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Special Purpose Application Reactors: Systems Integration Decision Support

Microreactors, also known as very-Small Modular Reactors (vSMRs) or Special Purpose Reactors (SPRs), are being considered for use in unique applications where other methods of megawatt level energy production are uneconomical or unavailable. For the purposes of this report, a microreactor is defined as meeting the following criteria: • Factory manufacturable • Easily transportable by truck, plane, train, and/or ship • Produce no more than 20 megawatt thermal (MWth) energy, in order to qualify as Hazard Category 2 under 10CFR830 • Maintains neutronic simplicity, allowing semi- or fully-autonomous operation In the United States (U.S.), there are two general types of microreactors under development: high-temperature gas reactors (HTGRs) and heat pipe reactors. These two reactor types may, due to their differences in working fluid temperatures, be more effectively suited to different use cases. Additionally, they will require different interfacing components depending on their intended use. This report identifies a broad range of use cases for micro-reactors. Additionally, multiple reactor technologies are examined with an emphasis on the reactor sub-systems and interconnections. Using this information, a decision framework is developed to guide future efforts.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Development of an Automated Decision-Making Tool for Supervisory Control System

This technical report was generated as a product of the Supervisory Control for Multi-Modular Small Modular Reactor (SMR) Plants project within the Instrumentation, Control and Human-Machine Interface technology area under the Advanced Small Modular Reactor (AdvSMR) Research and Development Program of the US Department of Energy. The report documents the definition of strategies, functional elements, and the structural architecture of a supervisory control system for multi-modular AdvSMR plants. This research activity advances the state of the art by incorporating real-time, probabilistic-based decision-making into the supervisory control system architectural layers through the introduction of a tiered-plant system approach. The report provides background information on the state of the art of automated decision-making, including the description of existing methodologies. It then presents a description of a generalized decision-making framework, upon which the supervisory control decision-making algorithm is based. The probabilistic portion of automated decision-making is demonstrated through a simple hydraulic loop example.

42 ENGINEERING↗

Public Views on Nuclear Facility Siting and Radioactive Waste Management in the United States: Methodology and Response Reference Report for the 2017 Energy and Environment Survey

This report presents the questions and responses to a nationwide survey of U.S. residents taken May 2017. The focus of the 2017 survey is public perceptions on (1) nuclear energy and new nuclear technology, such as small modular reactors; (2) the risks and benefits of nuclear energy; (3) different options for managing spent nuclear fuel (SNF); and (4) the role of individuals, organizations and elected officials in locating SNF facilities. Highlights of survey results related to nuclear energy found the U.S. public (1) wants nuclear energy to continue to contribute 20% of total electrical power produced for the next 20 years; (2) is more supportive of adding new reactors at existing sites than at new sites; and (3) is more supportive of small modular reactors than traditional reactors. Highlights related to risks and benefits of nuclear energy include (1) the public believes that terrorist attacks are the greatest threat and that energy independence is the greatest benefit; (2) 56% of the public believes that nuclear power has both high benefit and high risk; (3) the 2011 accident at the Fukushima nuclear facility continues to negatively influence the perceived benefits and risks of nuclear energy except for those who perceive nuclear power as low risk and high benefit; (4) building new reactors is supported by those who perceive nuclear power as low risk and high benefit, and opposed by those who perceive it as high risk and low benefit. Highlights related to SNF management options include (1) support remains higher for permanent disposal than either consolidated interim storage or continued on-site storage; (2) support for private initiatives for interim storage in New Mexico and Texas are higher than generic storage and disposal options; (3) the general public is more likely to increase support if a majority of the local and state residents support an SNF facility. Highlights related to the role of individuals, organizations, and elected officials include (1) 54% of the public agrees that if an SNF facility is deemed safe by regulators, political considerations should not derail the location; (2) 47% of the US public agrees that non-governmental organizations should be able to provide input to locating an SNF facility; (3) 36% of the public believes that the federal government should lead siting of an SNF facility while 25% believes private companies should lead; (4) 81% of the public believes local residents and farmers/ranchers should have a say in the final decision on locating a SNF facility and 56% believe that they should have veto power; and (5) only 27% of the public believes the state governor should have a say and veto power.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

