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DOE OSTI2016

Modeling of a Passive Endothermic Reaction Cooling System

Passive reactor core cooling systems can mitigate the effects of station blackout conditions and avoid accidents similar to what occurred at the Fukushima Dai-ichi reactors. Advanced nuclear reactor designs such as the Advanced Heavy Water Reactor (AHWR) and others include passive cooling systems. The AHWR utilizes water storage tanks that can remove decay heat for more than 50 days after station blackout. While these systems do not rely on electricity, they do rely on passive valves to function during an emergency situation. The purpose of the present work is to determine the feasibility of a passive endothermic reaction cooling system (PERCS). The PERCS relies on an endothermic reaction that does not proceed until it reaches a threshold temperature. Thus, a vessel filled with chemicals that undergo an endothermic reaction can be directly connected to the reactor core without any valves and only react once accident conditions raise the temperature past the threshold temperature. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2015

Heat up and potential failure of BWR upper internals during a severe accident

In boiling water reactors, the steam dome, steam separators, and dryers above the core are comprised of approximately 100 tons of stainless steel. During a severe accident in which the coolant boils away and exothermic oxidation of zirconium occurs, gases (steam and hydrogen) are superheated in the core region and pass through the upper internals. Historically, the upper internals have been modeled using severe accident codes with relatively simple approximations. The upper internals are typically modeled in MELCOR as two lumped volumes with simplified heat transfer characteristics, with no structural integrity considerations, and with limited ability to oxidize, melt, and relocate. The potential for and the subsequent impact of the upper internals to heat up, oxidize, fail, and relocate during a severe accident was investigated. A higher fidelity representation of the shroud dome, steam separators, and steam driers was developed in MELCOR v1.8.6 by extending the core region upwards. This modeling effort entailed adding 45 additional core cells and control volumes, 98 flow paths, and numerous control functions. The model accounts for the mechanical loading and structural integrity, oxidation, melting, flow area blockage, and relocation of the various components. The results indicate that the upper internals can reach high temperatures during a severe accident; they are predicted to reach a high enough temperature such that they lose their structural integrity and relocate. The additional 100 tons of stainless steel debris influences the subsequent in-vessel and ex-vessel accident progression.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2017

Orificed Flow Results for a Two-Phase Natural Circulation Loop

Researchers at University of Wisconsin - Madison have constructed a 1/4 scaled experimental facility to study the operation of a two-phase natural circulation decay heat removal system. The Reactor Cavity Cooling System (RCCS) is a decay heat removal safety system designed for high temperature generation IV reactors. The system removes decay heat by encircling the outside of the walls of the pressure/containment vessel with cooling channels. These channels are filled with either air in an open loop configuration or water in a closed loop configuration. Radiative and convective heat loads from the pressure vessel induce a buoyancy driven flow in the risers. In accident conditions, the water-based system reaches saturation and the resulting steam is vented off further removing heat until the water inventory is depleted. The water based system has been under investigation at University of Wisconsin - Madison for variations in performance under various flow restriction cases.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2017

Heat up and failure of BWR upper internals during a severe accident

In boiling water reactors, the shroud dome, separators, and dryers above the core are made of approximately 100,000 kg of stainless steel. During a severe accident in which the coolant boils away and exothermic oxidation of zirconium occurs, gases (steam and hydrogen) are superheated in the core region and pass through the upper internals. In this scenario, the upper internals can also be heated by thermal radiation from the hot degrading core. Historically, models of the upper internals have been relatively simple in severe accident codes. The upper internals are typically modeled in MELCOR as two lumped volumes with simplified heat transfer characteristics and no structural integrity considerations, and with limited ability to oxidize, melt, and relocate. The potential for and the subsequent impact of the upper internals to heat up, oxidize, fail, and relocate during a severe accident was investigated. A higher fidelity representation of the shroud dome, steam separators, and steam driers was developed in MELCOR v1.8.6 by extending the core region upwards. The MELCOR modeling effort entailed adding 45 additional core cells and control volumes, 98 flow paths, and numerous control functions. The model accounts for the mechanical loading and structural integrity, oxidation, melting, flow area blockage, and relocation of the various components. Consistent with a previous study, the results indicate that the upper internals can reach high temperatures during a severe accident sufficient to lose their structural integrity and relocate. Finally, the additional 100 metric tons of stainless steel debris influences the subsequent in-vessel and ex-vessel accident progression.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2025

