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All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Improved Heat Exchanger Performance Achieved Through Improved Condensate Drainage - 18478

Hanford's 242-A Evaporator has had the mission of reducing waste volume in the Hanford Tank Farms since the late 1970's. At the very heart of the 242-A evaporation process is the E-A-1 reboiler, a heat exchanger which adds heat to the waste, and the vacuum condenser system, a three-stage condenser/ejector system which condenses the waste vapors and provides the vacuum at which the evaporation process operates. Over a long period of time, significant degraded performance has been observed in the heat exchangers in both the heating system and in the vacuum condensing system. This degraded performance was indicative of degraded conditions in equipment and/or piping associated with both the heating and vacuum condenser systems. However, in searching for degraded condition in these systems, 242-A engineering also discovered that both systems contained design errors in the condensate drainage associated with all heat exchangers in both systems. By addressing the condensate design errors in addition to correcting the degraded conditions, system performance was restored and even possibly improved. All heat exchangers in the heating and vacuum condensing systems are shell and tube type heat exchangers. Both have liquid on the tube side and vapor on the shell side (in the case of the reboiler the liquid is tank waste and the vapor is 10 lb heating steam, and in the case of the vacuum condensers the liquid is cooling water and the vapor is waste vapor). In both systems, engineering proposed that the vapor condensate was backing up into the heat exchanger, resulting in added poor performance. By focusing on basic heat exchanger system design, projects were initiated to redesign the condensate drainage of these systems and re-plumb the condensate drain lines. Data collected before and after the replumbing provides evidence that the corrected drain lines from the heat exchangers improved the performance of the system as a whole, and in the case of the reboiler, removed a potentially damaging steam condensate water hammer condition. (author)

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

Fuel Performance Simulation of Iron-Chrome-Aluminum Cladding during Steady-State LWR Operation

During a severe accident scenario, such as a loss-of-coolant accident, a nuclear reactor may lose its capacity to cool its fuel. The fuel rod temperatures begin to increases, quickly reaching temperatures where the coolant begins to evaporate. This lowers the coolant level in the core and eventually uncovers the fuel. Without sufficient heat transfer from the fuel rods to the coolant, the temperature of the fuel and cladding will increase dramatically. In the case of traditional zirconium-based alloys, as the temperature of the cladding reaches ∼1200 deg. C the zirconium rapidly begins to oxidize with H{sub 2}O in the coolant and steam. This oxidation reaction causes both thinning of the Zr-alloy cladding as the metal reacts and releases large amounts of H{sub 2} gas into the reactor pressure vessel. In order to increase the safety margin of LWR fuel in severe accident scenarios, several alternative cladding materials have been proposed to replace the currently used zirconium-based alloys. Of these materials, there is a particular focus on select iron-chrome-aluminum (FeCrAl) alloys because of slower oxidation Kinetics in high temperature steam than zirconium-based alloys. This increased oxidation resistance may provide additional coping time to mitigate any further damage resulting from an accident. However, by changing the cladding material, the neutronic characteristics of the core will change. FeCrAl alloys exhibit thermal neutron absorption cross sections nearly ∼12-16 x greater than Zr-alloys. This decreases the reactivity of the core, subsequently decreasing the potential cycle length for operation. To counteract this, fissile material may be added in the form of increased enrichment or increased fuel mass or the cladding thickness might be decreased, effectively reducing the neutronic penalty from using the alloy. In this study, all three changes are considered. To assess the viability of FeCrAl as an alternative cladding to Zr-alloys, engineering-scale fuel performance analysis is used to monitor the cladding and integral fuel rod behavior through its operation. Ongoing work is being performed to identify key parameters essential to modeling FeCrAl alloys for LWR applications and properly model fuel compliance and stress relief mechanisms for UO{sub 2} fuel. To improve the predictive capability of fuel performance codes for FeCrAl cladding under normal operation, high-fidelity mechanistic/physics-based material models are needed. This information can be generated and improved by access to experimental data on targeted FeCrAl alloys. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Control strategies for coupling thermal energy storage systems with small modular reactors - 188

