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DOE OSTI2023

Synthetic Electricity Market Data Generation and HERON Use Case Setup of Advanced Nuclear Reactors Coupled with Thermal Energy Storage Systems

This study evaluates and optimizes advanced nuclear reactors coupled with thermal energy storage (TES) systems in an Integrated Energy System (IES) architecture to enable advanced nuclear power plants (A NPP) to participate in multi-commodity markets, thus enhancing their economic competitiveness. Nuclear-TES coupling scenarios studied herein are designed attenuate the nuclear heat dynamics and defer energy delivery to a later time, enabling the nuclear reactor to continue operating at or near steady-state design conditions as usual while also enabling flexible generation. Three A-NPPs, namely, an advanced light-water reactor (A LWR), a high temperature gas-cooled reactor (HTGR) and a liquid-metal fast reactor (LMFR) were selected as the initial use cases for demonstrating the technoeconomic of thermally balanced energy storage coupling design for thermal power extraction. Each of the reactor technologies were evaluated in two different electricity markets. Stochastic optimization approach was adopted which included the evaluation of price signals from the Pennsylvania-New Jersey-Maryland (PJM) market, and Electric Reliability Council of Texas (ERCOT), using an autoregressive moving average (ARMA) model. Risk Analysis Virtual Environment (RAVEN) tool and its dispatch optimization plugin, the Holistic Energy Resource Optimization Network (HERON), were used to perform dispatch and capacity optimization, using the price data provided by the ARMA models. The results from the Nuclear-TES use cases will be used to design and characterize dynamic integrated system behavior and feedback.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS1998

Computational Modeling Program

An Integrated Product Team (IPT) has been formed at NASA Ames Research Center which has set objectives to investigate devices and processes suitable for meeting NASA requirements on ultrahigh performance computers, fast and low power devices, and high temperature wide bandgap materials. These devices may ultimately be sub-100nm feature-size. Processes and equipment must meet the stringent demands posed by the fabrication of such small devices. Until now, the reactors for Chemical Vapor Deposition (CVD) and plasma processes have been designed by trial and error procedures. Further, once the reactor is in place, optimum processing parameters are found through expensive and time-consuming experimentation. If reliable models are available that describe processes and the operation of the reactors, that chore would be reduced to a routine task while being a cost-effective option. The goal is to develop such a design tool, validate that tool using available data from current generation processes and reactors, and then use that tool to explore avenues for meeting NASA needs for ultrasmall device fabrication. Under the present grant, ARL/Penn State along with other IPT members has been developing models and computer code to meet IPT goals. Some of the accomplishments achieved during the first year of the grant are described in this report

Govindan, T. R.↗
DOE OSTI2020

Relevant Advanced Reactor Benchmarks for Nuclear Data Assessment

Advanced reactor concepts currently being developed throughout the industry are significantly different from light water reactor (LWR) designs with respect to geometry, materials, and operating conditions, and consequently, with respect to their reactor physics behavior. Given the limited operating experience with non-LWRs, the accurate simulation of reactor physics and the quantification of associated uncertainties are critical for ensuring that advanced reactor concepts operate within the appropriate safety margins. Nuclear data are a major source of input uncertainties in reactor physics analysis. As part of an ongoing project at Oak Ridge National Laboratory, the effects of nuclear data uncertainties on key figures of merit associated with advanced reactor safety are being assessed for selected advanced reactor technologies. Key nuclear data relevant for reactor safety analysis for each selected advanced reactor technology were identified in Phase 1, and their impact on important key figures of merit was assessed in Phase 2. This report describes the outcome of Phase 3. Available benchmarks and fuel irradiation data for use in evaluating the impact of uncertainties and gaps in nuclear data that impact reactivity control for advanced reactor designs through the fuel cycle were identified and assessed. Benchmarks were identified by searching (1) the Organisation for Economic Co-operation and Development (OECD)/Nuclear Energy Agency (NEA) International Criticality Safety Benchmark Evaluation Project (IRPhEP) handbook, (2) the OCED/NEA International Reactor Physics Experiment Evaluation Project (IRPhEP) handbook, (3) ongoing OECD/NEA benchmark activities, and (4) documentation in public literature. Relevant benchmarks were identified by selecting reactors with geometry, materials, and neutron energy spectra similar to those of selected advanced reactor technologies. This assessment identified six benchmarks, of which three are experimental and three are purely computational. One experimental and one computation benchmark contain depleted fuel; all others are limited to fresh fuel. This report provides short descriptions of the selected benchmarks along with the availability of measured data for comparison.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI1988

