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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 1,243 records · Page 69
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Combined curing as a novel approach to improve resistance of ultra-high performance concrete to explosive spalling under high temperature and its mechanical properties

Highlights: • The combined curing can significantly enhance the resistance of UHPC to explosive spalling under high temperature. • The combined curing can improve the mechanical properties of UHPC remarkably. • Both further hydration of residual cement and pozzolanic reaction are activated under a high temperature steam condition. - Abstract: The influence of a combined curing composed of precuring in hot water and heating in dry air on mechanical properties and explosive spalling of ultra-high performance concrete (UHPC) subjected to high temperature was investigated in this experiment. Meanwhile, the microstructure of UHPC was determined by means of X-ray diffraction (XRD) with Rietveld method, thermogravimetric analysis (TGA), scanning electron microscope (SEM) and mercury intrusion porosimetry (MIP). The results indicate that the combined curing improves the mechanical properties of UHPC remarkably. More importantly, it is quite efficient to enhance the resistance of UHPC to explosive spalling under high temperature, which should be a novel approach to prohibit spalling occurrence, different from adding polypropylene fiber. During dry air heating in the combined curing, a high temperature steam environment is established in the dense structure framework of UHPC formed in the preceding hot water curing. Both further hydration of residual cement and pozzolanic reaction of mineral admixtures can be activated during the dry air heating, leading to not only the formation of a large amount of dense hydrates (such as C-S-H gels, katoite, tobermorite and xonotlite, etc.), which accumulate in the aforementioned existing framework, but also the consumption of most of internal free water. Therefore, both mechanical properties of UHPC and its resistance to explosive spalling under high temperature can be enhanced significantly.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Strength and rupture geometry of un-irradiated C26M FeCrAl under LOCA burst testing conditions

Ferritic iron-chromium-aluminum (FeCrAl) alloys are an accident tolerant fuel candidate to replace the incumbent Zr-based claddings. Nuclear grade FeCrAl alloys are marked by superior high temperature mechanical behavior and exceptional steam oxidation resistance, both of which increase safety margins during accident scenarios. In the present study, the loss of coolant accident (LOCA) burst behaviors of three un-irradiated commercially fabricated cladding materials in a simulated LOCA environment were compared: (1) T35Y2, a 1 st generation nuclear grade FeCrAl, (2) C26M, a 2 nd generation nuclear grade FeCrAl, and (3) Zircaloy-2. Both FeCrAl alloys showed improved mechanical strength and steam oxidation resistance compared to Zircaloy-2. C26M claddings burst at significantly higher temperatures for all tested engineering hoop stresses, had limited ballooning, and demonstrated preferential fuel retention behavior in terms of burst opening area and length. High temperature tensile data for C26M is also presented. For both un-irradiated FeCrAl alloys, it was found that a distinct “threshold” burst stress signified the transition between small and large openings. Finally, higher threshold hoop stresses were associated with higher uniaxial strength for the FeCrAl alloys, indicating that tensile data, rather than creep data, could be useful for predicting rupture size and assessing fuel dispersal concerns.

36 MATERIALS SCIENCE↗
DOE OSTI2011

The use of commercial and industrial waste in energy recovery systems - A UK preliminary study

Highlights: > Commercial and industrial waste samples collected. > Samples analysed for calorific value, moisture, ash and elemental composition. > Values similar to those of municipal solid waste and refuse derived fuel. > Sampled waste could be used in current energy recovery systems with minimal retrofitting. > Sampled waste could account 6.5% towards the UK's 2020 renewable electricity target if all qualifying waste is used. - Abstract: With 2020 energy targets set out by the EU fast approaching, the UK is trying to source a higher proportion of its energy from renewable resources. Coupled with this, a growing population and increasing trends in consumer demand have resulted in national waste loads increasing. A possible solution to both issues is energy-from-waste (EfW) technologies. Many studies have focused on municipal solid waste (MSW) as a potential feedstock, but appear to overlook the potential benefits of commercial and industrial waste (C and IW). In this study, samples of C and IW were collected from three North West waste management companies and Lancaster University campus. The samples were tested for their gross and net calorific value, moisture content, ash content, volatile matter, and also elemental composition to determine their suitability in EfW systems. Intra-sample analysis showed there to be little variation between samples with the exception two samples, from waste management site 3, which showed extensive variation with regards to net calorific value, ash content, and elemental analysis. Comparisons with known fuel types revealed similarities between the sampled C and IW, MSW, and refuse derived fuel (RDF) thereby justifying its potential for use in EfW systems. Mean net calorific value (NCV) was calculated as 9.47 MJ/kg and concentrations of sulphur, nitrogen, and chlorine were found to be below 2%. Potential electrical output was calculated using the NCV of the sampled C and IW coupled with four differing energy generation technologies. Using a conventional incinerator with steam cycle, total electrical output was calculated as 24.9 GWh, based on a plant operating at 100,000 tpa. This value rose to 27.0 GWh when using an integrated gasification combined cycle. A final aspect of this study was to deduce the potential total national electrical output if all suitable C and IW were to be used in EfW systems. Using incineration coupled with a steam turbine, this was determined to be 6 TWh, 1.9% of the national demand thereby contributing 6.5% towards the UK's 2020 renewable electricity target.