The contribution of intermittent (renewable) energy sources such as wind and solar continues to increase as renewables improve in both efficiency and price-point. However, the variability of renewables generates additional challenges for the electric grid in the form of rapidly varying electric loads. Proposed options for accommodating this load have included operating nuclear reactors in a load follow mode, or operating the reactor at or near steady state and bypassing steam directly to the condenser. Both of these strategies result in lost energy potential. In addition to lost energy potential, load follow operation can result in increased stress on the fuel and other mechanical components. A more attractive approach is to operate the reactor at or near steady state and bypass excess steam to a thermal energy storage system. The thermal energy can then be recovered, either for electric generation during periods of peak electric demand, or for use in ancillary applications such as desalination and hydrogen production. Sensible Heat Thermal Energy Storage is a mature technology currently used in solar energy systems. This research focuses on the design and coupling of such a system to Small Modular Reactors (SMRs), typical of Integral Pressurized Water Reactor (IPWR) designs currently under development. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Electric actuator modification on a turbine driven feed pump at Calvert Cliffs nuclear power plant - 296

Feedwater systems are a leading contributor to unit SCRAMs as indicated by the 2015 INPO data, where 9 of the 51 SCRAMs recorded were a result of feedwater issues. Digital feedwater control systems and enhanced feed pump control modifications can greatly improve performance and control of the plant feedwater systems and help drive down the potential for reactivity management events and unit SCRAMs. Westinghouse has a strong background in the implementation of digital feedwater control systems, and in planning for a recently completed feedwater control system upgrade at Calvert Cliffs, the Westinghouse team worked with Calvert Cliffs to design a solution to a number of common operational issues. Primarily, improving feedwater control over the entire operating range, eliminating single points of vulnerability (SPV) in the feedwater/feed pump systems and eliminating operational issues with the feed pump steam admission valve hydraulic control oil system. A robust first-of-a-kind design and implementation program was undertaken to split the existing hydraulic valve gear actuator with separate electric actuators. Westinghouse worked with Calvert Cliffs and Emerson Process Management (EPM) to install and commission General Electric (GE) feed pump turbines with separate electric actuators for the high-pressure (HP) and low-pressure (LP) steam admission valves. This successful mechanical and control modification was a first for both the nuclear and non-nuclear power generation industries. Westinghouse is working with a number of other utilities to implement similar modifications for PWR and BWR feed pump applications. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Assessment of Reactor BN-350 Impact on the Environment During Its Operation Period

Fast neutron reactor BN-350 was designed for dual purposes: steam production and plutonium generation. The steam produced in the reactor was used to generate electricity (150 MW) and desalination of seawater (up to 120 000 tons/day). The BN-350 reactor began operating in July 1973 with a design lifetime of 20 years. After the end of the reactor design lifetime in July 1993, its further operation was managed on the basis of annual decisions taken to extend the operating life for one year. The reactor was in downtime beginning in December 1998 and on 22 April 1999 the RK Government Resolution no. 456 'About decommissioning of the BN-350 reactor in Aktau, Mangistau region' was adopted. Thus, the total actual reactor operational period was more than 25 years. In the framework of work under the approved 'Plan of priority measures for decommissioning of the BN-350 reactor in Aktau', the specialists of the Institute of Nuclear Physics (INP) and the Mangystau Nuclear Power Complex performed a radiation survey of the plant area, the sanitary protection zone and reactor BN-350 monitoring area. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Study of Reactor BN-350 Wastewater Impact on the Environment

The fast neutron reactor BN-350 had a dual purpose: steam production and plutonium accumulation. The steam produced in the reactor was used to generate electricity (up to 150 MW) and desalination of seawater (up to 120,000 tons/day). The BN-350 reactor became operational in July 1973; its design operation lifetime was 20 years. After expiration of the reactor design lifetime in July 1993, its subsequent operation was continued according to the annual decisions on extension of the reactor operation, by one year. The reactor has been idle since December 1998. On April 22, 1999 Resolution No. 456 'About decommissioning of the BN-350 reactor in Aktau, Mangistau region' was adopted by the RK Government. Thus, the total operation period of the reactor was more than 25 years. One of the important issues related to the long-term operation of a radiation-hazardous facility, such as the BN-350 reactor, is to assess its impact on the environment. The necessity and urgency of detailed studies of the radiation situation is explained by the location of the Caspian Sea and Aktau city in the area of the BN-350 reactor's potential impact. This paper describes the preliminary results of the evaluation of the BN-350 reactor's effect on the level of radionuclide contamination of surface water, including the Caspian Sea via the mechanism of radionuclide discharge in wastewater. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Estimation of Core Recovery Time with TRACE

