Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “SMALL MODULE REACTOR”

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.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 91 records · Page 5
All sourcesNASA NTRSDOE OSTI
NASA NTRS2007

Design of a Resistively Heated Thermal Hydraulic Simulator for Nuclear Rocket Reactor Cores

A preliminary design study is presented for a non-nuclear test facility which uses ohmic heating to replicate the thermal hydraulic characteristics of solid core nuclear reactor fuel element passages. The basis for this testing capability is a recently commissioned nuclear thermal rocket environments simulator, which uses a high-power, multi-gas, wall-stabilized constricted arc-heater to produce high-temperature pressurized hydrogen flows representative of reactor core environments, excepting radiation effects. Initially, the baseline test fixture for this non-nuclear environments simulator was configured for long duration hot hydrogen exposure of small cylindrical material specimens as a low cost means of evaluating material compatibility. It became evident, however, that additional functionality enhancements were needed to permit a critical examination of thermal hydraulic effects in fuel element passages. Thus, a design configuration was conceived whereby a short tubular material specimen, representing a fuel element passage segment, is surrounded by a backside resistive tungsten heater element and mounted within a self-contained module that inserts directly into the baseline test fixture assembly. With this configuration, it becomes possible to create an inward directed radial thermal gradient within the tubular material specimen such that the wall-to-gas heat flux characteristics of a typical fuel element passage are effectively simulated. The results of a preliminary engineering study for this innovative concept are fully summarized, including high-fidelity multi-physics thermal hydraulic simulations and detailed design features.

Litchford, Ron J.↗
NASA NTRS1972

Completely modular thermionic reactor ion propulsion system /trips/.

Description of a nuclear reactor powered ion propulsion system which is an advanced completely modularized system lending itself to development of prototype and/or flight type components without the need for complete system tests until late in the development program. This modularity is achieved in all of the subsystems and components of the electric propulsion system, including the thermionic fuel elements, the heat rejection subsystem (head pipes), the power conditioning modules, and the ion thrusters. The use of heat pipes for cooling the collectors and as a means of heat transport to the radiator allows early prototype or flight configuration testing of a small module of the heat rejection subsystem as opposed to full scale liquid metal pumps and radiators in a large vacuum chamber. The power conditioner (p/c) is arranged in modules with passive cooling which allows complete prototype testing. The ion engines are typically matched with one or more p/c modules and are the same size for any power level propulsion system of interest.

Peelgren, M. L.↗
DOE OSTI2023

SPC-70486 Rev 0 Stirling Engine Prototype Performance Specification

Idaho National Laboratory (INL) is developing a small microreactor to produce electrical power utilizing a small nuclear core and Stirling engines under the Microreactor Application Research Validation and Evaluation (MARVEL) program. The purpose of this performance specification is to define requirements to build and test the Stirling Engine prototype, including interfacing with Kinnetech in the development/test of the engines. This specification includes prototyping components only and are not for use in the reactor. Prototype testing is a preliminary test to be conducted with conceptual equipment from which a final design will be developed. The results of the testing defined herein will not be used as a qualification test. The Stirling Engines have an engine controller module (QEC). Currently, the QEC doesn’t have a means to stop the engine. An analog controlled engine braking system must be developed and tested with the engine prototype. Any conflict between this specification and any supplementary specifications in the procurement documents requires written clarification from the Contractor prior to proceeding with any work. Any deviation from the procurement documents requires approval by the Contractor with the change request process.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
DOE OSTI2020

Imaging Based Detector with Efficient Scintillators for Neutron Diffraction Measurements

The Anger Camera developed by the detector group at the Oak Ridge National Laboratory was utilized for the present work for its unique advantage of employing multiple modules to obtain large active measurement area for detecting diffracted/scattered thermal neutrons. Considering the relatively small flux associated with diffracted/scattered neutrons, suitable efficiency with high spatial resolution is a requirement for utilizing two-dimensional imaging detectors. The potential to implement pulse shape (in addition to pulse height) discrimination-based scintillators further enhances the ability to detect diffracted neutrons with improved signal to noise ratio. In this paper, initial results associated with 6Li glass-based scintillator will be presented. The authors explored the feasibility of using this system to detect and quantify diffraction peaks and peak shifts at the Neutron Residual Stress Facility (NRSF2), High Flux Isotope Reactor (HFIR) in the Oak Ridge National Laboratory (ORNL). Suitable camera mounting and shielding had to be developed. Reference measurements using polycrystalline powders with known atomic planar spacing will be discussed along with measurement settings associated with expected resolution for peak shift measurements. Initial results are promising and demonstrate that a suitable scintillation-based neutron detecting system is viable for residual stress-based diffraction measurements. Small area detectors are also feasible with suitable consideration to scattering volume and distance to detector.