Final Prototype Microreactor Transportation Safety Program

This report fulfills the fiscal year 2025 M2 Milestone M2AT-25PN0802042, Final Microreactor Transportation Safety Program Planning Framework. Microreactors are compact reactors capable of producing less than 50 megawatts of electrical energy. Typically, these reactors are factory-fabricated and designed to be easily transportable by truck, rail, vessel, or air. Microreactor designs often assume that the unit can be transported containing either unirradiated or irradiated fuel. Interest in microreactors is driven by several factors, including the need to generate power at remote locations, military installations, and facilities such as data centers, and in areas recovering from natural disasters.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2019

First Step in Validating the Radiological Inventories of the Current French Nuclear Fleet - 19000

To anticipate the evolution of the mechanical characteristics of its PWR vessels, EDF developed a vessels fluence monitoring program. This program was based on vessel samples and neutron dosimeters deployed between the fuel assemblies and vessels. These samples and dosimeters are regularly analyzed in laboratories. To calculate the radioactive inventories, we have developed a calculation scheme which includes a 3D mapping of the neutron flux, the calculation of the radiological inventories of the structures of interest, and finally the validation of the results. The validation needs samples to propose calculation/measurement ratios. Unfortunately before the final shut down, the internal and vessel samplings are not possible. To anticipate the decommissioning of the current French nuclear fleet, this paper shows how the samples of the vessels fluence monitoring program can be used to validate the calculated radiological inventories. Using state-of-the-art simulations (i.e. 3D geometric model, 3D neutron sources, 3D neutron propagation, operational fuel loading coupled with effective power distributions) we calculated the radiological inventories of some vessel fluence monitoring program samples. The comparisons between calculated and measured results were done on 'standard chemical elements'. To qualify a chemical element as 'standard', many criteria must be satisfied such as a high concentration level, a low uncertainty about concentration, activation into an expanded cross section range, production of radionuclides having significant half-lives and finally no difficulties measuring the radionuclides produced. The retained 'standard' is Fe which produces, under neutrons flux, Fe-55. The comparisons show that the calculations slightly overestimate the measurements. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2018

Status of Metallic Structural Materials for Molten Salt Reactors

In the US and around the world, significant interest surrounds development and deployment of nuclear systems using molten salts to transfer heat. Several configurations are being explored, including the use of fluoride salts either as the primary coolant of a reactor with solid fuel or with the fissile material dissolved in the salt. Molten Salt Reactors (MSRs) with fluoride salt typically have a thermal neutron spectrum. There are also fast-spectrum reactors that could use chloride salts to transfer heat. As a result of operation of the Molten Salt Reactor Experiment (MSRE) at Oak Ridge National Laboratory during the 1960s, a system with uranium fuel dissolved in a fluoride-salt medium is the most well-established technology. A nickel alloy with about 6%Cr and molybdenum as a solid solution strengthening agent was developed and deployed for all structural applications in this reactor. The alloy was commercialized as Hastelloy N. During these experiments and in post-decommissioning characterization of material behavior, it was determined that the most significant challenges for structural materials are embrittlement, from helium introduced by transmutation of Ni, and corrosion and grain boundary embrittlement from the fission product tellurium. The MSRE had an outlet temperature of approximately 650°C. The mechanical properties of Hastelloy N are not sufficient to support long-term operation of an MSR above this temperature. Qualification of a material for use in the ASME Boiler and Pressure Vessel Code, Section III, “Rules for Construction of Nuclear Facility Components,” Division 5, “High Temperature Reactors,” will facilitate licensing with the Nuclear Regulatory Commission. Hastelloy N has not been qualified for use in nuclear construction, and significant additional characterization would be required for Code qualification. Given that this alloy is susceptible to He embrittlement and has limited high-temperature strength, it is not recommended that Code qualification be pursued. Instead, it is recommended that a systematic development program be initiated to develop new nickel alloys that contain a fine, stable dispersion of intermetallic particles to trap He at the interface between the matrix and particle and with increased solid solution strengthening from addition of refractory elements. Extensive screening of attractive alloy compositions for elevated temperature strength, microstructural stability, weldability, and resistance to He embrittlement (characterized using ion implantation) will lead to an alloy down-selection for commercialization and Code qualification.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

