Dissimilar Metal Bonding of Corrosion-Resistant Cladding to Structural Materials in Next-Generation Nuclear Reactors
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36 MATERIALS SCIENCE↗
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Dissimilar Metal Bonding of Corrosion-Resistant Cladding to Structural Materials in Next Generation Nuclear Reactors
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Cladding, corrosion resistance, molten salts, molt↗
Competitiveness Assessment of Decarbonizing Electricity and Process Heat Supply to a Campus with a Small Nuclear Reactor
This paper analyzes the competitiveness of siting a small nuclear reactor to support decarbonization of sites requiring tens of MW of electricity and/or process heat to support centralized heating and cooling system. This paper focuses on campuses as representative of sites with collections of buildings and research facilities with decarbonization needs represented by buildings heating, and electricity consumption by electrical loads which may include cooling via chilled water (e.g., for air conditioning and to cool down computer clusters). A nuclear reactor can be considered to decarbonize a site’s high-temperature steam generation used mostly for building heating needs, climate control, and hot water, by supplying process heat capabilities, while electricity decarbonization would be achieved mostly by the grid. However, a secondary application can be considered to maximize reactor utilization and avoid ramping down the reactor if the steam demand varies significantly throughout the year. Chilled water generation through steam-driven systems was identified as an attractive secondary option for the site analyzed, due to potential for plant design simplification, while electricity generation could be considered as well to reduce electricity purchases for a wider range of site applications. For a campus with peak 60MW thermal power demand, a small nuclear reactor with similar thermal power rating would almost eliminate CO2 emissions from steam generation and reduce electricity imports for chilled water production. A preliminary techno-economic feasibility study shows that a small nuclear reactor design that is optimized to support process heat can represent an economically feasible option when compared with other decarbonization alternatives.
Stauff, Nicolas E.↗
New Czech R and D Program on Fluoride Salt-cooled Nuclear Reactor Systems
A technology of nuclear reactor systems with liquid molten salt fuel has been investigated in the Czech Republic since 1999. The original effort came from the national Partitioning and Transmutation concept based on the subcritical Accelerator Driven System for incineration of transuranium elements with liquid fluoride fuel and pyrochemical partitioning fuel cycle technology. After 2005, the original R and D intentions were gradually converted to classical Molten Salt Reactor (MSR) technology and to thorium - uranium fuel cycle. The aim of this choice was to contribute to the development of an advanced nuclear technology, which can minimize environmental impact of nuclear power, save the natural resources and which has some potential to be deployed also in a non-superpower country. The basic technological development of selected areas of MSR and Th - U fuel cycle technology was realized in the frame of the SPHINX project solved in 2005 - 2012. In addition to the domestic activities of the MSR technology development, in 2012 the Czech Ministry of Industry and Trade and the US - Department of Energy concluded an agreement about the collaborative R and D on Molten Salt Reactor and Fluoride-salt-cooled High-temperature Reactor (FHR) technologies. These facts and results created the principal background for the present national program on the MSR/FHR technology development which was approved by the Ministry of Industry and Trade in 2016 and is granted by the Technological Agency of the Czech Republic.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Ultrasonic thermometry for nuclear reactors.
Ultrasonic thermometry for nuclear reactors based on temperature dependence of sound velocity in solids, discussing simulation experiments
Carnevale, E. H.↗
Dynamic friction and wear of a solid film lubricant during radiation exposure in a nuclear reactor
The effect of nuclear reactor radiation on the performance of a solid film lubricant was studied. The film consisted of molybdenum disulfide and graphite in a sodium silicate binder. Radiation levels of fast neutrons (E or = 1 MeV) were fluxed up to 3.5 times 10 to the 12th power n/sq cm-sec (intensity) and fluences up to 2 times 10 to the 18th power n/sq cm (total exposure). Coating wear lives were much shorter and friction coefficients higher in a high flux region of the reactor than in a low flux region. The amount of total exposure did not affect lubrication behavior as severely as the radiation intensity during sliding.
