Search

The Nuclear Science User Facilities (NSUF) is the U.S. Department of Energy Office of Nuclear Energy’s only designated nuclear energy user facility. Its mission is to provide nuclear energy researchers access to world-class capabilities and to facilitate the advancement of nuclear science and technology. This mission is supported by providing access to state-of-the-art experimental irradiation testing, postirradiation examination facilities, and high-performance computing capabilities, as well as technical and scientific assistance for the design and execution of projects. As part of an NSUF project, low dose silicon carbide monitors were irradiated in the Belgian Reactor 2 and were then evaluated both at the SCK•CEN and at the Idaho National Laboratory (INL) High Temperature Test Laboratory to determine their peak temperature achieved during irradiation. The technical significance of this work was that the monitors were irradiated to a dose that was significantly less than recommended in published literature. This article will discuss the evaluation process, the irradiation test, and the performance of the low dose silicon carbide temperature monitors.

Proton conducting ceramics have been considered as a strategic material system used in solid oxide fuel cells and solid oxide electrolysis cells for electricity generation and hydrogen production. The high proton conductivity of protonic electrolyte and rapid development of robust electrode materials have significantly promoted the reduction of feasible operating temperature to intermediate temperature range (400~600 °C), which can attribute to favorable technical and economic advantages when they are compared with conventional oxide-ion conducting electrolyte based solid oxide cell system. In the past few years, Idaho National Laboratory has been devoted to developing advanced electrolyte and electrode materials R&D, diverse manufacturing capability, and expanding evaluation facility for these applications, aiming to further improve performance, stability and manufacturing cost. There are some remarkable accomplishments achieved on electrode engineering, electrolyte optimization, new material down-selection and large-scale cell fabrication in Idaho National Laboratory.

The Transient Reactor Test (TREAT) facility, located at Idaho National Laboratory, restarted transient operations in 2018 following an extended shutdown. It is of interest to establish a methodology and capability to obtain an accurate estimate of the total number of fissions produced in a fissionable test item during a transient at TREAT. Uranium wires were irradiated in TREAT as part of a transient prescription test program, and gamma-ray spectrometry was performed on the wires following irradiation using a high-purity germanium detector. Many fission products are useful for estimating the number of fissions produced in a sample using gamma-ray spectrometry; at TREAT with the time periods used for analysis, the isotopes of interest include 95Zr, 95Nb, 103Ru, 140Ba, and 140La. The number of fissions per gram of 235U determined from these measurements establishes an estimate for future experiments to be performed in the core when a similar configuration is used with a similar transient prescription.

Idaho National Laboratory (INL) is a nuclear research facility located in southeastern Idaho, USA; over the course of its operational history, INL has operated 52 reactors and 1 reactor fuel reprocessing facility. To determine the extent to which previous nuclear operations at INL have impacted local environmental carbon-14 (14C) concentrations, tree and soil samples from the INL desert and surrounding areas were collected, combusted, and analyzed by Accelerator Mass Spectrometry (AMS). Transport models of the plumes from the Advanced Test Reactor Facility (ATR) and the Idaho Chemical Processing Plant (ICPP) suggest the historic annual integrated plume distributions from each of these source terms was most likely in the northeast-southwest direction, with very little ground contact from stack emissions in the immediate vicinity of the facilities and maximum estimated ground contact arising ∼400–1000 m to the northeast/southwest from each facility. 14C data from annual growth ring data from trees located immediately adjacent to the ATR, INL’s Central Facilities Area (CFA), and Mud Lake, Idaho (ML) are in statistical agreement with regional nuclear weapons testing fallout backgrounds. Surface soil samples taken near a low level radioactive disposal facility and downwind from the ICPP show percent modern carbon (pMC) values ranging from 28 ± 2 to 92 ± 4, suggesting a mixture of aged and modern background carbon containing materials. Taken together, these data suggest that 14C in the INL region is predominantly derived from a mixture of aged and modern background sources (e.g. nuclear weapons testing fallout), with insignificant contributions from INL source terms.

The Transient Reactor Test (TREAT) facility resides at the Materials and Fuels Complex (MFC) at Idaho National Laboratory (INL). The TREAT reactor is currently undergoing design and engineering studies for its conversion from a high enriched uranium (HEU) to a low-enriched uranium (LEU) core. The conceptual design of the LEU fuel element identified two main design differences compared with the HEU fuel element; namely, it will contain four times more fissionable material in its graphite matrix and distinct nuclear-grade Zirconium alloy, as Zircaloy-3 was used in the HEU fuel assembly and is not commercially available currently. These design changes may impact the magnitude of chemical interaction between fuel and cladding materials during physical contact under expected TREAT operation conditions and, therefore, was evaluated through a combination of experimental testing and thermodynamic modeling in order to determine implications for the fuel assembly. In this study, two potential cladding material types, Zircaloy-4 or Zr-1Nb alloys, were evaluated, and it was found for both material types that the extent of interaction and specific chemical reactions are minimal and no detrimental effect on the overall cladding properties is observed. The resulting interaction layer of 3–6 μm was measured after a 2-week exposure at 820 °C. The thermodynamic analysis was extended to temperatures beyond the TREAT reactor operation and accident conditions in order to give some insight that may be of interest for other reactor systems as the High Temperature Gas Reactors (operation above 1000 °C) and for Nuclear Reactor Severe Accident phenomenology study where the UO2 fuel could reach temperatures over 2800 °C and melt.

Surface contamination was quantified over a distributed source of activated potassium bromide from three detonations of Radiological Dispersal Devices (RDDs) at the Idaho National Laboratory Radiological Response Training Range, with a maximum sampled area of 19,900 m2, to provide a baseline comparison with other rapid, remote mapping methods. Measurements were obtained with a cerium bromide sensor collimated to a field of view of 3.14 m2, using lead shielding, and towed behind a ground vehicle. Sensor response correction factors for activated potassium bromide were calculated through simulation with SWORD to obtain activity per meter-squared. Continuous maps were produced by interpolating coverage from lawnmower raster scans. Radiological data was overlaid with aerial imagery from an automated unmanned aerial vehicle flight to provide contextual geological information relative to contamination levels. The contamination distribution measurements will be compared to unmanned aerial vehicle methods in future work.