Extended-MHD simulations of disruption mitigation via massive gas injection in SPARC

Recent developments to the M3D-C1 code enable higher fidelity modeling of disruptions, and can be applied in the design verification of reactor-scale tokamaks. Among these new capabilities is a method to mesh conducting vessel structures such as coils and passive plates, packing of the toroidal mesh around gas injectors, as well as anisotropic resistivity inside the vessel structures. We present extended-magnetohydrodynamic (MHD) simulations of disruption mitigation via massive gas injection (MGI) in SPARC. The goal of this study is to inform the disruption mitigation layout of SPARC and aid in the design of an effective gas injector configuration. Fully three-dimensional simulations with M3D-C1 are carried out for various injector configurations with the primary goal of determining the effect of different MGI parameters on heat loads and vessel forces. The simulations include a model for impurity ionization, recombination, advection and radiation, as well as spatially resolved conducting structures around the plasma. A localized mixture of deuterium and neon with a small toroidal and poloidal width is injected in up to six locations. We demonstrate that M3D-C1 can model a rapid shutdown via MGI using narrow and more realistic gas plumes than in previous simulations. As a result of the q = 1 surface in the SPARC baseline case a sawtooth is observed early in the simulations. Despite the sawtooth and the onset of edge MHD instabilities, the impurity distribution remains localized around the injector locations, but enables a radiative shutdown of the plasma. We find that using the maximum of six gas injectors results in a lower peaking factor and leads to a more even distribution of radiation toroidally than using two injectors.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2016

Progress Report on Alloy 617 Notched Specimen Testing

Creep behavior of Alloy 617 has been extensively characterized to support the development of a draft Code Case to qualify Alloy 617 in Section III division 5 of the ASME Boiler and Pressure Vessel Code. This will allow use of Alloy 617 in construction of nuclear reactor components at elevated temperatures and longer periods of time (up to 950°C and 100,000 hours). Prior to actual use, additional concerns not considered in the ASME code need to be addressed. Code Cases are based largely on uniaxial testing of smooth gage specimens. In service conditions, components will generally be under multi axial loading. There is also the concern of the behavior at discontinuities, such as threaded components. To address the concerns of multi axial creep behavior and at geometric discontinuities, notched specimens have been designed to create conditions representative of the states that service components experience. Two general notch geometries have been used for these series of tests: U notch and V notch specimens. The notches produce a tri axial stress state, though not uniform across the specimen. Characterization of the creep behavior of the U notch specimens and the creep rupture behavior of the V notch specimens provides a good approximation of the behavior expected of actual components. Preliminary testing and analysis have been completed and are reported in this document. This includes results from V notch specimens tested at 900°C and 800°C. Failure occurred in the smooth gage section of the specimen rather than at the root of the notch, though some damage was present at the root of the notch, where initial stress was highest. This indicates notch strengthening behavior in this material at these temperatures.

36 MATERIALS SCIENCE↗

DOE OSTI2018

Notch Effects on the Short Term Rupture Properties of Alloy 617 Base Metal and Weldments