Seals, Matthew↗
DOE OSTI2017

Fuel pin zoning to improve the accuracy of the palisades spent fuel pool criticality analysis

Most PWR spent fuel pool criticality analyses assume that the burnup is uniformly distributed among the fuel pins in the assembly. This assumption is accurate for PWR cores that have the typical small inter-assembly gap. The Palisades reactor core, however, is different in that it uses cruciform control rods in between fuel assemblies. Every interior assembly has two sides having a wide water gap and two sides having a narrow water gap. After depleting under these conditions, the pins near the wide water gap will have a higher burnup than the pins in the center. If the burnup is distributed uniformly among the pins, the resulting k calculation will be conservative but excessively so. The excess conservatism was found to be on the order of 1% in k. For this analysis, SCALE 6.1 depletion with TRITON is used followed by the CSAS5 module for the Palisades pool analysis. Several burnup zoning arrangements (uniform enrichment) were investigated ranging from 3 zones to as many as 19 zones. The 19 zone analysis agreed well with a CASMO analysis where every pin was its own depletion zone. It was found that 8 fuel zones gave essentially the same result as 19 zones with less complexity. The depletion calculation using 8 fuel zones instead of a single zone is more complicated but the resulting k calculation no longer has the excessive conservatism seen in a one zone model. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Decrypting bacterial polyphenol metabolism in an anoxic wetland soil

Abstract Microorganisms play vital roles in modulating organic matter decomposition and nutrient cycling in soil ecosystems. The enzyme latch paradigm posits microbial degradation of polyphenols is hindered in anoxic peat leading to polyphenol accumulation, and consequently diminished microbial activity. This model assumes that polyphenols are microbially unavailable under anoxia, a supposition that has not been thoroughly investigated in any soil type. Here, we use anoxic soil reactors amended with and without a chemically defined polyphenol to test this hypothesis, employing metabolomics and genome-resolved metaproteomics to interrogate soil microbial polyphenol metabolism. Challenging the idea that polyphenols are not bioavailable under anoxia, we provide metabolite evidence that polyphenols are depolymerized, resulting in monomer accumulation, followed by the generation of small phenolic degradation products. Further, we show that soil microbiome function is maintained, and possibly enhanced, with polyphenol addition. In summary, this study provides chemical and enzymatic evidence that some soil microbiota can degrade polyphenols under anoxia and subvert the assumed polyphenol lock on soil microbial metabolism.

59 BASIC BIOLOGICAL SCIENCES↗
NASA NTRS1990

Fluid dynamics and low gravity effects of chemical vapor deposition

Based on the comparison between experimental data and numerical results for the growth of GaAs from TMGa, it was shown that 3D simulations are necessary to simulate rectangular CVD reactors even when operated under subcritical (Ra) conditions. The important points found are summarized in the three attached reprints. The experimental studies of mixed convection in horizontal channels have shown three regimes of high Ra (22,220) number flows. At Re = 18.5, the rolls develop very quickly, significantly modulating the axial velocity even before it reaches the beginning of the hot plate. A few centimeters downstream, the velocities become asymmetric about the vertical centerplane and at x = 12 cm, become unsteady. These asymmetries were predicted theoretically, but experimental evidence has not been published prior to this work. At Re = 36, the axial velocity is only slightly modified at x = 0. Although the flow remains steady and symmetric about the vertical centerplane, there is a small spatial oscillation in the velocities over the length of the channel. The period of this oscillation was around 5 cm. At Re = 54, the longitudinal rolls developed smoothly over a length of 30 cm, with no asymmetries, unsteadiness, or spatial oscillations. Comparison of numerical simulations of these flows to experiments has revealed the importance and difficulty of setting proper thermal boundary conditions on the sidewalls. Calculated flows and experimentally measured flows showed very similar profiles, but at different axial locations, with the rolls developing more rapidly in the experiments. This is directly attributable to partially conducting sidewalls of the apparatus being hotter in the entrance section than the adiabatic walls of the simulations. A thorough comparison of the experimental data and numerical results for a variety of sidewall boundary conditions is in preparation.