29 ENERGY PLANNING, POLICY, AND ECONOMY↗
DOE OSTI2012

Novel rotating field probe for inspection of tubes

Inspection of steam generator tubes in nuclear power plants is extremely critical for safe operation of the power plant. In the nuclear industry, steam generator tube inspection using eddy current techniques has evolved over the years from a single bobbin coil, to rotating probe coil (RPC) and array probe, in an attempt to improve the speed and reliability of inspection. The RPC probe offers the accurate spatial resolution but involves complex mechanical rotation. This paper presents a novel design of eddy current probes based on rotating fields produced by three identical coils excited by a balanced three-phase supply. The sensor thereby achieves rotating probe functionality by electronic means and eliminates the need for mechanical rotation. The field generated by the probe is largely radial that result in induced currents that flow circularly around the radial axis and rotating around the tube at a synchronous speed effectively producing induced eddy currents that are multidirectional. The probe will consequently be sensitive to cracks of all orientations in the tube wall. The finite element model (FEM) results of the rotating fields and induced currents are presented. A prototype probe is being built to validate simulation results.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2013

Eddy current NDE performance demonstrations using simulation tools

To carry out performance demonstrations of the Eddy-Current NDE processes applied on French nuclear power plants, EDF studies the possibility of using simulation tools as an alternative to measurements on steam generator tube mocks-up. This paper focuses on the strategy led by EDF to assess and use code{sub C}armel3D and Civa, on the case of Eddy-Current NDE on wears problem which may appear in the U-shape region of steam generator tubes due to the rubbing of anti-vibration bars.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Eddy current signal comparison for tube identification

Inspection of nuclear power plant steam generator tubes is required to justify continued safe plant operation. The steam generators consist of thousands of tubes with nominal diameters of 15 to 22mm, approximately 1mm wall thickness, and 20 to 30m in length. The tubes are inspected by passing an eddy current probe through the tubes from tube end to tube end. It is critical to know exactly which tube identification (row and column) is associated with each tube's data. This is controlled by a precision manipulator that provides the tube ID to the eddy current system. Historically there have been some instances where the manipulator incorrectly reported the tube ID. This can have serious consequences including lack of inspection of a tube, or if a pluggable indication is detected, the tube is likely to be mis-plugged thereby risking a primary to secondary leak.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2016

Simulated performance of biomass gasification based combined power and refrigeration plant for community scale application

Thermal performance analysis and sizing of a biomass gasification based combined power and refrigeration plant (CPR) is reported in this study. The plant is capable of producing 100 kWe of electrical output while simultaneously producing a refrigeration effect, varying from 28-68 ton of refrigeration (TR). The topping gas turbine cycle is an indirectly heated all-air cycle. A combustor heat exchanger duplex (CHX) unit burns producer gas and transfer heat to air. This arrangement avoids complex gas cleaning requirements for the biomass-derived producer gas. The exhaust air of the topping GT is utilized to run a bottoming ammonia absorption refrigeration (AAR) cycle via a heat recovery steam generator (HRSG), steam produced in the HRSG supplying heat to the generator of the refrigeration cycle. Effects of major operating parameters like topping cycle pressure ratio (r{sub p}) and turbine inlet temperature (TIT) on the energetic performance of the plant are studied. Energetic performance of the plant is evaluated via energy efficiency, required biomass consumption and fuel energy savings ratio (FESR). The FESR calculation method is significant for indicating the savings in fuel of a combined power and process heat plant instead of separate plants for power and process heat. The study reveals that, topping cycle attains maximum power efficiency of 30%in pressure ratio range of 8-10. Up to a certain value of pressure ratio the required air flow rate through the GT unit decreases with increase in pressure ratio and then increases with further increase in pressure ratio. The capacity of refrigeration of the AAR unit initially decreases up to a certain value of topping GT cycle pressure ratio and then increases with further increase in pressure ratio. The FESR is found to be maximized at a pressure ratio of 9 (when TIT=1100°C), the maximum value being 53%. The FESR is higher for higher TIT. The heat exchanger sizing is also influenced by the topping cycle pressure ratio and GT-TIT.