Nuclear Safety [Vol. 29, No. 3, July-September 1988]

Nuclear Safety is a review journal that covers significant developments in the field of nuclear safety. Its scope includes the analysis and control of hazards associated with nuclear energy, operations involving fissionable materials, and the products of nuclear fission and their effects on the environment. Primary emphasis is on safety in reactor design, construction, and operation; however, the safety aspects of the entire fuel cycle, including fuel fabrication, spent-fuel processing, nuclear waste disposal, handling of radioisotopes, and environmental effects of these operations, are also treated. Table of Contents for this issue follows. GENERAL SAFETY CONSIDERATIONS: 259 Fifteenth Water Reactor Safety Information Meeting by E. G. Silver; CONTROL AND INSTRUMENTATION: 284 Reliability Technology to Improve and/or Maintain Emergency Diesel Generator Performance by S. Karimian and J. H. Taylor, 293 A Noise Diagnostics System for Operator Advice by G. Hessel, P. Liewers, P. Schumann, W. Schmitt, and F.-P. Weiss; PLANT SAFETY FEATURES: 307 A Passive Containment System for Advanced Light-Water Reactors by O. B. Falls, Jr., and F. W. Kleimola; ENVIRONMENTAL EFFECTS: 318 Data Base Construction for a Computerized Radiological Risk Investigation System by L. M. Hively, J. E. Nyquist, J. L. Bledsoe, and A. L. Sjoreen, 326 Erratum to "Radiation Hormesis and Nuclear Safety," Vol. 29, No. 1; OPERATING EXPERIENCES: 327 Operational Safety Experience and Passive Safety Testing at the Fast Flux Text Facility by Q. L. Baird, J. L. Rathbun, D. D. Stepnewski, R. L. Stover, and A. E. Waltar, 344 Backfilling of Independent Residual Heat Removal Systems in West Germany and Switzerland by G. Eckert and Y. Salomon, 353 Reactor Shutdown Experience Compiled by J. W. Cletcher, 356 Selected Safety-Related Events Compiled by G. A. Murphy, 363 Operating U.S. Power Reactors Compiled by E. G. Silver; RECENT DEVELOPMENTS: 384 General Administrative Activities Compiled by E. G. Silver, 390 Reports, Standards, and Safety Guides by D. S. Queener, 395 Status of Power-Reactor Projects Undergoing Licensing Review Compiled by E. G. Silver, 400 Proposed Rule Changes as of Mar. 31,1988; ANNOUNCEMENTS: 283 International ENS/ANS Conference on Thermal Reactor Safety "NUCSAFE 88", 317 Northwestern University Short Course on Radiation Safety, 407 Second International Seminar on Small- and Medium-Sized Nuclear Reactors, 407 International Workshop on New Developments in Occupational Dose Control and ALARA Implementation at Nuclear Power Plants and Similar Facilities, 408 Fourth International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-4), 404 The Authors.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Design, Development, and Operation of an Integrated Fluidized Carbon Capture Unit Using Polyethylenimine Sorbents

