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At least 325 records · Page 18

Techno-Economic Assessment of the Factory Production of Small Modular Reactor

The demand for electricity continues to grow around the world while pressures remain high to reduce carbon emissions. Nuclear energy provides a strong and reliable baseload power component without a high output of greenhouse gases. Global nuclear energy capacity is increasing with 62 reactors under construction in 14 countries. Nuclear energy will be crucial as energy demands are met with a sustainable and diverse energy portfolio. Small modular reactors (SMRs) are small scale (less than 300 MWe) reactors that incorporate the same advanced technologies included in recent LR nuclear reactor designs. SMRs are designed to include the latest reactor technology including passive safety features. SMRs offer economic advantages over LRs such as lower up-front capital and easier integration into the existing power grid. The modular characteristic of SMRs would allow incremental additions to power plants years after its completion. The cost of constructing SMR power plants will determine their future in the energy industry. Three SMR reactors are in advanced stages of design in the United States; the Westinghouse 225 MWe, mPower 180 MWe and nuScale 45 MWe. SMRs are expected to have lower capital expenses of reactor components due to factory production. Current estimates predict these savings to reduce SMR capital 30-40%. Existing studies have estimated SMR capital using economies of scale and holistic discounts applied to an overall capital expense figure, but do not include the expense of an SMR factory facility. To better understand the actual costs associated with construction of an SMR power plant, capital components will be broken down according to its place in the construction phase. The expense of a factory facility will also be included with SMR capital expenses to determine its significance and benefits to overall capital expenses. Results from this study show that SMRs can compete with LRs in terms of startup costs. The capital cost of building a factory for SMRs was previously overlooked for SMR capital estimates. The annualized capital expenses expected for an SMR factory are very small, but offer significant discounts to the overall capital costs for SMRs. Case 3 shows that factory production can reduce SMR capital expenses to the range of 30%-40% when compared to Case 1 in this study. Factory savings of this magnitude would require factory production of several SMR reactor and generator components including the reactor pressure vessel, control rods, steam generation, and chemical and volume control system, containment, condenser and turbo-generator. SMRs will also benefit from lower indirect expenses due to a much smaller construction duration. It is expected that SMRs will have capital costs competitive with LRs. The differences between Case 2 and Case 3 in this study show that SMRs are heavily dependent on the cost reductions coming from factory production. The factory facility represents a small increase in capital expense, but accounts for a large reduction of cost for factory components. SMRs have the potential to meet growing energy demands in the United States and around the world, but their economic viability is dependent on a factory facility that will lower the capital expenses of reactor and generator components. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Synthesis of supported silver nano-spheres on zinc oxide nanorods for visible light photocatalytic applications

Highlights: • Synthesis of supported Ag NPs on ZnO nanorods using open vessel microwave reactor. • Use of the Ag/ZnO NPs as an efficient visible light photocatalyst. • Complete degradation of methylene blue in 1 h with 0.5 g/L Ag/ZnO NPs. - Abstract: We report the synthesis of silver (Ag) nano-spheres (NS) supported on zinc oxide (ZnO) nanorods through two step mechanism, using open vessel microwave reactor. Direct reduction of ZnO from zinc nitrates was followed by deposition precipitation of the silver on the ZnO nanorods. The supported Ag/ZnO nanoparticles were then characterized by electron microscopy, X-ray diffraction, FTIR, photoluminescence and UV–vis spectroscopy. The visible light photocatalytic activity of Ag/ZnO system was investigated using a test contaminant, methylene blue (MB). Almost complete removal of MB in about 60 min for doses higher than 0.5 g/L of the Ag/ZnO photocatalyst was achieved. This significant improvement in the photocatalytic efficiency of Ag/ZnO photocatalyst under visible light irradiation can be attributed to the presence of Ag nanoparticles on the ZnO nanoparticles which greatly enhances absorption in the visible range of solar spectrum enabled by surface plasmon resonance effect from Ag nanoparticles.

