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All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Preliminary Results from Electric Arc Furnace Off-Gas Enthalpy Modeling

This article describes electric arc furnace (EAF) off-gas enthalpy models developed at Oak Ridge National Laboratory (ORNL) to calculate overall heat availability (sensible and chemical enthalpy) and recoverable heat values (steam or power generation potential) for existing EAF operations and to test ORNL s new EAF waste heat recovery (WHR) concepts. ORNL s new EAF WHR concepts are: Regenerative Drop-out Box System and Fluidized Bed System. The two EAF off-gas enthalpy models described in this paper are: 1.Overall Waste Heat Recovery Model that calculates total heat availability in off-gases of existing EAF operations 2.Regenerative Drop-out Box System Model in which hot EAF off-gases alternately pass through one of two refractory heat sinks that store heat and then transfer it to another gaseous medium These models calculate the sensible and chemical enthalpy of EAF off-gases based on the off-gas chemical composition, temperature, and mass flow rate during tap to tap time, and variations in those parameters in terms of actual values over time. The models provide heat transfer analysis for the aforementioned concepts to confirm the overall system and major component sizing (preliminary) to assess the practicality of the systems. Real-time EAF off-gas composition (e.g., CO, CO2, H2, and H2O), volume flow, and temperature data from one EAF operation was used to test the validity and accuracy of the modeling work. The EAF off-gas data was used to calculate the sensible and chemical enthalpy of the EAF off-gases to generate steam and power. The article provides detailed results from the modeling work that are important to the success of ORNL s EAF WHR project. The EAF WHR project aims to develop and test new concepts and materials that allow cost-effective recovery of sensible and chemical heat from high-temperature gases discharged from EAFs.

Electric Arc Furnace↗
DOE OSTI2014

Oxycombustion in transport oxy-combustor

A pressurized transport oxy-combustor with different configurations is disclosed. Substantially pure oxygen is fed to the transport oxy-combustor under pressure to combust fossil fuels, generating steam for power generation. The end product is the flue gas containing substantially pure CO2 after moisture condensation. The low excess oxygen necessary to achieve complete combustion in the combustor is scavenged by adding another fuel so that substantially all oxygen fed to the combustor is completely consumed. The capability to operate the transport oxy-combustor as a circulating fluidized bed combustor at very high solids circulation rates makes it unnecessary to use recycled CO2 or flue gas as a means to moderate and control the combustion temperature. The temperature in the combustor is effectively controlled by relatively cooler circulating solids that enter the combustion zone. A small amount of CO2 is recycled for aeration and to convey solids fuel to the combustor.

Vimalchand, Pannalal↗
DOE OSTI2016

CFD Predictions of Severe Accident Natural Circulation Flows in a Combustion Engineering PWR

Computational fluid dynamics (CFD) is used to predict natural circulation flows between a reactor vessel upper plenum and a steam generator during specific severe accidents in a pressurized-water reactor (PWR). The predictions are based on a model for a simplified Combustion Engineering (CE) design. This work is part of a larger study to investigate the potential for severe accident induced steam generator (SG) tube failures in PWRs being conducted by the U.S. Nuclear Regulatory Commission Office of Nuclear Regulatory Research. The CFD model builds upon the lessons learned from previous studies of a generic Westinghouse PWR under similar conditions. The CE SG is significantly larger than the Westinghouse design and geometric differences are expected to reduce the mixing and increase the potential thermal challenge to the SG tubes. The CFD predictions are utilized to improve the basis for simplified flow and mixing models applied in one-dimensional severe accident codes. Parameters are determined to support system code modeling of hot leg and SG inlet plenum mixing, hot leg and SG tube bundle flow rates, and tube inlet temperatures. This updated modeling uses boundary conditions obtained from system analysis code predictions of severe accident behavior in a CE PWR. Model refinements include an improved tube bundle representation and improvements to the thermal boundary conditions. Results indicate less mixing and higher tube inlet temperatures for the CE design compared to previous Westinghouse studies. (author)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Solar Steam on Demand