In this paper, we present the design, development, and operation of a reactor system for CO 2 capture. Modifications were implemented to address differences in sorbent from 180 μm Geldart group B to 115 μm Geldart group A material; operational issues were discovered during experimental trials. The major obstacle in system operation was the ability to maintain a constant circulation of a solid sorbent stemming from this change in sorbent material. The system consisted of four fluid beds, through which a polyamine impregnated sorbent was circulated and adsorption, preheat, regeneration, and cooling processes occurred. Pressure transducers, thermocouples, gas flow meters, and gas composition instrumentation were used to characterize thermal, hydrodynamic, and gas adsorption performance in this integrated unit. A series of shakedown tests were performed and the configuration altered to meet the needs of the sorbent performance and achieve desired target capture efficiencies. Methods were identified, tested, and applied to continuously monitor critical operating parameters including solids circulation rate, adsorbed and desorbed CO 2 , solids inventories, and pressures. The working capacity and CO 2 capture efficiency were used to assess sorbent performance while CO 2 closure was used to define data quality and approach to steady-state. Testing demonstrated >90% capture efficiencies and identified the regenerator to be the process step limiting throughput. Sorbent performance was found to be related to the reactant stoichiometry. Finally, a stochastic model with an exponential dependence on the relative CO 2 /amine concentration was used to describe 90% of the variance in the data.

01 COAL, LIGNITE, AND PEAT↗
DOE OSTI2021

Characterization of Safeguards Implementation Schedules for Advanced Reactors

Advanced reactor designs differ from the traditional light-water reactor designs and operational requirements in nearly every facet, including the heat transfer fluids, operating temperatures and pressures, and fuel cycle requirements. To account for these differences, significant effort will be required to ensure appropriate safeguards implementation for US-based reactors intended for global deployment. This is the second in a series of three reports that examine facets of this issue. The first paper (ORNL/TM-2021/2142) examines potential construction schedules for advanced reactors to create a baseline understanding of the timeframes involved. This second report characterizes International Atomic Energy Agency (IAEA) safeguards implementation schedules for potential advanced reactors. Finally, the third report (ORNL/TM-2021/2208) examines existing gaps for safeguards implementation at advanced reactors.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Inadvertent actuation block valve for a small modular nuclear reactor

An inadvertent actuation block valve includes inlet and outlet orifices being in selective fluid communication via a chamber. A disc is disposed within the chamber and a bellows is configured to contract at a predetermined pressure differential between reactor fluid entering a reference pressure orifice and control fluid entering the inlet orifice. When the bellows contracts, the disc engages the outlet orifice and isolates fluid communication between the inlet and outlet orifices. The inadvertent actuation block valve prevents inadvertent opening of an emergency core cooling valve when a reactor is at operating pressure that is above the predetermined set pressure range. The inadvertent actuation block valve permits the emergency cooling valves to open and to remain open when reactor pressure is below the predetermined set pressure range. The inadvertent actuation block valve does not impede long term emergency cooling that occurs when the reactor is at low pressure.

O'Brien, Hugh S.↗
DOE OSTI2020

Progress in extending high poloidal beta scenarios on DIII-D towards a steady-state fusion reactor and impact of energetic particles

