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DOE OSTI2016

Temperature and neutron flux dependence of in-reactor creep for cold-worked Zr-2.5Nb - 2016-0042

Over the past 10 years, internally pressurized capsules made from Zr-2.5Nb tubing have been irradiated in the National Research Universal (NRU) reactor at Chalk River Laboratories at temperatures of 280, 320, and 340 deg. C and dose rates between 3 X10{sup 15} n. m{sup -2}. s{sup -1} and 2 X10{sup 17} n. m{sup -2}. s{sup -1} (E>1 MeV). Periodic gaging has been used to assess the primary and secondary (steady-state) creep behavior. The objective of this detailed and controlled experiment was to determine, for the first time, the creep and microstructure evolution in Zr-2.5Nb tubing over a wide range of irradiation conditions for fast neutron fluxes applicable to a CANDU pressure tube. Similar but 'accelerated' creep experiments have been conducted in the Osiris test reactor at fast neutron fluxes of approximately 1.8 X10{sup 18} n. m{sup -2}. s{sup -1} (E>1 MeV), much greater than the neutron fluxes in the NRU reactor. Although accelerated tests in high-flux reactors such as Osiris provide information on irradiation creep, they do not represent the neutron flux conditions applicable to a power reactor. Tests covering power reactor operating conditions are needed to develop models for in-reactor creep of pressure tubes under the appropriate conditions. The data from the NRU reactor are compared with results from creep capsules with similar starting microstructures but irradiated in the Osiris reactor. The results show that the steady-state diametral and axial creep rates have a complex dependence on stress, temperature, and fast neutron flux. Data from out-reactor creep tests on unirradiated and pre-irradiated creep capsules that show the effect of prior irradiation on creep are also reported. The results are discussed in terms of a combination of creep mechanisms involving dislocation glide and mass transport. (authors)

36 MATERIALS SCIENCE↗

DOE OSTI2019

Low cost modular plasma system for reforming of natural gas

The initial aim of this project was to develop a low-cost compact, modularly scalable, point of use, on demand production system for the reforming of natural gas (NG) into products such as acetylene, hydrogen and other related hydrocarbons. Based on a pulsed dielectric barrier discharge (DBD), the fast rising voltage pulses produced by the system causes the dielectric breakdown of the gas to lag behind the rapid voltage increase so that a voltage greater than the static breakdown voltage of the gas, called an overvoltage, is briefly attained. The century long commercial use of the DBD discharge to push the highly endothermic oxygen to ozone reaction served as an additional motivation. The project meets the recommendations of the National Research Council 2007 report on plasma science for a greater investment in low temperature plasma science and engineering. An important finding is that the method investigated has a high specific energy requirement (SER) of the order of 1,000 to 5,000 kWh per kg of hydrogen, and a corresponding low efficiency of only ~0.1 to 0.5% of converting electrical into chemical (ΔH, ΔG) energy. This compares unfavorably with 52 kWh/kgH 2 required for NG thermal plasma reforming, or a pulsed spark discharge, Lotfalipour et. al, and Gao et. al. (For comparison electrolysis of water into hydrogen and oxygen gas requires only 45 to 80kWh/kgH 2 ). This project garnered data with different reactor configurations, materials, gases and mixtures including NG, methane, hydrogen, propane, argon, nitrogen and air, gap sizes ranging from 0.5 to 4.1 mm, electrode areas from 5 to 460 cm, voltages to 27 kV, pulse frequencies to 20 kHz and peak currents into the hundreds of amperes, temperatures to 625 °C and pressures from 15 mbar to 2 bar. An SRS 300 residual gas analyzer (RGA) yielded quadrupole mass spectrometry (QMS) data on the reformate. To analyze the data we extended the public domain software OpenChrom®, Wenig, to QMS data, yielding accurate identification and quantitative data on reformate products. This extension of OpenChrom® is applicable beyond this project and is a significant contribution to the general community of researchers. The documentation is available on the internet, see Whitlock, and a copy has been appended to this report. The main product of methane and natural gas plasma reforming is hydrogen, constituting ~90% of the products, followed by 1% to 10% ethane, propane, ethylene, acetylene and isobutene. Neopentane, propene, butane and pentane constitute less than 1 % of product. Adding a nickel catalyst increases the selectivity for hydrogen. Carbon nanosheets (graphene), a product of commercial interest, were synthesized by plasma reforming of acetylene at low pressure. The reformate analysis shows hat saturated and branched hydrocarbons are favored compared with unsaturated and linear hydrocarbons. For example, ethane, propane, iso-butane, 2-methyl butane, and 2-methyl pentane are produced in higher amounts compared with unsaturated and linear components. At low pressure acetylene and ethylene production rate is higher and propane production lower. We developed a recirculating system in which hydrogen is removed through a permeable membrane as fast as it is being produced by the reactor and the remaining products are recirculated at a high flow rate of several liters per minute. This keeps the concentration of H 2 in the reactor low, while achieving a total output concentration of 79%, high enough for pressure swing adsorption to be used for further purification. The recirculating system could be used with other gas reforming technologies. We report on four different reactors,1) a spark discharge reactor, 2) a standard DBD cylindrical reactor, 3) a small DBD reactor where catalyst powder can be quickly loaded onto a carrier to investigate the effect of the plasma on the catalyst and the effectiveness of the of the catalyst in promoting gas reformation and 4) a configurable DBD reactor to study the performance effects of dielectric selection, electrode gap distance, and electrode/dielectric interface on plasma reactor operation. This project achieved a commercializable 1kW plasma system including reactors a plasma pulse generator and a gas supply module.