Alloy 617 is currently undergoing approval for inclusion in the ASME Boiler and Pressure Vessel Code in Section III Division 5 for use in high temperature nuclear reactors. Prior to use in construction of these reactors, however, it must be approved by the Nuclear Regulatory Commission. Several concerns relating to the effects of geometric discontinuities, such as notches and multi-axial stress states (as opposed to uniaxial stress states typically present in laboratory run creep tests), must first be addressed. Alloy 617 has been found to be notch-strengthening in short term (1000-2500 hour) creep testing for both base metal and weld metal and will fail preferentially in a straight gauge section rather than at a sharp notch, provided that the applied stresses are similar in both the straight gauge and notch sections. Additionally, specimens designed with semi-circular notches (or U notches, as they are referred to in this report) to impose multi-axial stress states on the specimen during creep-testing were found to either maintain similar creep-rupture-lives as uniaxial stress specimens tested at similar conditions, or have creep-rupture-lives that were significantly longer. This work shows that notches and multi-axial stress states do not appear to raise concerns with relation to the use of design rules created for Alloy 617 using smooth, straight-gauge test specimens and uniaxial creep tests.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2019

Development of the Technical Basis of a Unified Viscoplastic Model of 316H Stainless Steel for Incorporation into ASME Division 5

This report describes the development of a unified viscoplastic model describing the deformation of 316 stainless steel. The model was developed for incorporation into Section III, Division 5 of the ASME Boiler and Pressure Vessel Code covering the design and fabrication of Class A high temperature nuclear reactor components. The current version of the Code does not provide reference inelastic models, leaving the responsibility of defining an adequate model up to either the plant owner or designer. This approach represents a substantial barrier to using the ASME design by inelastic analysis methods. As historical practice demonstrates these methods produce much more efficient designs, compared to the design by elastic analysis methods, eliminating this barrier could result in safer, more effective, and more efficient structural designs thereby lowering the cost of future advanced reactors. This report summarizes the process of gathering experimental data, calibrating the candidate inelastic model against that data, and validating the model. The report also summarizes work on a new method of calculating the sensitivities of history dependent nonlinear constitutive models that was used to calibrate the 316H model using gradient-based optimization.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2019

Initial study of notch sensitivity of Grade 91 using mechanisms motivated crystal plasticity finite element method

This report describes the development of a crystal plasticity finite element method (CPFEM) model for the multiaxial creep rupture of Grade 91 steel. The report demonstrates that core blocks in heat pipe reactor designs will see significantly different stress states than those typically found in conventional advanced reactor components where structures can often be represented as thin-walled pressure vessels. Current design practice uses effective stress measures calibrated to biaxial creep test data to account for the effect of stress multiaxiality on rupture life. Because these models rely on biaxial data they may not be accurate for the truly 3D, triaxial states found in heat pipe microreactor core blocks. A physically-based model, like the CPFEM model developed here, may be able to more accurately represent creep rupture under triaxial stresses when compared to conventional, empirical methods. This report describes the development, implementation, and validation of such a model and then applies the model to predict the effect of stress triaxiality on rupture in Grade 91 steel. The model results are used to assess the effective stress measures currently used in high temperature design codes and makes specific recommendations on improving these effective stresses to better account for the effect of 3D stress states on rupture in Grade 91. The validation and then implementation of these recommendations in design codes and standards could lead to safer, more effective heat pipe microreactor core block designs.

36 MATERIALS SCIENCE↗

DOE OSTI2025

Cold Spray Cobalt Magnetostrictive Electromagnetic Acoustic Transducers for High Temperature Structure Monitoring

The Department of Energy’s Advanced Sensors and Instrumentation program seeks to develop and qualify advanced sensors for the nuclear industry. Reliable high temperature and high radiation sensors for detection and characterization of structural flaws in pipes, vessels, and structurally critical components is a weakness for both conventional light water reactors with coolant T-hot approaching 350oC, and for advanced reactors with T-hot temperatures in excess of 500 to 800oC. Magnetostrictive Electromagnetic Acoustic Transducers using a cold spray cobalt coating have been proposed as a sensor design that can withstand these kinds of temperatures and radiation levels to serve as online sensors to detect flaws before cracks, pits, or erosion/corrosion flaws progress to through-wall failures. This report tests cold spray cobalt as part of a magnetostrictive EMAT for high temperature service. Cobalt is known to have strong magnetostrictive properties however the effect of cold spray application is not well studied. This program was surprised to discover that cold sprayed cobalt exhibited little or no magnetostrictive behavior until it was thermally annealed. Following annealing to 650oC however, cold spray cobalt did exhibit a magnetostrictive response. Work to date prior to this milestone report publication showed that magnetostrictive EMAT was successfully tested to 400oC with an alnico permanent magnet. The program plans to extend testing with an electromagnet to higher temperatures. This follow-on work will be reported under subsequent publications or as a revision to this report.

