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All sourcesNASA NTRSDOE OSTI
DOE OSTI2012

Code Qualification of Structural Materials for AFCI Advanced Recycling Reactors

This report summarizes the further findings from the assessments of current status and future needs in code qualification and licensing of reference structural materials and new advanced alloys for advanced recycling reactors (ARRs) in support of Advanced Fuel Cycle Initiative (AFCI). The work is a combined effort between Argonne National Laboratory (ANL) and Oak Ridge National Laboratory (ORNL) with ANL as the technical lead, as part of Advanced Structural Materials Program for AFCI Reactor Campaign. The report is the second deliverable in FY08 (M505011401) under the work package 'Advanced Materials Code Qualification'. The overall objective of the Advanced Materials Code Qualification project is to evaluate key requirements for the ASME Code qualification and the Nuclear Regulatory Commission (NRC) approval of structural materials in support of the design and licensing of the ARR. Advanced materials are a critical element in the development of sodium reactor technologies. Enhanced materials performance not only improves safety margins and provides design flexibility, but also is essential for the economics of future advanced sodium reactors. Code qualification and licensing of advanced materials are prominent needs for developing and implementing advanced sodium reactor technologies. Nuclear structural component design in the U.S. must comply with the ASME Boiler and Pressure Vessel Code Section III (Rules for Construction of Nuclear Facility Components) and the NRC grants the operational license. As the ARR will operate at higher temperatures than the current light water reactors (LWRs), the design of elevated-temperature components must comply with ASME Subsection NH (Class 1 Components in Elevated Temperature Service). However, the NRC has not approved the use of Subsection NH for reactor components, and this puts additional burdens on materials qualification of the ARR. In the past licensing review for the Clinch River Breeder Reactor Project (CRBRP) and the Power Reactor Innovative Small Module (PRISM), the NRC/Advisory Committee on Reactor Safeguards (ACRS) raised numerous safety-related issues regarding elevated-temperature structural integrity criteria. Most of these issues remained unresolved today. These critical licensing reviews provide a basis for the evaluation of underlying technical issues for future advanced sodium-cooled reactors. Major materials performance issues and high temperature design methodology issues pertinent to the ARR are addressed in the report. The report is organized as follows: the ARR reference design concepts proposed by the Argonne National Laboratory and four industrial consortia were reviewed first, followed by a summary of the major code qualification and licensing issues for the ARR structural materials. The available database is presented for the ASME Code-qualified structural alloys (e.g. 304, 316 stainless steels, 2.25Cr-1Mo, and mod.9Cr-1Mo), including physical properties, tensile properties, impact properties and fracture toughness, creep, fatigue, creep-fatigue interaction, microstructural stability during long-term thermal aging, material degradation in sodium environments and effects of neutron irradiation for both base metals and weld metals. An assessment of modified versions of Type 316 SS, i.e. Type 316LN and its Japanese version, 316FR, was conducted to provide a perspective for codification of 316LN or 316FR in Subsection NH. Current status and data availability of four new advanced alloys, i.e. NF616, NF616+TMT, NF709, and HT-UPS, are also addressed to identify the R&D needs for their code qualification for ARR applications. For both conventional and new alloys, issues related to high temperature design methodology are described to address the needs for improvements for the ARR design and licensing. Assessments have shown that there are significant data gaps for the full qualification and licensing of the ARR structural materials. Development and evaluation of structural materials require a variety of experimental facilities that have been seriously degraded in the past. The availability and additional needs for the key experimental facilities are summarized at the end of the report. Detailed information covered in each Chapter is given.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Variable Nuclear Electricity with Base-Load Reactor, Closed Brayton Power Cycle and Firebrick Resistance-Heated Energy Storage (Fires)

