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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 865 records · Page 48
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Antineutrino analysis for continuous monitoring of nuclear reactors: Sensitivity study

This paper explores the various contributors to uncertainty on predictions of the antineutrino source term which is used for reactor antineutrino experiments and is proposed as a safeguard mechanism for future reactor installations. The errors introduced during simulation of the reactor burnup cycle from variation in nuclear reaction cross sections, operating power, and other factors are combined with those from experimental and predicted antineutrino yields, resulting from fissions, evaluated, and compared. The most significant contributor to uncertainty on the reactor antineutrino source term when the reactor was modeled in 3D fidelity with assembly-level heterogeneity was found to be the uncertainty on the antineutrino yields. Using the reactor simulation uncertainty data, the dedicated observation of a rigorously modeled small, fast reactor by a few-ton near-field detector was estimated to offer reduction of uncertainty on antineutrino yields in the 3.0–6.5 MeV range to a few percent for the primary power-producing fuel isotopes, even with zero prior knowledge of the yields.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Physical particularities of nuclear reactors using heavy moderators of neutrons

In nuclear reactors, thermal neutron spectra are formed using moderators with small atomic weights. For fast reactors, inserting such moderators in the core may create problems since they efficiently decelerate the neutrons. In order to form an intermediate neutron spectrum, it is preferable to employ neutron moderators with sufficiently large atomic weights, using {sup 233}U as a fissile nuclide and {sup 232}Th and {sup 231}Pa as fertile ones. The aim of the work is to investigate the properties of heavy neutron moderators and to assess their advantages. The analysis employs the JENDL-4.0 nuclear data library and the SCALE program package for simulating the variation of fuel composition caused by irradiation in the reactor. The following main results are obtained. By using heavy moderators with small neutron moderation steps, one is able to (1) increase the rate of resonance capture, so that the amount of fertile material in the fuel may be reduced while maintaining the breeding factor of the core; (2) use the vacant space for improving the fuel-element properties by adding inert, strong, and thermally conductive materials and by implementing dispersive fuel elements in which the fissile material is self-replenished and neutron multiplication remains stable during the process of fuel burnup; and (3) employ mixtures of different fertile materials with resonance capture cross sections in order to increase the resonance-lattice density and the probability of resonance neutron capture leading to formation of fissile material. The general conclusion is that, by forming an intermediate neutron spectrum with heavy neutron moderators, one can use the fuel more efficiently and improve nuclear safety.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2013

Recovery and Blend-Down Uranium for Beneficial use in Commercial Reactors - 13373

In April 2001 the Department of Energy (DOE) and the Tennessee Valley Authority (TVA) signed an Interagency Agreement to transfer approximately 33 MT of off-specification (off-spec) highly enriched uranium (HEU) from DOE to TVA for conversion to commercial reactor fuel. Since that time additional surplus off-spec HEU material has been added to the program, making the total approximately 46 MT off-spec HEU. The disposition path for approximately half (23 MT) of this 46 MT of surplus HEU material, was down blending through the H-canyon facility at the Savannah River Site (SRS). The HEU is purified through the H-canyon processes, and then blended with natural uranium (NU) to form low enriched uranium (LEU) solution with a 4.95% U-235 isotopic content. This material was then transported to a TVA subcontractor who converted the solution to uranium oxide and then fabricated into commercial light water reactor (LWR) fuel. This fuel is now powering TVA reactors and supplying electricity to approximately 1 million households in the TVA region. There is still in excess of approximately 10 to 14 MT of off-spec HEU throughout the DOE complex or future foreign and domestic research reactor returns that could be recovered and down blended for use in either currently designed light water reactors, ∼5% enriched LEU, or be made available for use in subsequent advanced 'fast' reactor fuel designs, ∼19% LEU. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Transition Analysis of Promising U.S. Future Fuel Cycles Using ORION

