear fuels, materials, instrumentation, and sensors in high flux environments that are necessary to license the next generation of advanced reactor concepts. To support the VTR design process, uncertainties associated with the nuclear data has been propagated through the reactor core neutronics calculation to global parameters of interest, such as the core multiplication factor, kinetic parameters, and various reactivity feedback coefficients, following the sensitivity based uncertainty propagation approach. By folding the sensitivity coefficients, separately computed by the generalized perturbation theory code PERSENT and Monte Carlo code Serpent 2, with the variance-covariance matrices from COMMARA-2.0, we obtain the reaction-wise, isotope-wise, and overall uncertainties for each response of interest due to nuclear data uncertainty. With Serpent 2, the statistical error of the uncertainty is obtained by propagating the statistical error of the sensitivity coefficients through the same process using a newly developed uncertainty propagation method. From both codes, the overall top uncertainty contributors are found to be the cross section of Fe-56 elastic scattering, Na-23 elastic scattering, and U 238 inelastic scattering. The large contributions of the Fe-56 elastic scattering cross sections to global parameters are due to its relatively large relative uncertainty of 5–10% in nuclear data and the large volume of Fe-containing reflector assemblies in the fairly compact VTR core design. Both codes agreed well for the overall uncertainty estimates of all responses of interest, except the delayed neutron fraction, prompt neutron generation time, and the coolant density feedback coefficient, where Serpent 2 yielded a much larger value than PERSENT due to the large statistical error of sensitivity coefficients. The calculated uncertainties are also compared to those associated with other SFR cores. Another outcome of this study is a variance-covariance matrix of reactivity coefficients, which can be used in the subsequent uncertainty propagation to the system level to investigate the impact of identified uncertainties on system responses in the safety analysis.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Pressure Drop in Seven-Pin Wire-Wrapped Rod Bundle for the Sodium Cartridge Loop in Versatile Test Reactor

This work studies the hydrodynamics of the seven pin wire-wrapped rod bundle in the sodium cartridge loop for the Versatile Test Reactor (VTR) through scaled water experiments and computational fluid dynamics (CFD) simulations. Here, the scaling analysis is first performed to demonstrate the hydrodynamic similarity between water and sodium flows at the same Reynolds number Re. A separate-effects test facility is designed and constructed based on the scaling analysis. Detailed experimental data on the pressure drop covering a wide range of Re values (1165 to 27 689) are obtained, which are used to evaluate existing correlations for friction factor and to benchmark CFD simulations. The experimentally determined friction factors f Exp agree well with the Upgraded Cheng and Todreas Detailed Correlation and Pacio-Chen-Todreas Detailed Model within ±7% but are significantly underpredicted by Rehme's correlation by 25%. Various CFD near-wall treatment methods are tested using ANSYS Fluent and evaluated by experimental data. It is found that when the recommended wall y + values are met, most of the near-wall treatment methods can give accurate friction factor predictions. The resolved near-wall method (y + ~ 1) with the Shear Stress Transport k - ω turbulence model and the scalable wall functions (y + > 12) with the realizable k - ε turbulence model can predict f Exp within ±5% The standard wall functions (y + > 30) and nonequilibrium wall functions (y + > 30) with the realizable k - ε model can predict f Exp within ±10%.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS2014

Utilizing ARC EMCS Seedling Cassettes as Highly Versatile Miniature Growth Chambers for Model Organism Experiments

The aim of our ground testing was to demonstrate the capability of safely putting specific model organisms into dehydrated stasis, and to later rehydrate and successfully grow them inside flight proven ARC EMCS seedling cassettes. The ARC EMCS seedling cassettes were originally developed to support seedling growth during space flight. The seeds are attached to a solid substrate, launched dry, and then rehydrated in a small volume of media on orbit to initiate the experiment. We hypothesized that the same seedling cassettes should be capable of acting as culture chambers for a wide range of organisms with minimal or no modification. The ability to safely preserve live organisms in a dehydrated state allows for on orbit experiments to be conducted at the best time for crew operations and more importantly provides a tightly controlled physiologically relevant growth experiment with specific environmental parameters. Thus, we performed a series of ground tests that involved growing the organisms, preparing them for dehydration on gridded Polyether Sulfone (PES) membranes, dry storage at ambient temperatures for varying periods of time, followed by rehydration. Inside the culture cassettes, the PES membranes were mounted above blotters containing dehydrated growth media. These were mounted on stainless steel bases and sealed with plastic covers that have permeable membrane covered ports for gas exchange. The results showed we were able to demonstrate acceptable normal growth of C.elegans (nematodes), E.coli (bacteria), S.cerevisiae (yeast), Polytrichum (moss) spores and protonemata, C.thalictroides (fern), D.discoideum (amoeba), and H.dujardini (tardigrades). All organisms showed acceptable growth and rehydration in both petri dishes and culture cassettes initially, and after various time lengths of dehydration. At the end of on orbit ISS European Modular Cultivation System experiments the cassettes could be frozen at ultra-low temperatures, refrigerated, or chemically preserved before being returned to Earth for analyses. Our results suggest that with protocol modifications and future verification testing we can utilize the versatile EMCS to conduct tightly controlled experiments inside our culture cassettes for a wide variety of organisms. These physiological experiments would be designed to answer questions at the molecular level about the specific stress responses of space flight.