09 BIOMASS FUELS↗
DOE OSTI2025

Optimal Economic Dispatch and Load-Following Strategies for Nuclear Integrated Energy Systems

The need for distributed and adaptable energy resources that can handle the growing unpredictability in both supply and demand is rising as the power system continues to modernize. In order to satisfy those needs and maintain grid resilience, nuclear power plants can dynamically control their output, despite typically being used as baseload generators. By incorporating energy storage and renewable energy sources, nuclear integrated energy systems are designed to satisfy the electrical and thermal demands of different end-user applications while ensuring flexible power operation. These systems generate revenue by participating in both wholesale and ancillary services electricity markets, as well as commodity markets for various byproducts generated from coupled industrial processes. This study addresses the economic dispatch efficiency of a tightly coupled nuclear integrated energy system comprising a gigawatt-scale light water reactor, commercialized in the U.S., a high-temperature steam electrolysis unit, a district heating network, and specified electrical loads. To demonstrate the nuclear power plant’s flexibility within the day-ahead unit commitment and economic dispatch framework, while maintaining equilibrium even during periods of refueling outages, this paper develops a mixed-integer linear programming framework that models the subsystems and components of its nuclear steam supply system. A systematic comparative analysis of flexible versus baseload nuclear power plant operation under varying levels of renewable energy integration indicates that flexible operation enhances system profitability by more than 18% while also increasing energy storage utilization, improving reactor responsiveness to load fluctuations, and allowing for greater participation across numerous electricity markets.

29 ENERGY PLANNING, POLICY, AND ECONOMY↗
DOE OSTI2024

Optimization of District Heating Network Parameters in Steady-State Operation

Here we examine the modeling, simulation, and optimization of district heating systems, which are widely used for thermal transport using steam or hot water as a carrier. We propose a generalizable framework to specify network models and scenario parameters, and develop an optimization method for evaluating system states including pressures, fluid flowrates, and temperatures throughout the network. The network modeling includes pipes, thermal plants, pumps, and passive or controllable loads as system components. We propose basic models for thermodynamic fluid transport and enforce the balance of physical quantities in steady-state flow over co-located outgoing and return networks. We formulate an optimization problem with steam and hot water as the outgoing and return carriers, as in legacy twentieth century systems. The physical laws and engineering limitations are specified for each component type, and the thermal network flow optimization problem is formulated and solved for a realistic test network under several scenarios.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗
DOE OSTI2013

Calculation of the dose load during dismantling of large components in the process of decommissioning of nuclear installations

The paper is focused on the decommissioning process from the point of view of steam generator dismantling. A brief description of the steam generator (used in nuclear power plants with VVER 440 type reactor) and used computer code VISIPLAN 3D ALARA are given. The main part deals with the created model and dismantling strategy together with variable input parameters - decay time and decontamination. The obtained results - external exposure of workers and the influence of time and pre-dismantling decontamination - are studied. Also detailed analyses of every dismantling step considered are presented. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Treatment of irradiated graphite from French Bugey reactor