Following the severe accident in Fukushima Daiichi Nuclear Power Plant in Japan, safety related concerns about the zirconium based claddings used in current light water reactors have increased. High corrosion resistance and low neutron absorption cross section make zirconium a suitable cladding material for normal nuclear plant operational conditions. However, for nuclear accident scenarios when safety functions of the plant are lost, Zr claddings undergo exothermic reaction with steam and exacerbate the accident conditions. The exothermic reaction increases the temperature of the cladding and accelerates core melting in severe accident progression. And the generation of explosive hydrogen gas accumulates and challenges the structural integrity. The accident tolerant fuel (ATF) concept is the fuel and cladding system that gives a longer time margin, compared to standard Zr claddings, to cope with the accident when core cooling systems are lost in a nuclear power plant. The ATF cladding should have slower oxidation kinetics and less hydrogen generation than Zr to alleviate consequences of cladding interaction with steam at high temperatures. The maximum time available to mitigate progression of severe accident is an important measure to evaluate benefits of ATF claddings. For this work, the maximum time available to mitigate progression of severe accident is selected as the time at which cladding melting reaction starts. Melting reaction is selected as an upper limit because of the degradation of coolable geometry. Maintaining the coolable geometry is a regulation criteria for both loss of coolant (LOCA) analyses and outside the LOCA analyses. Once the melting of the cladding starts, it exacerbates the deformation of the fuel assembly and hottest rods within the core cannot be cooled by water reflood and results in fuel melting progression. The aim of this work is to evaluate time margin gain of ATF cladding in a severe accident scenario of LOCA without safety injections and station blackout considering oxidation kinetics only. Also it is assumed that safety injections are recovered before melting of the claddings start and the sensitivity of safety injection actuation time on core recovery is analyzed. The core recovery time is the time when makeup water from safety injections quench the fuel rods and halt the progression of the severe accident. This time is estimated by initiating safety injections at different times and observing the hot rod behavior which represents 1/5. of the core in this work. It is considered that core is not recoverable when 20% of the core is damaged.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Data-driven Quasi-static Surrogate Models for Nuclear-powered Integrated Energy Systems

The integration of nuclear power into energy systems presents a promising avenue to address the growing global energy demands while minimizing greenhouse gas emissions. In this paper, we introduce a data-driven quasi-static surrogate model for nuclear-powered Integrated Energy Systems (IES) that comprises various components, including a small modular reactor (SMR), steam manifold, balance of plant (BOP), high-temperature steam electrolysis (HTSE), and district heating (DH) system. Traditional physics-based models for these components often entail significant computational resource and time consumption, necessitating the development of efficient surrogate models. The development of a complete surrogate model for the IES involves the creation of individual surrogate models for each component, leveraging machine learning techniques and simulated data. These isolated surrogate models are subsequently integrated, enabling a holistic view of the IES and reducing the computational burden associated with detailed physics-based simulations. This paper outlines the development process, validation, and the performance evaluation of the surrogate models. The exceptional performance, with low root-mean-squared errors and R-squared scores of at least 99.8% across all individual surrogate models, underscores their accuracy and practical applicability. These results demonstrate the potential of these models to expedite the analysis of nuclear-powered IES, offering insights that can shape future research and development efforts.