The initial phase of Plutonium-238 (Pu-238) production for radioisotope thermal generation is described here in detail. Two dosimeters with/without cadmium sleeve containing Neptunium-237 (Np-237) and other neutron sensors were inserted into the low power Advanced Test Reactor Critical (ATRC) for Pu-238 production testing. The gamma-ray energy measurements from Np-238, the short-lived intermediate product, confirmed that sizable amount of Pu-238 can be produced in the full power Advanced Test Reactor (ATR). The Pu-238 production determined from the irradiation experiment was in accord with the modelling predictions. Detailed studies of the Au/Cu sensors for thermal and epithermal flux analysis and their consistency for sensors in different locations, bare and in Cd-sleeve, provided confidence in the Pu-238 production data.

Thermocouples are generally used to provide real-time temperature indications in instrumented tests performed at materials and test reactors. Melt wires or paint spots are often included in such tests as an independent technique of detecting peak temperatures incurred during irradiation. In addition, less expensive static capsule tests, which have no leads attached for real-time data transmission, often rely on melt wires and paint spots as a post-irradiation technique for peak temperature indication. Unfortunately, these techniques are limited in that they can only detect whether a single temperature is or is not exceeded. Silicon carbide (SiC) monitors are advantageous because a single monitor can be used to determine the peak temperature reached within a relatively broad range (200 – 800°C). Although the use of SiC monitors was proposed more than five decades ago, the ultimate performance limits of this technique are not fully understood. The Nuclear Science User Facilities (NSUF) is the United States Department of Energy Office of Nuclear Energy's only designated nuclear energy user facility. Its mission is to provide nuclear energy researchers access to world-class capabilities and to facilitate the advancement of nuclear science and technology. This mission is supported by providing access to state-of-the-art experimental irradiation testing, post irradiation examination facilities, and high performance computing capabilities as well as technical and scientific assistance for the design and execution of projects. As part of an NSUF project, low dose silicon carbide monitors were irradiated in the Belgian Reactor 2 and were then evaluated both at the SCK•CEN and at Idaho National Laboratory’s High Temperature Test Laboratory to determine their peak temperature achieved during irradiation. The technical significance of this work was that the monitors were irradiated to a dose that was significantly less than recommended in published literature. This paper will discuss the evaluation process, the irradiation test, and the performance of the low dose silicon carbide temperature monitors.

Traditional methods of manufacturing nuclear fuel are labor intensive and costly, generally including extrusion or sintering processes and the use of substantially pure materials. The US Department of Energy is working to develop accident tolerant nuclear reactor fuels as alternatives to Zircaloy-4-uranium dioxide (UO2) fuel systems. This could increase existing safety margins for nuclear plants while improving nuclear plant performance and decreasing maintenance and operational costs. Idaho National Laboratory has worked with Westinghouse to develop a new process called additive manufacturing as an Alternative Fabrication Technique (AMAFT) centered on the production of one such accident tolerant fuel, uranium silicide (U3Si2). A previous feasibility study performed with surrogate material demonstrates on a manual benchtop experimental setup that U3Si2 fuel pellets can be manufactured by selective laser melting (SLM). This report provides preliminary concepts to aid in process scale-up, including improvements that an automated system can provide for optimized and repeatable material properties. Automation and robotics provide further acceleration of commercialization with the potential to reduce manufacturing time, costs, and decreased material waste while decreasing or even eliminating direct handling of radioactive materials.

The present lack of a domestic fast neutron flux irradiation capability combined with continued development of fast reactor fuels in the U.S. motivated an innovative engineering solution to utilize a unique neutron flux tailoring capability in the Advanced Test Reactor at the Idaho National Laboratory. To achieve the objectives of the fast reactor fuel irradiation tests, the incident neutron flux was hardened substantially by placing fueled irradiation capsules inside specially designed cadmium shrouds. Use of cadmium prevents thermal neutrons from reaching the fuels being tested and alleviates the plutonium self-shielding that would normally arise during irradiations of high density, highly enriched fuels in a thermal neutron spectrum. The present paper illustrates the profound effect this engineered solution has on the efficacy of the experiments. Based on the comparison of post-irradiation measurements of the columnar grain region in fast reactor mixed-oxide fuels with fuel performance calculations, it is demonstrated that thermal conditions achieved in these cadmium-shrouded fuel experiments are substantially prototypic of a sodium fast reactor and are suitable for concept-screening tests supporting development of new fast reactor fuels. It is also shown that if the experiments were conducted in an unmodified ATR neutron spectrum, gross plutonium self-shielding would cause a strong depression of the fission power at the fuel centerline preventing fuel restructuring, a hallmark feature of mixed oxide fuel behavior under fast reactor conditions. Recognizing the need for testing metallic fuels for fast reactors, the impact of neutron flux energy spectrum on the radial temperature distribution in metallic fuel is investigated. It is shown that the use of Cd shrouds allows to attain radial temperature distributions nearly identical to those that exist in an SFR.

A conceptual low-enrichment uranium (LEU) fuel design has been developed for the Advanced Test Reactor (ATR) at Idaho National Laboratory. The ATR is currently fueled with a high-enrichment fuel but is slated to be converted to LEU under programs led by the National Nuclear Security Administration of the U.S. Department of Energy. A conceptual LEU fuel design, the Enhanced LEU Fuel (ELF), has been developed assuming power peaking control through the use of variable fuel meat thicknesses and no use of burnable poison. In initial work, this design was shown to satisfy performance requirements for ATR operation. Following these design calculations, a safety analysis process was initiated to demonstrate that the ELF design would successfully meet safety limits for postulated accident conditions. Those calculations, performed using RELAP5 and ATR-SINDA, require physics analysis to provide spatial power distributions and kinetics parameters for various core operations configurations. This article descri...

The PHENIX Project is a 6-year U.S./Japan bilateral, multi-institutional collaboration program for the Technological Assessment of Plasma Facing Components for DEMO Reactors. The goal is to address the technical feasibility of helium-cooled divertor concepts using tungsten as the armor material in fusion power reactors. The project specifically attempts to (1) improve heat transfer modeling for helium-cooled divertor systems through experiments including steady-state and pulsed high-heat-load testing, (2) understand the thermomechanical properties of tungsten metals and alloys under divertor-relevant neutron irradiation conditions, and (3) determine the behavior of tritium in tungsten materials through high-flux plasma exposure experiments. The High Flux Isotope Reactor and the Plasma Arc Lamp facility at Oak Ridge National Laboratory, the Tritium Plasma Experiment facility at Idaho National Laboratory, and the helium loop at Georgia Institute of Technology are utilized for evaluation of the respo...