In 2008, following the general French plan for nuclear waste management, Electricite de France attempted to find for irradiated graphite an alternative solution to direct storage at the low-activity long-life storage center in France managed by the national agency for wastes (ANDRA). EDF management requested that its engineering arm, EDF CIDEN, study the graphite treatment alternatives to direct storage. In mid-2008, this study revealed the potential advantage for EDF to use a steam reforming process known as Thermal Organic Reduction, 'THOR' (owned by Studsvik, Inc., USA), to treat or destroy the graphite matrix and limit the quantity of secondary waste to be stored. In late 2009, EDF began a test program with Studsvik to determine if the THOR steam reforming process could be used to destroy the graphite. The program also sought to determine if the graphite could be treated to release the bulk of activity while minimizing the gasification of the bulk mass of the graphite. In October 2009, tests with non-irradiated graphite were completed and demonstrated destruction of a graphite matrix by the THOR process at satisfactory rates. After gasifying the graphite, focus shifted to the effect of roasting graphite at high temperatures in inert gases with low concentrations of oxidizing gases to preferentially remove volatile radionuclides while minimizing the graphite mass loss to 5%. A radioactive graphite sleeve was imported from France to the US for these tests. Completed in April 2010, 'Phase I' of testing showed that the process removed >99% of H-3 and 46% of C-14 with <6% mass loss. Completed in September 2011, 'Phase II' testing achieved increased removals as high as 80% C-14. During Phase II, it was also discovered that roasting in a reducing atmosphere helped to limit the oxidation of the graphite. Future work seeks to explore the effects of reducing gases to limit the bulk oxidation of graphite. If the graphite could be decontaminated of long-lived radionuclides up to 95% for C-14 while minimizing mass loss to <5%, this would minimize the volume of any secondary waste streams and potentially lower the waste class of the larger bulk of graphite. Alternatively, if up to 95% decontamination of C-14 is achieved, the graphite may be completely gasified which could result in lower disposal. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2014

RELAP5-3D Transient Modelling for NGNP Integrated Plant

The High-Temperature Gas-cooled Reactor (HTGR) is designed with outlet temperatures ranging between 750°C and 800°C. These high outlet temperatures enhance the power production efficiency and facilitate a variety of industrial applications. The objective of this study is to understand the response of the primary system to potential transients in the secondary system. For this analysis, the transient condition originates in the Intermediate Heat Exchanger (IHX) or Steam Generator (SG) of the HTGR-integrated plant. The transients analysed are: a loss of pressure; loss of feedwater flow; inadvertent closure of main steam valve; decrease in returning gas temperature and heat load step change. The results show a large dependence on the negative reactivity added to the fuel as a function of increased temperature. The returning gas temperature decrease transient resulted in the highest fuel temperature (1361°C). Fuel temperature was shown to be less than the 1600°C fuel limit for each case analysed.

42 ENGINEERING↗
DOE OSTI2018

Global Value Chain and Manufacturing Analysis on Geothermal Power Plant Turbines

In this study, we have undertaken a robust analysis of the global supply chain and manufacturing costs for components of Organic Rankine Cycle (ORC) Turboexpander and steam turbines used in geothermal power plants. We collected a range of market data influencing manufacturing from various data sources and determined the main international manufacturers in the industry. The data includes the manufacturing cost model to identify requirements for equipment, facilities, raw materials, and labor. We analyzed three different cases; 1) 1 MW geothermal ORC Turboexpander 2) 5 MW ORC Turboexpander 3) 20 MW geothermal Steam Turbine

15 GEOTHERMAL ENERGY↗
DOE OSTI2016

Mechanical behavior at high temperatures of highly oxygen- or hydrogen-enriched α and prior-β phases of Zirconium alloys - 2016-0063

During a hypothetical loss-of-coolant accident, zirconium alloy fuel claddings can be loaded by internal pressure and exposed to steam at high temperatures (HTs) up to 1,200 deg. C and then cooled and quenched in water. A significant fraction of the oxygen that reacts with the cladding during HT oxidation diffuses beneath the oxide through the metallic substrate. This induces a progressive transformation of the metallic β{sub Zr}-phase layer into an intermediate layer of α{sub Zr}(O) phase containing up to 7 weight % (wt.%) oxygen. Furthermore, in some specific conditions, the cladding may rapidly absorb a significant amount of hydrogen during steam exposition at HTs. As a β{sub Zr} stabilizer, hydrogen would mainly diffuse and concentrate up to several thousands of parts per million by weight (wppm) into the inner β{sub Zr}-phase layer. This study intends to provide new and more comprehensive data on the HT mechanical behavior of the α{sub Zr}(O) and (prior-)β{sub Zr} phases containing high contents of oxygen and hydrogen, respectively. Model samples produced from M5{sup R} and Zircaloy-4 cladding tubes homogeneously charged in oxygen (less than or equal to 6 wt.%) and hydrogen (less than or equal to 3,000 wppm), respectively, were prepared. Their mechanical behavior was determined under vacuum between 800 and 1,100 deg. C for the oxygen-enriched α{sub Zr} phase and in air between 700 and 20 deg. C after cooling from the β{sub Zr} temperature domain for the hydrogen-enriched (prior-)β{sub Zr} phase. The α{sub Zr} phase was substantially strengthened and embrittled by oxygen. The contribution of the α{sub Zr}(O) layer to the HT creep behavior of an oxidized fuel cladding tube subjected to internal pressure is evaluated by finite element analysis. Mechanical strength and ductility of the model (prior-)β{sub Zr} phase appear to be affected by hydrogen contents of 2000-3000 wppm in ways that depend on temperature. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2016