INTRODUCTION: This Report summarizes the research and development project performed by Sunvapor and the subrecipient National Renewable Energy Laboratory (NREL) during the period of 10/01/2018-9/30/2021. The work was aimed at the economic integration of solar steam with an industrial process, and in particular, the advantages that a novel type of thermal energy storage and collector design could bring to the system. The completed work includes experimental research on the storage material, design of the collector, engineering of a complete solar steam plant, and economic analysis. The plant was, in the end, not constructed, due to COVID-19 impacts to the project host. PURPOSE: The purpose of the project was to test the hypothesis that a levelized cost of heat (LCOH) of 2¢/kWh delivered to a steam-consuming process, including energy discharged from storage, could be feasibly achieved. PROJECT OBJECTIVES: For Budget Period 1 our objectives were to engineer a prototype industrial solar steam generation system, and design and test a Latent Heat Energy Storage (LHES) lab-scale system. For Budget Period 2 our objectives were to construct and test the solar steam prototype plant, and to build and factory-test a full LHES module. For Budget Period 3 our objective was to integrate the storage module with the solar steam generator, show successful long duration energy tests of the combined system, and use the results to project the LCOH of delivered steam to a second-generation plant. PROJECT OUTCOMES: Various phase change materials (PCMs) for the LHES were tested for compatibility with industrial steam uses as measured by phase change temperature, stability of latent heat of fusion, and their corrosive effects on steel (the heat exchanger material). Two of these materials (sodium formate and sodium/ potassium) showed potential to meet the technical criteria. The estimated cost of a complete LHES module with these materials did not meet economic targets. The developmental collector design builds from previous work to exploit the potential cost advantages of a spaceframe of lumber construction. Design improvements were achieved in assembly efficiency and structural performance. A complete engineering package was completed for the solar steam system without storage that included the developmental as well as commercial arrays. The plant was permitted for construction. The storage CAPEX for a reference plant was estimate to be greater than $\$ 57$/kWh. The LCOH of the system without storage and with developmental collectors of lumber construction was projected to be 2¢/kWh, assuming a FCR of 8.2% over thirty years. PROJECT MILESTONES: The Milestones fully achieved in the first Budget Period includes obtaining a Letter of Intent from an industrial steam-consuming project host, issuance of a construction permit, and corrosion resistance. The Milestone that was partly achieved was the degradation in the latent heat of fusion. Milestones associated with Budget Periods 2 and 3, related to the construction and operation of the plant were not achieved as a result of the host shutting down the plant due to COVID-19 impacts. CONCLUSIONS: Sunvapor was able to secure an agreement with a host to build a fully engineered and permitted solar steam plant that met the host’s economic goals. As a basis for comparison, the solar field design was comprised of eight commercial collector arrays and one developmental Green Parabolic Trough Collector (GPTC) array. The projected LCOH with a solar field entirely populated with a second generation GPTC met the goal of 2¢/kWh. The experimental program concerning the PCM indicated the feasibility of meeting their technical requirements for industrial steam uses. The cost of the complete LHES exceeded its $\$ 14$/kWh target, and therefore the inclusion of LHES could not be economically justified. Due to COVID-19 impacts on the host, the solar steam plant was not constructed.

14 SOLAR ENERGY↗
DOE OSTI2025

Cost analysis of alternative large-scale high-temperature solid oxide electrolysis hydrogen production facilities

We extend our past cost analysis of gigawatt-scale solid oxide electrolysis (SOE) facilities that produce high purity hydrogen gas from water by estimating construction and operating costs for three new alternative design cases: (1) offsite feed steam generation; (2) near-atmospheric pressure (NAP) stack; and (3) onsite electric boiler feed steam generation. Pressure effects on hydrogen electrode-(cathode-)supported SOE cell (SOEC) stack performance are estimated for the same assumed cell and stack construction and used to determine facility-wide stack capital costs for achieving a fixed H2 production at different pressures. Modular balance of plant (BOP) process equipment capital costs are estimated for each new alternative design case using our past equipment sizing, design, and cost data and scaling relationships. Furthermore, we update BOP equipment sizing and design for the NAP case using Aspen®. Vendor quotes for electric boilers are used to estimate costs for the electric boiler design case. Factory and onsite assembly and installation costs for SOEC stacks and BOP equipment are calculated using our past simplified first-principles approach. First-of-a-kind (FOAK) and N th -of-a-kind (NOAK) production maturity cost estimates are included for all results. The case with NAP stacks offers the lowest facility total capital cost (TCC, ~23% lower than base) while use of small electric boilers requires the highest TCC (~3% higher than base). H 2 production prices decrease from the base of ~$\$2.17$ /kgH 2 to ~$\$1.92$/kgH 2 for 1 GW e DC SIP facilities utilizing NAP stacks supplied by offsites steam situated in large modules and blocks for $\$0.030$/kWh e and $0.009/kWh t prices for electricity and thermal energy, respectively. We report all costs in 2021 US dollars.