The US Department of Energy Office of Fuel Cycle Technologies performed an evaluation and screening (E&S) study of nuclear fuel cycle options to help prioritize future research and development decisions. Previous work for this E&S study focused on establishing equilibrium conditions for analysis examples of 40 nuclear fuel cycle evaluation groups (EGs) and evaluating their performance according to a set of 22 standardized metrics. Following the E&S study, additional studies are being conducted to assess transitioning from the current US fuel cycle to future fuel cycle options identified by the E&S study as being most promising. These studies help inform decisions on how to effectively achieve full transition, estimate the length of time needed to undergo transition from the current fuel cycle, and evaluate performance of nuclear systems and facilities in place during the transition. These studies also help identify any barriers to achieve transition. Oak Ridge National Laboratory (ORNL) Fuel Cycle Options Campaign team used ORION to analyze the transition pathway from the existing US nuclear fuel cycle—the once-through use of low-enriched-uranium (LEU) fuel in thermal-spectrum light water reactors (LWRs) —to a new fuel cycle with continuous recycling of plutonium and uranium in sodium fast reactors (SFRs). This paper discusses the analysis of the transition from an LWR to an SFR fleet using ORION, highlights the role of lifetime extensions of existing LWRs to aid transition, and discusses how a slight delay in SFR deployment can actually reduce the time to achieve an equilibrium fuel cycle.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Serrated Yielding in Alloy 709 at 500 deg. C

Gen-IV nuclear power reactors are being designed to improve the efficiency, safety, reliability, and lifetime in comparison with the current fleet of nuclear power reactors. Hence, candidate structural materials for Gen-IV reactors need to have superior mechanical properties to withstand harsh reactor conditions including higher operating temperature, higher radiation dose, and extremely corrosive environment. Sodium Fast Reactor (SFR) is one of the Gen-IV reactor concepts and it has a high technology readiness level for deployment. It uses liquid sodium as coolant and fast neutron spectrum to induce fission with core outlet temperature of no higher than 550 deg. C. In general, austenitic stainless steels have an excellent corrosion resistance and good creep properties due to high chromium and nickel contents. Alloy 709 is an austenitic stainless steel with base Fe-20Cr-25Ni stabilized with nitrogen and strengthened with niobium. Its excellent mechanical properties relative to similar steels (304 and 316) including high strength at elevated temperature, corrosion resistance, thermal stability, sodium compatibility and creep properties make it an excellent candidate material for the nth-of-a-kind SFR structural applications. Understanding the plastic deformation behavior at service temperature for Alloy 709 is essential to predict the performance of the material under the operating conditions. In this study, the effects of the temperature and strain-rate on the uniaxial quasi-static tensile plastic deformation of Alloy 709 are discussed with special emphasis on dynamic strain aging and work hardening behavior.

36 MATERIALS SCIENCE↗
DOE OSTI2015

Multiple recycle of REMIX fuel at VVER-1000 operation in closed fuel cycle

The basic features of loading the VVER-1000 core with a new variant of REMIX fuel (REgenerated MIXture of U–Pu oxides) are considered during its multiple recycle in a closed nuclear fuel cycle. The fuel composition is produced on the basis of the uranium–plutonium regenerate extracted at processing the spent nuclear fuel (SNF) from a VVER-1000, depleted uranium, and the fissionable material: {sup 235}U as a part of highly enriched uranium (HEU) from warheads superfluous for defense purposes or {sup 233}U accumulated in thorium blankets of fusion (electronuclear) neutron sources or fast reactors. Production of such a fuel assumes no use of natural uranium in addition. When converting a part of the VVER-1000 reactors to the closed fuel cycle based on the REMIX technology, the consumption of natural uranium decreases considerably, and there is no substantial degradation of the isotopic composition of plutonium or change in the reactor-safety characteristics at the passage from recycle to recycle.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Uncertainty analysis of a SFR core with sodium plenum

The new concepts of Sodium-cooled Fast Reactors have to reach the Generation IV safety objectives. In this regard the Sodium Void Effect has to be minimized for the future projects of large-size SFR as well as the uncertainties on it. The Inst. of Radiological Protection and Nuclear Safety (IRSN) as technological support of French public authorities is in charge of safety assessment of operating and under construction reactors, as well as future projects. In order to state about the safety of new SFR designs the IRSN must be able to evaluate core parameters and their uncertainties. In this frame a sensitivity and uncertainty study has been performed to evaluate the impact of nuclear data uncertainty on sodium void effect, for the benchmark model of large SFR BN-800. The benchmark parameters (effective multiplication factor and sodium void effect) have been evaluated using two codes, the deterministic code ERANOS and the Monte Carlo code SCALE, while the S/U analysis has been performed only with SCALE. The results of the these studies point out the most relevant cross section uncertainties that affect the SVE and how efforts should be done in increasing the existing nuclear data accuracies. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