Accident tolerant fuel and cladding materials are being investigated to provide a greater resistance to fuel degradation, oxidation and melting if long-term cooling is lost in a Light Water Reactor (LWR) following an accident such as a Station Blackout (SBO) or Loss of Coolant Accident (LOCA). Researchers at UW-Madison are analyzing an SBO sequence and examining the effect of a loss of auxiliary feedwater (AFW) with the MELCOR systems code. Our research work considers accident tolerant cladding materials (e.g., FeCrAl alloy) and their effect on the accident behavior. We first gathered the physical properties of this alternative cladding material via literature review and compared it to the usual zirconium alloys used in LWRs. We then developed a model for the Surry reactor for a Short-term SBO sequence and examined the effect of replacing FeCrAl for Zircaloy cladding. The analysis uses MELCOR, Version 1.8.6 YR, which is developed by Idaho National Laboratory in collaboration with MELCOR developers at Sandia National Laboratories. This version allows the user to alter the cladding material considered, and our study examines the behavior of the FeCrAl alloy as a substitute for Zircaloy. Our benchmark comparisons with the Sandia National Laboratory’s analysis of Surry using MELCOR 1.8.6 and the more recent MELCOR 2.1 indicate good overall agreement through the early phases of the accident progression. When FeCrAl is substituted for Zircaloy to examine its performance, we confirmed that FeCrAl slows the accident progression and reduce the amount of hydrogen generated. Our analyses also show that this special version of MELCOR can be used to evaluate other potential ATF cladding materials, e.g., SiC as well as innovative coatings on zirconium cladding alloys, Cr, ZrSi, TiAlC, and TiSiC. This ATF performance analysis platform will also be used to support experimental work underway in our Integrated Research Project.

The tristructural isotropic (TRISO) fuel particle is arguably the most robust nuclear fuel form ever developed. TRISO fuels have been realized in high temperature gas-cooled reactors (HTGRs) and have been proposed for many other nuclear energy applications including light water reactors, microreactors, nuclear thermal propulsion, and salt-cooled reactors. Significant data exist for steady-state irradiation testing and out-of-pile core conduction cooldown testing of TRISO fuel relevant to conventional HTGR applications. However, there is a lack of in-pile transient test data, especially related to fast transients, for advanced applications of TRISO fuel particles. This review article outlines the proposed advanced applications of TRISO particles, the existing in-pile transient and accident testing data, and highlights the potential value of additional in-pile transient test data for these new applications. Specifically, this review identifies the need for in-pile testing data to support licensing of TRISO fuel in light water reactors, microreactors, nuclear thermal propulsion, and salt-cooled reactor applications. In advanced non-HTGR TRISO applications the temperature transients may occur much faster than in HTGR depressurized loss-of-forced cooling (DLOFC) accidents (e.g. 0.05 °C/s in a DLOFC versus up to 1000 °C/s in an advanced TRISO application). Generally, historical tests from the literature indicate significant fuel particle failure at energy depositions greater than 1400 J/g-fuel, mostly due to kernel melting. Fuel compact graphite matrix failure was observed above 2300 J/g-fuel in historical tests of compacts. Failure thresholds and mechanisms may be different for particles with fuel materials different from uranium dioxide (such as uranium oxycarbide and uranium nitride), particles with different irradiation histories, and particles with different structure (layers and relative thickness). Although kernel melting was the dominant failure mechanism during historical tests, other failure mechanisms may be possible. One example is thermal stresses between coating layers during quick increases in temperature which could result in crack propagation and delamination of coating layers. Another example is matrix failure due to stresses, for example a graphite matrix in conventional HTGR-like fuel or silicon carbide matrix in fully ceramic microencapsulated fuel. The transient reactor test facility (TREAT) at Idaho National Laboratory is an ideal location for these fuel tests to be carried out in the United States.

The Advanced Gas Reactor-5/6/7 (AGR-5/6/7) experiment is the fourth and final experiment in the AGR experiment series and will serve as the formal fuel qualification test for the TRISO fuels under development by the U.S. Department of Energy. Certain locations in this experiment reach temperatures higher than any of the previous AGR tests, up to 1500°C. Such extreme temperatures create unique challenges for thermocouple-based temperature measurements. High-temperature platinum-rhodium thermocouples (Types S, R, and B)and tungsten-rhenium thermocouples (Type C) suffer rapiddecalibration due to transmutation of the thermoelements fromneutron absorption. For lower temperature applications, previousexperience with Type K thermocouples in nuclear reactors haveshown that they are affected by neutron irradiation only to alimited extent. Similarly, Type N thermocouples, which are morestable than Type K at high temperatures, are only slightly affectedby neutron fluence. Until recently, the use of these nickel-basedthermocouples was limited when the temperature exceeds 1050°Cdue to drift related to phenomena other than nuclear irradiation.Recognizing the limitations of existing thermometery to measuresuch high temperatures, the sponsor of the AGR-5/6/7 experimentsupported a development and testing program for thermocouplescapable of low drift operation at temperatures above 1100°C. High Temperature Irradiation Resistant Thermocouples (HTIR-TCs)based on molybdenum/niobium thermoelements have been underdevelopment at Idaho National Laboratory (INL) since circa 2004. A step change in accuracy and long-term stability of thisthermocouple type has been achieved as part of the AGR-5/6/7thermometry development program. Additionally, long-termtesting (9000+ hrs) at 1250°C of the Type N thermocouplesutilizing a customized sheath developed at the University ofCambridge has been completed with low drift results. Both theimproved HTIR and the Cambridge Type N thermocouple typeshave been incorporated into the AGR-5/6/7 test, which beganirradiation in February 2018 in INL’s Advanced

A project to measure neutron capture cross sections of a number of actinides in a reactor environment by Accelerator Mass Spectrometry (AMS) at the ATLAS facility of Argonne National Laboratory is underway. This project will require the precise and accurate measurement of produced actinide isotopes in many (>30) samples irradiated in the Advanced Test Reactor at Idaho National Laboratory with neutron fluxes having different energy distributions. The AMS technique at ATLAS is based on production of highly-charged positive ions in an electron cyclotron resonance (ECR) ion source followed by acceleration in the ATLAS linac and mass-to-charge ( m / q ) measurement at the focus of the Fragment Mass Analyzer. Laser ablation was selected as the method of feeding the actinide material into the ion source because we expect it will have higher efficiency and lower chamber contamination than either the oven or sputtering techniques, because of a much narrower angular distribution of emitted material. In addition, a new multi-sample holder/changer to allow quick change between samples and a computer-controlled routine allowing fast tuning of the accelerator for different beams, are being developed. An initial test run studying backgrounds, detector response, and accelerator scaling repeatability was conducted in December 2010. The project design, schedule, and results of the initial test run to study backgrounds are discussed.