Power cycle choices are determined by markets and available technologies. Two changes are occurring in the electricity sector. First, the world is beginning to move away from fossil fuels because of concerns about climate change. What replaces fossil fuels for variable electricity production? Second, there have been extraordinary technological advances in gas turbine technology. The changes in market requirements and technology suggest the need to reexamine nuclear power cycles. One option is to couple high-temperature base-load nuclear reactors to closed Brayton cycles (helium, nitrogen, etc.) with Firebrick Resistance-Heated Energy Storage (FIRES). The reactor operates at base-load but not the power cycle. When electricity prices are low, firebrick is electrically heated to high temperatures using base-load electricity and/or electricity from the grid. The gas turbine can use nuclear heat for base-load operation and produce added peak electricity by sending the compressed heated gas through hot firebrick after nuclear heating to further raise power cycle gas temperatures and power output from the turbine-creating a nuclear plant with a thermodynamic topping (NUTOP) cycle. If such cycles can be developed, they would increase plant revenue and enable a low-carbon electricity grid. Such cycles can be coupled to Fluoride-salt-cooled High-temperature Reactors (FHRs), molten salt reactors (MSRs), high-temperature gas-cooled reactors (HTGRs), sodium-cooled fast reactors (SFRs) and lead-cooled fast reactors (LFRs). We discuss (1) market changes, (2) previous work on Nuclear Air-Brayton Combined cycles (NACC) with base-load reactors and variable power to the electricity grid and (3) options for closed Brayton cycles with variable power while the reactor operates at base load. Market (customer) requirements ultimately drive selection of power cycles. A low-carbon electricity grid requires dispatchable nuclear electricity to replace the traditional role of fossil fuels in the electricity grid. While reactors can be operated at part load, this is expensive for a capital-intensive low-operating-cost technology. Brayton power cycles with FIRES have the potential capability to provide that variable electricity while the reactor operates at base load to (1) improve economics and (2) enable a zero-carbon electrical grid. Open Brayton cycles have been investigated. Work has now started on developing closed Brayton cycles with FIRES. Only limited work has been done to date has been done for closed cycles with FIRES. (authors)

42 ENGINEERING↗
DOE OSTI2017

NBSR reactor control room upgrade - 215

The maintenance and modernization of the National Institute of Standards and Technology (NIST) Center for Neutron Research (NCNR) reactor, known as the NBSR, is an ongoing process to endure until the eventual decommissioning of the reactor. As part of the renovation effort, the reactor control system is to be upgraded. Following some initial outsourcing efforts, the NBSR reactor control system upgrade has been restructured to be implemented by in-house means. The beginning stage of the upgrade involved the development of a simulator environment based on a PC system and digitization of the majority of nuclear reactor monitoring data. The NBSR reactor surveillance data consists of approximately thirteen hundred input/output points, including analog/digital sensor signals, signal conditioners, and relay logic. The aim of the current phase is to implement a new reactor data acquisition and display system (RDS) and lay out the infrastructure for the upgrade of the reactor control system (RCS). Key design criteria are to implement a reliable, redundant, diverse, modular and candid system where any complex digital manipulations are taken outside of the control room. Basically, the data analytics and custom codes are separated from reactor operation control and decision making logic. Furthermore, the new RDS is designed such that it could be maintained internally by NCNR personnel. The RDS is unique in design and modularity, minimizing custom codes and testing, making it ideal for research reactors. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Modular high temperature gas reactor core modeling with RELAP5-3D/PHISICS – Optimization schemes for load following

The objective of this study is to develop a 3D/NK model to study the load following operation of a Modular High-Temperature Gas-Cooled Reactor (MHTGR). The selected design is the 350 MW prismatic, graphite moderated, helium cooled thermal reactor based on the MHTGR-350 transient benchmark led by the High-Temperature Gas-Cooled Reactor (HTGR) Methods Core Simulation Group at Idaho National Laboratory (INL) in the framework of the Organization for Economic Cooperation and Development Nuclear Energy Agency (OECD/NEA). This kind of reactor usually reacts very slowly to the perturbation of the core parameters due to the large amount of graphite in the core. This kind of behavior promotes the reactor stability but on the other hand limits the reactor load-following operability. Recently advances in gas reactor designs have made load-following in gas reactors a feasible and achievable goal. Modern nuclear reactors (such as the MHTGR-350) are designed to withstand the structural stresses associated with load-following. Operating the reactor in load-following mode will inevitably reduce the load factor. Although a higher load factor means more revenue and the best usage of the fuel, following the variable energy demand will increase the maximum achievable fraction of generated nuclear power, being no more limited to the base load power generation. In general, if low impact on material aging and the safe operability are demonstrated the plant economics will be minimally affected. In order to ensure that the system can be safely operated in a load-following mode an extensively study has been carried out. Some tests with linear coolant mass flow rate change demonstrated that the reactor behavior is suitable for operation in load following mode. This assumption has been confirmed by a 4 days load following transient test in which the reactor supplied the requested power with a negligible error.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Development of Heat Pipe Reactor Modeling in SAM