03 NATURAL GAS↗

DOE OSTI2020

Fission Gas Measurements in Spent Fuel: Literature Review

The fission-derived noble gases, xenon(Xe) and krypton(Kr), have long been proposed as a tool for supporting nuclear safeguards and declarations verification. A vast body of literature outlines how these non-reactive fission gases released during reactor operation and shutdowns correlate to burnup level, power history, other reactor specifications and reprocessing activities. However, irradiation conditions can vary appreciably both radially and axially within a fuel pin.

07 ISOTOPE AND RADIATION SOURCES↗

DOE OSTI2022

Using Chemical Reactor Models to Predict Fluidized Bed Pyrolysis Yields of Biomass Feedstocks

A detailed biomass pyrolysis kinetics scheme is implemented in reduced-order reactor models to predict chemical species yields from a fluidized bed reactor. The biomass composition in terms of cellulose, hemicellulose, lignins, and extractives are determined for several biomass feedstocks. Model results are compared to yields measured from a two-inch diameter bench-scale fluidized bed reactor operating at fast pyrolysis conditions. The reduced-order chemical reactor models compare favorably with the experimental measurements and capture trends in product yields due to biomass compositional effects such as high ash content. This work offers a computationally inexpensive approach to predict the quality of biomass pyrolysis products in a timely manner. Source code for the reactor models along with a biomass composition web tool are made available online for future scientific research efforts.