This project performed thermodynamic, kinetic, and mineral structural studies on the reactivity of phyllosilicate Fe(II/III) with metal-reducing bacteria, and with two important poly-valent DOE contaminants (chromium and uranium) that show high mobility in their oxidized state. We focused on Fe-bearing phyllosilicates because these are important components of the reactive, fines fraction of Hanford, Oak Ridge, and Idaho National Laboratory sediments. Iron-bearing phyllosilicates strongly influence the redox state and mobility of Cr and U because of their limited hydraulic conductivity, high specific surface area, and redox reactivity. This was a collaborative project between Penn State (W.D. Burgos – PI), Miami University (H. Dong – Co-PI), and Argonne National Laboratory (K. Kemner and M. Boyanov – Co-PIs). Penn State and Miami University were funded together but separately from ANL. This report summarizes research findings and publications produced by Penn State and Miami University.

The Transient Reactor Test (TREAT) that resides at the Materials and Fuels Complex (MFC) at Idaho National Laboratory (INL), first achieved criticality in 1959, and successfully performed many transient tests on nuclear fuel until 1994 when its operations were suspended [1]. Resumption of operations at TREAT was approved in February 2014 to meet the U.S. Department of Energy (DOE) Office of Nuclear Energy’s objectives in transient testing of nuclear fuels. The National Nuclear Security Administration’s is converting TREAT from its existing highly enriched uranium (HEU) core to a new core containing low enriched uranium (LEU) (i.e., U-235< 20% by weight) [1]. The results of the FSAR for TREAT can be found in [2]. The TREAT Conversion project is currently progressing with conceptual design phase activities. Dimensional stability of the fuel element assemblies, predictable fuel can oxidation and sufficient heat conductivity by the fuel blocks are some of the critical performance requirements of the new LEU fuel. Furthermore to enable the design team to design fuel block and can specifications, it is amongst the objectives to evaluate TREAT LEU fuel and cladding material’s chemical interaction. This information is important to understand the viability of Zr-based alloys and fuel characteristics for themore » fabrication of the TREAT LEU fuel and cladding. Also, it is very important to make the right decision on what type of nuclear fuel will be used at TREAT. In particular, one has to consider different oxides of uranium, and most importantly, UO2 vs U3O8 [1]. In this report, the results are documented pertaining to the choice mentioned above (UO2 vs U3O8). The conclusion in favor of using UO2 was made based on the analysis of historical data, up-to-date literature, and self-consistent calculations of phase equilibria and thermodynamic properties in the U-O and U-O-C systems. The report is organized as follows. First, the criteria that were used to make the choice are analyzed. Secondly, existing historical data and current literature were reviewed. This analysis was supplemented by the construction and examination of the U-O and U-O-C phase diagrams at pressure close to negligent, thereby mimicking the conditions in which nuclear fuel is supposed to function inside the zirconium-based cladding in the reactor. Finally, our conclusion in favor of the UO2 down selection was summarized and explained in the last Section of this document.« less

Accident tolerant fuels (ATF) are being developed in response to the Fukushima Daiichi accident in Japan. One of the options being pursued is U-Si fuels, such as the U3Si2 and U3Si5 compounds, which benefit from high thermal conductivity (metallic) compared to the UO2 fuel (semi-conductor) used in current Light Water Reactors (LWRs). The U-Si fuels also have higher fissile density. In order to perform meaningful engineering scale nuclear fuel performance simulations, the material properties of the fuel, including the response to irradiation environments, must be known. Unfortunately, the data available for U-Si fuels are rather limited, in particular for the temperature range where LWRs would operate. The ATF HIP is using multi-scale modeling and simulations to address this knowledge gap. Even though Density Functional Theory (DFT) calculations can provide useful answers to a subset of problems, they are computationally too costly for many others, including properties governing microstructure evolution and irradiation effects. For the latter, semi-empirical potentials are typically used. Unfortunately, there is currently no potential for the U-Si system. In this brief report we present initial results from the development of a U-Si semi-empirical potential based on the Modified Embedded Atom Method (MEAM). The potential should reproduce relevant partsmore » of the U-Si phase diagram as well as defect properties important in irradiation environments. This work also serves as an assessment of the general challenges associated with the U-Si system, which will be valuable for the efforts to develop a U-Si Tersoff potential undertaken by Idaho National Laboratory (also part of the ATF HIP). Going forward the main potential development activity will reside at INL and the work presented here is meant to provide input data and guidelines for that activity. The main focus of our work is on the U3Si2 and U3Si5 compounds, because they are the main nuclear fuel candidates. U3Si5 is derived from USi2 in the AlB2 structure by creating 1/6 vacant sites on Si sublattice. The ordering of these vacancies will not be studied in any detail here.« less

One of the primary INL missions is to support development of advanced fuels with the goal of creating reactor fuels that produce less waste and are easier to store. The Advanced Fuel Cycle Initiative (AFCI) Glovebox in the Fuel Manufacturing Facility (FMF) is used for several fuel fabrication steps that involve transuranic elements, including americium. The AFCI glove box contains equipment used for fuel fabrication, including an arc melter – a small, laboratory-scale version of an electric arc furnace used to make new metal alloys for research – and an americium distillation apparatus. This overview summarizes key findings related to the investigation into the releases of airborne radioactivity that occurred in the AFCI glovebox room in late August and early September 2014. The full report (AFCI Glovebox Radiological Release – Evaluation, Corrective Actions and Testing, INL/INL-15-36996) provides details of the identified issues, corrective actions taken as well as lessons learned

An initial and preliminary analysis to determine the operating conditions for the Next Generation Nuclear Plant intermediate heat exchanger (IHX) that will transfer heat from the reactor primary system to demonstration hydrogen production plants was discussed in Chap. 3. In this chapter, we will expand on that point. The Department of Energy, under the leadership of the Idaho National Laboratory, is currently investigating two primary options for the production of hydrogen using a high-temperature reactor as the power source. These options are high-temperature electrolysis (HTE) and SI thermochemical hydrogen production processes. However, since the SI process relies entirely on process heat from the reactor, while the HTE process relies primarily on electrical energy, with only a small amount of process heat required, the design of the IHX is dictated by the SI process heat requirements. Therefore, the IHX operating conditions were defined assuming 50 MWt would be available for the production of hydrogen using the SI process.

