A domain-specific analysis system for examining nuclear reactor simulation data for light-water and sodium-cooled fast reactors

Building a new generation of fission reactors in the United States presents many technical and regulatory challenges. One important challenge is the need to share and present results from new high-fidelity, high-performance simulations in an easily usable way. Since modern multiscale, multi-physics simulations can generate petabytes of data, they will require the development of new techniques and methods to reduce the data to familiar quantities of interest (e.g., pin powers, temperatures) with a more reasonable resolution and size. Furthermore, some of the results from these simulations may be new quantities for which visualization and analysis techniques are not immediately available in the community and need to be developed. This paper describes a new system for managing high-performance simulation results in a domain-specific way that naturally exposes quantities of interest for light water and sodium-cooled fast reactors. It describes requirements to build such a system and the technical challenges faced in its development at all levels (simulation, user interface, etc.). An example comparing results from two different simulation suites for a single assembly in a light-water reactor is presented, along with a detailed discussion of the system's requirements and design.Comment: Article on NiCE's Reactor Analyzer. 23 pages. Keywords: modeling, simulation, analysis, visualization, input-outpu

Elements of nuclear safety – Research reactors

This publication gives a global overview of the diversity and complementarity of research reactors, some of which have been or are still being used to conduct experiments that are essential for the development and operation of nuclear power reactors, including in relation to safety issues. This work highlights the many uses of these reactors, which have very different designs, use highly varied quantities of radioactive substances with varying levels of risk for safety and radiation protection, and which — in many cases because they are old or have been shut down — require appropriate measures to control the ageing or obsolescence of some of their equipment, as well as, on an organisational and human level, to ensure that they continue to be operated safely. For some research reactors, safety and radiation protection aspects must be considered, taking into account that two types of operators are present at the same time within these reactors: reactor operating personnel and operators in charge of experimental devices using neutrons from the reactor for fundamental or applied research purposes. There are two specific chapters on the safety standards established under the aegis of the IAEA for research reactors and on serious accidents, notably those involving criticality and reactivity, in research reactors. The second part of the work focuses on French research reactors, including the regulations and official documents applicable to these reactors, on lessons learned in France from significant events and accidents — as well as abroad, such as the Fukushima Daiichi nuclear power plant accident in 2011 — on the consideration of reactivity accidents in the design of French research reactors, and on the ten-yearly safety reviews carried out in France

Army Officer Corps Science, Technology, Engineering and Mathematics (STEM) Foundation Gaps Place Countering Weapons of Mass Destruction (CWMD) Operations at Risk – Part 1

This is the first of three articles from the authors describing the risk to Joint Operations incurred by an Army that is vulnerable to the STEM challenges faced in a great power competition involving CWMD operations. In this article, we describe the problem. In articles two and three of the series, we will elaborate on the problem utilizing the Joint Publication 3-0 as our guide and recommend solutions to address this gap

Materials for Sustainable Nuclear Energy: A European Strategic Research and Innovation Agenda for All Reactor Generations

Nuclear energy is presently the single major low-carbon electricity source in Europe and is overall expected to maintain (perhaps eventually even increase) its current installed power from now to 2045. Long-term operation (LTO) is a reality in essentially all nuclear European countries, even when planning to phase out. New builds are planned. Moreover, several European countries, including non-nuclear or phasing out ones, have interests in next generation nuclear systems. In this framework, materials and material science play a crucial role towards safer, more efficient, more economical and overall more sustainable nuclear energy. This paper proposes a research agenda that combines modern digital technologies with materials science practices to pursue a change of paradigm that promotes innovation, equally serving the different nuclear energy interests and positions throughout Europe. This paper chooses to overview structural and fuel materials used in current generation reactors, as well as their wider spectrum for next generation reactors, summarising the relevant issues. Next, it describes the materials science approaches that are common to any nuclear materials (including classes that are not addressed here, such as concrete, polymers and functional materials), identifying for each of them a research agenda goal. It is concluded that among these goals are the development of structured materials qualification test-beds and materials acceleration platforms (MAPs) for materials that operate under harsh conditions. Another goal is the development of multi-parameter-based approaches for materials health monitoring based on different non-destructive examination and testing (NDE&T) techniques. Hybrid models that suitably combine physics-based and data-driven approaches for materials behaviour prediction can valuably support these developments, together with the creation and population of a centralised, “smart” database for nuclear materials