Jacobson, T. P.↗
The Effect of Point Reactor Kinetics Solvers on the Uncertainties of Nuclear Reactor Simulations
In this study, the transient response of a reactor in the case of reactivity insertion with and without reactivity feedback was explored. The uncertainties resulting from using various numerical solvers to simulate the reactor response using the point reactor kinetics model were investigated.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Systems and methods for dismantling a nuclear reactor
Systems and methods for dismantling a nuclear reactor are described. In one aspect the system includes a remotely controlled heavy manipulator ("manipulator") operatively coupled to a support structure, and a control station in a non-contaminated portion of a workspace. The support structure provides the manipulator with top down access into a bioshield of a nuclear reactor. At least one computing device in the control station provides remote control to perform operations including: (a) dismantling, using the manipulator, a graphite moderator, concrete walls, and a ceiling of the bioshield, the manipulator being provided with automated access to all internal portions of the bioshield; (b) loading, using the manipulator, contaminated graphite blocks from the graphite core and other components from the bioshield into one or more waste containers; and (c) dispersing, using the manipulator, dust suppression and contamination fixing spray to contaminated matter.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Development of Refractory Alloys and Refractory Coatings for Advanced Nuclear Reactors
The next generation of nuclear reactors will benefit from materials that enable operation at higher temperatures (>500°C), higher irradiation doses (up to 200 displacements per atom (dpa)), and the use of more corrosive and reactive coolants. This work package represents the first experimental steps towards a longer-term effort to develop refractory materials for nuclear energy applications which will enable operation under these conditions. Specifically, this work package focuses on additive manufacturing of refractories as both a refractory liner coating deposited onto the interior surface of a metallic tubular backbone and as bulk refractory alloys. During fiscal year 23 (FY23), several refractory metal coating systems and bulk alloys were examined and selected using a decision criteria matrix. The refractory metal coating systems included molybdenum, tungsten, and zirconium as refractory coatings on backbones of either carbon-carbon (C/C) or silicon carbide-silicon carbide (SiC/SiC) ceramic matrix composites. The bulk refractory alloys included C-103, WTa, and WNiFe as bulk alloys. During FY24 additional bulk refractory alloys and metallic backbones were evaluated using the decision criteria matrix based on input from the AMMT leadership team. These included 316 SS and 316H SS for the metallic backbones and Mo-La, Ta, and Nb1Zr as bulk refractory alloys. The primary focus for the FY24 effort was placed on establishing the capabilities to deposit refractory coatings based on the results of the scoring in the decision criteria matrix and finalizing the additively manufactured TZM studies which were incorporated into the AMMT program from the microreactor program. Further efforts were dedicated to establishing the capabilities to additively manufacture down-selected bulk refractory alloys.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Long-Term Stewardship Optimization for a Decommissioned Nuclear Reactor - 19579
The decommissioned Hallam Nuclear Power Facility (Hallam), located in Lancaster County, Nebraska, approximately 19 miles south of Lincoln, was developed as a prototype to investigate the technical and economic feasibility of a sodium-cooled, graphite-moderated nuclear reactor. As part of its Power Demonstration Program, the US Atomic Energy Commission (US AEC), a predecessor agency of the US Department of Energy (US DOE), operated the Hallam reactor from 1962 to 1964. The radiological surveillance of the Hallam site was established when the operating agreement between AEC and the Nebraska Public Power District was terminated. The US DOE Chicago Operations Office and the Nebraska Department of Health (NDH) agreed to install a shallow groundwater monitoring system as part of the environmental surveillance program. Surveillance at the site was scheduled to end in fiscal year 2005, with no further activities planned after that. At that time, the plan was to transition the site to the Nebraska Public Power District with future use of the facility to remain restricted. US DOE has title to and responsibility for the entombed radioactive materials. A surveillance and monitoring program was initiated in 1970 by NDH and was funded by US AEC. It included the analysis of samples from deep production wells (groundwater from the regional aquifer at depths greater than 55 meters [180 feet]) at the Sheldon Station. NDH was concerned in 1990 about the possibility of shallow groundwater coming in contact with the buried radiological materials along the buried walls of the reactor. Subsequently, DOE and NDH agreed to further characterize hydrologic conditions and establish a monitoring program in the shallow, perched groundwater zones. The current monitoring program focuses on the shallow, perched groundwater. Today, decades since facility decommissioning, site inspections and groundwater monitoring continue at this Cold War-era legacy site. The US DOE Office of Legacy Management (LM) monitors groundwater as a best management practice in response to a request from the Nebraska Department of Health and Human Services (DHHS). In 2006, LM recommended discontinuing groundwater monitoring because analytical results since 1970 demonstrated there had been no impact to shallow, perched groundwater and no current or anticipated unacceptable risk to human health and the environment. The State of Nebraska did not concur with the recommendation to stop monitoring but did agree to a reduction in sampling and analysis from once a year to once every 2 years--a schedule that began in 2008. An updated assessment of the groundwater monitoring effort was issued in 2016. The entire body of water-quality