36 MATERIALS SCIENCE↗

CTF Theory Manual

Coolant-Boiling in Rod Arrays|Two Fluids (COBRA-TF) is a thermal/ hydraulic (T/H) simulation code designed for light water reactor (LWR) vessel analysis. It uses a two-fluid, three-field (i.e. fluid film, fluid drops, and vapor) modeling approach. Both sub-channel and 3D Cartesian forms of 9 conservation equations are available for LWR modeling. The code was originally developed by Pacific Northwest Laboratory in 1980 and had been used and modified by several institutions over the last few decades. COBRA-TF also found use at the Pennsylvania State University (PSU) by the Reactor Dynamics and Fuel Management Group (RDFMG) and has been improved, updated, and subsequently re-branded as CTF. As part of the improvement process, it was necessary to generate sufficient documentation for the open-source code which had lacked such material upon being adopted by RDFMG. This document serves mainly as a theory manual for CTF, detailing the many two-phase heat transfer, drag, and important accident scenario models contained in the code as well as the numerical solution process utilized. Coding of the models is also discussed, all with consideration for updates that have been made when transitioning from COBRA-TF to CTF. Further documentation outside of this manual is also available at RDFMG which focus on code input deck generation and source code global variable and module listings.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Effects of ATR-2 Irradiation to High Fluence on Nine RPV Surveillance Materials

The reactor pressure vessel (RPV) in a light-water reactor (LWR) represents the first line of defense against a release of radiation in case of an accident. Thus, regulations that govern the operation of commercial nuclear power plants require conservative margins of fracture toughness, both during normal operation and under accident scenarios. In the unirradiated condition, the RPV has sufficient fracture toughness such that failure is implausible under any postulated condition, including pressurized thermal shock (PTS) in pressurized water reactors (PWR). In the irradiated condition, however, the fracture toughness of the RPV may be severely degraded, with the degree of toughness loss dependent on the radiation sensitivity of the materials. The available embrittlement predictive models and our present understanding of radiation damage are not fully quantitative, and do not treat all potentially significant variables and issues, particularly considering extension of operation to 80y.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Summary of LWRS Research in Addressing RPV Research Gaps in NRC EMDA Report

Reactor Pressure Vessels (RPVs) are critical components in nuclear reactors, housing the reactor core and coolant under extreme conditions of temperature, pressure, and radiation. These harsh environments contribute to the degradation of RPV materials over time, presenting challenges for extending reactor operations beyond their original design lifespans. The NRC Expanded Materials Degradation Assessment (EMDA) report volume 3 have been instrumental in guiding research to support extending the operational life of light water reactors (LWRs) up to 80 years. It provides a comprehensive framework to address technical challenges related to aging and degradation mechanisms in RPVs. The LWRS program has played a key role in advancing this research, supporting projects such as the UCSB ATR-2 Experiment, material testing from Zion and Palisades reactors, and the development of advanced mini-compact tension testing techniques. These efforts have been crucial in identifying and addressing gaps in our understanding of RPV aging, contributing to the successful subsequent license renewals of eight LWR units in the U.S. The EMDA report volume 3, built on the Phenomena Identification and Ranking Table (PIRT) analysis from earlier versions of the EPRI Materials Degradation Matrix (MDM) and Issue Management Tables (IMTs), provides a detailed assessment of RPV degradation mechanisms. However, as EPRI has updated the MDM and IMT, it is important to revisit research priorities and methodologies to reflect these changes. The revised MDM and IMT may introduce new factors affecting long-term RPV performance and safety, highlighting the need for continued research and updated guidance to ensure the reliable and safe operation of reactors beyond 80 years.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Passive filtration of air egressing from nuclear containment

A nuclear reactor includes a reactor core comprising fissile material disposed in a reactor pressure vessel. A radiological containment contains the nuclear reactor. A containment compartment contains the radiological containment. A heat sink includes a chimney configured to develop an upward-flowing draft in response to heated fluid flowing into a lower portion of the chimney. A fluid conduit is arranged to receive fluid from the containment compartment and to discharge into the chimney. A filter may be provided, with the fluid conduit including a first fluid conduit arranged to receive fluid from the containment compartment and to discharge into an inlet of the filter, and a second fluid conduit arranged to receive fluid from an outlet of the filter and to discharge into the chimney. As the draft is developed passively, there is no need for a blower or pump configured to move fluid through the fluid conduit.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Automated X-Ray Diffraction of Irradiated Materials