Current Research in Nuclear Reactor Technology in Brazil and Worldwide

The aim of this book is to disseminate state-of-the-art research and advances in the area of nuclear reactors technology. The book was divided in two parts.Topics discussed in the first part of this compilation include: experimental investigation and computational validation of thermal stratification in PWR reactors piping systems, new methods in doppler broadening function calculation for nuclear reactors fuel temperature, isothermal phase transformation of uranium-zirconium-niobium alloys for advanced nuclear fuel, reactivity Monte Carlo burnup simulations of enriched gadolinium burnable poison for PWR fuel, utilization of thermal analysis technique for study of uranium-molybdenum fuel alloy, probabilistic safety assessment applied to research reactors, and a review on the state-of-the art and current trends of next generation reactors. The second part includes: thermal hydraulics study for a ultra high temperature reactor with packed sphere fuels, benefits in using lead-208 coolant for fast reactors and accelerator driven systems, nuclear power as a basis for future electricity production in the world: Generation III and IV reactors, nanostructural materials and shaped solids for improvement and energetic effectiveness of nuclear reactors safety and radioactive wastes, multilateral nuclear approach to nuclear fuel cycles, and a cold analysis of the Fukushima accident

Description of new meso-scale models and their implementation in fuel performance codes

This deliverable illustrates the new (2.0) version of the SCIANTIX meso-scale code, developed within Task 5.2 of the PATRICIA Project, highlighting first the code structure and its numerical features. Then, the SCIANTIX models for various physics involved in the inert gas behaviour are described in detail along with their corresponding separate-effect validation database and validation results. The coupling of SCIANTIX with integral, pin-level fuel performance codes is also introduced, presenting the different strategy and interface details for the coupling with the TRANSURANUS and GERMINAL fuel performance codes. Finally, conclusions and future perspectives are provided, mentioning several envisaged developments targeted in the framework of multiple research initiatives at a European and international level, and outlining the strategy foreseen for further developments of the code (in both its stand-alone and coupled fashion)