Nyce, Thomas A.↗
DOE OSTI2026

Modeling Flow in a Microreactor Core: From High-Fidelity CFD to Subchannel Analysis

The accurate prediction of pressure drop in tightly packed, low–Reynolds number (Re) bare rod bundles is essential for the thermal-hydraulic design of the Microreactor Applications Research Validation and EvaLuation (MARVEL) reactor and other microreactor concepts. However, existing friction factor correlations, particularly the upgraded Cheng-Todreas (UCTD) correlation, have limited validation for the small pitch-to-diameter ratios (P/Ds) and transitional flow conditions characteristic of these systems. In this work, we perform high-fidelity large-eddy simulations (LESs) of both an infinite bare rod bundle and a finite bare 37-pin scalloped rod bundle across the range 1000≤ 𝑅𝑒 ≤5500. The simulations reveal strong gap vortex–driven transitional behavior and indicate that the UCTD may underpredict the friction factor by up to 28% at 𝑃/𝐷 = 1.05. Using the LES-calculated pressure drops, we formulate a new friction factor correlation that follows the Cheng-Todreas functional form but is calibrated for low-Re and tightly packed geometries representative of a MARVEL-like reactor. The correlation is implemented in the MOOSE (Multiphysics Object-Oriented Simulation Environment) subchannel module and compared against both the LES and UCTD predictions. Across all subchannel types, the proposed model reduces the streamwise velocity differences from as high as 44% (UCTD) to below 9%, and decreases the pressure gradient differences from 13% to 25% (UCTD) to 0.7% to 7% relative to the LES results. These results suggest that the new correlation has the potential to improve the pressure drop and flow field predictions for such geometries, highlighting the importance of high-fidelity simulations in supporting microreactor thermal-hydraulic model development and motivating future pressure drop experiments for compact rod bundles to further validate these findings.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
DOE OSTI2016

Development of plasma bolometers using fiber-optic temperature sensors

Measurements of radiated power in magnetically confined plasmas are important for exhaust studies in present experiments and expected to be a critical diagnostic for future fusion reactors. Resistive bolometer sensors have long been utilized in tokamaks and helical devices but suffer from electromagnetic interference (EMI). Results are shown from initial testing of a new bolometer concept based on fiber-optic temperature sensor technology. A small, 80 μm diameter, 200 μm long silicon pillar attached to the end of a single mode fiber-optic cable acts as a Fabry–Pérot cavity when broadband light, λ{sub o} ∼ 1550 nm, is transmitted along the fiber. Changes in temperature alter the optical path length of the cavity primarily through the thermo-optic effect, resulting in a shift of fringes reflected from the pillar detected using an I-MON 512 OEM spectrometer. While initially designed for use in liquids, this sensor has ideal properties for use as a plasma bolometer: a time constant, in air, of ∼150 ms, strong absorption in the spectral range of plasma emission, immunity to local EMI, and the ability to measure changes in temperature remotely. Its compact design offers unique opportunities for integration into the vacuum environment in places unsuitable for a resistive bolometer. Using a variable focus 5 mW, 405 nm, modulating laser, the signal to noise ratio versus power density of various bolometer technologies are directly compared, estimating the noise equivalent power density (NEPD). Present tests show the fiber-optic bolometer to have NEPD of 5-10 W/m{sup 2} when compared to those of the resistive bolometer which can achieve <0.5 W/m{sup 2} in the laboratory, but this can degrade to 1-2 W/m{sup 2} or worse when installed on a tokamak. Concepts are discussed to improve the signal to noise ratio of this new fiber-optic bolometer by reducing the pillar height and adding thin metallic coatings, along with improving the spectral resolution of the interrogator.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2024