On the use of moderating material to enhance the feedback coefficients in SFR cores with high minor actinide content

The use of fine distributed moderating material to enhance the feedback effects and to reduce the sodium void effecting sodium cooled fast reactor cores is described. The influence of the moderating material on the neutron spectrum, the power distribution, and the burnup distribution is shown. The consequences of the use of fine distributed moderating material into fuel assemblies with fuel configurations foreseen for minor actinide transmutation is analyzed and the transmutation efficiency is compared. The degradation of the feedback effects due to the insertion of minor actinides and the compensation by the use of moderating materials is discussed. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2013

Properties measurements of (U{sub 0.7}Pu{sub 0.3})O{sub 2-x} in PO{sub 2}-controlled atmosphere

The investigation of physical properties of uranium and plutonium mixed oxide (MOX) fuels is important for the development of fast reactor fuels. It is well known that MOX is a nonstoichiometric oxide, and the physical properties change drastically with the Oxygen-to-Metal (O/M) ratio. A control technique for O/M ratio was established for measurements of high temperature properties of uranium and plutonium mixed oxide fuels. Sintering behavior, thermal expansion and O/M change of (U{sub 0.7}Pu{sub 0.3})O{sub 2.00} and (U{sub 0.7}Pu{sub 0.3})O{sub 1.99} were investigated in PO{sub 2}-controlled atmosphere which was controlled by H{sub 2}/H{sub 2}O gas system. Sintering behavior changed drastically with O/M ratio, and shrinkage of (U{sub 0.7}Pu{sub 0.3})O{sub 2.00} was faster and more advanced at lower temperatures as compared with (U{sub 0.7}Pu{sub 0.3})O{sub 1.99}. Thermal expansion was observed to be slightly increased with decreasing O/M ratio. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Nuclear Material Characterization Using High-Energy X-rays at BNL Synchrotrons: From Reactor Steels and Molten Salts to Large Hadron Collider Novel Materials

New and improved materials, composites, and super-alloys capable of withstanding the anticipated extreme states of fusion reactors, fission reactors (i.e., high-temperature fast reactors) and multi-MW particle accelerators, with the latter facing similarly extreme states to reactors under energetic protons, are continuously sought. Materials structures like improved reactor steels, super-alloys, and novel composites are continuously being explored to help meet both the challenge of the higher-demand environments and longevity under extreme conditions. Greater fluxes of irradiating species (primarily fast neutrons and energetic protons), extreme temperatures, and aggressively corrosive environments make up the new cocktail of operating conditions these materials must withstand. A challenge in characterizing the effects under intense irradiation, high temperatures, and stresses is the establishment of the link between lattice-induced damage and phase transformation under particle irradiation and macroscopic physical properties, which ultimately determine performance in the real environment. With ever increasing irradiation fluxes, dramatic changes in the material properties are expected to occur stemming from the evolution or development of new microstructures.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Calculation of Neutron Damage to Sodium Coolant Inlet Plenum

The coolant inlet plenum is a life limiting component of a Sodium-cooled Fast Reactor (SFR). The inlet plenum is a fixed reactor component which receives sodium from the cold side of the sodium pool redan and directs it into the inlet nozzles of the fuel assemblies. Neutron leakage from the reactor core causes irradiation damage to the steel components of the inlet plenum leading to a reduction in strength. The life of the reactor is essentially tied to the acceptable strength degradation of the inlet plenum. Therefore, typical SFR designs employ a neutron reflector/shield as an integral feature of the fuel assembly. The role of the axial reflector/shield is to: 1) direct fast neutrons back to the core, and 2) attenuate the neutron flux to minimize neutron interaction with the inlet plenum. The longer the reflector and the greater the share of steel within the reflector design, the greater the attenuation. The axial reflector is heated by gamma rays from the core, and must have sufficient coolant penetrations to ensure adequate cool-ability. Also, the sodium coolant must pass through the reflector before reaching the fuel rods above it. The present work investigates varying reflector geometries in support of the Versatile Test Reactor (VTR), a SFR research reactor planned to restart fast reactor irradiation capabilities in the U.S. DOE laboratory complex. This work will demonstrate that the neutron damage to the inlet plenum is relatively insensitive to the number and arrangements of coolant channel penetrations through the reflector.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2018