09 BIOMASS FUELS↗

DOE OSTI2016

Regulatory Basis for Dry Spent Nuclear Fuel Storage Canister Inspections

Nuclear power plants in the United States began moving Spent Nuclear Fuel (SNF) from wet storage to dry storage in Independent Spent Fuel Storage Installations (ISFSIs) in 1986 at Surry station in Virginia. Dry storage represented an innovative solution to the problem created by the U.S. government's failure to fulfill its obligation to begin removing SNF from power plant sites. Today, about 70 ISFSIs have been placed into operation on the sites of current, or former nuclear power plants. There are currently over 2,200 SNF dry storage systems (DSSs) in service at these locations. Thirteen of these ISFSIs are located at sites with no operating reactor, and this number is due to increase in the coming years. Operating nuclear power stations use ISFSI storage to maintain wet storage capacity for recently discharged SNF. Permanently shutdown nuclear power plants use ISFSI storage to allow decommissioning and dismantlement of the power plant structures, systems, and components. The U.S. Nuclear Regulatory Commission (NRC) regulates the licensing, construction, and operation of ISFSIs under Title 10 of the U.S. Code of Federal Regulations, Energy, Part 72, Licensing Requirements for the Independent Storage of Spent Nuclear Fuel, High-Level Radioactive Waste, and Reactor-Related Greater than Class-C Waste (10 CFR 72). When the regulations in 10 CFR 72 were created around 1980, 20 years was chosen by the NRC as the license term for a specific ISFSI license and later, for a DSS Certificate of Compliance (CoC). Twenty years was, at the time, believed to be a sufficient time period for on-site interim dry storage of spent nuclear fuel before the U.S. government began moving the spent fuel away from the plant site for final disposal. Little, if any consideration was given to providing a regulatory framework for extending ISFSI licenses and DSS CoCs beyond 20 years at that time. However, shortly after the turn of the century, with still no federal program for removing SNF from reactor sites on the near-term horizon it became apparent that most DSSs would be required to remain in service at the ISFSIs beyond the initial 20-year license term. Operation beyond the original license term requires the specific license or the DSS CoC (used at a general license ISFSI) to be renewed. The requirements for renewal of a specific ISFSI license or a DSS CoC include addressing age-related degradation of structures, systems, and components that perform or support safety functions. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2026

Full-core high-burnup BWR LOCA fuel performance analysis and FFRD susceptibility

The susceptibility of the boiling water reactor (BWR) Limerick Unit 1 to fuel fragmentation, relocation, and dispersal during a postulated large-break loss-of-coolant accident (LBLOCA) was calculated using a multiphysics framework. The simulations include full-core, rod-resolved neutronic, thermal hydraulic, and fuel performance models using the VERA, TRACE, and BISON codes. This work focused on the transient BISON simulations, which include both the normal operation and LBLOCA periods in the same simulations. Cladding integrity was assessed using two correlations that are included with BISON. make page break Several new BWR-specific features were recently added to BISON. This work represents the first time these features have been included in a core-scale set of simulations. This study hence evaluates the performance of these new models for an operating reactor with realistic operating conditions. Simulation results showed that cladding integrity was maintained (i.e., no rods burst). Finally, future work to improve BWR and PWR predictions using this framework is suggested.

BISON↗

DOE OSTI2024

Methane partial oxidation under periodic reaction conditions on Pt/Al 2 O 3

The increasing interest in utilizing methane, the primary component of natural gas, for chemical production has spurred research into methane partial oxidation (MPO) as an alternative to traditional steam methane reforming (SMR). MPO has lower energy requirements and potential for carbon capture, making it an attractive option for hydrogen production. Challenges remain, however, such as carbon deposition leading to degradation and achieving high hydrogen selectivity. Here, the impact of periodic reactor operation on MPO over a Pt/Al 2 O 3 catalyst was studied, primarily via varying reactor inlet compositions. Experiments were conducted using periodic operation strategies to assess the influence of changing reactant inlet concentrations on hydrogen formation during MPO. The results suggest that cycling between mixtures with low and high oxygen content can lead to transient hydrogen formation rates that surpass those achieved at steady state. Control experiments and density functional theory (DFT) calculations show that enhanced hydrogen formation can be attributed to the reaction between CO with hydroxyl groups at the metal and alumina support interface. Furthermore, this work underscores the critical role of surface coverages at the metal–support interface and suggests avenues for future exploration, including alternative support materials with higher OH mobility and changes in the cycling scheme to enhance catalyst performance under periodic conditions.