To prepare for steady-state operation of future fusion reactors (e.g. the International Thermonuclear Experimental Reactor and China Fusion Engineering Test Reactor (CFETR)), experiments on DIII-D have extended the high poloidal beta ( β P ) scenario to reactor-relevant edge safety factor q 95 ~ 6.0, while maintaining a large-radius internal transport barrier (ITB) using negative magnetic shear. Excellent energy confinement quality ( H 98y2 > 1.5) is sustained at high normalized beta ( β N ~ 3.5). This high-performance ITB state with Greenwald density fraction near 100% and q min ≥ 3 is achieved with toroidal plasma rotation V tor ~ 0 at ρ ≥ 0.6. This is a key result for reactors expected to have low V tor . At high β P (≥1.9), large Shafranov shift can stabilize turbulence leading to a high confinement state with a low pedestal and an ITB. At lower β P (<1.9), negative magnetic shear in the plasma core contributes to turbulence suppression and can compensate for reduced Shafranov shift to continue to access a large-radius ITB and excellent confinement with low V tor , consistent with the results of gyrofluid transport simulations. These high- β P cases are characterized by weak/no Alfvén eigenmodes (a.e.) and classical fast-ion transport. At high density, the fast-ion deceleration time decreases and Δ β fast is lower; these reduce a.e. drive. The reverse-shear Alfvén eigenmodes are weaker or stable because the negative magnetic shear region is located at higher radius, away from the peaked fast-ion profile. Resistive wall modes can be a limitation at simultaneous high β N , low internal inductance, and low rotation. Analysis suggests that additional off-axis external current drive could provide a more stable path at reduced q 95 . Based on a DIII-D high- β P plasma with large-radius ITB, two scenarios are proposed for CFETR Q = 5 steady-state operation with ~1 GW fusion power: a lower-$l_i$($l_i$ ~ 0.66) and a higher-$l_i$($l_i$ ~ 0.75) case. Using a Landau closure model, multiple energetic particle (EP) effects on the a.e. stability are analyzed modifying the growth rate of the a.e.s triggered by the neutral-beam-injection EPs and alpha particles, although the stabilizing/destabilizing effect is weak for the cases analyzed. The stabilizing effects of the combined EP species β , energy, and density profile in CFETR need further investigation.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
NASA NTRS1975

Design, fabrication, and operation of capsules for the irradiation testing of candidate advanced space reactor fuel pins

Fuel irradiation experiments were designed, built, and operated to test uranium mononitride (UN) fuel clad in tungsten-lined T-111 and uranium dioxide fuel clad in both tungsten-lined T-111 and tungsten-lined Nb-1% Zr. A total of nine fuel pins was irradiated at average cladding temperatures ranging from 931 to 1015 C. The UN experiments, capsules UN-4 and -5, operated for 10,480 and 10,037 hr, respectively, at an average linear heat generation rate of 10 kW/ft. The UO2 experiment, capsule UN-6, operated for 8333 hr at an average linear heat generation rate of approximately 5 kW/ft. Following irradiation, the nine fuel pins were removed from their capsules, externally examined, and sent to the NASA Plum Brook Facility for more detailed postirradiation examination. During visual examination, it was discovered that the cladding of the fuel pin containing dense UN in each of capsules UN-4 and -5 had failed, exposing the UN fuel to the NaK in which the pins were submerged and permitting the release of fission gas from the failed pins. A rough analysis of the fission gas seen in samples of the gas in the fuel pin region indicated fission gas release-to-birth rates from these fuel pins in the range of .00001.

Thoms, K. R.↗
DOE OSTI2017

Neutron-Gamma flux and Dose calculations in a Pressurized Water Reactor (PWR)

The Institute of Radiological Protection and Nuclear Safety (IRSN) acts as technical support to the French nuclear safety authorities. In this context, one of its missions is to improve nuclear safety through the enhancement of knowledge regarding operating nuclear power reactors. The issues concerning the aging and the characterisation of the reactor vessel require knowledge of the neutron and gamma radiation environment on the vessel and beyond. Since few years, ENEA (Energia Nucleare e Energie Alternative, i.e. Nuclear Energy and Alternative Energy agency) and IRSN collaborate on Monte-Carlo simulations and Variance Reduction Techniques (VRT), which are used to determine neutron and gamma spectra in and out of the core. Additionally, IRSN participates in the DISCOMS project, which stands for 'Distributed Sensing for Corium Monitoring and Safety', aiming at examining the status of the third barrier of confinement, in order to define possible mitigation strategies in case of a severe accident, namely: i) reactor pressure vessel breakthrough and corium relocation outside the vessel, ii) reactor pit concrete erosion and iii) corium cooling. One of the IRSN tasks in the frame of the DISCOMS project is to provide the radiation level in terms of neutron and gamma doses and flux energy spectra at various locations. This paper summarizes the study performed on the radiation transport in the Tihange-I Belgian nuclear reactor and contributes to several issues requiring the knowledge of the radiation environment in and outside the reactor vessel. This paper gives an overview of the work presented at the ICRS-13 and RPSD-2016 conference (for more details see the proceedings paper)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Metabolic multireactor: Practical considerations for using simple oxygen sensing optodes for high-throughput batch reactor metabolism experiments