08 HYDROGEN↗
DOE OSTI2023

Thermal Extraction and Delivery from Advanced Reactor Systems

This document presents preliminary thermodynamic analysis of thermal extraction and delivery from advanced nuclear reactors. Light-water, high temperature gas, sodium fast, and molten salt reactors are templated and analyzed for thermal extraction from nominal steam production conditions. Uncontrolled, controlled, and partial admission steam extraction methods are described and analyzed to show the differences in system impact from the various methods. Thermal delivery modeling results show comparisons that have become possible through the computational capabilities of the thermal delivery model. Due to the significant degrees of freedom of the problem, specific cases are used to compare the thermodynamic costs of heat transport using various heat transfer fluids. All this work is done using Modelica models, allowing for the future application of models in dynamic analyses beyond the steady-state calculations presented in this report. Appendices describe additional capabilities and methods that can help users analyze complex problems using HYBRID models developed in Modelica.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Developing a Novel Ultrafine Coal Dewatering Process

Dewatering fine coal is needed in many applications but has remained a great challenge. The hydrophobic-hydrophilic separation (HHS) method is a powerful technology to address this problem. However, organic solvents in solvent-coal slurries produced during HHS must be recovered for the method to be economically viable. Here, the experimental studies of recovering solvents from pentane-coal and hexane-coal slurries by combining liquid-solid filtration and in-situ vaporization and removing the solvent by a carrier gas (i.e., drying) are reported. The filtration behaviors are studied under different solid mass loading and filtration pressure. It is shown that using pressure filtration driven by 20 psig nitrogen, over 95% of solvents by mass in the slurries can be recovered, and filtration cakes can be formed in 60 s. The drying behavior was studied using nitrogen and steam at different temperatures and pressures. It is shown that residual solvents in filtration cakes can be reduced below 1400 ppm within 10 s by 15 psig steam superheated to 150 degree C, while other parameter combinations are far less effective in removing solvents. Physical processes involved in drying and the structure of solvent-laden filtration cakes are analyzed in light of these results.

01 COAL, LIGNITE, AND PEAT↗
DOE OSTI2024

Optimal operations of a nuclear-based integrated energy system: A mixed integer program approach

A nuclear-based integrated energy system (IES), consisting of multiple carbon-free energy generation and conversion technologies to meet heterogeneous end-use demands, offers a promising approach to decarbonizing the U.S. economy. Operating such an IES is challenging due to its complexity and the diverse end-use demands, such as heating and electricity. This paper aims to address the optimal operation of an IES composed of a small modular reactor (SMR), a steam manifold, a balance of plant (BOP), a high-temperature steam electrolysis (HTSE) system, a district heating (DH) network, and electrical grids. We formulate the system’s operation as a mixed integer linear programming (MILP) problem to maximize net revenues from sales of electricity and hydrogen. To evaluate the efficacy of the proposed model, we conduct a 24-hour simulation considering day-ahead (DA) electricity prices from CAISO and a varying DH demand profile with hourly resolution. The simulation results show that our model effectively optimizes the operation by selling electricity during on-peak periods and purchasing electricity for hydrogen production during off-peak periods, while satisfying operating constraints within the IES.

08 HYDROGEN↗
DOE OSTI2024

Site Integration and Regulatory Considerations for a Nuclear Power Plant Colocated with Industrial Facilities: Colocation Studies for a Petroleum Refinery, Methanol Plant, and Wood Pulp Plant