Synchrotron-based X-ray diffraction (XRD) and small-angle Xray scattering (SAXS) characterization techniques used on unirradiated and irradiated reactor pressure vessel steels yield large amounts of data. Machine learning techniques, including PCA, offer a novel method of analyzing and visualizing these large data sets in order to determine the effects of chemistry and irradiation conditions on the formation of radiation induced precipitates. In order to run analysis on these data sets, preprocessing must be carried out to convert the data to a usable format and mask the 2-D detector images to account for experimental variations. Once the data has been preprocessed, it can be organized and visualized using principal component analysis (PCA), multi-dimensional scaling, and k-means clustering. In conclusion, from these techniques, it is shown that sample chemistry has a notable effect on the formation of the radiation induced precipitates in reactor pressure vessel steels.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

Development of HELIOS/BIPR/PARCS/MCNP6 Computation Route for WWER RPV Neutron Fluence Analysis and Validation Against Ex-Vessel Detector Measurement Data

Here, a computational route was developed for precise calculation of fast neutron fluence on a WWER-type reactor pressure vessel (RPV). The method is based on the transfer of neutronics data from HELIOS-2 lattice calculations and nodal diffusion neutronics data (power, density, and temperature) from BIPR7.1 and PARCS 3.36/PATHS core calculations into a three-dimensional (pinwise axially distributed) fixed neutron source for modeling of transport of fast neutrons from the reactor core to the outer surface of the RPV using MCNP6.2. Validation of the proposed computational method was carried out based on comparative analysis of MCNP6.2-predicted and neutron dosimetry–measured reaction rates [ 54 Fe(n,p) 54 Мn, 93 Nb(n,n') 93 mNb, and 58 Ni(n,р) 58 Со] on the outer surface of the Armenian Nuclear Power Plant (ANPP) Unit 2 RPV. Validation revealed that the MCNP6.2-predicted fast neutron fluence results are very sensitive to the ENDF-B neutron data. Particularly, MCNP6.2 with ENDF/B-VIII.0 significantly underpredicts (20% to 30%) fast neutron fluence while using ENDF/B-VII.1 data overpredicts it. Adding revised beta-released evaluations of 54 Fe, 56 Fe, 57 Fe, and 16 O from the International Nuclear Data Evaluation Network (INDEN) to ENDF/B-VIII.0 allows one to obtain reasonable agreement with measurement results for all types of measured reaction rates.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Infrastructure development for radioactive materials at the NSLS-II

The X-ray Powder Diffraction (XPD) Beamline at the National Synchrotron Light Source-II is a multipurpose instrument designed for high-resolution, high-energy X-ray scattering techniques. In this article, the capabilities, opportunities and recent developments in the characterization of radioactive materials at XPD are described. The overarching goal of this work is to provide researchers access to advanced synchrotron techniques suited to the structural characterization of materials for advanced nuclear energy systems. XPD is a new beamline providing high photon flux for X-ray Diffraction, Pair Distribution Function analysis and Small Angle X-ray Scattering. The infrastructure and software described here extend the existing capabilities at XPD to accommodate radioactive materials. Such techniques will contribute crucial information to the characterization and quantification of advanced materials for nuclear energy applications. We describe the automated radioactive sample collection capabilities and recent X-ray Diffraction and Small Angle X-ray Scattering results from neutron irradiated reactor pressure vessel steels and oxide dispersion strengthened steels.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

Mechanical and Chemical Properties of Harvested Hypalon Cable Jacket Subjected to Accelerated Thermal Aging