MULTI-PHYSICS DESIGN AND ANALYSES OF LONG LIFE REACTORS FOR LUNAR OUTPOSTS

Future human exploration of the solar system is likely to include establishing permanent outposts on the surface of the Moon. These outposts will require reliable sources of electrical power in the range of 10\u27s to 100\u27s of kWe to support exploration and resource utilization activities. This need is best met using nuclear reactor power systems which can operate steadily throughout the long ~27.3 day lunar rotational period, irrespective of location. Nuclear power systems can potentially open up the entire lunar surface for future exploration and development. Desirable features of nuclear power systems for the lunar surface include passive operation, the avoidance of single point failures in reactor cooling and the integrated power system, moderate operating temperatures to enable the use of conventional materials with proven irradiation experience, utilization of the lunar regolith for radiation shielding and as a supplemental neutron reflector, and safe post-operation decay heat removal and storage for potential retrieval. In addition, it is desirable for the reactor to have a long operational life. Only a limited number of space nuclear reactor concepts have previously been developed for the lunar environment, and these designs possess only a few of these desirable design and operation features. The objective of this research is therefore to perform design and analyses of long operational life lunar reactors and power systems which incorporate the desirable features listed above. A long reactor operational life could be achieved either by increasing the amount of highly enriched uranium (HEU) fuel in the core or by improving the neutron economy in the reactor through reducing neutron leakage and parasitic absorption. The amount of fuel in surface power reactors is constrained by the launch safety requirements. These include ensuring that the bare reactor core remains safely subcritical when submerged in water or wet sand and flooded with seawater in the unlikely event of a launch abort accident. Increasing the amount of fuel in the reactor core, and hence its operational life, would be possible by launching the reactor unfueled and fueling it on the Moon. Such a reactor would, thus, not be subject to launch criticality safety requirements. However, loading the reactor with fuel on the Moon presents a challenge, requiring special designs of the core and the fuel elements, which lend themselves to fueling on the lunar surface. This research investigates examples of both a solid core reactor that would be fueled at launch as well as an advanced concept which could be fueled on the Moon. Increasing the operational life of a reactor fueled at launch is exercised for the NaK-78 cooled Sectored Compact Reactor (SCoRe). A multi-physics design and analyses methodology is developed which iteratively couples together detailed Monte Carlo neutronics simulations with 3-D Computational Fluid Dynamics (CFD) and thermal-hydraulics analyses. Using this methodology the operational life of this compact, fast spectrum reactor is increased by reconfiguring the core geometry to reduce neutron leakage and parasitic absorption, for the same amount of HEU in the core, and meeting launch safety requirements. The multi-physics analyses determine the impacts of the various design changes on the reactor\u27s neutronics and thermal-hydraulics performance. The option of increasing the operational life of a reactor by loading it on the Moon is exercised for the Pellet Bed Reactor (PeBR). The PeBR uses spherical fuel pellets and is cooled by He-Xe gas, allowing the reactor core to be loaded with fuel pellets and charged with working fluid on the lunar surface. The performed neutronics analyses ensure the PeBR design achieves a long operational life, and develops safe launch canister designs to transport the spherical fuel pellets to the lunar surface. The research also investigates loading the PeBR core with fuel pellets on the Moon using a transient Discrete Element Method (DEM) analysis in lunar gravity. In addition, this research addresses the post-operation storage of the SCoRe and PeBR concepts, below the lunar surface, to determine the time required for the radioactivity in the used fuel to decrease to a low level to allow for its safe recovery. The SCoRe and PeBR concepts are designed to operate at coolant temperatures ≤ 900 K and use conventional stainless steels and superalloys for the structure in the reactor core and power system. They are emplaced below grade on the Moon to take advantage of the regolith as a supplemental neutron reflector and as shielding of the lunar outpost from the reactors\u27 neutron and gamma radiation. The SCoRe and PeBR concepts are designed with cores divided into six and three sectors, respectively. The sectors are thermally and neutronically coupled but hydraulically decoupled. Each sector is served by a separate cooling loop(s), with independent energy conversion and heat rejection radiator panels. This combination of a sectored core and power system integration with multiple loops avoids single point failures in reactor cooling, energy conversion, and heat rejection. These unique attributes would allow the reactor power system to continue operation, but at lower power, in the event one of the sectors experiences a Loss of Coolant (LOC) or a Loss of Cooling (LOCo). The power system could thus maintain a vital supply of electrical power to the lunar outpost for crew life support. The performed multi-physics design and performance analyses of the SCoRe show that increasing the diameter of the UN fuel rods increases the core excess reactivity. The larger diameter rods, however, increase the NaK-78 coolant flow bypass near the walls of the core sectors. Scalloping the dividing walls of the core sectors produces a more even flow distribution. The use of 151Eu spectral neutron poison additive to the UN fuel ensures subcriticality during a water submersion accident, for the compact SCoRe core, with the highest excess reactivity and lowest mass. The redesigned Solid-Core Sectored Compact Reactor (SC-SCoRe) with a monolithic solid core of Oxide Dispersion Strengthened Molybdenum (ODS-Mo) achieves a long operational life of 21 years at a nominal power of 1,000 kWth. The high thermal conductivity ODS-Mo core structure allows the reactor to continue safe operation in the event that one of the core sectors experiences a LOC of LOCo. The ODS-Mo solid core readily conducts heat generated in that sector to the adjacent core sectors, to be removed by the flowing liquid metal coolant. The SCoRe power system with SiGe energy conversion is fully passive and load following. In addition, the decay heat is removed safely and passively following shutdown of the reactor at end of life. Neutronics and analyses show that the PeBR can achieve a very operational life of 66 years at a nominal thermal power of 471 kWth by fueling the reactor core on the Moon. This full power operational life is beyond what is possible with a reactor fueled at launch like the SC-SCoRe. Neutronics safety analysis of the fuel pellets transport canisters for the PeBR shows that they are made highly subcritical in a water submersion accident. This is achieved using a combination of favorable geometry and neutron absorbers. The DEM fuel loading simulation of the PeBR on the lunar surface demonstrates that the PeBR core sectors can be successfully fueled in lunar gravity. Post-operation storage analyses of the SC-SCoRe and PeBR concepts show that the radioactivity in the fuel decays away to a sufficiently low level within 300 years. The radiation field around the post-operational reactor at that time is low enough to allow humans to safely handle and retrieve the cores, for potential recovery of the valuable quantities of 235U remaining in the fuel