data continues to demonstrate that there have been no negative impacts to the shallow, perched groundwater from radioactive materials entombed at the site 46 years ago. LM, with concurrence from Nebraska DHHS, changed the sampling frequency, effective in 2018, as follows: once every 5 years from 2021 to 2041 and once every 10 years from 2041 to 2071. In summary, LM remains committed to its primary mission to protect human health and the environment at each of its 92 legacy sites. With results from groundwater sampling events demonstrating compliance for 48 years, the Hallam site was able to set an LM benchmark to optimize long-term stewardship and define an end-state strategy for other decommissioned reactor facilities across the DOE complex. Continued collaboration with local communities, stakeholder, and regulators often leads to outcomes of sound protectiveness without compromise to safety or compliance, as demonstrated by the optimization of the long-term stewardship at the Hallam site. (authors)
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Accelerated development of Zr-containing new generation ferritic steels for advanced nuclear reactors
The mission of the Nuclear Energy Enabling Technologies (NEET) program is to develop crosscutting technologies for nuclear energy applications. Advanced structural materials with superior performance at elevated temperatures are always desired for nuclear reactors, which can improve reactor economics, safety margins, and design flexibility. They benefit not only new reactors, including advanced light water reactors (LWRs) and fast reactors such as the sodium-cooled fast reactor (SFR) that is primarily designed for management of high-level wastes, but also life extension of the existing fleet when component exchange is needed. Developing and utilizing the modern materials science tools (experimental, theoretical, and computational tools) is an important path to more efficient alloy development and process optimization. The ultimate goal of this project is, with the aid of computational modeling tools, to accelerate the development of Zr-bearing ferritic alloys that can be fabricated using conventional steelmaking methods. The new alloys are expected to have superior high-temperature creep performance and excellent radiation resistance as compared to Grade 91. The designed alloys were fabricated using arc-melting and drop-casting, followed by hot rolling and conventional heat treatments. Comprehensive experimental studies have been conducted on the developed alloys to evaluate their hardness, tensile properties, creep resistance, Charpy impact toughness, and aging resistance, as well as resistance to proton and heavy ion (Fe 2+ ) irradiation.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Review and Modeling of Integrated Energy Systems with Nuclear Reactor Coupled Desalination and District Heating
Detailed reviews of a past advanced nuclear reactor based integrated energy system, as well as other nuclear reactor and fossil fuel based integrated energy systems have been performed for this work. Review of the utilization of heat from nuclear reactors for various applications and cogeneration has been done. The heat can be utilized by extraction of the steam from the turbine while the steam is still at a desired temperature. While use of nuclear process heat for district heating in countries like Finland, France, China, Poland, and elsewhere is discussed, more focus of the review has been given on nuclear desalination processes. Integrated energy systems (IES) where distinct types of reactors like PWR, BWR, sodium cooled fast reactor, heavy water reactor and other advanced reactors are coupled with various nuclear desalination processes like multi-effect distillation (MED), multi-stage flashing (MSF) and reverse osmosis (RO) methods have been discussed. The nuclear desalination plant at Aktau has been discussed in more detail due to its decades of successful operation. The IES of the Aktau plant coupled with 5-effect MED desalination plant has been taken as a reference for modeling the Open Modelica (OM) based IES of this work. Here, the OM IES model shows good agreement with the MED plant output of Aktau and can be extended for future applications of IES.
42 ENGINEERING↗
Improved Nuclear Reactor and Shield Mass Model for Space Applications
New technologies are being developed to explore the distant reaches of the solar system. Beyond Mars, solar energy is inadequate to power advanced scientific instruments. One technology that can meet the energy requirements is the space nuclear reactor. The nuclear reactor is used as a heat source for which a heat-to-electricity conversion system is needed. Examples of such conversion systems are the Brayton, Rankine, and Stirling cycles. Since launch cost is proportional to the amount of mass to lift, mass is always a concern in designing spacecraft. Estimations of system masses are an important part in determining the feasibility of a design. I worked under Michael Barrett in the Thermal Energy Conversion Branch of the Power & Electric Propulsion Division. An in-house Closed Cycle Engine Program (CCEP) is used for the design and performance analysis of closed-Brayton-cycle energy conversion systems for space applications. This program also calculates the system mass including the heat source. CCEP uses the subroutine RSMASS, which has been updated to RSMASS-D, to estimate the mass of the reactor. RSMASS was developed in 1986 at Sandia National Laboratories to quickly estimate the mass of multi-megawatt nuclear reactors for space applications. In response to an emphasis for lower power reactors, RSMASS-D was developed in 1997 and is based off of the SP-100 liquid metal cooled reactor. The subroutine calculates the mass of reactor components such as the safety systems, instrumentation and control, radiation shield, structure, reflector, and core. The major improvements in RSMASS-D are that it uses higher fidelity calculations, is easier to use, and automatically optimizes the systems mass. RSMASS-D is accurate within 15% of actual data while RSMASS is only accurate within 50%. My goal this summer was to learn FORTRAN 77 programming language and update the CCEP program with the RSMASS-D model.