For nuclear power plants (NPPs) that are considering second license renewal for their operation beyond 60 years, knowledge of long-term operation, condition monitoring, and viability for the reactor components including reactor pressure vessel, concrete structures, and cable systems is essential. Such knowledge will provide NPPs with a game plan to predict performance and to estimate costs that are associated with monitoring or replacement programs for the affected systems. For cable systems that encompass a wide variety of materials, manufacturers, and in-plant location, accelerated aging of harvested cable jacket and insulation can provide insight as to remaining useful life and methods for monitoring. Accelerated thermal aging in air at temperatures between 80°C and 120°C was carried out on a multi-conductor control rod cable that had been in-service for over 30 years and was made by Boston Insulated Wire with Hypalon cable jacket and ethylene-propylene rubber insulation. From elongation at break (EAB) measurements and supporting Arrhenius analysis of the jacket material, an activation energy of 97.84 kJ/mol was estimated and the time to degradation, as represented by 50% EAB at the expected maximum operating temperature of 45°C, was estimated as consistent with 80 years. These values were slightly below previous measurements and could either be attributed to in-service degradation or variations in material properties from a given manufacturer. In addition, results from indenter modulus measurements and Fourier transform infrared spectroscopy suggest possible markers that could be beneficial to monitor cable conditions.

cable aging, cable insulation, Hypalon, second lic↗

Irradiation of ultrasonic sensors and adhesive couplants for application in light water reactor primary loop piping and components

The Electric Power Research Institute (EPRI) Nuclear Sector and US Department of Energy Light Water Reactor Sustainability Program are committed to engaging in research and development endeavors to address materials aging issues specific to long term operation of light water power reactors. To this effect, EPRI launched an industry initiative to develop nondestructive evaluation systems for online monitoring of existing cracks in light water reactor primary coolant loop piping and components. One of the goals of this initiative is to develop a sensor system (or systems) that can determine nondestructively if cracks are growing or arrested and, in the case of the former, to characterize their growth rates. A missing component of this initiative is an experimental assessment of how sensors and adhesive couplants will perform in service when exposed to chronic energetic neutron radiation, particularly at the primary coolant loop hot and cold leg dissimilar metal welds, which join the primary loop piping to the reactor pressure vessel and reside in the vicinity of the reactor core. The objective of this experimental study was to determine how ultrasonic transducers and adhesive couplants perform when exposed to irradiation in a test reactor to simulate and accelerate in-service exposure. Further, to achieve this objective, the signal stability of piezoelectric transducers and performance of adhesive couplants as a function of accumulated fast neutron fluence were characterized by collecting ultrasonic data in-situ during irradiation. Of particular interest were the ultrasonic signal quality and time decay of the amplitude of acoustic reflections as a function of fast neutron fluence. The results of the study showed that, of the 8 transducer/substrate sample assemblies tested, only 3 generated usable ultrasonic signals through the conclusion of the irradiation campaign. It was found that high temperature epoxy tends to ultrasonically couple the sensors to the substrates better than three types of refractory ceramic cements studied, as is supported by post irradiation examination. The results obtained through this experimental study will be utilized in the achievement of the overall goal of development of a sensor system to perform online monitoring of primary loop components.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Initial Results from the High Fluence UCSB ATR-2 Irradiation