Tritium Module for Calculating the Behavior of Tritium in a Loop of a Reactor Installation with Sodium Coolant

A description of the models and the results of tests of the Tritium module, which is intended for modeling the transport of tritium in the loops of fast reactors, are presented. The characteristics features of the module are alienability, functional relation with other modules of the code, and calculation of model coefficients as a function of the temperature of the coolant and surface of the loops. The module makes it possible to perform calculations of the behavior of tritium, including transport along loops, penetrability through the channel walls, run-off in a cold trap, and flow through leaks in the reactor installation.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Materials, Turbomachinery and Heat Exchangers for Supercritical CO2 Systems

The objective of this project is to produce the necessary data to evaluate the performance of the supercritical carbon dioxide cycle. The activities include a study of materials compatibility of various alloys at high temperatures, the heat transfer and pressure drop in compact heat exchanger units, and turbomachinery issues, primarily leakage rates through dynamic seals. This experimental work will serve as a test bed for model development and design calculations, and will help define further tests necessary to develop high-efficiency power conversion cycles for use on a variety of reactor designs, including the sodium fast reactor (SFR) and very high-temperature gas reactor (VHTR). The research will be broken into three separate tasks. The first task deals with the analysis of materials related to the high-temperature S-CO 2 Brayton cycle. The most taxing materials issues with regard to the cycle are associated with the high temperatures in the reactor side heat exchanger and in the high-temperature turbine. The system could experience pressures as high as 20MPa and temperatures as high as 650°C. The second task deals with optimization of the heat exchangers required by the S-CO 2 cycle; the S-CO 2 flow passages in these heat exchangers are required whether the cycle is coupled with a VHTR or an SFR. At least three heat exchangers will be required: the pre-cooler before compression, the recuperator, and the heat exchanger that interfaces with the reactor coolant. Each of these heat exchangers is unique and must be optimized separately. The most challenging heat exchanger is likely the pre-cooler, as there is only about a 40°C temperature change but it operates close to the CO{sub 2} critical point, therefore inducing substantial changes in properties. The proposed research will focus on this most challenging component. The third task examines seal leakage through various dynamic seal designs under the conditions expected in the S-CO 2 cycle, including supercritical, choked, and two-phase flow conditions.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

ASME Code Considerations for the Compact Heat Exchanger

The mission of the U.S. Department of Energy (DOE), Office of Nuclear Energy is to advance nuclear power in order to meet the nation's energy, environmental, and energy security needs. Advanced high temperature reactor systems such as sodium fast reactors and high and very high temperature gas-cooled reactors are being considered for the next generation of nuclear reactor plant designs. The coolants for these high temperature reactor systems include liquid sodium and helium gas. Supercritical carbon dioxide (sCO₂), a fluid at a temperature and pressure above the supercritical point of CO₂, is currently being investigated by DOE as a working fluid for a nuclear or fossil-heated recompression closed Brayton cycle energy conversion system that operates at 550°C (1022°F) at 200 bar (2900 psi). Higher operating temperatures are envisioned in future developments. All of these design concepts require a highly effective heat exchanger that transfers heat from the nuclear or chemical reactor to the chemical process fluid or the to the power cycle. In the nuclear designs described above, heat is transferred from the primary to the secondary loop via an intermediate heat exchanger (IHX) and then from the intermediate loop to either a working process or a power cycle via a secondary heat exchanger (SHX). The IHX is a component in the primary coolant loop which will be classified as "safety related." The intermediate loop will likely be classified as "not safety related but important to safety." These safety classifications have a direct bearing on heat exchanger design approaches for the IHX and SHX. The very high temperatures being considered for the VHTR will require the use of very high temperature alloys for the IHX and SHX. Material cost considerations alone will dictate that the IHX and SHX be highly effective; that is, provide high heat transfer area in a small volume. This feature must be accompanied by low pressure drop and mechanical reliability and robustness. Classic shell and tube designs will be large and costly, and may only be appropriate in steam generator service in the SHX where boiling inside the tubes occurs. For other energy conversion systems, all of these features can be met in a compact heat exchanger design. This report will examine some of the ASME Code issues that will need to be addressed to allow use of a Code-qualified compact heat exchanger in IHX or SHX nuclear service. Most effort will focus on the IHX, since the safety-related (Class A) design rules are more extensive than those for important-to-safety (Class B) or commercial rules that are relevant to the SHX.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Characterization of BOR-60 Irradiated 14YWT-NFA1 Tubes