Balance of plant (BOP) process equipment↗
DOE OSTI2017

Visualization of flow interaction in a channel flow produced by a pair of helically coiled rods

Helically coiled tubes have been considered in the design of steam generators or heat exchangers of next generation nuclear power plants. Its compactness, efficiency, and the capability to absorb thermal expansion are desirable for nuclear reactor designs. However, the complex flow generated by the helically coiled geometry can result in flow induced vibration (FIV). In order to predict the FIV, there have been efforts using Computational Fluid dynamics (CFD)4, 5 to be coupled with multi-mechanics simulations. However, it is required to validate CFD to have confidence in capturing the flow structure characteristics of the transient complex flow behavior and fluctuation. In the present study, a simplified helically coiled steam generator (HCSG) design experiment was built acquiring data for CFD validation purpose. It is not perfect representation of the multiple rods, however, the current test facility produced transient data of the flow interaction between the rods and flow fluctuations near the rods. Particle Image Velocimetry (PIV) was applied to visualize the turbulent flow in a channel with slant rods. The two dimensional turbulent flow fields were analyzed. The velocity components were defined as below. u = U + u' (1) v = V + v' (2) where u is the lateral velocity on x-axis, v is the vertical velocity on y-axis, U and V are the mean velocities, and u' and v' are the fluctuation. Reynolds number, Re, was calculated using the rod diameter, d = 15.9 mm, and the upstream approach velocity, V{sub 0} = 0.5 m/s. Re = V{sub 0} d/η (3) where η is the kinematic viscosity of water. The experiment was conducted at Re ≅ 9,000 considering the currently reasonable computing power. Flow visualization in a flow channel produced by helically coiled rods was performed using PIV. The mean velocity field and streamlines were presented in this paper. The vortical flow structures in the cavity between rods were observed. Because of the asymmetric geometry, the flow profile was skewed. Additional measurements and transient analysis will be conducted to produce benchmark data for FIV study using CFD. Also, a test facility with a 5-rod heat exchanger design is operating to produce more data as a further work.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

All Nuclear Hybrid Superheater Design Method

The 'All Nuclear Hybrid' (ANH) concept combines Light Water (LWR) and Generation IV (Gen IV) reactor technologies to produce superheated steam. LWR technology characteristically produces steam temperatures of ∼300 C. Process steam demands such as refineries and hydrogen production can require steam temperatures that exceed 600 C. Therefore, an additional heat source must be added to provide the heat to increase or 'boost' the temperature of the steam generated from by the LWR. A recent study demonstrated that application of the ANH concept to a refinery optimized the use of nuclear technologies and maximized carbon dioxide emission reductions. The objective of the present work is to size the heat exchangers (Superheaters) required to apply the ANH concept to a Westinghouse AP1000 LWR to produce 675 C superheated steam. Two GenIV reactor concepts are considered to provide the additional heat source and high primary coolant temperatures required to produce the superheated steam: a High temperature Gas Reactor (HTGR) using helium as primary coolant, and a Molten Salt reactor. The details of the GenIV reactor designs are not the focus of this study; however, the primary coolant thermal conditions from these concepts are utilized directly in the superheater sizing calculations. Also, a tube and shell superheater design is assumed for the superheater due to its extensive use in the nuclear and other industries. The ANH concept generates superheated steam by first utilizing LWR technology to accomplish the phase change from liquid water to steam within the steam generator. This is a well-established technology with several decades of successful safe operation. Next, the concept utilizes GenIV reactor technology to superheat the steam. Since both the working fluids in the superheater are single phase, the heat transfer and pressure drop calculations required to size the superheater can be performed without complex two-phase computer analyses (i.e., RELAP5). This study presents an approach for the sizing the ANH superheater using an iterative LMTD heat exchanger design method. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Power Generation Cycle with RELAP-7