Glass, Samuel W.↗

NASA NTRS2013

Nuclear Thermal Rocket Element Environmental Simulator (NTREES) Upgrade Activities

A key technology element in Nuclear Thermal Propulsion is the development of fuel materials and components which can withstand extremely high temperatures while being exposed to flowing hydrogen. NTREES provides a cost effective method for rapidly screening of candidate fuel components with regard to their viability for use in NTR systems. The NTREES is designed to mimic the conditions (minus the radiation) to which nuclear rocket fuel elements and other components would be subjected to during reactor operation. The NTREES consists of a water cooled ASME code stamped pressure vessel and its associated control hardware and instrumentation coupled with inductive heaters to simulate the heat provided by the fission process. The NTREES has been designed to safely allow hydrogen gas to be injected into internal flow passages of an inductively heated test article mounted in the chamber.

Emrich, William↗

DOE OSTI2016

Core Design Characteristics of the Fluoride Salt-Cooled High Temperature Demonstration Reactor

Fluoride salt-cooled high temperature reactors (FHRs) are a promising reactor technology option with significant knowledge gaps to implementation. One potential approach to address those technology gaps is via a small-scale demonstration reactor with the goal of increasing the technology readiness level (TRL) of the overall system for the longer term. The objective of this paper is to outline a notional concept for such a system, and to address how the proposed concept would advance the TRL of FHR concepts.Development of the proposed FHR Demonstration Reactor (DR) will enable commercial FHR deployment through disruptive and rapid technology development and demonstration. The FHR DR will close remaining gaps to commercial viability. Lower risk technologies are included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated within an acceptable budget and schedule. Important capabilities that will be demonstrated by building and operating the FHR DR include core design methodologies; fabrication and operation of high temperature reactors; salt procurement, handling, maintenance, and ultimate disposal; salt chemistry control to maximize vessel life; tritium management; heat exchanger performance; pump performance; and reactivity control. The FHR DR is considered part of a broader set of FHR technology development and demonstration efforts, some of which are already underway. Nonreactor test efforts (e.g., heated salt loops or loops using simulant fluids) can demonstrate many technologies necessary for commercial deployment of FHRs. The FHR DR, however, fulfills a crucial role in FHR technology development by advancing the technical maturity and readiness level of the system as a whole.

Brown, Nicholas↗

DOE OSTI2019

A User Guide to PARET/ANL

PARET was originally created in 1969 at what is now Idaho National Laboratory (INL), to analyze reactivity insertion events in research and test reactor cores cooled by light or heavy water, with fuel composed of either plates or pins. The use of PARET is also appropriate for fuel assemblies with curved fuel plates when their radii of curvatures are large with respect to the fuel plate thickness. The PARET/ANL version of the code has been developed at Argonne National Laboratory (ANL) under the sponsorship of the U.S. Department of Energy/NNSA since the inception of the Reactor Conversion Program. Since Reduced Enrichment for Research and Test Reactors (RERTR) began in 1978, PARET/ANL has been benchmarked to experimental data including SPERT testing, and used to determine the expected transient behavior of a large number of reactors regardless of enrichment both inside and outside the Reactor Conversion Program. This document provides the pertinent information for the use of PARET/ANL Version 7.6. PARET/ANL models the various fueled regions of a reactor core as channels. Each of these channels consists of a single flat fuel plate/pin (including cladding and, optionally, a gap) with water coolant on each side. In slab geometry the coolant channels for a given fuel plate are of identical dimensions (mirror symmetry), but they can be of different thickness in each channel. There can be many channels, but each channel is independent and coupled only through reactivity feedback effects to the whole core. The time-dependent differential equations that represent the system are replaced by an equivalent set of finite-difference equations in space and time, which are integrated numerically. PARET/ANL uses fundamentally the same numerical scheme as RELAP5 for the time-integration of the point-kinetics equations. The one-dimensional thermal-hydraulic model includes temperature-dependent thermal properties of the solid materials, such as heat capacity and thermal conductivity, as well as the transient heat production and heat transfer from the fuel meat to the coolant. Temperature- and pressure-dependent thermal properties of the coolant such as enthalpy, density, thermal conductivity, and viscosity are also used in determining parameters such as friction factors and heat transfer coefficients. The code first determines the steady-state solution for the initial state. Then the solution of the transient is obtained by integration in time and space. Multiple heat transfer, DNB and flow instability correlations are available. The code was originally developed to model reactors cooled by an open loop, which was adequate for rapid transients in pool-type cores. An external loop model appropriate for Miniature Neutron Source Reactors (MNSR’s) was also added to PARET/ANL to model natural circulation within the vessel, heat transfer from the vessel to pool and heat loss by evaporation from the pool. PARET/ANL also contains models for decay heat after shutdown, control rod reactivity versus time or position, time-dependent pump flow, and loss-of-flow event with flow reversal as well as logic for trips on period, power, and flow. Feedback reactivity effects from coolant density changes and temperature changes are represented by tables. Feedback reactivity from fuel heat-up (Doppler Effect) is represented by a four-term polynomial in powers of fuel temperature. Photo-neutrons produced in beryllium or in heavy water may be included in the point-kinetics equations by using additional delayed neutron groups.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2022