This research explores the colocation of nuclear power plants (NPPs) with industrial applications. Three existing industrial sites were considered to demonstrate the siting process and illuminate technological gaps for future work. The three applications demonstrated for colocation here are a petroleum refinery, a methanol production plant, and a pulp and paper plant. This study uses a modified version of the EPRI siting criteria to explore the geological and demographic characteristics of the location of the current industrial site, as well as exploring external hazards from the industrial plant and its surrounding land use. Data was collected from public databases to estimate site characteristics. We then discuss how the site characteristics may impact the ability to colocate an NPP with an industrial application. The application site and 5 additional sites were explored for each application to give a general indication of the siting implications for an NPP in each area. The hazards for each industrial application was also explored to determine how colocation may impact reactor safety. The following gaps have been identified and should be explored in future research on colocation of NPPs with petroleum refineries, methanol plants, and pulp and paper plants: - There is a variety of industrial use, hazards, and pipelines in the surrounding area. A more thorough review of these hazards should be considered for colocation. - In general, the whole region around some applications seems to have softer soil, with implications for large site preparation costs. Further site investigations should prioritize looking into the geotechnical conditions. - Applications along coastlines are susceptible to flooding and hurricanes. The benefits of colocation should be weighed against the potential design implications. - The benefits of natural gas pipeline infrastructure in place should be explored further. If heat supply from the NPP is not required or not feasible due to the distance between the NPP and the application, there may be an opportunity to supply hydrogen to the plant through an existing pipeline. - Because there are several collocated industrial plants in the regions for the refinery and methanol plant, the benefits of sharing resources from the NPP should be explored further. This may open up additional sites for colocation. The following knowledge gaps were identified for the colocation of NPPs with these three industries, and industrial applications in general. These gaps are: - While the STAND tool contains many important characteristics for the reactor siting process, it is not calibrated for the colocation of NPPs with industrial facilities. - There are aspects of both the NPP and industrial application that need to be quantified for a siting analysis. Particularly, we need to understand the water intake requirements for NPPs and each application. - Further work may focus on adapting the STAND site comparison methodology to comparison of sites for co-location. This will involve using the data documented in this report as a starting point and performing a comprehensive and quantitative comparison. - Without spending significant resources, it would be impossible to gather data for each site to evaluate all aspects of siting. One approach to finding data and understanding its implications to siting is looking at FSARs for existing plants. For example, most sites considered in this study have small Vs30 values, indicating soft soil. However, there are NPPs located in the vicinity of most of the sites (e.g., Waterford Steam Electric Station near New Orleans) and reviewing available site characteristics and geotechnical data for these NPPs, might provide further information for siting. - The siting analysis in this study indicates that colocation of the NPP with the industrial site could be difficult based on external hazards, cooling requirements, weather, or population. We need to determine the impact of distance between the two facilities on cost and quality of energy transport. - This study did not touch on socioeconomic impacts for NPP colocation with industrial facilities. The input-output analysis methodology could be applied to the communities referenced in this study to determine the socioeconomic impact of these projects. - Similarly, the impacts of colocation on emergency planning was not explored in this study. The impacts on emergency planning infrastructure are somewhat related to the socioeconomic impacts, and could be explored using a similar methodology. - This study also did not address physical and cybersecurity, which will be important aspects of co-location [ref] . Cybersecurity will be important, regardless of the distance, but physical security will be important if the facilities are located very closely. Physical security might also be important for the steam lines between the plants, unless they are determined to be non-safety significant. - In many site l

08 HYDROGEN↗
DOE OSTI2023

Scaling for Nuclear Reactor System: Overview

A scaling analysis is essential for predicting the behavior of a reactor system under various operating conditions, as well as for ensuring safe and reliable experimentation and adequate dataset generation for supporting the system code model development and assessment in accordance with USNRC guidelines 1.203. Testing and modeling can be carried out at various scales, from small test units to full-scale reactors, to validate the predictions produced by the analysis. However, due to the cost and challenges associated with full-scale testing, the nuclear industry uses scaled test facilities—integral effect test (IET) and separate effect test (SET)—to identify potential design issues and optimize the design to ensure regulatory compliance for safer operation and anticipated accident management. In general, IET facilities are non-nuclear thermal hydraulics test facilities. In contrast, SET facilities can be non-nuclear testing (e.g., steam generator, containment, pump system SETs) and nuclear testing (mostly fuel-material irradiation testing). IETs are used to study the system-level thermal-hydraulic behavior of the reactor coolant system (RCS) under different anticipated operational occurrences (AOOs) and design basis accidents (DBAs) such as loss-of-coolant-accident (LOCA) and loss-of-power (i.e., station blackout) and loss-of-flow (i.e., failure of RCS pump). Several LOCA scenarios could happen in reactor DBAs, such as main steam line break (MSLB), SG tube rapture (SGTR), break in direct vessel injection (DVI), and other small breaks LOCA. Scaling analysis determines the appropriate size and operating conditions for the test facility. This involves analyzing the system's thermal-hydraulic behavior at different scales and identifying the key parameters that affect the thermal-hydraulic behavior of the reactor. These key parameters can be identified so that the scaled facility can replicate the expected conditions in a prototypical setting. The scaled facility can simulate the same flow rates, pressure gradients, heat transfer coefficients, and fluid properties as the prototype. Scaling analysis is used to ensure that the results from these tests can be scaled up to predict the behavior of a full-scale prototype with an acceptable range of scaling distortions.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Modeling Advanced Strippers for CO2 Capture from Gas-Fired Power Plants using Aqueous Piperazine