A primary focus of the Department of Energy, Office of Nuclear Energy Light Water Reactor Sustainability Program (LWRSP) is the safe operation of nuclear reactors to extended lifetimes of up to 80 years. One of the primary reactor components of concern is the reactor pressure vessel (RPV), which is increasingly embrittled over time due to damage from high energy neutrons created in the reactor core. Radiation damage leads to the formation of nm-scale precipitates and solute-defect clusters, which result in an increase in the steel's ductile to brittle transition temperature. Transition temperature shifts (TTS) are used to quantify and regulate embrittlement. Extended life operation will subject some RPVs to higher neutron fluences than previously experienced by in-service vessels. Thus, robust prediction methods must be developed that ensure the safety of our nation's fleet of reactors. Current models underpredict embrittlement data from high flux test reactor irradiations at extended life fluences. It has been hypothesized that one reason for this underprediction is the formation of so-called 'Late- Blooming Phase' Mn-Ni-Si precipitates, which form and grow only at high fluence and that are currently not included in regulatory models. Recent research has demonstrated the existence of high volume fractions of intermetallic Mn-Ni-Si precipitate (MNSP) phases at high fluence and have generated significant insight into their character. However, a number of questions remain regarding MNSPs, such as: 1) what alloys are most susceptible to MNSP formation; 2) how is MNSP formation affected by flux (φ), fluence (φt) and irradiation temperature (T{sub irr}); and, 3) how to accurately predict TTS under extended life low flux service conditions? The fact that no two RPVs are exactly the same, both in terms of the steel composition and vessel service conditions, makes these questions particularly difficult to answer. Further, understanding the effect of flux differences between test reactor irradiations and in-vessel service is absolutely critical. Hence, physical models must be developed that predict TTS as a function of alloy solute content and vessel service conditions. A large-scale irradiation designed to develop and validate an extended life embrittlement prediction model was recently completed in the Advanced Test Reactor (ATR) at the Idaho National Laboratory as part of the DOE sponsored Nuclear Scientific User Facilities. This experiment, called the UCSB ATR-2 irradiation, is a collaboration between Oak Ridge National Lab (ORNL) and the University of California, Santa Barbara (UCSB) and is now primarily supported by the DOE Light Water Reactor Sustainability Program. The experiment is designed to investigate the single and combined effects of neutron φ, φt, T{sub irr} and alloy chemistry on precipitate formation, hardening and embrittlement. While postirradiation examinations (PIE) is ongoing, and is expected to take several years to complete, the first preliminary results are presented below. It should also be noted that Rolls-Royce has been a collaborator in UCSB ATR-2 from the start, and there are several other national and international institutions supporting the PIE. (authors)

36 MATERIALS SCIENCE↗

A BWR Fuel Assembly Concept with Partial Segregation of Fertile Material

Small modular reactors (SMRs) have been a hot topic in recent years due to their own set of advantages over conventional large light water reactors (LWRs). One of the driving forces for SMR development is the significant reduction in the up-front cost and enhanced inherent safety features. The reduced safety concerns are attributed to two main aspects, decay heat and source term. By taking the advantages of the two-phase flow driving head, which allows a much smaller and simpler reactor pressure vessel compared to the integral PWRs, a 50 MWe BWR-based SMR called the Novel Modular Reactor (NMR-50) is currently being developed at Purdue University. For this NMR-50, an optimized long-life core design was developed, which achieves a 10.2-year cycle length under current industrial practices and constraints. Due to the reduced operating power density, however, it was evident that there is a significant room to increase the fuel linear power. Therefore, core designs of partial segregation of fertile material within a fuel assembly (FA) were investigated to extend the fuel cycle length further. A heterogeneous fertile material loading (HFL) within a fuel assembly was developed based on the existing industry UO{sub 2} once-through fuel cycle option for the purpose of near-term deployment and maximum retrofitability. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Nuclear reactor support and seismic restraint with in-vessel core retention cooling features

A nuclear reactor including a lateral seismic restraint with a vertically oriented pin attached to the lower vessel head and a mating pin socket attached to the floor. Thermally insulating materials are disposed alongside the exterior surface of a lower portion of the reactor pressure vessel including at least the lower vessel head.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Nonlinear Ultrasonic Techniques to Monitor Radiation Damage in RPV and Internal Components