Model Organism Experiments↗
DOE OSTI2018

Evaluation of Control Rod Designs for the Versatile Test Reactor

Different fractions of boron carbide were investigated for the control rod assemblies of the Versatile Test Reactor (VTR). Reducing the number of rods per assembly increases the overall volume fraction of B4C, thus increasing reactivity worth. Doing so also reduces the melting margin inside the pellets and the reactivity worth loss by EOL. As a result the control rod layouts with the most (CR61) and least (CR7) number of rods were deemed less favorable for each of those two reasons. Overall, CR19 was found to have sufficient safety margins with the second best reactivity worth. The design only required a 44% 10B enrichment to reach the conservative margins put forward. The arrangement is recommended for both the control and safety rods of the VTR.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Fuel Loading Optimization and Planning for the Versatile Test Reactor

The Versatile Test Reactor (VTR) mission is to enable accelerated testing of advanced reactor fuels and materials required for advanced reactor technologies. In service of this duty, the goal of the design is to maximize the peak fast flux and the experimental volume exposed to a fast flux above 4.0 × 1015n/cms s. Delivering the desired neutron fluence for each experiment requires the capability to optimize the fuel loading to achieve a specified cycle length and flux, while also satisfying other criteria such as thermal hydraulic design constraints. To achieve the desired conditions and cycle length, several measures may be taken, including optimization of the fuel loading pattern, moving partially-burned assemblies back and forth between the core and the in-vessel storage, and potentially shuffling fuel within the core. Optimizing in a "brute-force" manner by solving a wholecore transport equation for hundreds of configurations would be inefficient. A sufficiently refined DIF3D-VARIANT solution for the whole core requires on the order of 1 hour. Instead, we seek an approximate evaluation of the parameters for many variations on a base core configuration without directly solving a diffusion or transport equation for each.

fuel loading optimization↗
DOE OSTI2019

ANS Summary of Safety Design Strategy for the Versatile Test Reactor

This paper describes the unique challenges with developing and documenting the Safety Design Strategy for the Versatile Test Reactor in the following key areas: 1. Identification of the key safety goals in the VTR facility design and operation that are to ensure adequate protection of the public, workers, and environment as required by 10 CFR 830, “Nuclear Safety Management.” These goals are fulfilled by an approach that applies defense-in-depth (DID) by applying various layers (or levels) of protection to prevent the release of radioactive and hazardous materials. 2. Early identification of the hazards that are normally associated with a large fast reactor facility can result from postulated failure conditions in one or more of the reactor systems or from operational errors. 3. Identification of the key decisions that must be made during the initial project life cycle that will affect the eventual design and construction of the VTR and enable clarity in safety expectations. 4. Identification of the applicable regulatory framework requirements for demonstrating compliance with safety basis requirements in 10 CFR 830 Subpart B for the VTR. 5. Application of the proposed Licensing Modernization Program (LMP) recommendations for advanced non-LWR reactors as outlined in Nuclear Energy Institute (NEI)-18-04, “Risk-Informed Performance-Based Guidance for Non-Light Water Reactor Licensing Basis Development.”

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Preliminary Evaluation of Fuel Management Strategies for the Versatile Test Reactor (VTR)

Different fuel management approaches have been investigated for the Versatile Test Reactor (VTR). The main design variables were the number of batches, the number of (orifice) zones, and the loading pattern within the core. The optimization criteria are maximizing the fast flux at the centermost position, minimizing the peak assembly power, and reducing fuel consumption. Design constraints were mainly driven by material limits such as peak clad temperature, and maximum fuel burnup. The exploratory study determined that active fuel shuffling only brought very limited benefits. A three-orifice region was found to provide enough flexibility to ensure thermal criteria were met. Lastly, a flexible batch loading termed ‘5-6-8’ was identified as a promising management strategy from a resource utilization standpoint.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

TRANSIENT ANALYSES OF THE VERSATILE TEST REACTOR USING RELAP5-3D

The DOE Office of Nuclear Energy established the Versatile Test Reactor (VTR) program in February 2017 in response to reports outlining the need for a fast spectrum test reactor. In this framework, Idaho National Laboratory (INL), as the proposed hosting site for the VTR, is expanding its capabilities leveraging on its staff nuclear expertise on experiments design and reactors operation, codes development and assessment, advanced safety analyses. One of the key points of this strategy is to develop capabilities necessary to perform transient analyses in order to understand the plant behavior during upset conditions and appropriately quantify margins. In this context system thermal-hydraulic models of the VTR are being developed using the SAS4A/SASSYS-1 code developed at Argonne National Laboratory as well as the INL code RELAP5-3D. The former provides reference analyses whereas the latter could provide an independent assessment of the reference analyses if deemed necessary by the project. This paper focuses on the RELAP5-3D modeling, highlighting the different solutions adopted for correctly analyzing sodium fast reactor behavior. After showing the modeling activities, we present a set of reference transient analyses for protected and unprotected transients, highlighting the intrinsic safety characteristics of the VTR design.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