Fast Neutron Irradiation of a Multichannel JFET-Based Optical Encoder

Modern electrical components are susceptible to damage from high levels of radiation and extreme temperatures found near reactors in terrestrial nuclear power plants and in aerospace applications. Radiation-hardened electronics are being developed, largely for the aerospace industry, but they sometimes rely on application-specific, small-batch semiconductor fabrication processes. These processes tend to be prohibitively expensive to develop and maintain outside major industrial facilities or governmental agencies. Recently, commercially available, nonradiation-rated junction-gate field-effect transistors (JFETs) were shown to maintain their functionality at gamma doses exceeding 1 MGy, suggesting that nonrated, commercially available electrical components could be used to develop systems that are tolerant to ionizing radiation. However, gamma ray survival is not indicative of neutron dose survival, and few studies characterize JFETs under neutron irradiation. To address this knowledge gap, a JFET-based analog multiplexer and optical pulsewidth modulation (PWM) encoder was developed and irradiated using a 252 Cf source to 1.6×10 13 n/cm 2 . The multiplexed optical encoder (MOE) system maintained functionality throughout testing and showed little evidence of radiation effects. These results indicate that circuitry tolerant to fast neutron damage can be developed using low-cost, nonradiation-rated, commercially available JFETs, which could provide a lower production cost alternative to specialized semiconductor processes when designing and building electronics better able to survive neutron irradiation.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2012

Benefits of the delta K of depletion benchmarks for burnup credit validation

Pressurized Water Reactor (PWR) burnup credit validation is demonstrated using the benchmarks for quantifying fuel reactivity decrements, published as 'Benchmarks for Quantifying Fuel Reactivity Depletion Uncertainty,' EPRI Report 1022909 (August 2011). This demonstration uses the depletion module TRITON available in the SCALE 6.1 code system followed by criticality calculations using KENO-Va. The difference between the predicted depletion reactivity and the benchmark's depletion reactivity is a bias for the criticality calculations. The uncertainty in the benchmarks is the depletion reactivity uncertainty. This depletion bias and uncertainty is used with the bias and uncertainty from fresh UO{sub 2} critical experiments to determine the criticality safety limits on the neutron multiplication factor, k{sub eff}. The analysis shows that SCALE 6.1 with the ENDF/B-VII 238-group cross section library supports the use of a depletion bias of only 0.0015 in delta k if cooling is ignored and 0.0025 if cooling is credited. The uncertainty in the depletion bias is 0.0064. Reliance on the ENDF/B V cross section library produces much larger disagreement with the benchmarks. The analysis covers numerous combinations of depletion and criticality options. In all cases, the historical uncertainty of 5% of the delta k of depletion ('Kopp memo') was shown to be conservative for fuel with more than 30 GWD/MTU burnup. Since this historically assumed burnup uncertainty is not a function of burnup, the Kopp memo's recommended bias and uncertainty may be exceeded at low burnups, but its absolute magnitude is small. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Steam Generator Model Design Parameter Sensitivity Study Using Advanced Optimization Tools

This study focuses on design parameter sensitivity studies pertaining to a steam generator (SG) model, using both Python and machine-learning tools. The SG model is a mathematical representation (including fluid flow and heat transfer equations/models/correlations) of a steam-generating unit in a pressurized water reactor (PWR)-type small modular reactor (SMR) system. Design studies involve changing the model’s input design parameters (e.g., temperature, pressure, mass flow rate) to observe the resulting effects on the output of the system (e.g., heat transfer coefficient [HTC], Nusselt number, heat transfer performance). Sensitivity studies analyze the degree to which system output and/or desired parameters (e.g., HTC or heat transfer performance) are sensitive to changes in input parameters. By using machine-learning tools such as the Risk Analysis Virtual Environment (RAVEN) developed at Idaho National Laboratory (INL), detailed design parametric sensitivity studies and model optimization were performed. Six input parameters—namely, the pressure, temperature, and mass flow rate for the inlet of the primary-side (hot fluid) and secondary-side (cold fluid) of the SG—were randomly perturbed via RAVEN’s Monte Carlo Sampler module, using uniform distributions (±1% relative changes). The analysis results give valuable insights into SG system performance and optimization, and provide justification for researching optimized sensor placement to effectively monitor and obtain experimental data.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Steam Generator Model Design Parameter Sensitivity Study Using Advanced Optimization Tools