System Analysis Module (SAM) is under development at Argonne National Laboratory as a modern system-level modeling and simulation tool for advanced non-light water reactor safety analyses. It utilizes the object-oriented application framework MOOSE to leverage the modern software environment and advanced numerical methods available in PETSc. The capabilities of SAM are being extended to enable the transient modeling, analysis, and design of various advanced nuclear reactor systems. This report presents the development of new capabilities for modeling the heat pipe type reactor systems. The need for power at remote locations away from a reliable electrical grid is an important niche for nuclear energy. Heat pipe-cooled fast-spectrum nuclear reactors are well suited for these applications. The key feature of the heat pipe reactors is the use of heat pipes for heat removal from the reactor core. The heat pipe makes use of the phase change of the working fluid and transports a large amount of heat from the evaporator to the condensation end with very small temperature drops. In contrast to the traditional nuclear reactor system that makes use of pumped loop for extracting the thermal power, the heat pipe reactors make use of hundreds of heat pipes for removing the thermal power (including the decay heat) passively. This could potentially significantly improve the reliability and safety of the reactor systems. The essential part in the analysis of a heat pipe type reactor is the modeling of heat transport inside the heat pipe. The capability of SAM is extended in this work to enable the modeling of the conventional heat pipe. Two alternative modeling options, 2D-RZ Heat conduction and 3D-1D coupling, are developed, which are both based on the assumption that the heat pipe vapor core can be modeled as a superconductor of extremely high thermal conductivity. Both modeling options are verified with a simplified thermal resistance model. The heat pipes are also coupled with the fuel region of a prototype micro reactor under both normal and off-normal conditions. It is confirmed that both modeling options can correctly model the heat transport in the heat-pipe reactors.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Extension and Demonstration of NEAMS Multiphysics Tools to Lead-Cooled, Sodium-Cooled, and Molten Salt Fast Reactor Applications

The SHARP toolkit is a high-fidelity reactor simulation tool developed under the U.S. Department of Energy, Office of Nuclear Energy Advanced Modeling and Simulation (NEAMS) Campaign. SHARP toolkit is comprised of the neutronics module PROTEUS thermal hydraulics module Nek5000, and structural mechanics module Diablo. During FY17 and FY18, the PROTEUS and Nek5000 components of SHARP were applied to solve challenging sodium-cooled fast reactor (SFR) problems. In particular, selected hot channel factors (HCF) for a prototype metal-fueled SFR design (the AFR-100) were analyzed in high fidelity, and the “SHARP zooming capability” for SFRs was developed and demonstrated to reduce computational expense for full core problems in cases where detailed data is needed in selected fuel assemblies. After the previous success applying SHARP to challenging SFR problems, the focus in FY19 expanded to additional fast reactor applications: lead cooled fast reactors (LFR), sodium cooled fast reactors (SFR), and fast molten salt reactors (MSR). The specific technical tasks were (1) assessment of hot channel factors for LFR, for which no data currently exists, (2) implementation of a gamma transport capability in the PROTEUS solvers to extend the zooming capability to non-fueled assemblies, (3) demonstration of zooming capability in assemblies of the Versatile Test Reactor (VTR), and (4) development of a new coupled capability for simulating fast MSRs, specifically modeling precursor transport inside and outside the core. The progress on tasks (1) and (3) are discussed primarily in this document. Tasks (2) and (4) are briefly mentioned in this document, and the interested reader is referred to separate reports for detailed information. First-of-a-kind hot channel factor (HCF) estimation for LFR with high fidelity codes (PROTEUS/Nek5000) was successfully demonstrated in this study. Selected HCF were computed and compared with SFR data (AFR-100, EBR-II). The findings confirm that different reactor types, design parameters and uncertainties lead to different HCFs. Careful estimation of HCF for a specific design is necessary to obtain appropriate HCFs. In addition to improvement in HCF accuracy, high fidelity tools generate data to help the designer better understand the mechanism of the impact from these uncertainties. For example, the impact of cladding thickness manufacturing tolerance resulted in non-intuitive effects in the corner pins of the LFR assembly. This procedure of computing HCF using high fidelity models shows promise and flexibility for being repeated for any arbitrary reactor of choice, since these tools are targeted to solve many reactor types and geometries. Progress was made towards extending the previously demonstrated SHARP zooming capability to non-fueled SFR assemblies. In particular, a gamma transport capability was implemented in both high fidelity PROTEUS solvers in order to accurately account for heat deposition caused by gamma particles, which accounts for ~10% of total core power. Neutronics verification cases were carried out for a candidate Versatile Test Reactor (VTR) design using the new gamma transport capability in PROTEUS. Comparisons were made with continuous energy MCNP calculations and shown to agree well. The models for the full core design with heterogeneous control and fuel assemblies is in progress and will be completed in FY20. In the fast molten salt reactor (MSR) domain, a new capability was developed to model the flow of neutron precursors both inside and outside of the core, which impacts the effective delayed neutron fraction of the system. The PROTEUS-SN and PROTEUS-NODAL solvers were coupled to Nek5000 in order to simulate precursor flow. The workflows were demonstrated on test problems. The fast MSR modeling capability was supported under two other work packages, and comprehensive details are included in a separate report.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Zero Power Reactor Database (ZPRD) Development Plan