We present a system for carrying out small batch reactor oxygen consumption experiments on water and sediment samples for environmental questions. In general, it provides several advantages that can help researchers achieve impactful experiments at relatively low costs and high data quality. In particular, it allows for multiple reactors to be operated and their oxygen concentrations to be measured simultaneously, providing high throughput and high time-resolution data, which can be advantageous. Most existing literature on similar small batch-reactor metabolic studies is limited to either only a few samples, or only a few time points per sample, which can restrict the ability for researchers to learn from their experiments. The oxygen sensing system is based very directly on the work of Larsen, et al. [2011], and similar oxygen sensing technology is widely used in the literature. As such we do not delve deeply into the specifics of the fluorescent dye sensing mechanism. Instead, we focus on practical considerations. We describe the construction and operation of the calibration and experimental systems, and answer many of the questions likely to come up when other researchers choose to build and operate a similar system themselves (questions we ourselves had when we first built the system). In this way, we hope to provide an approachable and easy to use research article that can help other researchers construct and operate a similar system that can be tailored to ask their own research questions, with a minimum of confusion and missteps along the way.

3D printing↗
DOE OSTI1991

Nuclear Safety [Vol. 32, No. 1, January-March 1991]

Nuclear Safety is a review journal that covers significant developments in the field of nuclear safety. Its scope includes the analysis and control of hazards associated with nuclear energy, operations involving fissionable materials, and the products of nuclear fission and their effects on the environment. Primary emphasis is on safety in reactor design, construction, and operation; however, the safety aspects of the entire fuel cycle, including fuel fabrication, spent-fuel processing, nuclear waste disposal, handling of radioisotopes, and environmental effects of these operations, are also treated. Table of Contents for this issue follows. GENERAL SAFETY CONSIDERATIONS: 1 The Safety Review and Approval Process for Space Nuclear Power Sources, G. L. Bennett; 19 Report on the American Nuclear Society International Topical Meeting: "The Safety, Status, and Future of Non-Commercial Reactors and Irradiation Facilities", E. G. Silver; 35 Book Review: Fission Product Transport Processes in Reactor Accidents, Proceedings to the International Centre tor Heat and Mass Processes T. S. Kress; 38 Fast Reactor Technology in the 1990s: A Summary of the 1990 International Fast Reactor Safety Meeting, A. E. Levin; ACCIDENT ANALYSIS: 56 Effects of Chemical Phenomena on LWR Severe Accident Fission Product Behavior, A. P. Malinauskas and T. S. Kress; CONTROL AND INSTRUMENTATION: 65 Technical Note: Safety Parameter Display Systems—10 Years Later, R. J. Eckenrode; 68 Potential Application of Neural Networks to the Operation of Nuclear Power Plants, R. E. Uhrig; DESIGN FEATURES: 80 Twenty-First DOE/NRC Nuclear Air-Cleaning Conference, R. R. Bellamy, D. W. Moeller, and M. W. First; 91 Impact of an Apparent Radiation Embrittlement Rate on the Life Expectancy of PWR Vessel Supports, R. D. Cheverton, G. C. Robinson, W. E. Pennell, and R. K. Nanstad; ENVIRONMENTAL EFFECTS: 103 Technical Note: The Impact of Offsite Factors on the Safety Performance of Small Nuclear Power Plants, Yu. D. Baranaev and A. N. Viktorov; WASTE AND SPENT FUEL MANAGEMENT: 109 Activities Related to Waste Management, Compiled by E. G. Silver; OPERATING EXPERIENCES: 118 Reactor Shutdown Experience, Compiled by J. W. Cletcher; 121 Selected Safety-Related Events, Compiled by G. A. Murphy; 123 Operating U.S. Power Reactors, Compiled by E. G. Silver; RECENT DEVELOPMENTS: 140 General Administrative Activities, Compiled by E. G. Silver; 150 Reports, Standards, and Safety Guides, D. S. Queener; 155 Status of Power-Reactor Licensing Activities, Compiled by E. G. Silver; 157 Proposed Rule Changes as of Sept. 30, 1990; ANNOUNCEMENTS: 79 MIT Offers Summer Program on Nuclear Power Reactor Safety; 108 Harvard School of Public Health Offers Several Short Courses; 139 Short Course and Workshop on Nuclear Criticality Safety at University of New Mexico; 160 SCK/CEN Announces Training Course on Emergency Planning and Response; 161 The Authors; 164 Indexes to Nuclear Safety, Volume 31