This study focuses on design parameter sensitivity studies pertaining to a steam generator (SG) model, using both Python and machine-learning tools. The SG model is a mathematical representation (including fluid flow and heat transfer equations/models/correlations) of a steam-generating unit in a pressurized water reactor (PWR)-type small modular reactor (SMR) system. Design studies involve changing the model’s input design parameters (e.g., temperature, pressure, mass flow rate) to observe the resulting effects on the output of the system (e.g., heat transfer coefficient [HTC], Nusselt number, heat transfer performance). Sensitivity studies analyze the degree to which system output and/or desired parameters (e.g., HTC or heat transfer performance) are sensitive to changes in input parameters. By using machine-learning tools such as the Risk Analysis Virtual Environment (RAVEN) developed at Idaho National Laboratory (INL), detailed design parametric sensitivity studies and model optimization were performed. Six input parameters—namely, the pressure, temperature, and mass flow rate for the inlet of the primary-side (hot fluid) and secondary-side (cold fluid) of the SG—were randomly perturbed via RAVEN’s Monte Carlo Sampler module, using uniform distributions (±1% relative changes). The analysis results give valuable insights into SG system performance and optimization, and provide justification for researching optimized sensor placement to effectively monitor and obtain experimental data.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2026

Extension of Clad Damage Propagation Model for Fission Gas Dispersal and Two-Phase Flow Effects in MOOSE SubChannel Module

This report presents an extension of the Clad Damage Propagation (CDAP) model implemented in the MOOSE SubChannel Module (SCM) to capture post-failure fission-gas dispersal and two-phase flow effects in sodium-cooled fast reactor assemblies. The extended model tracks discharged gas axially and radially, computes channel-averaged flow quality and void fraction using a Lockhart–Martinelli framework, evaluates two-phase frictional pressure-drop multipliers, determines inlet mass-flow degradation under fixed core pressures, and applies an intensified-void-based heat-transfer degradation to affected fuel pins. Radial plume expansion is parameterized using mineral-oil jet experiments mapped to sodium conditions via Reynolds–Weber similarity. Implementation details are documented, along with the new methods and user inputs needed to control plume mapping and two-phase behavior. Demonstration simulations for 19- and 37-pin bundles show that breach size and inlet velocity strongly influence propagation potential: small breaches (≤0.5 mm) produce limited degradation while larger breaches (~1 mm) can drive oscillatory temperature spikes and enhanced failure propagation, especially at higher velocities. These results demonstrate that the extended CDAP model provides a more complete framework for quantifying cladding damage propagation and evaluating propagation potential in transient scenarios. The approach remains computationally efficient, consistent with subchannel-level analysis, yet incorporates sufficient physics to bridge localized post-failure effects with bundle- and assembly-scale degradation.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS1989

M.I.N.G., Mars Investment for a New Generation: Robotic construction of a permanently manned Mars base

A basic procedure for robotically constructing a manned Mars base is outlined. The research procedure was divided into three areas: environment, robotics, and habitat. The base as designed will consist of these components: two power plants, communication facilities, a habitat complex, and a hangar, a garage, recreation and manufacturing facilities. The power plants will be self-contained nuclear fission reactors placed approx. 1 km from the base for safety considerations. The base communication system will use a combination of orbiting satellites and surface relay stations. This system is necessary for robotic contact with Phobos and any future communication requirements. The habitat complex will consist of six self-contained modules: core, biosphere, science, living quarters, galley/storage, and a sick bay which will be brought from Phobos. The complex will be set into an excavated hole and covered with approximately 0.5 m of sandbags to provide radiation protection for the astronauts. The recreation, hangar, garage, and manufacturing facilities will each be transformed from the four one-way landers. The complete complex will be built by autonomous, artificially intelligent robots. Robots incorporated into the design are as follows: Large Modular Construction Robots with detachable arms capable of large scale construction activities; Small Maneuverable Robotic Servicers capable of performing delicate tasks normally requiring a suited astronaut; and a trailer vehicle with modular type attachments to complete specific tasks; and finally, Mobile Autonomous Rechargeable Transporters capable of transferring air and water from the manufacturing facility to the habitat complex.