BISON analyses of TRISO fuel performance, its dependence on time-at-temperature, and possible implications for fuel design and qualification

The Advanced Gas Reactor Fuel Development and Qualification (AGR) program has established a substantial technical foundation to support private entry into the U.S. high-temperature gas-cooled reactor market. However, emerging tristructural isotropic (TRISO)-fueled reactor applications include small modular reactors and microreactors with longer fuel residence times, which may expose fuels to higher time-at-temperature (TAT) values than were explored by the AGR program. Increased TAT could affect diffusive and thermomechanical behaviors such as Pd penetration, fission gas release, creep, and fission product transport. In this work, we applied multiscale best-estimate BISON fuel performance modeling to assess these effects within a representative design space based on the AGR-5/6/7 experiment and analyzed trends in predicted particle and compact fuel performance metrics with possible implications for near-term fuel design and qualification. BISON unambiguously predicted that TRISO fuel performance is sensitive to TAT. Increasing TAT was not predicted to increase the magnitude of failure-inducing tangential stresses in particle coating layers. Predictions obtained using a mechanistic model for Pd penetration indicated that penetration depth does not depend strongly on TAT. While these observations suggest that AGR testing provides a conservative upper bound for the steady-state operation of TRISO particles at lower powers and higher residence times, BISON also predicted that the release of poorly retained Ag would increase with TAT. Because these analyses applied models to extrapolate beyond the available experimental data, the authors recommend performing targeted experiments to confirm these predictions. Nevertheless, these predictions may provide reactor developers with enough confidence to make near-term design decisions associated with the potential fuel performance trade-offs of increasing TAT.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Global Trends of ASME "N-Stamp" Certifications for Nuclear Component Vendors

Historically, nuclear component manufacturing vendors, from small businesses through large conglomerates, have felt compelled to obtain an American Society for Mechanical Engineers (ASME) Nuclear Certification, known colloquially as an "N-stamp", to assure supply chain quality standards that will be acceptable to regulators and safety concerns. Since the N-stamp quality standard is a U.S.-origin code, combined with the apparent decline in the U.S. nuclear industry alongside the growth of the Asian nuclear industry, there is the question of whether the rest of the world, including new entrants to the nuclear industry, also regard N-stamp as a needed certification. This study addresses this question through analysis of the entire N-stamp database of holders, and former holders, of N-stamp certificates of all types and for all regions worldwide from 1989-2020 (the dates available in the database). From this 30 years of data, we find that actually U.S.-based vendors still consistently obtain the largest number of N-stamps worldwide over all time periods, but also find that the countries participating in the N-stamp certification process has broadly expanded beyond just North America, Japan, S. Korea and Western Europe (the primary N-stamp recipients before the mid-2000's). We produced global heats maps and bar charts to illustrate our findings, as well as further investigation into why the data shows changes over time and region. We note that nuclear entities involved with Soviet-type reactors do not participate in the N-stamp process, but instead pursue the Russian version PNAE, which is substantially similar to the ASME code. We conclude that at least from the N-stamp database, the United States nuclear component manufacturing industry is alive and well, although there have been some consolidations, and that the ASME N-stamp appears to still be a valued certificate worldwide, including in China which now ranks second only to the United States in obtaining N-stamp certificates in recent years. We further note that the vendors of new reactor types, in particular High Temperature Gas-Cooled Reactors (HTGRs) and Small Modular Reactors (SMRs), are actively engaged with ASME (and other U.S.-based nuclear standards bodies such as the American Nuclear Society and Nuclear Energy Institute) to coordinate updates to the ASME N-Stamp criteria to ensure applicability of the code for these new designs. Implications of these findings include the following: The global use of the U.S.-origin N-stamp certification supports the view that, despite the decline of the U.S. nuclear industry, the United States remains an esteemed global leader in the area of nuclear safety. As the U.S. Government works to revitalize the U.S. nuclear industry, especially in the area of exports, it may be beneficial to leverage the global standing of the N-stamp certification. The findings indicate that the N-stamp database would be a useful tool for the U.S. Government to use to track the growth of the civil nuclear industry in foreign countries, under certain circumstances. The Excel-format N-stamp database produced as part of this study may be a useful tool for this purpose. N-stamp data may be a useful tool for foreign governments to use to identify nuclear manufacturers within their own country, especially to identify "targets" for outreach on nuclear export control compliance. The U.S. Government could carry this message to foreign partners during bilateral engagements or as part of Nuclear Suppliers Group (NSG) discussions on industry outreach.