03 NATURAL GAS↗

DOE OSTI2017

Evaluation of Iodine Remediation Technologies in Subsurface Sediments: Interim Status Report

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2017

Evaluation of Iodine Remediation Technologies in Subsurface Sediments (Interim Status Report)

Isotopes of iodine were generated during plutonium production from nine production reactors at the U.S. Department of Energy Hanford Site. The long half-life 129 I generated at the Hanford Site during reactor operations was 1) stored in single-shell and double-shell tanks, 2) discharged to liquid disposal sites (e.g., cribs and trenches), 3) released to the atmosphere during fuel reprocessing operations, or 4) captured by off-gas absorbent devices (silver reactors) at chemical separations plants (PUREX, B-Plant, T-Plant, and REDOX). Releases of 129 I to the subsurface have resulted in several large, though dilute, plumes in the groundwater, including the plume in the 200-UP-1 operable unit. There is also 129 I remaining in the vadose zone beneath disposal or leak locations. Because 129 I is an uncommon contaminant, relevant remediation experience and scientific literature are limited. Iodine cannot be degraded/transformed to a less toxic form. Therefore, the primary metric for remediation is changing the fate and transport properties in a way that works to decrease radioiodine concentrations in groundwater. To this end, targeted laboratory experiments were conducted to provide proof-of-principle screening of remedial options and to evaluate scale-up and performance elements. The evaluation of different iodine remediation technologies will occur over multiple years and fall into three different categories: 1) geochemical/biological sequestration, 2) microbial volatilization, and 3) enhanced pump-and-treat. This report provides status on two in situ sequestration methods (apatite and iron oxides) and two methods examining enhanced pump-and-treat.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2018

ATR PRA Presentation for MEXT visit - additional slides

Using the ATR PRA and Risk Monitor for risk-informed decision making as it pertains to online maintenance during reactor operation.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2023

Assessment of Core Physics Characteristics of Extended Enrichment and Higher Burnup LWR Fuels using the Polaris/PARCS Two-Step Approach (Vol. 2: BWR Fuel)

Nuclear fuel vendors and utilities are currently investigating changes to fuel contents and fuel designs for more economical and safer reactor operations. Extending cycle lengths beyond 18-month cycles for pressurized water reactors (PWRs) and 24-month cycles for boiling water reactors (BWRs) requires extending fuel enrichments beyond the current 5 wt % 235U limit. Therefore, low-enriched uranium plus (LEU+) fuel is expected to be used in current light-water reactor fleets in the near term. LEU+ is a subset of high-assay low-enriched uranium (HALEU) and is a term to describe fuel enrichments above 5% up to 10%. A series of studies were conducted at Oak Ridge National Laboratory (ORNL) to compare low-enriched uranium (LEU) with LEU+ fuel with respect to isotopic fuel content, lattice parameters, and core physics to identify any challenges in operation, storage, and transportation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2021

Versatile semi-batch apparatus for manometric measurement of gas–solid reaction rates

In this study, we describe the design, construction, operation, and performance of a simple and versatile semi-batch reactor that is especially useful for measurement of gas/solid reaction rates at pressures in the range of 1 mTorr to 1500 Torr. The reactor operates by repeatedly imposing small AC modulations of reactant gas pressure on top of a much larger DC pressure background. Based on the rate of pressure relaxation following each AC pulse, the reaction rate is determined. Our design is characterized by modular construction from off-the-shelf, ultra-high-vacuum-compatible components, which facilitate easy retrofitting and adaptation to a range of experimental conditions. Automated experiment control and data acquisition is accomplished via a custom National Instruments © LabView virtual instrument. Data analysis is automated using a custom series of Mathworks © Matlab scripts. We demonstrate reactor performance through measurements of hydrogenation kinetics for a composite H 2 getter material consisting of 1,4-bis(phenylethynyl)benzene mixed with a palladium/carbon catalyst.

47 OTHER INSTRUMENTATION↗

DOE OSTI2023

Nuclear Material Control & Accounting for Pebble Bed Reactors (FY 2023 Summary Report)