Robb, Kevin↗
Systems and methods for managing shared-path instrumentation and irradiation targets in a nuclear reactor
Systems and methods permit discriminate access to nuclear reactors. Systems provide penetration pathways to irradiation target loading and offloading systems, instrumentation systems, and other external systems at desired times, while limiting such access during undesired times. Systems use selection mechanisms that can be strategically positioned for space sharing to connect only desired systems to a reactor. Selection mechanisms include distinct paths, forks, diverters, turntables, and other types of selectors. Management methods with such systems permits use of the nuclear reactor and penetration pathways between different systems and functions, simultaneously and at only distinct desired times. Existing TIP drives and other known instrumentation and plant systems are useable with access management systems and methods, which can be used in any nuclear plant with access restrictions.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Corrosion Resistance and Mechanical Stability of Nickel Alloys in Molten-Salt Nuclear Reactors
Choosing structural materials for molten-salt nuclear reactors is a high-priority problem. The results making it possible to choose the basic structural material for a molten-salt nuclear reactor are colligated in the present article. Forced and natural circulation setups operating under reactor and laboratory conditions with molten metal-fluoride salts with different compositions have been developed at the National Research Center Kurchatov Institute: LiF–NaF–BeF{sub 2}–PuF{sub 3}, LiF–BeF{sub 2}–UF{sub 4}, and LiF–BeF{sub 2}–ThF{sub 4}–UF{sub 4}. Corrosion tests were conducted with domestically produced corrosion-resistant steel (12Kh18N10T, EP-164) and specially developed nickel-based alloys (KhN80M-VI, KhN80MTYu, KhN80MTV, et al.) as well as the alloy Hastelloy (USA), MONICR (Czechoslovakia) and EM-721 (France) in the working temperature range 600–800°C and mechanical loads up to 80 MPa on the samples.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
An analysis of the ''laminar instability'' problem in gas-cooled nuclear reactor passages.
Instability of gas-cooled nuclear reactor passages in steady laminar flow studied by time-dependent analysis
LAMINAR FLOW↗
A brief history of design studies on innovative nuclear reactors
In a short period after the success of CP1, many types of nuclear reactors were proposed and investigated. However, soon only a small number of reactors were selected for practical use. Around 1970, only LWRs with small number of CANDUs were operated in the western world, and FBRs were under development. It was about the time when Apollo moon landing was accomplished. However, at the same time, the future of human being was widely considered pessimistic and Limits to Growth was published. In the end of 1970’s the TMI accident occurred and many nuclear reactor contracts were cancelled in USA and any more contracts had not been concluded until recent years. From the reflection of this accident, many Inherent Safe Reactors (ISRs) were proposed, though none of them were constructed. A common idea of ISRs is smallness of their size. Tokyo Institute of Technology (TokyoTech) held a symposium on small reactors, SR/TIT, in 1991, where many types of small ISRs were presented. Recently small reactors attract interest again. The most ideas employed in these reactors were the same discussed in SR/TIT. In 1980’s the radioactive wastes from fuel cycle became a severe problem around the world. In TokyoTech, this issue was discussed mainly from the viewpoint of nuclear transmutations. The neutron economy became inevitable for these innovative nuclear reactors especially small long-life reactors and transmutation reactors.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Additive Manufacturing in the Nuclear Reactor Industry
This article discusses the application of additive manufacturing in the nuclear reactor industry.Additive manufacturing for nuclear applications increased significantly in the late 2010s due to rapid progress in technology, particularly with methods using metals and ceramics.Because these manufacturing technologies are uniquely data-rich, they allow for an advanced understanding of materials, and they offer the potential to predict part performance based on build data. The inherent characteristics of additive manufacturing technologies allow for rapid prototyping and geometric freedom, making it possible to implement an accelerated agile design process for advanced nuclear reactor applications. Predictive high-fidelity multiphysics simulations are crucial to fully leverage this geometric flexibility. Qualification and regulator acceptance ultimately determine the breadth and scope of applications for these technologies. The US Department of Energy Office of Nuclear Energy Transformational Challenge Reactor program integrates many of these elements to accelerate the deployment of additive manufacturing technologies to industry.
Betzler, Benjamin↗