Past sodium-cooled fast reactors (SFR) were built with an active experimental program in place to support the design and development work. Most of the experimental facilities in the United States that were important for SFR design were shutdown in the 1980s and 1990s. Reactor licensing and construction requires any reactor design to be verified against existing reactor facilities or experimental measurements. With the absence of those experimental facilities, modern SFR projects must rely on historical measurements to demonstrate that the engineering modeling software and data being used for the new reactor design work are reliable. There has been a considerable push in the last 6 years by both DOE and commercial companies to obtain historical experimental measurements that are relevant for SFRs, in particular those with features that are important for the new reactor designs of interest. The zero power reactor experiments carried out at Argonne National Laboratory’s critical facilities (ZPR-3, ZPR-6, ZPR-9, and ZPPR) from the 1950s to the 1980s are some of the best reactor physics experiments on SFR technology that are available today. Of particular interest today are the ZPPR-15 measurements done at the ZPPR facility for the Integral Fast Reactor project in the 1980s as they are in line with most commercial and DOE interests today. In the past 10 years, the measurements done on ZPPR-15 have been processed into both Monte Carlo (MCNP) and deterministic models (MC2-3 and DIF3D) useable for validating the engineering modeling software for key parts of the SFR design work. To achieve this, a detailed model description must be created for the experiment and the experimental measurement that the engineering modeling software is to reproduce. Then, an assessment of the uncertainty on the measured quantity which considers all of the sources of uncertainty in defining the model must be obtained and documented. The models created for ZPPR-15 provide the best validation basis available today for neutronics modeling software. Reference 2 is a good resource to understand how these models were built and how the uncertainties on the measured quantities were derived. The intention of the Zero Power Reactor Database (ZPRD), hosted at frdb.ne.anl.gov, is to make available the experimental measurements and models that have been constructed to-date. Though ZPPR-15 measurements are the primary data requested for validation needs, other measurements on ZPPR, ZPR-6, and ZPR-9 in support of the Clinch River Breeder Reactor (CRBR) and Fast Test Reactor (FFTF) should also be considered important for future software validation needs. In this manuscript, the details of available measurements on ZPR-3, ZPR-6, ZPR-9, and ZPPR facilities are summarized, and a general organization of the web interface is displayed. Many of the documents associated with the measurements are export controlled information so access to the database will also have to be controlled.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Multimetallic Layered Composites (MMLCs) for Rapid, Economical Advanced Reactor Deployment (Final Report)