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Assessment of Microreactor Safety Analysis Challenges and Recommendations for Utilization of the Comprehensive Reactor Analysis Bundle

To enable the broad deployment of microreactors in fundamentally new application regimes (i.e., mobile and autonomous operations), their safety must be indisputable in terms of possessing inherent resistance to severe offsite dose consequences. Therefore, mechanistic beyond-design-basis event source term calculations that demonstrate a sufficiently large margin of safety will be required to accommodate these new application regimes, which have no history of commercial regulation. Even for traditional reactor operation configurations, safety analysis expertise and familiarity with accident phenomena and conditions in microreactors—specifically those with heat pipe primary cooling arrangements—are lacking compared with other advanced reactor concepts and small modular reactors. Recently, modeling and simulation tools to account for unique heat pipe design aspects have been developed by Sandia National Laboratories with MELCOR and by the US Department of Energy’s (DOE’s) Office of Nuclear Energy Advanced Modeling and Simulation Program with BlueCRAB. However, further demonstration and assessment of potential knowledge gaps are needed to support these codes’ broad usage by the microreactor community. Through the DOE Microreactor Program, an initial assessment of these two tools and guidance on how an evaluation model could be constructed was performed and is reported herein. Moreover, a proposed approach for demonstrating an evaluation model using these two tools is outlined.

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Coronado Generating Station Repowering Evaluation - Siting Evaluation

The purpose of this initial siting evaluation is to consider the suitability of the Salt River Project (SRP) owned land at/near Coronado Generating Station (CGS) for a nuclear generating station (including advanced light water reactors, light water small modular reactors, and non-light-water designs1) and to assess if there are any exclusionary or avoidance factors associated with coal to nuclear repowering at CGS. This initial siting evaluation considers several characteristics (e.g., environmental conditions, seismic concerns, site footprint, water use, etc.) and highlights favorable/preferred characteristics as well as potential risks to feasibility. The results are intended to inform SRP on the strengths and weaknesses associated with the CGS site and inform the selection of candidate nuclear technologies for the CGS site that align with SRP’s mission and business objectives. This report relies on industry-recognized siting guidance, including Electric Power Research Institute’s (EPRI’s) Siting Guide (Reference 1) 2 and Nuclear Regulatory Commission’s (NRC’s) Regulatory Guide (RG) 4.7 (Reference 3), as well as nuclear domain expertise within the Gateway for Accelerated Innovation in Nuclear (GAIN), MPR Associates, Inc. (MPR), and the Idaho National Laboratory (INL).