Effects of chemistry and microstructure on corrosion performance of Zircaloy-2-based BWR cladding - 2016-0066

In boiling water reactors (BWRs), Zircaloy-2 is typically the material of choice for fuel cladding. With increased focus on fuel performance and reliability, consistent cladding corrosion performance under varying BWR conditions has become an acknowledged requirement for modern BWR fuel. To support optimization of corrosion performance, modifications to alloy chemistry and manufacture conditions can be made, supported by ex-reactor corrosion testing and demonstrated through in-reactor operation. In this work, the combined effects of alloy chemistry and fabrication process were investigated. The effect of alloy chemistry is often complicated by variation within a production ingot. In this work, additional measurements were intentionally taken at the tube-shell stage (before cold pilgering) so that chemistry more representative of the final cladding could be obtained and the effect of iron, nickel, and tin could be examined more closely. Although chemistry range of Zircaloy-2 was the primary focus, the effect of iron above the ASTM limit in a new alloy was also explored. Three types of cladding fabrication processes were considered. Two types included a heat treatment applied after the first cold pilgering and annealing, and the treatment was limited to the outer portion of the cladding and differed in the temperature of the heat treatment; the third type had no special heat treatments after the first cold pilgering and annealing. The weight gain of the three types of cladding with varying compositions under two corrosion test conditions (400 deg. C steam and 410/520 deg. C two-stage steam) are discussed in terms of the initial second-phase particle sizes obtained from transmission electron microscopy examinations. Based on poolside inspections and hot-cell examinations, the in-reactor corrosion performance of the two cladding types up to typical discharge exposures are discussed in the terms of corrosion margin, microstructure evolution, and correlation with ex-reactor corrosion test data. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2016

Post-quench ductility of zirconium alloy cladding materials - 2016-0088

The ductility of zirconium-based alloys after steam oxidation at high temperatures, followed by quenching in water, is of significant interest for the nuclear industry. High-temperature steam oxidation and quenching of nuclear fuel cladding occurs during postulated loss-of-coolant accidents. The industry is currently establishing in-reactor operation limits for peak cladding temperature and time at temperature based on the measured ductile-to-brittle transition of different cladding materials, both in the as-fabricated condition and using hydrogen charging as a surrogate for in-reactor operation. Aiming to study possible differences in post-quench ductility caused by differences in chemical composition, extensive testing on Zircaloy-4, ZIRLO{sup R}, and Optimized ZIRLO{sup TM} cladding materials as a function of hydrogen content, extent of oxidation, and peak cladding temperature was performed. The results showed similar ductile-to-brittle transition for the alloys studied despite differences in chemical composition, with a decrease in the extent of oxidation to reach the ductile to brittle transition with increasing hydrogen content. Microstructural characterization was performed to investigate the evolution of the oxygen-enriched layer that forms underneath the oxide, both during the time at high temperatures and during cooling. Significantly different morphologies of the oxygen-enriched layer were observed with different cooling rates, particularly for alloys containing niobium. Possible mechanisms for the development of the oxygen-enriched layer are discussed. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2011

SOLIDIFICATION OF THE HANFORD LAW WASTE STREAM PRODUCED AS A RESULT OF NEAR-TANK CONTINUOUS SLUDGE LEACHING AND SODIUM HYDROXIDE RECOVERY