Amos, Jeff↗
DOE OSTI2012

An improved energy-collapsing method for core-reflector modelization in SFR core calculations using the PARIS platform

In the framework of the ASTRID project, sodium cooled fast reactor studies are conducted at CEA in compliance with GEN IV reactors criteria, particularly for safety requirements. An improved safety requires better calculation tools to obtain accurate reactivity effects (especially sodium void effect) and power map distributions. The current calculation route lies on the JEFF3.1.1 library and the classical two-step approach performed with the ECCO module of the ERANOS code system at the assembly level and the Sn SNATCH solver - implemented within the PARIS platform - at the core level. 33-group cross sections used by SNATCH are collapsed from 1968-group self-shielded cross-section with a specific flux-current weighting. Recent studies have shown that this collapsing is non-conservative when dealing with core-reflector interface and can lead to reactivity discrepancies larger than 500 pcm in the case of a steel reflector. Such a discrepancy is due to the flux anisotropy at the interface, which is not taken into account when cross sections are obtained from separate fuel and reflector assembly calculations. A new approach is proposed in this paper. It consists in separating the self-shielding and the flux calculations. The first one is still performed with ECCO on separate patterns. The second one is done with SNATCH on a 1D traverse, representative of the core-reflector interface. An improved collapsing method using angular flux moments is then carried out to collapse the cross sections onto the 33-group structure. In the case of a simplified ZONA2B 2D homogeneous benchmark, results in terms of k{sub eff} and power map are strongly improved for a small increase of the computing time. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2014

Update on Small Modular Reactors Dynamics System Modeling Tool -- Molten Salt Cooled Architecture

The Small Modular Reactor (SMR) Dynamic System Modeling Tool project is in the third year of development. The project is designed to support collaborative modeling and study of various advanced SMR (non-light water cooled) concepts, including the use of multiple coupled reactors at a single site. The objective of the project is to provide a common simulation environment and baseline modeling resources to facilitate rapid development of dynamic advanced reactor SMR models, ensure consistency among research products within the Instrumentation, Controls, and Human-Machine Interface (ICHMI) technical area, and leverage cross-cutting capabilities while minimizing duplication of effort. The combined simulation environment and suite of models are identified as the Modular Dynamic SIMulation (MoDSIM) tool. The critical elements of this effort include (1) defining a standardized, common simulation environment that can be applied throughout the program, (2) developing a library of baseline component modules that can be assembled into full plant models using existing geometry and thermal-hydraulic data, (3) defining modeling conventions for interconnecting component models, and (4) establishing user interfaces and support tools to facilitate simulation development (i.e., configuration and parameterization), execution, and results display and capture.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

Assessment of NuScale SMR Steam Heat Augmentation for Chemical Plant Decarbonization

Nearly 50% of the total energy consumed by the industrial sector in the United States is used to produce process steam with natural gas and coal-fired boilers1 . This project conducts a technoeconomic assessment of a NuScale Small Modular Reactor (SMR) coupled with a chemical plant as an Integrated Energy System (IES) where nuclear produces steam and electric power to meet the requirements of a large chemical plant. In a 2020 study, ORNL evaluated the feasibility of using advanced SMRs, including the NuScale design, to supply energy to the Eastman Chemical Plant. However, since that report was published, NuScale received NRC approval for its uprated 77 MWe design with 56% more power and has also introduced a high-temperature, high pressure, steam heat-augmentation system, a key focus of the new study. The new study also benefits from revised capital costs, a 10-day refueling outage time, reduced plant staffing, higher capacity factors, and a site boundary Emergency Planning Zone methodology. The study consists of a techno-economic assessment of two possible energy sources (nuclear and natural gas) in a number of steam and power generation configurations (NuScale Power Modules (NPMs), boilers and combinations of both) to satisfy the steam and power demand with the most reliable and cost competitive system. A total of 2,947.3 klb/hr of steam and 72.5 MWe of electricity are required for the demonstration case. A range of scenarios and solutions are explored, from a 12-NPM plant (3,000 MWth)—with excess capacity and redundancy, capable of supplying a significant amount of extra power to the grid—to a 4-NPM (1,000 MWth) plant—supplemented with existing boilers or grid power for redundancy. The study uses historical steam and power data from a chemical plant and examines the sensitivity to natural gas and grid power cost variations. Scenarios with up to two times gas and electricity costs were considered. Profitability in a 60-year time horizon was analyzed, consistent with NuScale’s design life specification. A steady-state site integration and reliability analysis was performed, and trade-offs were identified.

20 FOSSIL-FUELED POWER PLANTS↗