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NuScale Power Module Instrumentation

The safe and efficient operation of any nuclear power plant (NPP) depends on accurate and timely measurement of the primary system temperature, pressure, level, flow, and neutron flux. Integral pressurized water reactor (iPWR) or small modular reactors (SMRs) present unique challenges to instrumentation and control (I&C) sensors and their maintenance strategies. More specifically, the calibration and response time of these sensors may have to be verified periodically to ensure that they maintain their required degree of accuracy and speed of response. A review of technical reports and safety analysis documentation of the NuScale Power Module (NPM) by Analysis and Measurement Services Corporation (AMS), together with NuScale engineers, confirms that the I&C sensors within the NPM will need to be testable as installed to verify their static and dynamic performance at plant operating conditions. Sensor placement and installation as well as the process conditions expected in natural circulation integral SMRs like the NPM are very different from those in large scale NPPs with conventional primary system loop piping. In particular, reactor coolant system (RCS) flows are more complex and flow rates much lower, the average containment temperature is higher, nuclear radiation levels in some areas are greater, and the accessibility of sensors for hands-on maintenance is very limited. These challenges among others related to I&C sensors can be overcome by adapting existing I&C sensor test methods, developing new techniques for non-conventional I&C sensors, and incorporating online monitoring (OLM) technologies into the I&C architecture of the plant to verify the calibration and response time of sensors within the NPM prior to startup, during operation, and during subsequent refueling outages. These methods and technologies will be demonstrated and validated through a research and development (R&D) grant recently awarded to AMS by the U.S. Department of Energy (DOE) to facilitate timely deployment of the first NuScale SMR in the United States and to enable efficient I&C maintenance strategies during the life of the plant. Lastly, this paper presents the R&D plan for the DOE project with a focus on temperature measurement needs of NuScale.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

Steam Generator Model Design Parameter Sensitivity Study Using Advanced Optimization Tools

This study focuses on design parameter sensitivity studies pertaining to a steam generator (SG) model, using both Python and machine-learning tools. The SG model is a mathematical representation (including fluid flow and heat transfer equations/models/correlations) of a steam-generating unit in a pressurized water reactor (PWR)-type small modular reactor (SMR) system. Design studies involve changing the model’s input design parameters (e.g., temperature, pressure, mass flow rate) to observe the resulting effects on the output of the system (e.g., heat transfer coefficient [HTC], Nusselt number, heat transfer performance). Sensitivity studies analyze the degree to which system output and/or desired parameters (e.g., HTC or heat transfer performance) are sensitive to changes in input parameters. By using machine-learning tools such as the Risk Analysis Virtual Environment (RAVEN) developed at Idaho National Laboratory (INL), detailed design parametric sensitivity studies and model optimization were performed. Six input parameters—namely, the pressure, temperature, and mass flow rate for the inlet of the primary-side (hot fluid) and secondary-side (cold fluid) of the SG—were randomly perturbed via RAVEN’s Monte Carlo Sampler module, using uniform distributions (±1% relative changes). The analysis results give valuable insights into SG system performance and optimization, and provide justification for researching optimized sensor placement to effectively monitor and obtain experimental data.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Steam Generator Model Design Parameter Sensitivity Study Using Advanced Optimization Tools

This study focuses on design parameter sensitivity studies pertaining to a steam generator (SG) model, using both Python and machine-learning tools. The SG model is a mathematical representation (including fluid flow and heat transfer equations/models/correlations) of a steam-generating unit in a pressurized water reactor (PWR)-type small modular reactor (SMR) system. Design studies involve changing the model’s input design parameters (e.g., temperature, pressure, mass flow rate) to observe the resulting effects on the output of the system (e.g., heat transfer coefficient [HTC], Nusselt number, heat transfer performance). Sensitivity studies analyze the degree to which system output and/or desired parameters (e.g., HTC or heat transfer performance) are sensitive to changes in input parameters. By using machine-learning tools such as the Risk Analysis Virtual Environment (RAVEN) developed at Idaho National Laboratory (INL), detailed design parametric sensitivity studies and model optimization were performed. Six input parameters—namely, the pressure, temperature, and mass flow rate for the inlet of the primary-side (hot fluid) and secondary-side (cold fluid) of the SG—were randomly perturbed via RAVEN’s Monte Carlo Sampler module, using uniform distributions (±1% relative changes). The analysis results give valuable insights into SG system performance and optimization, and provide justification for researching optimized sensor placement to effectively monitor and obtain experimental data.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Design of a High-Assay Low-Enriched Uranium Tri-Structural Isotropic Critical Experiment for Advanced Reactor Validation