Idaho National Laboratory (INL) operates a mass separator that is currently producing high purity isotopes for use as internal standards for high precision isotope dilution mass spectrometry (IDMS). In 2008, INL began the revival of the vintage 1970s era instrument. Advancements thus far include the successful upgrading and development of system components such as the vacuum system, power supplies, ion-producing components, and beam detection equipment. Progress has been made in the separation and collection of isotopic species including those of Ar, Kr, Xe, Sr, and Ba. Particular focuses on ion source improvements and developments have proven successful with demonstrated output beam currents of over 10 μA 138Ba and 350 nA 134Ba from a natural abundance Ba source charge (~2.4 % 134Ba). In order to increase production and collection of relatively high quantities (mg levels) of pure isotopes, several improvements have been made in ion source designs, source material introduction, and ion detection and collection. These improvements have produced isotopes of high purity (>98 %) and in quantities in the tens of micrograms per run. The instrument and results for pure isotope production for IDMS standards will be presented.

The amount of plutonium discarded as wastes from the U.S. nuclear weapons complex appears to be significantly greater than the U.S. Department of Energy's 1996 declaration of its plutonium holdings. This is due to in part to improved radioactive waste characterization and the disposal of plutonium residues originally intended for use in weapons. The Hanford site in Washington State has the largest quantity of plutonium wastes, which pose potentially serious human risks to ground water and the near shore the Columbia River. The department should revise its accounting for plutonium, and take steps to remove plutonium discarded to the environment at Hanford, as it is required to do at Idaho National Laboratory.

Sodium-bearing waste (SBW) containing high concentration of 90Sr was accidentally released to the vadose zone at the Idaho Nuclear Technology and Engineering Center, Idaho National Laboratory, Idaho Falls, ID, in 1972. To investigate the transport and fate of the 90Sr through this 137-m-thick, heterogeneous, variably saturated subsurface, we conducted a two-dimensional numerical modeling using TOUGHREACT under different assumed scenarios (low permeability of an entire interbed or just its surface) for the formation of perched water whose presence reflects the unique characteristics of the geologic materials and stratification at the study site. The results showed that different mechanisms could lead to different flow geometries. The assumption of low permeability for the entire interbed led to the largest saturated zone area and the longest water travel time (55 vs. 43 or 44 yr in other scenarios) from the SBW leakage to the groundwater table. Simulated water travel time from different locations on the land surface to the groundwater aquifer varied from 80 yr. The results also indicated that different mechanisms may lead to differences in the peak and travel time of a small mobile fraction of Sr. The effective distribution coefficient and retardation factor for Sr2+ would change more thanmore » an order of magnitude for the same material during the 200-yr simulation period because of large changes in the concentrations of Sr2+ and competing ions. Understanding the migration rate of the mobile Sr2+ is necessary for designing long-term monitoring programs to detect it.« less

Tritium behaviors in the very high temperature gas reactor (VHTR)/high temperature steam electrolysis (HTSE) system have been analyzed by the TPAC developed by Idaho National Laboratory (INL). The reference system design and conditions were based on the indirect parallel configuration between a VHTR and a HTSE. The analyses were based on the SOBOL method, a modern uncertainty and sensitivity analyses method using variance decomposition and Monte Carlo method. A total of 14 parameters have been taken into account associated with tritium sources, heat exchangers, purification systems, and temperatures. Two sensitivity indices (first order index and total index) were considered, and 15,360 samples were totally used for solution convergence. As a result, important parameters that affect tritium concentration in the hydrogen product have been identified and quantified with the rankings. Several guidelines and recommendations for reducing modeling uncertainties have been also provided throughout the discussions along with some useful ideas for mitigating tritium contaminations in the hydrogen product.

A tritium permeation analyses code (TPAC) has been developed at Idaho National Laboratory (INL) by using MATLAB SIMULINK package for analysis of tritium behaviors in the VHTR integrated with hydrogen production and process heat application systems. The modeling is based on the mass balance of tritium-containing species and hydrogen (i.e., HT, H 2 , HTO, HTSO 4 , and TI) coupled with a variety of tritium source, sink, and permeation models. The code includes: (1) tritium sources from ternary fission and neutron reactions with 6 Li, 7 Li 10 B, and 3 He; (2) tritium purification system; (3) leakage of tritium with coolant; (4) permeation through pipes, vessels, and heat exchangers; (5) electrolyzer for high temperature steam electrolysis (HTSE); and (6) isotope exchange for SI process. Verification of the code has been performed by comparisons with the analytical solutions, the experimental data, and the benchmark code results based on the Peach Bottom reactor design. The results showed that all the governing equations are well implemented into the code and correctly solved. This paper summarizes all the background, the theory, the code structures, and some verification results related to the TPAC code development at INL.

As part of a process initiated through the Advanced Test Reactor (ATR) National Scientific User Facility (NSUF) program to make Silicon Carbide (SiC) temperature monitors available for experiments, a capability was developed at the Idaho National Laboratory (INL) to complete post-irradiation evaluations of these monitors. INL selected the resistance measurement approach for detecting peak irradiation temperature from SiC temperature monitors. To demonstrate this new capability, comparison measurements were completed by INL and Oak Ridge National Laboratory (ORNL) on identical samples subjected to identical irradiation conditions. Results reported in this paper indicate that the resistance measurement approach yields similar peak irradiation temperatures if appropriate equipment is used and appropriate procedures are followed.

A global collaboration is currently developing solvent extraction separations for the nuclear fuel cycle of the future. The goal is to recover fissionable material for recycle, mitigate proliferation concerns, and mitigate the environmental impact of nuclear waste disposal. Relying on selective metal complexing agents, the radiation stability of these solvent extraction ligands will determine the efficiency and recycle lifetime of any solvent intended for use in this high-radiation environment. This paper reviews work at the Idaho National Laboratory regarding the radiation chemistry of nuclear solvent extraction ligands, with particular emphasis on the reactions of nitrogen-centered radicals.

A new research effort aimed at increasing the catalog of boron delivery agents for boron neutron capture therapy (BNCT) has been undertaken by the International Institute of Nano and Molecular Medicine (IINMM) at the University of Missouri. The MU Research Reactor (MURR) and Idaho National Laboratory (INL) have constructed a facility for small animal BNCT experiments. The facility incorporates silicon and bismuth single crystal filters to produce a thermal neutron beam. The thermal beam has a measured thermal flux of 8.8 × 108 n cm−2 s−1 and a gold cadmium ratio of 130. The neutron and photon dose rates were measured using paired ion chambers.