The U.S. Department of Energy (DOE), Office of River Protection (ORP), is responsible for the remediation and stabilization of the Hanford Site tank farms, including 53 million gallons of highly radioactive mixed wasted waste contained in 177 underground tanks. The plan calls for all waste retrieved from the tanks to be transferred to the Waste Treatment Plant (WTP). The WTP will consist of three primary facilities including pretreatment facilities for Low Activity Waste (LAW) to remove aluminum, chromium and other solids and radioisotopes that are undesirable in the High Level Waste (HLW) stream. Removal of aluminum from HLW sludge can be accomplished through continuous sludge leaching of the aluminum from the HLW sludge as sodium aluminate; however, this process will introduce a significant amount of sodium hydroxide into the waste stream and consequently will increase the volume of waste to be dispositioned. A sodium recovery process is needed to remove the sodium hydroxide and recycle it back to the aluminum dissolution process. The resulting LAW waste stream has a high concentration of aluminum and sodium and will require alternative immobilization methods. Five waste forms were evaluated for immobilization of LAW at Hanford after the sodium recovery process. The waste forms considered for these two waste streams include low temperature processes (Saltstone/Cast stone and geopolymers), intermediate temperature processes (steam reforming and phosphate glasses) and high temperature processes (vitrification). These immobilization methods and the waste forms produced were evaluated for (1) compliance with the Performance Assessment (PA) requirements for disposal at the IDF, (2) waste form volume (waste loading), and (3) compatibility with the tank farms and systems. The iron phosphate glasses tested using the product consistency test had normalized release rates lower than the waste form requirements although the CCC glasses had higher release rates than the quenched glasses. However, the waste form failed to meet the vapor hydration test criteria listed in the WTP contract. In addition, the waste loading in the phosphate glasses were not as high as other candidate waste forms. Vitrification of HLW waste as borosilicate glass is a proven process; however the HLW and LAW streams at Hanford can vary significantly from waste currently being immobilized. The ccc glasses show lower release rates for B and Na than the quenched glasses and all glasses meet the acceptance criterion of < 4 g/L. Glass samples spiked with Re{sub 2}O{sub 7} also passed the PCT test. However, further vapor hydration testing must be performed since all the samples cracked and the test could not be performed. The waste loading of the iron phosphate and borosilicate glasses are approximately 20 and 25% respectively. The steam reforming process produced the predicted waste form for both the high and low aluminate waste streams. The predicted waste loadings for the monolithic samples is approximately 39%, which is higher than the glass waste forms; however, at the time of this report, no monolithic samples were made and therefore compliance with the PA cannot be determined. The waste loading in the geopolymer is approximately 40% but can vary with the sodium hydroxide content in the waste stream. Initial geopolymer mixes revealed compressive strengths that are greater than 500 psi for the low aluminate mixes and less than 500 psi for the high aluminate mixes. Further work testing needs to be performed to formulate a geopolymer waste form made using a high aluminate salt solution. A cementitious waste form has the advantage that the process is performed at ambient conditions and is a proven process currently in use for LAW disposal. The Saltstone/Cast Stone formulated using low and high aluminate salt solutions retained at least 97% of the Re that was added to the mix as a dopant. While this data is promising, additional leaching testing must be performed to show compliance with the PA. Compressive strength tests must also be performed on the Cast Stone monoliths to verify PA compliance. Based on testing performed for this report, the borosilicate glass and Cast Stone are the recommended waste forms for further testing. Both are proven technologies for radioactive waste disposal and the initial testing using simulated Hanford LAW waste shows compliance with the PA. Both are resistant to leaching and have greater than 25% waste loading.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2011

An Evaluation of Possible Next-Generation High Temperature Molten-Salt Power Towers

Since completion of the Solar Two molten-salt power tower demonstration in 1999, the solar industry has been developing initial commercial-scale projects that are 3 to 14 times larger. Like Solar Two, these initial plants will power subcritical steam-Rankine cycles using molten salt with a temperature of 565 °C. The main question explored in this study is whether there is significant economic benefit to develop future molten-salt plants that operate at a higher receiver outlet temperature. Higher temperatures would allow the use of supercritical steam cycles that achieve an improved efficiency relative to today's subcritical cycle (~50% versus ~42%). The levelized cost of electricity (LCOE) of a 565 °C subcritical baseline plant was compared with possible future-generation plants that operate at 600 or 650 °C. The analysis suggests that ~8% reduction in LCOE can be expected by raising salt temperature to 650 °C. However, most of that benefit can be achieved by raising the temperature to only 600 °C. Several other important insights regarding possible next-generation power towers were also drawn: (1) the evaluation of receiver-tube materials that are capable of higher fluxes and temperatures, (2) suggested plant reliability improvements based on a detailed evaluation of the Solar Two experience, and (3) a thorough evaluation of analysis uncertainties.

14 SOLAR ENERGY↗