This report discusses the work done under the US Department of Energy NE-5 Advanced Reactor Safeguards and Security Program during FY 2023. It provides a summary of material control and accounting (MC&A) for pebble bed reactors (PBRs) and addresses some of the main challenges with current PBR MC&A approaches that will inform safeguards and security by design efforts. The efforts to date have focused on tristructural isotropic (TRISO) pebble fuel material accounting and control including working with partners in industry, loss and production of nuclear material as part of reactor operations, burnup modeling and measurements, uncertainty quantifications for such modeling and measurements, statistical approaches needed, and measurement methods. The unique fuel management and utilization in a PBR, where the fuel in spherical form is introduced and circulates through the reactor, poses special challenges for MC&A. This contrasts with traditional water-cooled reactors in which the fuel is contained in large assemblies and can be easily identified and counted. Even online fueled reactors, such as the CANDU reactors (none of which operate in the United States), are significantly different because the fuel is still contained in relatively large assemblies, is uniquely identified, and the number of assemblies that pass through the core on an annual basis is much fewer than the hundreds of thousands that circulate in a PBR, none of which are uniquely identified. Additionally, the nature of the TRISO fuel results in very low heavy metal loading with each pebble containing less than 10 g of uranium and on the order of less than 1 g of fissile material. This low fuel density and the robustness of the TRISO particles are major features of the TRISO fuel from a safety basis as each TRISO particle and pebble acts as a containment for the nuclear material and fission products during normal and accident conditions. This also results in very low plutonium loading per pebble during normal operations, which is on the order of 0.1 g at full burnup. A major feature of PBRs is that they will allow for significantly higher burnup, on the order of 160 GWd/THM compared to the burnup of traditional LWRs, which is on the order of 45 GWd/THM. This is achieved by monitoring the pebbles as they circulate through the reactor and allowing them to be reintroduced into the core until the desired burnup is achieved and they are removed from the reactor and enter the spent fuel storage areas.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2021

Yttrium Hydride Post Irradiation Examination Plan

The overarching goal of the Department of Energy Office of Nuclear Energy (DOE-NE) microreactor program is to develop technologies for the deployment of civilian microreactors by stakeholders1. Microreactors are expressed as advanced transportable nuclear reactors operating at low power (<20MWth) but high temperatures (>600°C), as well as plug-and-play and inherently safe designs. One prerequisite of a microreactor is the compactness, so that a truck can transport the reactor under safe conditions with the current road infrastructure1,2. The compactness of these reactors likely can be attainable by use of solid components for the essentials of the nuclear core, such as fuel, core heat removal components, reflectors, and moderators. Among these essentials, where fuel enrichment must remain < 20% to meet High Assay Low Enrichment Uranium criteria (HALEU), the largest contribution to the compactness is offered by use of solid moderators which benefit from light atomic weight elements, such as hydrogen, carbon, and beryllium2. Among these, hydrogen-bearing materials, such as metal hydrides, are superior to other options from the lowest critical mass standpoint. Noting that, factors other than critical mass should be considered for a specific reactor design. Yttrium- or zirconium-based metal hydrides have been down-selected due to their neutronic performance. In addition to the neutronic perspective, maintaining hydrogen within the metal hydride is important at the high operating temperatures proposed by advanced reactors. Yttrium hydride (YHx) is, therefore, a proposed moderator material that offers better hydrogen retention at higher operating temperatures than zirconium hydrides due to higher retention and thermal stability of hydrogen in the metal3. The irradiated materials properties of metal hydrides, in this case YHx, must be assessed for the qualification of these moderators. Material testing and inspection processes must illustrate that the effects of dimensional and property changes on thermophysical and mechanical properties do not cause any significant changes on the microreactor safety, and the moderating power is maintained within design limits. Thus, the effect of irradiation on the thermophysical and mechanical properties must be determined. This post-irradiation examination (PIE) plan specifically aims to determine these properties for YHx following Advanced Test Reactor (ATR) irradiation.