For the past several years, the Idaho National Laboratory and Ceramatec, Inc. have been studying the feasibility of high temperature solid oxide electrolysis for large-scale, nuclear-powered hydrogen production. Parallel to this effort, the INL and Ceramatec have been researching high temperature solid oxide co-electrolysis of steam/CO2 mixtures to produce syngas, the raw material for synthetic fuels production. When powered by nuclear energy, high temperature co-electrolysis offers a carbon-neutral means of syngas production while consuming CO2. The INL has been conducting experiments to characterize the electrochemical performance of co-electrolysis, as well as validate INL-developed computer models. An inline methanation reactor has also been tested to study direct methane production from co-electrolysis products. Testing to date indicate that high temperature steam electrolysis cells perform equally well under co-electrolysis conditions. Process model predictions compare well with measurements for outlet product compositions. The process appears to be a promising technique for large-scale syngas production.

Electrometallurgical treatment of spent nuclear fuel is performed in the Fuel Conditioning Facility (FCF) at the Idaho National Laboratory (INL) by electrochemically separating uranium from the fission products and structural materials in a vessel called an electrorefiner (ER). To continue processing without waiting for sample analyses to assess process conditions, an ER process model predicts the composition of the ER inventory and effluent streams via multi-component, multi-phase chemical equilibrium for chemical reactions and a numerical solution to differential equations for electro-chemical transport. The results of the process model were compared to the electrorefiner measured data.

As part of the spent fuel treatment program at the Idaho National Laboratory, a vacuum distillation process is being employed for the recovery of actinide products following an electrorefining process. Separation of the actinide products from a molten salt electrolyte and cadmium is achieved by a batch operation called cathode processing. A cathode processor has been designed and developed to efficiently remove the process chemicals and consolidate the actinide products for further processing. This paper describes the fundamentals of cathode processing, the evolution of the equipment design, the operation and efficiency of the equipment, and recent developments at the cathode processor. In addition, challenges encountered during the processing of irradiated spent nuclear fuel in the cathode processor will be discussed.

Mixed-oxide test capsules prepared with weapons-derived plutonium have been irradiated to a burnup of 50 GWd/MT. The mixed-oxide fuel was fabricated at Los Alamos National Laboratory by a master-mix process and has been irradiated in the advanced test reactor at the Idaho National Laboratory. Previous withdrawals of the same fuel have occurred at 9, 21, 30, and 40 GWd/MT. Oak Ridge National Laboratory manages this test series for the Department of Energy’s Fissile Materials Disposition Program. This paper describes the preparation of the mixed-oxide fuel, the equipment design, and the irradiation history of the test capsules, and discusses the significance of the more important observations of the post-irradiation examinations. Code predictions (FRAPCON-3 and TRANSURANUS) are presented and compared with available post-irradiation examination data for the highest and lowest powered mixed-oxide capsules. Fuel performance has been excellent and consistent with code predictions and with existing US and European experience.

The long-term durability of WAXFIX, a paraffin-based grout, was evaluated for in situ grouting of activated beryllium wastes in the subsurface disposal area (SDA), a radioactive landfill at the Radioactive Waste Management Complex, part of the Idaho National Laboratory (INL). The evaluation considered radiological and biological mechanisms that could degrade the grout using data from an extensive literature search and previous tests of in situ grouting at the INL. Conservative radioactive doses for WAXFIX were calculated from the "hottest" (i.e., highest-activity) Advanced Test Reactor beryllium block in the SDA. These results indicate that WAXFIX would not experience extensive radiation damage for many hundreds of years. Calculation ofradiation-induced hydrogen generation in WAXFIX indicated that grout physical performance should not be reduced beyond the effects ofradiation dose on the molecular structure. Degradation of a paraffin-based grout by microorganisms in the SDA is possible and perhaps likely, but the rate of degradation will be at a slower rate than found in the literature reviewed. The calculations showed the outer 0.46-m (18-in.) layer of each monolith, which represents the minimum expected distance to the beryllium block, was calculated to require 1000 to 3600 yr to be consumed. The existing data and estimations of biodegradation and radiolysis rates for WAXFIX/paraffin do not indicate any immediate problems with the use of WAXFIX for grouting beryllium or other wastes in the SDA.

A number of waste components in US defense high level radioactive wastes (HLW) have proven challenging for current Joule heated ceramic melter (JHCM) operations and have limited the ability to increase waste loadings beyond already realized levels. Many of these “troublesome" waste species cause crystallization in the glass melt that can negatively impact product quality or have a deleterious effect on melter processing. Recent efforts at US Department of Energy laboratories have focused on understanding crystallization behavior within HLW glass melts and investigating approaches to mitigate the impacts of crystallization so that increases in waste loading can be realized. Advanced glass formulations have been developed to highlight the unique benefits of next-generation melter technologies such as the Cold Crucible Induction Melter (CCIM). Crystal-tolerant HLW glasses have been investigated to allow sparingly soluble components such as chromium to crystallize in the melter but pass out of the melter before accumulating.The Hanford site AZ-101 tank waste composition represents a waste group that is waste loading limited primarily due to high concentrations of Fe2O3 (also with high Al2O3 concentrations). Systematic glass formulation development utilizing slightly higher process temperatures and higher tolerance to spinel crystals demonstrated that an increase in waste loading of moremore » than 20% could be achieved for this waste composition, and by extension higher loadings for wastes in the same group. An extended duration CCIM melter test was conducted on an AZ-101 waste simulant using the CCIM platform at the Idaho National Laboratory (INL). The melter was continually operated for approximately 80 hours demonstrating that the AZ-101 high waste loading glass composition could be readily processed using the CCIM technology. The resulting glass was close to the targeted composition and exhibited excellent durability in both the as poured state and after being slowly cooled according to the canister centerline cooling (CCC) profile. Glass formulation development was also completed on other Hanford tank wastes that were identified to further challenge waste loading due to the presence of appreciable quantities (>750 g) of plutonium in the waste tanks. In addition to containing appreciable Pu quantities, the C-102 waste tank and the 244-TX waste tank contain high concentrations of aluminum and iron, respectively that will further challenge vitrification processing. Glass formulation testing also demonstrated that high waste loadings could be achieved with these tank compositions using the attributes afforded by the CCIM technology.« less