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Nitrification in a completely stirred tank reactor treating the liquid phase of digestate: The way towards rational use of nitrogen

Highlights: • Nitrification of digestate liquid phase was simulated in a laboratory scale. • The attention was paid mainly to rationale use of nitrogen. • Nitrate was produced immediately after the process initiation. • Ammonia oxidation rate reached up to 0.59 kg/(m{sup 3} d). • The operation of the system was effectively optimized by pH control. - Abstract: The nitrification of the liquid phase of digestate (LPD) was conducted using a 5 L completely stirred tank reactor (CSTR) in two independent periods (P1 – without pH control; P2 – with pH control). The possibility of minimizing nitrogen losses during the application of LPD to the soil as well as during long-term storage or thermal thickening of LPD using nitrification was discussed. Moreover, the feasibility of applying the nitrification of LPD to the production of electron acceptors for biological desulfurization of biogas was assessed. Despite an extremely high average concentration of ammonia and COD in LPD reaching 2470 and 9080 mg/L, respectively, nitrification was confirmed immediately after the start-up of the CSTR. N-NO{sub 3}{sup −} concentration reached 250 mg/L only two days after the start of P1. On the other hand, P1 demonstrated that working without pH control is a risk because of the free nitrous acid (FNA) inhibition towards nitrite oxidizing bacteria (NOB) resulting in massive nitrite accumulation. Up to 30.9 mg/L of FNA was present in the reactor during P1, where the NOB started to be inhibited even at 0.15 mg/L of FNA. During P2, the control of pH at 7.0 resulted in nitrogen oxidation efficiency reaching 98.3 ± 1.5% and the presence of N-NO{sub 3}{sup −} among oxidized nitrogen 99.6 ± 0.4%. The representation of volatile free ammonia within total nitrogen was reduced more than 1000 times comparing with raw LPD under these conditions. Thus, optimum characteristics of the tested system from the point of view of minimizing the nitrogen losses as well as production of electron acceptors for the desulfurization of biogas were gained in this phase of reactor operation. Based on the results of the experiments, potential improvements and modifications of the tested system were suggested.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2025

A simplified approach for creep damage and remaining life calculation for materials surveillance in advanced reactors