Energy performance of 5 m piperazine was evaluated with an advanced stripper at the pilot scale, accounting for heat loss. Modeling studies indicated that heat loss at locations other than the heat source impacted the heat duty. For strippers with excess packing, the column was the most important source of heat loss, and values as low as 0.03 GJ/hr can cause pinch and reabsorption of CO2. Solvents like 5 m PZ are more affected by heat loss in the column due to a top-side temperature pinch at high lean loading. Heat loss impacts NGCC CO2 capture more than coal-based CO2 capture. Surface temperature measurements pilot plants showed that heat loss was 50-75% controlled by natural convection. Measured heat loss was correlated with steam flowrate and wind speed and ranged from 8126 to 219668 Btu/hr, representing 35% of the measured heat rate on average. Net heat duty linearly varied with CO2 removal at low lean loading and was similar for coal and NGCC conditions at fixed removal. At 90% removal, net heat duty was about 2.5 GJ/t with 4 and 12% CO2 in the flue gas. 7 At NGCC conditions, strippers with finite packing can benefit from a flashing feed to the top of the column by using a hot bypass. A flashing feed was linked to a reduction in irreversibility of the stripper from temperature driving forces. It also reduces heat duty by up to 6% at fixed packing height and reduces packing requirement for a fixed steam heater size compared to a warm bypass. Lean loading and lean solvent rate were effective handles to maximize profitability of a fixed stripper design at low gas price. High ambient temperature operation can benefit from low pressure stripping to about 0.18 lean loading at 150 ℃ to maintain the cyclic capacity at reduced rich loading. At low gas price, the capture plant was able to maximize profitability even with a high heat duty of 2.7-3.1 GJ/t at a lean loading of 0.2-0.22 mol/mol. Heat recovery by partial water vapor condensation in a gas-liquid exchanger can be replaced by a direct contact condenser (DCC). The DCC improved gas cooling and reduced heat duty compared to the CO2 exchanger at 0.2 lean loading. The DCC when used with a 150 ℃/5.5 bar stripper can reduce the cost of capture compared to the base case by $3- 4/tonne at 15 ft of packing (optimum) but increases cost of capture by $6/tonne with a 120 ℃/2 bar stripper and 10 ft of packing (optimum). The DCC can work effectively with a solvent with a high heat of absorption and thermal stability.

Suresh Babu, Athreya↗
DOE OSTI2024

Technology Feasibility of High-Temperature Heat Pumps for Industrial Drying Process in the US

High-temperature heat pumps are a prospective technology for electrifying and decarbonizing industrial drying processes. This study investigates the technological potential of HTHPs in replacing fossil-fuel burners or steam boilers in industrial spray dryers. The performance of HTHPs with three configurations was investigated, including high high-temperature heat pump boosted with an electric air heater, a steam-generating high-temperature heat pump with multi-stage mechanical vapor compressions, and transcritical high-temperature heat pumps. R1336mzz(Z) is used as the refrigerant in both subcritical and transcritical high-temperature heat pumps, and R717 is used as the working fluid in the mechanical vapor recompression and the heat transfer fluids. Thermodynamic models are built to evaluate the energy efficiency of HTHPs. A case study was carried out on an industrial spray dryer integrated with a waste heat recuperator, where the HTHPs are used to recover precooled air with a dew point of 40℃ and supply heat for drying air at 200℃. For a high-temperature heat pump boosted with an electric air heater, a higher system-level COP is achieved with a high supply temperature from high-temperature heat pump. For a high-temperature heat pump with mechanical vapor recompressions, larger temperature drops in the flashing tank provide higher heat capacities but almost constant values of the coefficient of performance. For a transcritical high-temperature heat pump, an optimum discharge pressure exists for the temperature profiles within the gas cooler and the coefficient of performance. The transcritical high-temperature heat pump provides the best energy efficiency among the three configurations. The theoretical results prove the technical feasibility of high-temperature heat pumps in industrial air-drying applications. .