99 GENERAL AND MISCELLANEOUS↗

DOE OSTI2022

PLN-6268 Yttrium Hydride Post Irradiation Examination Plan

99 GENERAL AND MISCELLANEOUS↗

DOE OSTI2024

Space Nuclear Power Autonomous Control Algorithm and Control Element Test Bed

Nuclear thermal rockets are currently NASA’s preferred option for use in a manned mission to Mars in the 2040s. The communication delay between an Earth ground station and a spacecraft heading toward Mars can be up to 20 min. Therefore, controlling the nuclear rocket engine would require either a full-time reactor operator on the mission or an autonomous control system for the reactor. The latter idea of making space nuclear reactors fully autonomous has drawn more interest from stakeholders, but such an autonomous control system must be rigorously tested and validated before it is certified for human use. The cost of a full ground test for a space nuclear reactor is tremendous, so a nonnuclear mock reactor test bed was created to test and validate control elements and control algorithms for space nuclear reactors. The test bed consists of control element hardware that inputs physical measurement data into a reactor emulator to produce the reactor’s performance under steady-state, transient, and fault conditions. The control element hardware consists of six full-sized control drums equipped with servo drives and motors and is instrumented with optical encoders, resolvers, and torque sensors for drum movement characterization. In addition to the drums, a two-phase flow loop was designed and built to mimic the valves and turbomachinery associated with the propellant flow through a nuclear thermal rocket engine; components such as pressure sensors, flow meters, thermocouples, and tachometers are instrumented throughout the loop to characterize the fluid flow, valve, and turbomachinery behavior of the system. The data from the physical hardware (e.g., drum position, propellant flow rates) are input to a nuclear reactor simulator to determine the actual nuclear reactor parameters, and the data are sent back to a control algorithm to complete the control loop. The ability to conduct numerous tests of the control systems and autonomous algorithms can help validate the instrumentation and control aspects for a space nuclear reactor for every possible fault situation.

Wilson, Brandon↗

DOE OSTI2012

Assessment of Feasibility of the Beneficial Use of Waste Heat from the Advanced Test Reactor

This report investigates the feasibility of using waste heat from the Advanced Test Reactor (ATR). A proposed glycol waste heat recovery system was assessed for technical and economic feasibility. The system under consideration would use waste heat from the ATR secondary coolant system to preheat air for space heating of TRA-670. A tertiary coolant stream would be extracted from the secondary coolant system loop and pumped to a new plate and frame heat exchanger, where heat would be transferred to a glycol loop for preheating outdoor air in the heating and ventilation system. Historical data from Advanced Test Reactor operations over the past 10 years indicates that heat from the reactor coolant was available (when needed for heating) for 43.5% of the year on average. Potential energy cost savings by using the waste heat to preheat intake air is $\$$242K/yr. Technical, safety, and logistics considerations of the glycol waste heat recovery system are outlined. Other opportunities for using waste heat and reducing water usage at ATR are considered.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2017

Conceptual Model of Iodine Behavior in the Subsurface at the Hanford Site

Isotopes of iodine were generated during plutonium production within the nine production reactors at the U.S. Department of Energy Hanford Site. The short half-life 131I that was released from the fuel into the atmosphere during the dissolution process (when the fuel was dissolved) in the Hanford Site 200 Area is no longer present at concentrations of concern in the environment. The long half-life 129I generated at the Hanford Site during reactor operations was (1) stored in single-shell and double-shell tanks, (2) discharged to liquid disposal sites (e.g., cribs and trenches), (3) released to the atmosphere during fuel reprocessing operations, or (4) captured by off-gas absorbent devices (silver reactors) at chemical separations plants (PUREX, B-Plant, T-Plant, and REDOX). Releases of 129I to the subsurface have resulted in several large, though dilute, plumes in the groundwater. There is also 129I remaining in the vadose zone beneath disposal or leak locations. The fate and transport of 129I in the environment and potential remediation technologies are currently being studied as part of environmental remediation activities at the Hanford Site. A conceptual model describing the nature and extent of subsurface contamination, factors that control plume behavior, and factors relevant to potential remediation processes is needed to support environmental remedy decisions. Because 129I is an uncommon contaminant, relevant remediation experience and scientific literature are limited. In addition, its behavior in subsurface is different from that of other more common and important contaminants (e.g., U, Cr and Tc) in terms of sorption (adsorption and precipitation), and aqueous phase species transformation via redox reactions. Thus, the conceptual model also needs to both describe known contaminant and biogeochemical process information and identify aspects about which additional information is needed to effectively support remedy decisions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