High-assay low-enriched uranium (HALEU) fuel is a key component of many small modular reactor designs. Critical experiments are an important way to understand the neutronic performance of systems by obtaining nuclear data validations through measurements. Data reduce uncertainty and risk by showing that systems respond as predicted to changes such as temperature, subsequently advancing the overall technology readiness level of the materials within. Numerous critical experiments have been performed at the National Criticality Experiments Research Center (NCERC) operated by Los Alamos National Laboratory at the Nevada National Security Site since it became operational in 2011. However, the first experiment with HALEU fuel did not occur until 2024. Through extensive engineering, the experiment described in this paper was successfully designed and executed for the Comet vertical lift assembly at NCERC to perform measurements with HALEU tri-structural isotropic fuel that will assist in validation of nuclear data and computational modeling of small modular reactors for years to come.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Engineered Solutions Group's SMR Containment Cable and Electrical Penetration Assembly System

GAIN Project CRADA Number NFE-21-08839, entitled “SMR Containment Cable and EPA System” was initiated by the partnership of Oak Ridge National Laboratory (ORNL) and Engineered Solutions Group (ESG) to test an ESG-designed Electrical Penetration Assembly and Containment Cabling System Qualified for not only legacy LWR designs, but also Small Modular Reactors designs currently being designed by several different suppliers. This project was undertaken to fill the equipment gap of EPA and Cabling Systems that require much more severe environmental requirements than legacy plant applications present due to their smaller containment volumes that result in high energy densities compared to legacy designs. This high energy density results in severe accident environments and more severe normal operating conditions as well. We developed two approaches to qualify Electrical Penetration Assembly (EPA) and Containment Cabling Systems for SMRs and Advanced Reactors. We take into consideration the more severe environmental parameters found with SMR designs. The system will need to meet a qualification test program addressing wear/cyclic aging, potential radiation exposure, thermal aging, vibration aging, thermal cycling, seismic qualification, electrical fault testing (per IEEE 317) and accident simulation. The equipment must meet the requirements of 10CFR50.49, GDC 50 in 10CFR50 Appendix A, and 10CFR50 Appendix J. NRC Regulatory Guides (RGs) identify an acceptable way of meeting regulatory requirements. RGs frequently endorse a standard for meeting these requirements. Specific to this review, equipment would be qualified in accordance with the following IEEE Standards. • IEEE 317-2013 (Electrical Penetration Assemblies), which is endorsed by RG 1.63, Rev. 3, • IEEE 323-2003 and the more current IEC/IEEE 60780-323 (Environmental Qualification of 1E Equipment) which is endorsed by RG 1.89. IEEE 323-2003 is endorsed by RG 1.209, • IEEE 344-2020 (Seismic Qualification. The 2013 version endorsed by RG 1.100, Rev. 4, with exceptions), • IEEE 383-2015 (Electrical Cables) (which is endorsed by RG 1.189 Rev. 4 and the -2003 version endorsed by RG 1.211 rev. 0), • IEEE 572-2019 (Electrical Connectors and Assemblies), which is endorsed by 1.156 Rev 1, and • [IEEE 1202 (endorsed by RG 1.189) would normally be applicable but the advanced cable designs are impervious to this cable flame test.] The primary goal of such a program is to provide an EPA design that can meet the qualification requirements for all legacy light water reactor plants currently operating as well as new plant designs including light water Small Modular Reactors. Thus, these requirements are applicable to plants licensed under 10CFR50 and 10CFR52. Other reactor designs may be evaluated, and this test system and qualification method applied to those applications if the requirements would satisfy the requirements of the intended plant. A secondary benefit of this work is to document some of the history and background in these requirements as there have been recent delays in an SMR licensing process due to NRC Requests for Additional Information in this subject matter area.