A User Guide to PARET/ANL

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2015

A User Guide to PARET/ANL (V.7.6) (Rev.1)

PARET was originally created in 1969 at what is now Idaho National Laboratory (INL), to analyze reactivity insertion events in research and test reactor cores cooled by light or heavy water, with fuel composed of either plates or pins. The use of PARET is also appropriate for fuel assemblies with curved fuel plates when their radii of curvatures are large with respect to the fuel plate thickness. The PARET/ANL version of the code has been developed at Argonne National Laboratory (ANL) under the sponsorship of the U.S. Department of Energy/NNSA, and has been used by the Reactor Conversion Program to determine the expected transient behavior of a large number of reactors. PARET/ANL models the various fueled regions of a reactor core as channels. Each of these channels consists of a single flat fuel plate/pin (including cladding and, optionally, a gap) with water coolant on each side. In slab geometry the coolant channels for a given fuel plate are of identical dimensions (mirror symmetry), but they can be of different thickness in each channel. There can be many channels, but each channel is independent and coupled only through reactivity feedback effects to the whole core. The time-dependent differential equations that represent the system are replaced by an equivalent set of finite-difference equations in space and time, which are integrated numerically. PARET/ANL uses fundamentally the same numerical scheme as RELAP5 for the time-integration of the point-kinetics equations. The one-dimensional thermal-hydraulic model includes temperature-dependent thermal properties of the solid materials, such as heat capacity and thermal conductivity, as well as the transient heat production and heat transfer from the fuel meat to the coolant. Temperature- and pressure-dependent thermal properties of the coolant such as enthalpy, density, thermal conductivity, and viscosity are also used in determining parameters such as friction factors and heat transfer coefficients. The code first determines the steady-state solution for the initial state. Then the solution of the transient is obtained by integration in time and space. Multiple heat transfer, DNB and flow instability correlations are available. The code was originally developed to model reactors cooled by an open loop, which was adequate for rapid transients in pool-type cores. An external loop model appropriate for Miniature Neutron Source Reactors (MNSR’s) was also added to PARET/ANL to model natural circulation within the vessel, heat transfer from the vessel to pool and heat loss by evaporation from the pool. PARET/ANL also contains models for decay heat after shutdown, control rod reactivity versus time or position, time-dependent pump flow, and loss-of-flow event with flow reversal as well as logic for trips on period, power, and flow. Feedback reactivity effects from coolant density changes and temperature changes are represented by tables. Feedback reactivity from fuel heat-up (Doppler Effect) is represented by a four-term polynomial in powers of fuel temperature. Photo-neutrons produced in beryllium or in heavy water may be included in the point-kinetics equations by using additional delayed neutron groups.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