This project focused on the development of multi-metallic layered composites (MMLCs) for advanced fission reactor technologies. There are many instances where one alloy or material simply cannot meet all the demands thrown at it by a reactor system, or cannot allow it to perform as strongly as one would like. Instead of focusing all our effort on developing one perfect alloy, we seek to leverage the design principle of “separation of functionality,” used in many other arenas in design, to boost performance beyond single alloys alone. One illustrative example shows the power of this approach for molten salt-cooled reactors: A three meter tall, three meter diameter reactor vessel made of Incoloy 800 was quoted at $\$$500k in 2018. A Hastelloy N vessel was quoted at $\$$5M. An MMLC vessel, in which a layer of Hastelloy N would be weld-overlaid onto Incoloy 800, was quoted at $\$$700k, and it would achieve the same performance. The potential economic gains of leveraging this approach are therefore substantial. At a minimum, each MMLC would contain one core structural layer and one coolant-facing corrosion-resistant layer. Sometimes, MMLCs required buffer layers, as the structural and corrosion-resistant layers were metallurgically incompatible. In other words, they didn’t always play nice, thus separating layers compatible with both functioned as intermediaries to keep the composite together. However, in doing so we inevitably produce new interfaces, where new issues can arise. Therefore, this project focused on what happens at these interfaces from a combination of high temperatures, irradiation, corrosion, and time. After all, a reactor makes money when it is operating, and outages of any kind erode its economic viability. First, we set out to experimentally prove that MMLCs for at least two advanced reactor systems can be made, today, in US domestic facilities. In this respect we were successful – one MMLC (a Ni-201/Incoloy 800H composite) was successfully made and drawn into two-inch coolant piping. Others were attempted, though new issues relating to cracking in vanadium layers for one and radiation damage performance of the corrosion-resistant layer in another prevented us from moving further in those specific arenas – these are engineering problems which deserve continued focus after this project. Additional experimental work focused on long-term corrosion testing of the outermost layers of the salt-cooled and liquid lead-cooled MMLC concepts, which would then be fed into predictions of how long the MMLCs could last. Next, computational (thermodynamics and atomistic) simulation studies studied how much we expect the interfaces to “blend,” due to the mixing action of neutron irradiation. This eats into both the margin for the structural layer of each MMLC, as dilution from the corrosion-resistant layer into the structural layer would decrease the total load-bearing capacity of an MMLC of finite size. On the other hand, dilution of the corrosion-resistant layer into the structural layer further reduced the margin of corrodible material, reducing the lifetime of the MMLC or necessitating extra thickness to be imparted to the MMLC to meet its functional requirements. Work here focused on irradiation-induced segregation to predict new phases which may embrittle the MMLCs, as well as quantifying irradiation-induced mixing at each interface. The results showed that mixing is expected, but it is both steady and therefore predictable, and not lifetime-limiting for most MMLC concepts – it simply has to be accounted for in calculations of reactor performance when utilizing an MMLC. Then, full-core simulations using the experimentally-derived corrosion data, the computationally discovered irradiation-induced mixing data (partially validated by experiment), and existing, benchmarked core designs for large and small sized reactor concepts (one salt-cooled, one lead-cooled) were conducted to quantify any expansion of reactor operating envelopes achieved by utilizing these MMLCs. This new framework, called REX (Reactor Envelope Expansion), incorporates a combination of core neutronics, thermal hydraulics, and the material performance data derived from this project to see how using an MMLC expands advanced fission reactor operating envelopes. It was discovered that in some cases, MMLC utilization does indeed increase the maximum operating temperatures and cycle lengths of reactor concepts, while in other cases it does not. Finally, our tech-to-market (T2M) strategy was not necessarily to create specific embodiments of MMLCs for immediate sale (because getting into the nuclear market is incredibly slow and laden with regulation, this is a long-term goal), but rather immediate stimulation of US industry using the design approach of MMLCs derived from this project. In this respect we were successful, as one of the PhD students funded on this project co-founded Allium Engineering, Inc., which created a stainless steel / low-alloy steel MMLC to function as chloride corrosion-resistant rebar for embedding into concrete structures. Allium Engineering continues to be successful, having recently opened their first factory as of this writing.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Heavy Water Moderated, Molten Uranium Cooled Reactor MCNP Model

Two of the major challenges facing the nuclear industry are the large amounts of nuclear waste produced by the current light water reactors and the risk of nuclear proliferation. Most proposed solutions for the waste problem involve fast breeder reactors, but that doesn't help with the nuclear proliferation problem. It is generally understood that a uranium fueled (U-238 + U-235) thermal reactor cannot be a breeder because there aren't enough neutrons. That analysis does not include the effects of fast fission of U-238. This is not surprising because the percentage of fast fissions in existing thermal reactor designs is low, around 3% (in Candu reactors) or less. Table I shows the neutrons involved in 1,000 fissions in a thermal reactor for a conversion ratio of 1.000 when the fuel is mostly U-238 and the fissile content is low enough so that most fast fissions are of U-238. The top section of the table is for U-235 in U-238, the bottom section is for Pu-239 in U-238. The bottom line in each section shows the number of neutrons available for all other losses (escape from the reactor, absorption in non fuel components, and absorption in xenon, samarium, and other fission products). If the percentage of fast fissions of U-238 is 10 or 15 % a thermal breeder may be feasible. This is largely independent of the percentage of fissile content of the fuel. If the reactor starting fissile content is mostly U-235, as the U-235 is replaced with Pu-239 the number of available neutrons increases even if the conversion ratio is only 1.000. The percentage of fast fissions in a reactor is a question of reactor geometry, as the minimum dimension of the fuel elements increases the average distance a fast neutron travels in the fuel before exiting increases and with it the probability of fast fission of U-238.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

A Summary of the Department of Energy’s Advanced Demonstration and Test Reactor Options Study