Wang, Pengtao↗
DOE OSTI1989

Nuclear Safety [Vol. 30, No. 3, July-September 1989]

Nuclear Safety is a review journal that covers significant developments in the field of nuclear safety. Its scope includes the analysis and control of hazards associated with nuclear energy, operations involving fissionable materials, and the products of nuclear fission and their effects on the environment. Primary emphasis is on safety in reactor design, construction, and operation; however, the safety aspects of the entire fuel cycle, including fuel fabrication, spent-fuel processing, nuclear waste disposal, handling of radioisotopes, and environmental effects of these operations, are also treated. Table of Contents for this issue follows. GENERAL SAFETY CONSIDERATIONS: 325 Safety of Framatome Advanced Nuclear Steam Supply Systems Designs by J. A. Charles and D. Lange, 333 Book Review of Nuclear Accidents: Intervention Levels for the Protection of the Public by H. B. Piper; ACCIDENT ANALYSIS: 335 Living PRA Computer Systems by S. C. Dinsmore and H.-P. Balfanz, 343 Summary of ICAP Assessments of RELAP5/MOD2 by W. E. Driskell and R. G. Hanson; CONTROL AND INSTRUMENTATION: 352 Thermal Performance Monitoring System at Maanshan Nuclear Power Plant by H.-J. Chao, Y.-P. Lin, G.-H. Jou, L.-Y. Liao, and Y.-B. Chen; DESIGN FEATURES: 358 Warning Systems for Nuclear Power Plant Emergencies by J. H. Sorensen and D. S. Mileti; WASTE AND SPENT FUEL MANAGEMENT: 371 Activities Related to Waste Management Compiled by E G. Silver; OPERATING EXPERIENCES: 382 Steam Generator Tube Performance: Experience with Water-Cooled Nuclear Power Reactors During 1985 by O. S. Tatone and R. L. Tapping, 400 Systems Interaction Analyses: Concepts and Techniques (Part II) by M. D. Muhlheim and G. A. Murphy, 413 Reactor Shutdown Experience Compiled by J. W. Cletcher, 416 Operating U.S. Power Reactors Compiled by E G. Silver; RECENT DEVELOPMENTS: 440 General Administrative Activities Compiled by E G. Silver, 460 Reports, Standards, and Safety Guides by D. S. Queener, 466 Status of Power-Reactor Licensing Activities Compiled by E G. Silver, 470 Proposed Rule Changes as of Mar. 31, 1989; ANNOUNCEMENTS: 334 Proceedings Published, 351 CEC Seminar on Methods and Codes for Assessing the Off-Site Consequences of Nuclear Accidents, 357 Short Course on Multiphase Flow and Heat Transfer: Bases and Applications in A: The Nuclear Power Industry B: The Process Industries, 357 International Conference on Probabilistic Safety Assessment and Management, 478 International Topical Meeting on the Safety, Status, and Future of Non-Commercial Reactors and Irradiation Facilities, 475 The Authors.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Synergistic heat pumped thermal storage and flexible carbon capture system

A power plant system is disclosed. The power plant system includes a combustor configured, a turbine configured to generate electricity, a heat exchanger and a steam turbine, a carbon capture system configured to remove at least a portion of carbon-based gasses from the flue gas downstream from the heat recovery steam generator, and a thermal storage system including a hot storage unit configured to store thermal energy at a hot temperature, the hot temperature greater than ambient temperature. The power plant is configured to operate in at least a first mode for storing thermal energy in the thermal storage system and a second mode for releasing the stored thermal energy from the thermal storage system and during the second mode, heat stored in the hot storage unit is transferred to the carbon capture system.

Bandhauer, Todd M.↗
DOE OSTI2025

A Rapid Procedure Development Approach for Multi-stage Validation of Novel Nuclear Power Plant Systems

The integration of renewable energy sources into the modern grid introduces variability in generation, challenging nuclear reactors to remain economically viable amid low wholesale electricity prices. To address this, Idaho National Laboratory?s Flexible Plant Operation and Generation program explores alternative revenue streams, including commoditizing excess thermal energy. A Thermal Power Dispatch (TPD) system enables steam extraction from the secondary loop for industrial uses like hydrogen production via high-temperature steam electrolysis.

99 - GENERAL AND MISCELLANEOUS↗