Balance of plant of a High-Temperature gas-cooled Reactor – Pebble bed Module (HTR-PM) is modeled using RELAP-7. The model includes the primary helium loop where the heat produced by the core is transferred to a helical-coil steam generator and the secondary loop where super-heated steam is fed to a turbine to extract power. The steam is then condensed to subcooled water and pumped back into the steam generator. Results demonstrate that RELAP-7 has the capability to model different regimes encountered in two-phase flow with wall boiling, superheated steam, subcooled liquid and condensation. RELAP-7 also has the capability of modeling different fluids with their own specific models in a very flexible way, without any code change. This also demonstrates the capability of modeling the transfer of fission power into electrical power.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS1974

Energy recovery from solid waste

A recent group study on the problem of solid waste disposal provided a decision making model for a community to use in determining the future for its solid waste. The model is a combination of the following factors: technology, legal, social, political, economic and environmental. An assessment of local or community needs determines what form of energy recovery is desirable. A market for low pressure steam or hot water would direct a community to recover energy from solid waste by incineration to generate steam. A fuel gas could be produced by a process known as pyrolysis if there is a local market for a low heating value gaseous fuel. Solid waste can also be used directly as a fuel supplemental to coal in a steam generator. An evaluation of these various processes is made.

Dalton, C.↗
DOE OSTI2016

Development of TDLAS for Water Vapor Concentration Measurement during VHTR Steam Ingress

Very-high-temperature reactors (VHTR) utilize a graphite-moderated, helium-cooled design that provides higher efficiencies and a low-carbon source of process heat (>700 deg. C) for many industrial applications. Safety concerns have led to extensive accident condition investigations. Steam generator heat-exchanger tube rupture accidents (SGTR) can lead to increased reactor core steam density, which causes neutron slowing-down and chemical corrosion of the graphite fuel elements and structural columns. It is important to be able to computationally model the propagation of steam throughout the lower plenum and nuclear reactor core, but experimental data is needed to validate such models. The primary objective of this project is to instrument a scaled down experiment of a VHTR designed by Ohio State University (OSU), with tunable diode laser absorption spectroscopy (TDLAS) diagnostics to measure steam concentration changes and map how water vapor propagates through the test facility's lower plenum and core. The basic principle of absorption spectroscopy is based on a molecular species ability to absorb photons of specific energies. A collimated laser beam of known energy is passed through a test section containing a sample gas. When a transition of the absorbing gas species resonates with the laser frequency, a certain amount of energy will be absorbed based on the rovibrational characteristics of the molecule. The intensity of the transmitted laser light can be measured by a detector after exiting the test section. The Beer-Lambert Law of absorbance is used to recover this information relating beam attenuation to the system dimensions, gas composition, pressure, and temperature. The line-strength is a function of temperature alone and relates the fraction of molecules in a given rovibrational state through the partition function. The partition function describes the statistical properties of a system in thermodynamic equilibrium where the states obey a Boltzmann distribution. Thus, the number density of molecules in a particular state can be determined from the total internal partition function at a particular temperature. The lineshape function of a given absorption transition is derived from the physical interaction between the absorbing species and is dependent on temperature and pressure. Random thermal motion of absorbing molecules result in broadening of the linewidth, called Doppler broadening, which follows a Gaussian distribution. Molecular collisions result in broadening and shifting of the linewidth, called pressure broadening, which follows a Lorentzian distribution. The overall lineshape results in a Voigt profile, which is a convolution of the Gaussian and Lorentzian distributions. The Voigt profile does not have a simple analytical form and is most often calculated numerically or approximated by complex functions. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Solar Process Heat Potential in California, USA

The cost for solar industrial process heat (SIPH) is quantified by defining the levelized cost of heat (LCOH). The state of California, in the United States offers a favorable environment for SIPH given its excellent insolation, industrial gas prices typically higher than the national average, and policies promoting solarthermal deployment. Given historically low gas prices, competing with natural gas remains the primary challenge to deployment. However, this study finds that the solar LCOH from Concentrating Solar Power (CSP) collectors for many regions in California is expected to be lower than the LCOH from natural gas, using a representative installed solar hardware price and the average price for industrial natural gas in California. Lastly, modifications are in progress to the parabolic trough and Direct Steam Generation (DSG) trough and Linear Fresnel models within NREL's System Advisor Model (SAM) to allow users to more easily predict performance for these industrial heat and steam-generation applications.