Tubes of FCRD 14YWT-NFA1 Alloy were placed in the BOR-60 reactor and irradiated under a fast flux neutron environment to two conditions: 7 dpa at 360-370 °C and 6 dpa at 385-430 °C. Small sections of the tube were cut and sent to UC Berkeley for nanohardness testing and focused ion beam (FIB) milling of TEM specimens. FIB specimens were sent back to LANL for final FIB milling and TEM imaging. Hardness data and TEM images are presented in this report. This is the first fast reactor neutron irradiated information on the 14YWT-NFA1 alloy.

36 MATERIALS SCIENCE↗
DOE OSTI2018

United States Nuclear Manufacturing Infrastructure Assessment

This report provides a review and assessment of the existing U.S. infrastructure capabilities to manufacture on a commercial scale, the major components and unique materials required by the various U.S. designs for advanced nuclear reactors and small modular reactors (SMRs). The approach was to compare the needs of SMRs and advanced reactors against the capability of U.S. manufacturers to ascertain gaps that could inhibit U.S. industrial support for deployment of SMRs or advanced reactors. This report focused on the reactor plant designs being developed by NuScale Power (the NuScale Power Module), TerraPower (both the Traveling Wave Reactor and the Molten Chloride Fast Reactor to a limited extent), and X-energy (the Xe-100). To gather the information, MPR visited each of these companies and selected candidate suppliers to tour their facilities and conduct in-person interviews. Information was also obtained by telephone interviews, internet research, and literature review of technical reports. The overarching conclusion from this study is that the U.S. has existing capability or can readily develop capability to support and sustain most, but not all, aspects of commercial deployment of advanced reactors and SMRs. The greatest concerns for capability gaps in the U.S. infrastructure for supply of components for SMRs and advanced reactors are associated with reactor pressure vessels, steam generators, and fuel fabrication. Foreign vendors can supply reactor pressure vessels and steam generators. Issues with fuel fabrication are associated with uranium enrichment and development of fabrication methods for the innovative fuel concepts planned for advanced reactors. The technical capabilities of U.S. manufacturers appear satisfactory with the possible exceptions noted above. However, the capacity of the existing U.S. manufacturing infrastructure may be insufficient for a high rate of construction. Manufacturers contacted as part of this study indicated a willingness to expand capacity if sufficient orders are in place. The review of U.S. infrastructure included a detailed review of capabilities in many different processes including forging, tube drawing, machining, fabrication, and testing. In addition, MPR researched the advanced manufacturing processes for additive manufacturing (i.e., 3D printing) and powder metallurgy - hot isostatic pressing. Reactor plant designers and candidate suppliers are planning to rely on traditional processes for the initial plants. If advanced manufacturing processes are developed and accepted by the nuclear industry, these processes may be used in the future. In the immediate term, ongoing research (much of which is funded by DOE) is necessary to develop and refine advanced manufacturing methods. This report includes several recommendations for DOE that will advance deployment of SMRs and advanced reactors from a manufacturing perspective and other areas identified during the course of this project.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Advancement of Supercritical Carbon Dioxide Technology through Round Robin Testing and Fundamental Modeling

Growing interest in supercritical carbon dioxide (sCO 2 ) cycles for advanced reactors is driving the need for corrosion data on candidate plant materials. The sCO 2 Brayton cycle is being considered for power conversion systems for a number of nuclear reactor concepts including the sodium fast reactor, fluoride salt-cooled high temperature reactor, high temperature gas reactor, and several types of small modular reactors. Multiple organizations have developed test facilities to address the corrosion data knowledge gap in high temperature, high pressure sCO 2 environments but, to date, there has been no formal test program among these organizations to validate the consistency of the data they produce. A demonstration of comparable and reproducible results enables a coordinated effort to explore the sCO 2 parameter space relevant to advanced reactor technology. This proposal establishes a round robin test plan for sCO 2 corrosion testing and the organization of a sCO 2 Materials Group to guide future materials testing directions. Furthermore, this proposal outlines efforts to elucidate the mechanisms of sCO 2 corrosion by performing identical tests in supercritical steam, testing model alloys with various composition and environmental conditions, accompanied by modeling efforts to study rate controlling mechanisms of corrosion and to help understanding the effects of impurities.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