High temperature reactor materials will experience a combination of mechanical degradation caused by creep and fatigue and environmental degradation caused by neutron irradiation and, in some cases, exposure to corrosive coolants. A materials surveillance program is one option for ensuring the safe, reliable operation of key reactor components under these conditions. Such a program would monitor the degradation in key material properties over time, using this data to predict how changes in the material properties affect components performance. This could then be used to make plant operational decisions and support eventual plant license extensions. Past work de-scribes such a material surveillance program based on passively actuated mechanical test articles that impose creep-fatigue type loading on test materials driven only by changes in temperature, for example those experienced by components under standard operating cycles. This report focuses on two aspects of the proposed material surveillance program: how to design the test articles to mimic the mechanical response of the corresponding component and how to use ex-situ test data to deter-mine the amount of damaged experienced by the test article (and hence the corresponding compo-nent) in service. Specifically, this report derives and validated simplified methods for both tasks, replacing earlier, more complicated approaches based on simulating the test article response and finding the best specimen design or current damage via complex numerical optimization. The sim-plified methods for each task developed here can be implemented in spreadsheet software and are simple enough for practical use in future operating plants. This report derives the methods (one for sizing and two options for damage inference), verifies the simplified approaches versus more so-phisticated methods, and compares the results of applying the simplified methods to previous re-sults using the complex numerical optimization approach. The report also provides a worked sam-pled problem applying the simplified techniques to a realistic high temperature reactor component.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

The Radiation Science and Engineering Center Utilizations and Future Developments at Penn State University

The Radiation Science and Engineering Center (RSEC) facilities at Penn State include the Penn State Breazeale Reactor (PSBR), gamma irradiation facilities (In-pool Irradiator and Dry Irradiator), Neutron Beam Laboratory, Hot Cells, Radio-nuclear Applications Laboratory (RAL), Radiochemistry Teaching Laboratory, Nuclear Security Education Laboratory, Subcritical Graphite Reactor Facility, and various Radiation Detection and Measurement Laboratories. The PSBR is a 1 MW TRIGA Mark III reactor with pulsing capabilities. A variety of dry tubes and fixtures are available in or near the core for irradiations. When the reactor core is placed next to the D{sub 2}O tank and graphite reflector assembly near the beam port locations, thermal neutron beams become available. In steady state operation at 1 MW, the thermal neutron flux is 1x10{sup 13} n/cm{sup 2} sec at the edge of the core and 3x10{sup 13} n/cm{sup 2} sec at the central thimble. The peak flux during a maximum pulse is ∼ 6x10{sup 16} n/cm{sup 2} sec with a pulse half width of ∼10 msec. The PSBR, the centerpiece of the RSEC, first went critical in 1955 and is the longest continuously operating university research reactor in the United States. The Nuclear Regulatory Commission renewed the PSBR's license for an additional 20 years in November 2009. A significant redesign of the core-moderator/beam port is currently underway to make full use of the PSBR's capabilities and to establish state-of-the-art neutron beam facilities.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2020

Molten Chloride Thermophysical Properties, Chemical Optimization, and Purification

Next-generation concentrating solar power (CSP) technology requires high-temperature heat-transfer fluids and thermal energy storage that can operate in the temperature range of 500°–720°C. This high temperature demands a new chemistry for the heat-transfer fluids that must have high-temperature thermal and chemical stabilities, high energy density, and low corrosion rate on metallic materials used in the heat exchangers, piping, and thermal energy storage tanks. A ternary MgCl 2 -KCl-NaCl chloride-salt system has been proposed to achieve such goals, given its high-temperature thermal stability. The chloride-salt-based systems have been used by the metallurgy industry as the electrolyte to electrolytically extract metallic magnesium. They have also attracted attentions from the nuclear industry for developing the molten salt fast reactors that can operate at close to atmospheric pressure and at higher temperature than traditional light-water reactors. However, the corrosion properties of the ternary salt are not well known. The corrosion mechanism is believed to be correlated to the presence of MgOHCl impurity at the operation temperature. MgOHCl is fundamentally caused by the hygroscopic nature of the MgCl 2 component. Therefore, the first key objective of this project is to understand the nature of the corrosiveness of the ternary chloride-salt system and its prevention by designing new lab-scale purification processes. Because of its importance, a working group called the Chloride Salt Collective across multiple U.S. Department of Energy national laboratories is established to gain experimental agreement on the corrosion behaviors. In addition, water generation during salt dehydration could lead to highly corrosive HCl/water mixture via a hydrolysis process between water and anhydrous MgCl 2 or MgCl 2 hydrates. MgOHCl will eventually be converted to MgO particulates regardless of the chemical path chosen (i.e., by thermal decomposition, chemical purification with an active metal, or electrochemical purification utilizing the reduction power of electron flows). Large MgO particles with high hardness can wear out bearing and bushing and even clog the delicate channels in the primary heat exchanger. Because of this, a second objective is to understand the chemistry behind salt purification and transfer the scientific understanding to a set of engineering know-how for large-scale salt melter design. Because there is limited information on the thermophysical properties and the sensitivity of these properties around the targeted salt composition, a third objective is to measure thermophysical properties relevant to system and component design and support Topic 1 Liquid Pathway development. For example, composition shift due to salt reaction with water can be a potential issue because it can lead to changes of thermophysical properties, such as (1) melting point (or freeze point), which can jeopardize safe operation of the supercritical CO 2 power cycle proposed for next-generation CSP, and (2) heat capacity and density, which can decrease the overall power-generation capability of the plants. The value of the thermal conductivity of the selected salt composition is also a critical criterion for the solar receiver design.