An assessment of advanced reactor technology options was conducted to provide a sound comparative technical context for future decisions by the U.S. Department of Energy (DOE) concerning these technologies. Strategic objectives were established that span a wide variety of important missions, and advanced reactor technology needs were identified based on recent DOE and international studies. A broad team of stakeholders from industry, academia, and government was assembled to develop a comprehensive set of goals, criteria, and metrics to evaluate advanced irradiation test and demonstration reactor concepts. Point designs of a select number of concepts were commissioned to provide a deeper technical basis for evaluation. The technology options were compared on the bases of technical readiness and the ability to meet the different strategic objectives. Using the study's evaluation criteria and metrics, an independent group of experts from industry, universities, and national laboratories scored each of the point designs. Pathways to deployment for concepts of varying technical maturities were estimated for the different demonstration systems with regard to cost, schedule, and possible licensing approaches. This study also presents the trade-offs that exist among the different irradiation test reactor options in terms of the ability to conduct irradiations in support of advanced reactor research and development and to serve potential secondary missions. The main findings of the study indicate the following: (1) for industrial process heat supply, a high-temperature gas-cooled reactor is the best choice because of the high outlet temperature of the reactor and its strong passive and inherent safety characteristics; (2) for resource utilization and waste management, a sodium-cooled fast reactor (SFR) is best because of the use of a fast flux to destroy actinides; (3) to realize the advantages of a promising but less-mature technology, a fluoride salt-cooled high-temperature reactor and a lead-cooled fast reactor fare about the same; (4) for fulfilling the needs of a materials test reactor, a SFR is considered best because of its ability to produce high fast flux, incorporate test loops, and provide additional large volumes for testing.

Advanced reactors↗
DOE OSTI2016

Heat Transfer in a Molten Natural Uranium Nuclear Reactor With Thermal Storage

Previous papers describe a method for controlling a thermal nuclear reactor by moderator displacement between the reactor core and the reflector. The low neutron loss and high control range of this method enable a design for a molten natural uranium fueled and cooled heavy water moderated reactor. The design is geometrically similar to CANDU (Canadian Deuterium Uranium) reactors but with fewer, larger, vertically oriented fuel tubes. The large fuel tubes have a mixed spectrum which is largely thermal at the periphery of the tubes with a very hard spectrum in the large central region of the tube. This configuration provides a fast fission rate of around 15% (mostly of U-238) and an initial conversion ratio of around 1.15. It is a hypothetical reactor because it is unclear what materials could actually be used to fabricate the fuel tubes and other primary coolant plumbing. The molten uranium serves both as the fuel and the primary coolant and as such is continuously circulated through the fuel tubes in the reactor core, the primary heat exchange, and back to the the reactor core. This means that at any point in time all the fuel has the same isotopic configuration no matter where it is in the reactor or heat exchanger. Over time the isotopic configuration changes, but it changes uniformly for all the fuel. Uranium melts at 1135 deg. C, so a proposed temperature range for the circulating molten uranium is 1200 deg. C at the exit of the heat exchanger and entry to the core fuel tubes and 1400 deg. C at the exit from the core fuel tubes and the entry into the heat exchanger. The advantages of this design include: 1. Reduced nuclear proliferation concern because of natural uranium fuel. 2. High conversion ratio and large fuel mass provides 30 to 100 year fuel life. 3. No heavy metal waste because fissile content increases over the fuel life. 4. High output temperatures are suitable for high efficiency turbine generators. This paper discusses the heat transfer in the molten natural uranium fueled reactor described above and in the cited paper in the simple case and when the reactor is combined with a thermal storage unit which is composed of ordinary salt (NaCl). The proposed molten uranium fueled and cooled nuclear reactor works particularly well when coupled with a molten salt heat storage unit, a process heat user, and high Efficiency air-Brayton cycle turbine generators which can also add energy to the hot air output of the tin/air heat exchanges by burning some natural gas or hydrogen. The combination can provide highly variable electrical output to the grid to accommodate large daily and seasonal mismatches between electrical demand and supply from variable sources such as wind and solar. Future work includes modeling of the temperatures throughout the molten salt heat storage unit under varying charge and discharge cycles.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2013

BDDR, a new CEA technological and operating reactor database

The new application BDDR (Reactor database) has been developed at CEA in order to manage nuclear reactors technological and operating data. This application is a knowledge management tool which meets several internal needs: -) to facilitate scenario studies for any set of reactors, e.g. non-proliferation assessments; -) to make core physics studies easier, whatever the reactor design (PWR-Pressurized Water Reactor-, BWR-Boiling Water Reactor-, MAGNOX- Magnesium Oxide reactor-, CANDU - CANada Deuterium Uranium-, FBR - Fast Breeder Reactor -, etc.); -) to preserve the technological data of all reactors (past and present, power generating or experimental, naval propulsion,...) in a unique repository. Within the application database are enclosed location data and operating history data as well as a tree-like structure containing numerous technological data. These data address all kinds of reactors features and components. A few neutronics data are also included (neutrons fluxes). The BDDR application is based on open-source technologies and thin client/server architecture. The software architecture has been made flexible enough to allow for any change. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