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Assessment of Regulatory Technology Gaps for Advanced Small Modular Sodium Fast Reactors

As part of an ongoing Regulatory Technology Development Plan (RTDP) for advanced small modular reactors, a review of the Sandia National Laboratories (SNL) report Sodium Fast Reactor Safety and Licensing Research Plan was conducted to identify technology-related barriers that would impact the licensing of a small sodium fast reactor (SFR), which could be resolved with technology development performed in conjunction with key additional regulatory interactions. While the general conclusions of the SNL report are still valid for small SFRs, a subsequent evaluation of the report recognized several areas that would benefit from regulatory clarification of requirements, as described in the following table.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

High Fidelity Digital Twins for BWRX-300 Critical Systems

This project developed and demonstrated digital-twin technology for advanced nuclear systems under the ARPA-E GEMINA program, with a focus on the BWRX-300 small modular reactor. The primary objective was to reduce operation and maintenance (O&M) costs through predictive health monitoring and model-based fault detection. The project combined MIT’s high-fidelity physics models and machine-learning and digital-twin algorithms, and GE Hitachi’s digital-twin algorithms and operational expertise to build physics-based digital twins capable of real-time system monitoring and maintenance scheduling.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Use of Solid Pyrex Rod for Conceptual Soluble Boron Free SMR

Various small modular reactor (SMR) designs have been proposed by the nuclear vendors and universities in order to address the energy needs for developing countries and remote communities. Most of them target for the long-term fuel cycle of 4 years or more to achieve high availability. In addition, soluble boron free operation is proposed to simplify the primary system, to enhance the safety characteristics by inherent negative moderator temperature coefficients, etc. The Pyrex rods are used to hold-down the excess reactivity in the core. In order to achieve the cycle length of 4 years, Solid Pyrex rods are used for Conceptual Soluble Boron Free SMR (CSBF-SMR) design. The Solid Pyrex rods which are one of conventional burnable absorbers (BAs) have advantages in the license and applicability into current design of fuel assembly. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

A BWR Fuel Assembly Concept with Partial Segregation of Fertile Material

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

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Simulations of a Helical Tube Bundle in Cross-Flow for Application to Flow-Induced Vibration

Helical tubes have been used in a number of systems, including the energy and chemical processing industries. Helical-Coil Steam Generators (HCSGs) in particular have special application in systems for which spatial constraints are important. Variants of these designs are notably included in a number of integral-vessel small modular reactor (SMR) designs which are currently in development by multiple companies. These designs typically have the primary side located within the vessel, with the secondary side flowing upward through the helical tubes. One complexity in implementing an HCSG within the reactor pressure vessel is that any potential flow-induced vibration (FIV) can impact reactor safety. In a typical power plant, heat exchanger tube failures caused by FIV can be costly, but generally do not have immediate impact on the core safety due to their relative isolation from the core area. In an integral-vessel design, a tube rupture could eject material from the secondary side into the primary flow, potentially damaging the core itself or other crucial components. Thus a higher degree of confidence in predicting and/or preventing FIV should be attained for implementing the helical tubes in these designs. A significant campaign is currently underway to address the topic of FIV in helical tube arrays. The U.S. Department of Energy, under the Nuclear Energy Advanced Modeling and Simulation (NEAMS) initiative, has deemed this a 'high-impact problem'. This will include both experimental and computational contributions from a consortium consisting of multiple universities, industry partners, and national laboratories, including Argonne National Laboratory. The primary industrial partner is NuScale, which will see direct benefits from this initiative to the development of their SMR and its component design. The primary focus of this paper is to establish some of the computational efforts that have been made to this stage and the initial benchmarking of computational tools. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