This activity is supported by the US Nuclear Energy Advanced Modeling and Simulation (NEAMS) Fuels Product Line (FPL) and aims at providing experimental data for the validation of the mesoscale simulation code MARMOT. MARMOT is a mesoscale multiphysics code that predicts the coevolution of microstructure and properties within reactor fuel during its lifetime in the reactor. It is an important component of the Moose-Bison-Marmot (MBM) code suite that has been developed by Idaho National Laboratory (INL) to enable next generation fuel performance modeling capability as part of the NEAMS Program FPL. In order to ensure the accuracy of the microstructure based materials models being developed within the MARMOT code, extensive validation efforts must be carried out. In this report, we summarize our preliminary synchrotron radiation experiments at APS to determine the grain size of nanograin UO2. The methodology and experimental setup developed in this experiment can directly apply to the proposed in-situ grain growth measurements. The investigation of the grain growth kinetics was conducted based on isothermal annealing and grain growth characterization as functions of duration and temperature. The kinetic parameters such as activation energy for grain growth for UO2 with different stoichiometry are obtained and compared with molecular dynamicsmore » (MD) simulations.« less

This activity is supported by the US Nuclear Energy Advanced Modeling and Simulation (NEAMS) Fuels Product Line (FPL). Two major accomplishments in FY 15 are summarized in this report: (1) implementation of the FASTGRASS module in the BISON code; and (2) a Xe implantation experiment for large-grained UO2. Both BISON AND MARMOT codes have been developed by Idaho National Laboratory (INL) to enable next generation fuel performance modeling capability as part of the NEAMS Program FPL. To contribute to the development of the Moose-Bison-Marmot (MBM) code suite, we have implemented the FASTGRASS fission gas model as a module in the BISON code. Based on rate theory formulations, the coupled FASTGRASS module in BISON is capable of modeling LWR oxide fuel fission gas behavior and fission gas release. In addition, we conducted a Xe implantation experiment at the Argonne Tandem Linac Accelerator System (ATLAS) in order to produce the needed UO2 samples with desired bubble morphology. With these samples, further experiments to study the fission gas diffusivity are planned to provide validation data for the Fission Gas Release Model in MARMOT codes.

This activity is supported by the US Nuclear Energy Advanced Modeling and Simulation (NEAMS) Fuels Product Line (FPL) and aims at providing experimental data for the validation of the mesoscale simulation code MARMOT. MARMOT is a mesoscale multiphysics code that predicts the coevolution of microstructure and properties within reactor fuel during its lifetime in the reactor. It is an important component of the Moose-Bison-Marmot (MBM) code suite that has been developed by Idaho National Laboratory (INL) to enable next generation fuel performance modeling capability as part of the NEAMS Program FPL. In order to ensure the accuracy of the microstructure based materials models being developed within the MARMOT code, extensive validation efforts must be carried out. In this report, we summarize our preliminary synchrotron radiation experiments at APS to determine the grain size of nanograin UO2. The methodology and experimental setup developed in this experiment can directly apply to the proposed in-situ grain growth measurements. The investigation of the grain growth kinetics was conducted based on isothermal annealing and grain growth characterization as functions of duration and temperature. The kinetic parameters such as activation energy for grain growth for UO2 with different stoichiometry are obtained and compared with molecular dynamicsmore » (MD) simulations.« less

This report provides a description of the simulation toolkit developed at Lawrence Livermore National Laboratory to support the design of nuclear safeguards experiments using grazing incidence multilayer mirrors in the energy band of uranium (U) and plutonium (Pu) emission lines. This effort was motivated by the data analysis of a scoping experiment at the Irradiated Fuels Examination Facility (IFEL) at Oak Ridge National Laboratory in FY13 and of a benchmark experiment at the Idaho National Laboratory (INL) in FY14 that highlighted the need for predictive tools built around a ray-tracing capability. This report presents the simulation toolkit and relevant results such as the simulated spectra for TMI, MOX, and ATM106 fuel rods based on spent fuel models provided by Los Alamos National Laboratory and for a virgin high 240Pu-content fuel plate, as well as models of the IFEL and INL experiments implemented in the ray tracing tool. The beam position and height were validated against the INL ~60 keV americium data. Examples of alternate configurations of the optics or experimental set-up illustrate the future use of the simulation suite to guide the next IFEL experimental campaign.

The US Department of Energy has presented a plan to use existing reactors at Oak Ridge National Laboratory (ORNL) and Idaho National Laboratory (INL) and processing facilities at ORNL, modified as needed, to produce Pu. The basic capabilities that need to be put into place to produce new Pu are (1) neptunium storage and transport, (2) target fabrication, (3) target irradiation, and (4) chemical processing of irradiated targets to recover Pu. Neptunium currently in storage at INL will be shipped to ORNL during CY 2015. The target design has progressed to a prototypic target design that is expected to be used for production. Initial chemical processing experiments have shown successful recovery of neptunium and plutonium, but overall product purity has not been as high as desired.

Four of the radioactive xenon isotopes ((131m)Xe, (133m)Xe, (133)Xe and (135)Xe) with half-lives ranging from 9 h to 12 days are produced from nuclear fission and can be detected from days to weeks following their production and release. Being inert gases, they are readily transported through the atmosphere. Sources for release of radioactive xenon isotopes include operating nuclear reactors via leaks in fuel rods, medical isotope production facilities, and nuclear weapons' detonations. They are not normally released from fuel reprocessing due to the short half-lives. The Comprehensive Nuclear-Test-Ban Treaty has led to creation of the International Monitoring System. The International Monitoring System, when fully implemented, will consist of one component with 40 stations monitoring radioactive xenon around the globe. Monitoring these radioactive xenon isotopes is important to the Comprehensive Nuclear-Test-Ban Treaty in determining whether a seismically detected event is or is not a nuclear detonation. A variety of radioactive xenon quality control check standards, quantitatively spiked into various gas matrices, could be used to demonstrate that these stations are operating on the same basis in order to bolster defensibility of data across the International Monitoring System. This paper focuses on Idaho National Laboratory's capability to produce three of the xenon isotopes in pure form and the use of the four xenon isotopes in various combinations to produce radioactive xenon spiked air samples that could be subsequently distributed to participating facilities.