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↗
DOE OSTI2016

Reactivity Coefficient Calculation for a Low-Power LEU Research Reactor

Core design studies on a 20-MW thermal power low enriched uranium (LEU) fueled research reactor are underway at the NIST Center for Neutron Research (NCNR) with the primary purpose of providing quality neutron beams for scientific research. The 'tank-in-pool' design pattern is adopted for the new reactor: A compact core composed of two horizontally split halves are surrounded by a cylindrical D{sub 2}O reflector tank, and the tank is itself surrounded by a large H{sub 2}O pool, which functions as thermal and biological shielding. The split core consists of 18 MTR-type fuel elements which are evenly distributed into two regions. The LEU fuel used in the study is the U{sub 3}Si{sub 2}-Al dispersion fuel with U-235 enrichment of 19.75%, the only NRC certified LEU fuel in the U.S. The core is cooled and moderated by H{sub 2}O, and separated from D{sub 2}O by zircaloy core boxes. Hafnium control blades are utilized as both criticality and safety control elements for the reactor. They are controlled by a mechanical driver located at the bottom of the core but with the fully withdrawn positions at the top of the core. In a standard reactor calculation, it is necessary to evaluate the reactivity coefficient with respect to some physical parameters such as fuel or moderator temperature. For nearly all reactor designs, negative reactivity coefficients are required for safety. As a light-water cooled thermal reactor, the new reactor is designed to work in an under-moderated mode with the most achievable reactivity. The core is required to have negative reactivity coefficients under any normal or abnormal circumstances such as over power transient, loss of coolant or reflector, etc. Therefore the reactivity coefficient addresses reactor safety concerns as well as provides valuable quantitative parameters for reactor safety analyses. In this summary, three typical reactivity coefficients for the low-power LEU core are calculated using the Monte-Carlo code MCNP-6: the moderator temperature coefficient (MTC), the fuel temperature coefficient (FTC), and the void coefficient (VC). A direct perturbation methodology is applied to calculate all these coefficients. The perturbed case is produced by manually perturbing a single physical property with respect to the reference case (the one with normal operational conditions), as the multiplication factor (i.e. k{sub eff}) of the perturbed case is calculated by MCNP. The corresponding reactivity coefficient is thereby determined by dividing the reactivity change over the associated parameter changes. Core design studies are underway at NIST to develop a low-power LEU fueled beam reactor to advance the neutron source capability at NCNR for the next century. The reactivity effect caused by moderator temperature, fuel temperature and void is investigated though a direct perturbation approach using the MCNP code. The MTC, FTC, and VC for a representative startup core is obtained and the negative effect characteristics of these coefficients are observed. The reactivity effects under several postulated abnormal scenarios are also investigated. The overall negative feedback to the reactivity at each situation is confirmed. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Feasibility study on ultralong-cycle operation and material performance for compact liquid metal-cooled fast reactors: a review work

This paper reviews the feasibility of ultralong-cycle operation on a compact liquid metal-cooled fast reactor (LMR) firstly by assessing the operation of a long-life fast reactor core and secondly by evaluating material performance in respect to both long-cycle operation and compact-size fast reactor. Many kinds of reactor concepts have been proposed, and LMR and small modular reactor (SMR) are the issued leading technologies for generation four (Gen-IV) reactor system development. The breed-and-burn strategy was proposed as a core burning strategy to operate a long cycle, and it has been evaluated in this paper with two reactor concepts: constant axial shape of neutron flux, nuclide densities, and power shape during life of energy and ultralong cycle fast reactor. In addition, Super-Safe, Small, and Simple and small modular fast reactor, compact LMR concepts, have been simulated to evaluate their long-life operation strategies. For the other practical issues, the materials for fuel, coolant, and structure have been identified and some of them are selected to have their performance optimized specifically for compact LMR with a long-cycle operation. It is believed that this comprehensive review will propose a proper direction for future reactor development and will be followed by the next step research for a complete reactor model with the other reactor components.

breed-and-burn↗
DOE OSTI2023

A reduced-order modeling of a tubular solar reactor for long duration thermochemical energy storage