14 SOLAR ENERGY↗
DOE OSTI2024

Leveraging Optimal Sparse Sensor Placement to Aggregate a Network of Digital Twins for Nuclear Subsystems

Nuclear power plants (NPPs) require continuous monitoring of various systems, structures, and components to ensure safe and efficient operations. The critical safety testing of new fuel compositions and the analysis of the effects of power transients on core temperatures can be achieved through modeling and simulations. They capture the dynamics of the physical phenomenon associated with failure modes and facilitate the creation of digital twins (DTs). Accurate reconstruction of fields of interest (e.g., temperature, pressure, velocity) from sensor measurements is crucial to establish a two-way communication between physical experiments and models. Sensor placement is highly constrained in most nuclear subsystems due to challenging operating conditions and inherent spatial limitations. This study develops optimized data-driven sensor placements for full-field reconstruction within reactor and steam generator subsystems of NPPs. Optimized constrained sensors reconstruct field of interest within a tri-structural isotropic (TRISO) fuel irradiation experiment, a lumped parameter model of a nuclear fuel test rod and a steam generator. The optimization procedure leverages reduced-order models of flow physics to provide a highly accurate full-field reconstruction of responses of interest, noise-induced uncertainty quantification and physically feasible sensor locations. Accurate sensor-based reconstructions establish a foundation for the digital twinning of subsystems, culminating in a comprehensive DT aggregate of an NPP.

42 ENGINEERING↗
DOE OSTI2017

Synthetic Graphene Oxide Leaf for Solar Desalination with Zero Liquid Discharge

Water vapor generation through sunlight harvesting and heat localization by carbon-based porous thin film materials holds great promise for sustainable, energy-efficient desalination and water treatment. However, the applicability of such materials in a high-salinity environment emphasizing zero-liquid-discharge brine disposal has not been studied. Here this paper reports the characterization and evaporation performance of a nature-inspired synthetic leaf made of graphene oxide (GO) thin film material, which exhibited broadband light absorption and excellent stability in high-salinity water. Under 0.82-sun illumination (825 W/m 2 ), a GO leaf floating on water generated steam at a rate of 1.1 L per m 2 per hour (LMH) with a light-to-vapor energy conversion efficiency of 54%, while a GO leaf lifted above water in a tree-like configuration generated steam at a rate of 2.0 LMH with an energy efficiency of 78%. The evaporation rate increased with increasing light intensity and decreased with increasing salinity. During a long-term evaporation experiment with a 15 wt % NaCl solution, the GO leaf demonstrated stable performance despite gradual and eventually severe accumulation of salt crystals on the leaf surface. Furthermore, the GO leaf can be easily restored to its pristine condition by simply scraping off salt crystals from its surface and rinsing with water. Therefore, the robust high performance and relatively low fabrication cost of the synthetic GO leaf could potentially unlock a new generation of desalination technology that can be entirely solar-powered and achieve zero liquid discharge.

42 ENGINEERING↗
DOE OSTI2019

Confinement effects on the solar thermal heating process of TiN nanoparticle solutions

We propose a theoretical approach to describe quantitatively the heating process in aqueous solutions of dispersed TiN nanoparticles under solar illumination. The temperature gradients of the solution with different concentrations of the TiN nanoparticles are calculated when confinement effects of the container on the solar absorption are taken into account. We find that the average penetration of solar radiation into the solution is significantly reduced upon increasing the nanoparticle concentration. At high concentrations, our numerical results show that photons are localized near the surface of the solution. Moreover, the heat energy balance equation at the vapor–liquid interface is used to describe the solar steam generation. The theoretical time dependence of temperature rise and vaporization weight losses is consistent with experiments. Our calculations give strong evidence that the substantially localized heating near the vapor–liquid interface is the main reason for the more efficient steam generation process by floating plasmonic membranes when compared to randomly dispersed nanoparticles. The validated theoretical model suggests that our approach can be applied towards new predictions and other experimental data descriptions.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2019