The objective of this research is to demonstrate that nonlinear ultrasonics (NLU) can be used to directly and quantitatively measure the remaining life in radiation damaged reactor pressure vessel (RPV) and internal components. Specific damage types to be monitored are irradiation embrittlement and irradiation assisted stress corrosion cracking (IASCC). Our vision is to develop a technique that allows operators to assess damage by making a limited number of NLU measurements in strategically selected critical reactor components during regularly scheduled outages. This measured data can then be used to determine the current condition of these key components, from which remaining useful life can be predicted. Methods to unambiguously characterize radiation related damage in reactor internals and RPVs remain elusive. NLU technology has demonstrated great potential to be used as a material sensor – a sensor that can continuously monitor a material’s damage state. The physical effect being monitored by NLU is the generation of higher harmonic frequencies in an initially monochromatic ultrasonic wave. The degree of nonlinearity is quantified with the acoustic nonlinearity parameter, β, which is an absolute, measurable material constant. Recent research has demonstrated that nonlinear ultrasound can be used to characterize material state and changes in microscale characteristics such as internal stress states, precipitate formation and dislocation densities. Radiation damage reduces the fracture toughness of RPV steels and internals, and can leave them susceptible to IASCC, which may in turn limit the lifetimes of some operating reactors. The ability to characterize radiation damage in the RPV and internals will enable nuclear operators to set operation time thresholds for vessels and prescribe and schedule replacement activities for core internals. Such a capability will allow a more clear definition of reactor safety margins. The research consists of three tasks: (1) materials sensing and monitoring; (2) physics-based materials and damage evolution modeling; and (3) remaining life estimation by integrating sensing, modeling and uncertainty.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Conformal core cooling and containment structure

A nuclear reactor includes a pressure vessel and a nuclear reactor core comprising fissile material disposed inside the pressure vessel at the bottom of the pressure vessel. A secondary core containment structure includes a containment basket comprising insulation with a maximum stable temperature of at least 2200K cladded by steel. The bottom of the pressure vessel and the nuclear reactor core are disposed inside the containment basket with the containment basket spaced apart from the bottom of the pressure vessel by a clearance gap. The containment basket may comprise zirconia insulation cladded by steel. In some embodiments the clearance gap between the containment basket and the bottom of the pressure vessel is no larger than one meter. The secondary core containment structure may further comprise conduits arranged to inject water into the clearance gap between the containment basket and the bottom of the pressure vessel.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Conformal core cooling and containment structure

A nuclear reactor includes a pressure vessel and a nuclear reactor core comprising fissile material disposed inside the pressure vessel at the bottom of the pressure vessel. A secondary core containment structure includes a containment basket comprising insulation with a maximum stable temperature of at least 2200K cladded by steel. The bottom of the pressure vessel and the nuclear reactor core are disposed inside the containment basket with the containment basket spaced apart from the bottom of the pressure vessel by a clearance gap. The containment basket may comprise zirconia insulation cladded by steel. In some embodiments the clearance gap between the containment basket and the bottom of the pressure vessel is no larger than one meter. The secondary core containment structure may further comprise conduits arranged to inject water into the clearance gap between the containment basket and the bottom of the pressure vessel.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Preliminary Primary System Thermal Fluids Analysis of a Horizontal Compact HTGR

The Horizontal Compact High Temperature Gas Reactor (HC-HTGR) is being designed by a multi-disciplinary team of nuclear, mechanical, and structural engineers, under the support of a DOE-NE Advanced Reactor Demonstration Program’s Advanced Reactor Concepts-20 (ARC-20) award. The objective of this ARC-20 project is to deliver a conceptual design for the proposed HC-HTGR in 3 years and support its commercialization as a safe, low-cost HTGR. Argonne National Laboratory (Argonne) is collaborating on the thermal hydraulic design and analysis of the HC-HTGR reactor pressure vessel internals. The scope of this work includes ensuring the reactor is able to maintain maximum core temperatures below designated safety thresholds during normal operation, shutdown, and accident conditions. This report documents the preliminary thermal hydraulic analysis of the HC-HTGR core design performed with a 1D fluid-3D solid coupled model built using the System Analysis Module (SAM). This assembly level model was utilized to inform the core assembly design, predict the temperature distribution in the peak power assembly including the peak fuel temperature, coolant channel outlet temperatures, and graphite temperature gradients. A key result of this analysis was the determination that the peak fuel temperature remains below the safety threshold of 1250°C. Additionally, this model was used to assess the assembly coolant channel and bypass flow mass flow rate distribution. Following the assembly level analysis, attention turned to the development of a full core reduced order model to be used to predict the core wide coolant flow distribution and to model certain operational and accidental transients.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