The storage of solar energy in a solid form, referred to as a “solar fuel”, can be achieved through a process known as endothermic solar thermochemistry. This process transforms the absorbed solar energy into a stable and retrievable form that can be stored for extended periods of time. This paper presents a low–order heat transfer model of a counter–current tubular falling bed reactor designed to produce thermally reduced magnesium manganese oxide pellets for long duration thermochemical energy storage. The energy required for the endothermic reduction was supplied by concentrated solar energy or renewable electricity via indirect heating of the gas and solid reactants flowing in a ceramic tube. The counter-current gas flow enhances the mixing of the solid particles with the heat recuperation zone, allowing the gas and particles to enter and exit the tubular reactor close to room temperature. Further, the reactor was vertically oriented and was heated circumferentially by an adjustable level heat flux along a finite segment of its length. The temperature distribution of the reactor in response to transient changes along the tube was modeled by considering conduction, convection, and radiation heat transfer. Governing equations for the heat transfer model were solved by discretizing the reactor tube into a finite number of control volumes and using an energy balance for the heat exchange between the reactor wall, gas, and particles within the control volume. The energy absorbed during this endothermic reaction was modeled numerically by fitting the data of the chemical conversion rate with the corresponding temperature of particles in the heating zone. The numerical model has been experimentally validated using a reactor prototype made of a 121.92 cm alumina tube heated by a 7kW electric tube–furnace. The alumina tube receives magnesium manganese oxide pellets of 3.66±0.516 mm in diameter from the top, and a counter–current gas flow from the bottom. The reactor wall temperature was monitored by six thermocouples installed along the reactor tube length. The experimental procedure was numerically simulated, and the temperature variation along the reactor tube was compared with a matrix of experimental runs for a range of particles mass flowrates (0.75–1.25g/s) and corresponding gas flowrates (36–65 SLPM). The reactor system was heated gradually from room temperature to a steady state temperature of 1673K, and then cooled down to room temperature. The heating and cooling processes were simulated, and the numerical and experimental results were compared throughout processes. The numerical model showed similar trends to the experimental results, with an error of 0.69 to 7.9% for the particle inlet and 0.7 to 7.9% for the gas inlet during steady-state operation. The proposed numerical model can be implemented as a simplified physical model to design a feedback control system to regulate reactor temperature.

14 SOLAR ENERGY↗
DOE OSTI2026

Depletion benchmark for a high-assay low-enriched uranium fuel experiment in the advanced test reactor

Reactor physics depletion benchmarks for high-assay low-enriched uranium (HALEU) fuel are limited in number. In particular, there is limited data for HALEU benchmarks for U-10Mo (uranium-10% molybdenum) plate fuel that is being developed for use in the United States’ high performance research reactors including the Advanced Test Reactor (ATR), Advanced Test Reactor Critical Facility (ATR-C), High Flux Isotope Reactor (HFIR), Massachusetts Institute of Technology Reactor (MITR), University of Missouri Research Reactor (MURR), National Bureau of Standards Reactor (NBSR). These six reactors currently operate with highly enriched uranium dispersed fuel in an aluminum matrix. In support of conversion to a HALEU fuel, qualification of U-10Mo formed into a monolithic foil is being performed. Fuel qualification involves irradiating fuel specimens in the ATR. The irradiation tests provide an opportunity to benchmark depletion capabilities of reactor physics codes in support of the ATR operation, as well as develop benchmarks that can be used by other institutions to benchmark other reactor physics codes. This paper documents the development of a benchmark model of the irradiation of the ATR Full-size plate In center flux trap Position 7 (AFIP-7) experiment using the depletion codes MC21 and Advanced Dimensional Depletion for Engineering of Reactors (ADDER).

Nielsen, Joseph W. [Idaho National Laboratory (INL↗
DOE OSTI2020

Computational fluid dynamics analysis of char conversion in Sandia’s pressurized entrained flow reactor

Design and analysis of practical reactors utilizing solid feedstocks rely on reaction rate parameters that are typically generated in lab-scale reactors. Evaluation of the reaction rate information often relies on assumptions of uniform temperature, velocity, and species distributions in the reactor, in lieu of detailed measurements that provide local information. This assumption might be a source of substantial error, since reactor designs can impose significant inhomogeneities, leading to data misinterpretation. Spatially resolved reactor simulations help understand the key processes within the reactor and support the identification of severe variations of temperature, velocity, and species distributions. In this work, Sandia’s pressurized entrained flow reactor is modeled to identify inhomogeneities in the reaction zone. Tracer particles are tracked through the reactor to estimate the residence times and burnout ratio of introduced coal char particles in gasifying environments. The results reveal a complex mixing environment for the cool gas and particles entering the reactor along the centerline and the main high-speed hot gas reactor flow. Furthermore, the computational fluid dynamics (CFD) results show that flow asymmetries are introduced through the use of a horizontal gas pre-heating section that connects to the vertical reactor tube. Computed particle temperatures and residence times in the reactor differ substantially from the idealized plug flow conditions typically evoked in interpreting experimental measurements. Furthermore, experimental measurements and CFD analysis of heat flow through porous refractory insulation suggest that for the investigated conditions (1350 °C, <20 atm), the thermal conductivity of the insulation does not increase substantially with increasing pressure.

47 OTHER INSTRUMENTATION↗