System and method for production of ultra-pure hydrogen from biomass

The disclosure provides a system and method for synthesizing ultra-pure hydrogen from biomass waste. The present invention comprises a gasifier, an oils and tars filter, a steam generator, a water gas shift reactor ("WGS"), a heat-exchange two-phase water separator, a scrubber, a hydrogen separator, and fluid conduits in fluid communication with the various system components, which together convert hydrocarbon-based biomass, e.g., woodchips, into ultra-pure hydrogen gas. Fluid conduits connect the gasifier and the steam generator, separately, to the WGS, the WGS to the two-phase separator, the two-phase separator to the scrubber, and the scrubber to the hydrogen separator, which further comprises an outlet port through which hydrogen gas may flow free of carbon monoxide. The hydrogen may flow to a device that utilizes hydrogen to generate energy, such as a hydrogen fuel cell or to an internal combustion engine.

Tawfik, Hazem↗
DOE OSTI2016

Overview of PGSFR Project

The long-term plan for advanced SFR development was authorized by Korean Atomic Energy Commission (KAEC) in 2008 and updated in 2011, which will be carried out toward the construction of a SFR prototype plant called PGSFR (Prototype Gen-IV Sodium-cooled Fast Reactor) by 2028. The primary purpose of PGSFR is to demonstrate the technology reducing radioactive waste of spent nuclear fuel from nuclear power plants (NPP) in Korea. There are 24 NPPs in operation nationwide covering 30% of domestic electricity demand. The spent nuclear fuel is currently stored in the on-site storage pool of NPPs. Then, it has become a critical issue to ensure technical options to handle spent nuclear fuel in the long term basis to secure the sustainable operation of NPPs. PGSFR is a pool type sodium-cooled fast reactor with the power of 150 MWe or 400 MWth. PGSFR uses metallic fuel to make use of inherent safety characteristics of metallic fuel. The spent nuclear fuel will be processed using pyro-process to produce U-TRU-Zr fuel for PGSFR. But the initial core shall be loaded by low enriched U-Zr fuels. PGSFR has two intermediate loops with two intermediate heat exchangers (IHX) for each loop to deliver the generated heat from the core to power conversion system. There a steam generators producing super-heated steam in each loop. There are four decay heat removal circuits; two passive and two active ones. The reactor building is built on a seismic isolation base mat which can significantly mitigate seismic vibration of major components in the reactor building. The reactor building has a containment structure to isolate the reactor building under specific accident conditions and also provide structural integrity at the time of aircraft impact accident. PGSFR project is the national development program of Korea to construct the first SFR in Korea by 2028. The primary goal of PGSFR is the demonstration of reduction of radioactive waste from spent nuclear fuel by incineration in PGSFR. SFRA will flexibly cope with the licensing environment to minimize the uncertainty in licensing and safety concerns. The successful construction of PGSFR will bring Korean nuclear industry a step closer to guarantee the sustainable operation of NPPs nationwide and finally most importantly will solve the spent nuclear fuel issue of Korea. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Economic solution for low carbon process heat: A horizontal, compact high temperature gas reactor