The Environmental Protection Agency (EPA) has teamed with Battelle Energy Alliance, LLC (BEA) at Idaho National Laboratory (INL) to facilitate further testing of geologic-fracture-identification methodology at a field site near the Monsanto Superfund Site located in Soda Springs, Idaho. INL has the necessary testing and technological expertise to perform this work. Battelle Memorial Institute (BMI) has engaged INL to perform this work through a Work for Others (WFO) Agreement. This study continues a multi-year collaborative effort between INL and EPA to test the efficacy of using field deployed Cr-39 radon in soil portals. This research enables identification of active fractures capable of transporting contaminants at sites where fractures are suspected pathways into the subsurface. Current state of the art methods for mapping fracture networks are exceedingly expensive and notoriously inaccurate. The proposed WFO will evaluate the applicability of using cheap, readily available, passive radon detectors to identify conductive geologic structures (i.e. fractures, and fracture networks) in the subsurface that control the transport of contaminants at fracture-dominated sites. The proposed WFO utilizes proven off-the-shelf technology in the form of CR-39 radon detectors, which have been widely deployed to detect radon levels in homes and businesses. In an existing collaborative EPA/INL studymore » outside of this workscope,. CR-39 detectors are being utilized to determine the location of active transport fractures in a fractured granitic upland adjacent to a landfill site at the Fort Devens, MA that EPA-designated as National Priorities List (NPL) site. The innovative concept of using an easily deployed port that allows the CR-39 to measure the Rn-222 in the soil or alluvium above the fractured rock, while restricting atmospheric Rn-222 and soil sourced Ra from contaminating the detector is unique to INL and EPA approach previously developed. By deploying a series of these inexpensive detector-casing combinations statistical samples of the Rn-222 flux can be measured, elucidating the most communicative fractures (i.e. fractures that are actively transporting water and gasses). The Rn-222 measurements can then be used as an input to create a more accurate conceptual model to be used for transport modeling and related cleanup activities. If the team’s approach is demonstrated to be applicable to a wide variety of rock types and soil conditions it might potentially offer significant cost saving without a reduction in data quality at Monsanto Superfund and other sites underlain by fracture-dominated bedrock.« less

The Toxic Substances Control Act, 40 CFR 761.65(a)(1) provides an exemption from the one year storage time limit for PCB/radioactive waste. PCB/radioactive waste may exceed the one year time limit provided that the provisions at 40 CFR 761.65(a)(2)(ii) and 40 CFR 761.65(a)(2)(iii) are followed. These two subsections require, (ii) "A written record documenting all continuing attempts to secure disposal is maintained until the waste is disposed of" and (iii) "The written record required by subsection (ii) of this section is available for inspection or submission if requested by EPA." EPA Region 10 has requested the Department of Energy (DOE) to submit an inventory of radioactive-contaminated PCB waste in storage at the Idaho National Laboratory (INL) for the previous calendar year. The annual inventory is separated into two parts, INL without Advanced Mixed Waste Treatment Project (AMWTP) (this includes Battelle Energy Alliance, LLC, CH2M-WG Idaho, LLC, and the Naval Reactors Facility), and AMWTP.

The Office of Material Management and Minimization Fuel Thermo-physical Characterization Project at Pacific Northwest National Laboratory (PNNL) is tasked with using PNNL facilities and processes to receive irradiated low enriched uranium-molybdenum fuel plate samples and perform analyses in support of the Office of Material Management and Minimization Reactor Conversion Program. This work is in support of the Fuel Development Pillar that is managed by Idaho National Laboratory. A key portion of the scope associated with this project was to measure the thermal properties of fuel segments harvested from plates that were irradiated in the Advanced Test Reactor. Thermal diffusivity of samples prepared from the fuel segments was measured using laser flash analysis. Two models, one developed by PNNL and the other developed by the Technische Universitat Munchen (TUM), were evaluated to extract the thermal diffusivity of the uranium-molybdenum alloy from measurements made on the irradiated, layered composites. The experimental data of the “TC” irradiated fuel segment was evaluated using both models considering a three-layer and five-layer system. Both models are in acceptable agreement with one another and indicate that the zirconium diffusion barrier has a minimal impact on the overall thermal diffusivity of the monolithic U-Mo fuel.

Radiometric and mass spectrometric analyses of Cs contamination in the environment can reveal the location of Cs emission sources, release mechanisms, modes of transport, prediction of future contamination migration, and attribution of contamination to specific generator(s) and/or process(es). The Subsurface Disposal Area (SDA) at Idaho National Laboratory (INL) represents a complicated case study for demonstrating the current capabilities and limitations to environmental Cs analyses. (137)Cs distribution patterns, (135)Cs/(137)Cs isotope ratios, known Cs chemistry at this site, and historical records enable narrowing the list of possible emission sources and release events to a single source and event, with the SDA identified as the emission source and flood transport of material from within Pit 9 and Trench 48 as the primary release event. These data combined allow refining the possible number of waste generators from dozens to a single generator, with INL on-site research and reactor programs identified as the most likely waste generator. A discussion on the ultimate limitations to the information that (135)Cs/(137)Cs ratios alone can provide is presented and includes (1) uncertainties in the exact date of the fission event and (2) possibility of mixing between different Cs source terms (including nuclear weapons fallout and a source of interest).

Traditional methods for measuring in-pile temperatures degrade above 1100°C. Hence, the Idaho National Laboratory (INL) initiated a project to explore the use of specialized thermocouples for high ...

The high-temperature characteristics of the modular helium reactor (MHR) make it a strong candidate for producing hydrogen using either thermochemical or high-temperature electrolysis (HTE) processes. Using heat from the MHR to drive a sulfur-iodine (SI) thermochemical hydrogen production process has been the subject of a U.S. Department of Energy sponsored Nuclear Engineering Research Initiative (NERI) project led by General Atomics, with participation from the Idaho National Laboratory (INL) and Texas A&M University. While the focus of much of the initial work was on the SI thermochemical production of hydrogen, recent activities included development of a preconceptual design for an integral HTE hydrogen production plant driven by the process heat and electricity produced by a 600 MW MHR. This paper describes ATHENA analyses performed to evaluate alternative primary system cooling configurations for the MHR to minimize peak reactor vessel and core temperatures while achieving core helium outlet temperatures in the range of 900–1000 °C that are needed for the efficient production of hydrogen using either the SI or HTE process. The cooling schemes investigated are intended to ensure peak fuel temperatures do not exceed specified limits under normal or transient upset conditions, and that reactor vessel temperatures do not exceed American Society of Mechanical Engineers (ASME) code limits for steady-state or transient conditions using standard light water reactor vessel materials. Preconceptual designs for SI and HTE hydrogen production plants driven by one or more 600 MW MHRs at helium outlet temperatures in the range of 900–1000 °C are described and compared. An initial SAPHIRE model to evaluate the reliability, maintainability, and availability of the SI hydrogen production plant is also described. Finally, a preliminary flowsheet for a conceptual design of an HTE hydrogen production plant coupled to a 600 MW modular helium reactor is presented and discussed.