14 SOLAR ENERGY↗
DOE OSTI2018

Alternate Design Concepts for the High Flux Isotope Reactor using Low-Enriched Uranium Fuel Systems: Scoping Analysis

The Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR) is a multi-mission research reactor dedicated mainly to isotope production, neutron scattering experiments, material irradiation, testing and activation. In the framework of non-proliferation policies, the international community is currently aiming to reduce or even eliminate, when and where possible, the use of Highly Enriched Uranium (HEU: 235 U/U ≥ 20 wt. %) in civilian applications. In that context, most research reactors worldwide still operating with HEU fuel are actively engaged in an effort to convert to Low Enriched Uranium fuel system (LEU: 235 U/U < 20 wt. %). Within the US, the High Performance Research Reactor (USHPRR) fleet (which includes HFIR) is expected to convert to LEU fuel using the UMo “monolithic” fuel system currently under development. ORNL is responsible for the development of a LEU design for HFIR. In the past years, Argonne National Laboratory (ANL) offered technical assistance to the ORNL LEU design team whenever deemed appropriate. Due to their complex nature, acceptable HFIR LEU designs identified so far, all based on the UMo monolithic fuel system, would require a substantial fabrication development effort before being proven commercially viable. In order to mitigate technical risks and reduce cost/schedule uncertainties, it appears appropriate to study alternate design solutions and alternate fuel systems as possible backup options for the HFIR conversion to LEU fuel. In that context, ANL has been exploring alternate design concepts for HFIR. Complementing the ORNL design activities, the goal is to find the best design solution for HFIR in term of cost, performance and safety. The present report summarizes the findings regarding two types of alternate conceptual design: U 3 Si 2 designs and un-contoured UMo design. Based on available tools and methods, preliminary analysis show that both concepts could probably meet performance and steady-state safety requirements but element fabrication would remain complex and would require a non-trivial fabrication development effort.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

NSUF Research Reactor Fitness Workshop Results

The Nuclear Science User Facilities has been working with the National Organization of Test, Research, and Training Reactors to collect input from the US university reactor community on their current and future needs. The effort started with a web survey in March to collect initial data. NSUF held a follow-up workshop in July at the Center for Advanced Energy Studies in Idaho Falls. We identified four areas of interest in the university research reactor community that can affect future operations and sustainability of the reactors: capital infrastructure, regulatory burden, staffing and knowledge transfer, and utilization and relevancy. Members of the community presented challenges and best practices from their facilities. Discussions were held to create and distill the list of challenges for each area. A working group was formed for each area to continue the discussion and prioritize the challenges. NSUF and the working groups intend to create a final recommendations report by the end of the year. This panel session will discuss the workshop and summarize the discussion in the four topic areas. The audience is encouraged to add to the discussion.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