In 2018, nuclear energy generated 55% of United States’ and one third of the world’s carbon free electricity, making nuclear energy a key tool in efforts to mitigate climate change before 2050. However, the current nuclear technology, light water reactors (LWRs), is limited to 300°C, so it cannot be used to decarbonize industrial process heat which accounts for 12% of US greenhouse gas emissions. High temperature gas reactors (HTGRs) can meet the high temperature demand with carbon free nuclear heat. The estimated cost of HTGRs, such as the Next Generation Nuclear Plant (NGNP), are even higher than state-of-the-art LWRs. In this paper, we expanded our nuclear cost estimating tool to include HTGRs and find that the NGNP overnight capital costs were 32% higher than an advanced LWR per unit capacity. The higher cost will naturally result in larger risk to cost overrun as recently experienced by larger LWRs in western nations. With a design-to-build mindset to minimize cost and construction risk, we introduce the horizontal, compact HTGR (HC-HTGR). The reactor core and steam generator are mounted horizontally on rails and in-line with one another, decreasing the size of the reactor building relative to the power capacity four times when compared to traditional HTGRs. The HC-HTGR reduced overnight civil structure costs by 42%, indirect costs by 38%, and total capital costs by 20% from NGNP. We discussed the required engineering of new systems for the HC-HTGR including vessel supports, the reactor cavity cooling system, and steam generator design. Finally, we estimated the fuel and operations costs of the HC-HTGR, and a survey of low-carbon industrial process heat technology showed the HC-HTGR can deliver a highly competitive levelized cost of heat in the range of $6.13–12.48/GJ.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Assessment on Transport of Radioactive Species from Nuclear Facilities - 17079

It is generally recognized that radioactive nuclides are most hazardous fission products that could be released in the event of reactor accidents. The reactor coolant system of a pressurized water reactor (PWR) nuclear plant exhibits radioactivity from various sources. These radiation sources consist of the fission products from fuel element and activated products from corrosion products. The complex infrastructure of the processes does not allow detailed measurement of activity until after the reactor is shutdown. All of the components that require dismantling will be radioactive which was caused by neutron activation during operation, contamination by corrosion products, contamination by damaged by fuel rods and leaks by equipment. Due mainly to the various operation modes and their aging, there are the increasing needs to carry out repair works that are not sufficiently provided for by the original design. Many difficulties are involved in these tasks, namely, high radiation doses, cramped areas, hard to reach components, coupled with stringent quality conditions. The need to simulate the behavior of corrosion products in a nuclear power plant environment has attracted many researchers' attention. It has performed the in-depth reviews of the various theories and physics associated with the corrosion products transportation, the laboratory loop testing, and the field measurements in operating plants. Before we start the decommissioning of nuclear power plants, workers, the public and the environment shall be properly protected from radiation hazards. For this purpose, a safety assessment should be carried out relating to the potential hazards that may occur during the decommissioning. The radiological hazards involved in the decommissioning activities also should be identified, and be taken into account in the safety assessment. The radiation conditions of the primary circuit are determined by the activity of corrosion and fission products that have accumulated on the internal surfaces of the components. Therefore the estimation of contamination on the components is very important and, may be the first step in the reactor operation and decommissioning of components. The predominant radionuclide in the PWR primary coolant is Co-60, although Co-58 contributes significantly to radioactivity level at the early stage of a reactor operation. Since Co-60 is produced from the in-core Co-58, which is the source of cobalt in a reactor coolant system, many researchers have investigated its transmutation in the core, its transport around the system, and its subsequent deposition on surfaces. The major parameter influencing the behavior of radioactivity, such as Co-60 in the coolant, is the corrosion product oxides. The objective of this work is to analyze the corrosion product activity transported in the primary circuit by using a multi-region model. This model, which is incorporated with the mass and activity transport between the dissolved corrosion products in the coolant and the surface, describes the specific activity of corrosion products in coolant and on the surfaces according to the operation modes as well as decommissioning stage. In order to evaluate the analysis of the corrosion products in an integral type reactor, a multi-region mass balance model is applied. The specific activities of the various species on the surface of core and steam generator throughout the effective full power hours, 15,000 hours, are simulated for the various modes of operations. As a result, the specific activities of corrosion products from the surface of primary circuit were shown as very low compared to those of other commercial nuclear power plants. In this simulated reactor, a significant reduction of specific activities from the activated corrosion products is obtained due to the introduction of titanium alloy to the steam generator tube and the avoidance of boric acid induced corrosion. From this result we conclude that we are able to predict the level of radioactive nuclides generated by corrosion products in a primary circuit according to the various modes of operations by using multi-region model. In this point of view it is concluded that this multi-region model can be used as an engineering tool for analysis of the corrosion products within the primary coolant for the nuclear power plants. It is also considered that it is possible to extend to a calculation tool for estimating the radiological hazards involved in the decommissioning activities in terms of safety assessment. (authors)

07 ISOTOPE AND RADIATION SOURCES↗