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All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

High-Burnup BWR LOCA Burst Analysis Using High-Fidelity Multiphysics Simulations

The US nuclear industry is looking to improve on the operating economics of the current fleet of light-water reactors (LWRs). One way of achieving this is by operating fuel to higher burnup. In pressurized water reactors (PWRs), relaxing the current burnup limit will allow for cycle length extensions and power uprates; in boiling water reactors (BWRs) it may allow for improved fuel utilization and reduced feed assemblies, as well as more efficient power uprates and increased capacity factors that will support the Administration’s Executive Order to facilitate 5 GW of power uprates at existing nuclear facilities. However, one of the key limitations to operating fuel to higher burnup is the risk of fuel fragmentation, relocation, and dispersal (FFRD). Recognizing the high interest in extending burnup limits, the US Nuclear Regulatory Commission (NRC) has issued Draft Regulatory Guide DG-1434, which defines an approach that would be acceptable to the NRC for addressing FFRD risk. The approach defined will require better understanding of the phenomena leading to FFRD as well as best-estimate simulation methods to understand FFRD risk in high-burnup cores. The Nuclear Energy Advanced Modeling and Simulation program is supporting the FFRD industry challenge problem through development of state-of-the-art, high-fidelity modeling and simulation LWR analysis capabilities; namely, the BISON fuel performance code and the VERA core simulator software. These tools, along with the US NRC TRACE system analysis code, have been utilized for analysis of FFRD risk in both PWR and BWR cores in recent years. The work documented in this report addresses the lack of high-fidelity research for BWRs and builds on a previous activity where the framework has been applied to Cycles 16 through 18 of Limerick Unit 1, a BWR/4, with introduction of 8 high-burnup lead use assemblies (HBLUAs) that were representative of the 8 HBLUAs loaded into Limerick Unit 2 in 2021. VERA was used in this previous activity to model rod-by-rod depletion in these cycles, and its solution was used to initialize a TRACE simulation of a large-break loss-of-coolant accident (LBLOCA) at the end of Cycle 18. In the work documented in this report, the TRACE model was improved by refining the core mesh and utilizing a new feature that allows for capturing the full 3D VERA power distribution in the model. This allows for a more detailed solution for setting BISON boundary conditions. Furthermore, the solutions from VERA and TRACE were used to set up and perform BISON simulations of about 1,000 rods sampled from the core, including all burnup levels. Utilizing two cladding burst models, it was shown that no fuel rods were predicted to burst during the postulated LBLOCA transient. Additionally, a sensitivity study was performed by artificially increasing linear heat rate during the postulated LBLOCA to identify parameters that correlate with rod burst susceptibility. Burnup, fission gas release, and hoop strain were all found to be positively correlated with rod burst susceptibility. Small-break loss-of-coolant accident (SBLOCA) analyses were also performed; these analyses predicted cladding temperature increases that were bounded by the LBLOCA cladding temperatures for all small break sizes studied for this plant. However, future refinements to the plant response assumptions during the SBLOCA could impact the predicted cladding response. Finally, a benchmark study was performed between CTF and TRACE for LOCA conditions to better qualify CTF for BWR LOCA modeling.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Survey of baffle plate samples, specimen machining, and hydrogen measurements

This report outlines preliminary results on the survey and characterization of irradiated baffle plate, baffle former, and flux thimble tube specimens as part of the ongoing investigation into irradiation-induced embrittlement in austenitic stainless steels at ambient (room temperature) conditions. The specimens originated from commercial pressurized water reactor (PWR) components, covering displacement damage doses ranging from approximately 0.065 to 75 dpa. Initial scanning electron microscopy (SEM) surveys revealed that specimen surfaces exhibited fine machining marks and in-service-formed oxide layers on the side surfaces of the analyzed specimens. The oxide layers revealed specific features resembling localized "pitting-like" corrosion. An additional set of specimens representing in-service oxidation is being preserved for future microstructure analysis work. As believed, this will provide additional insights into long-term material degradation in PRWs. Specimen machining challenges emerged due to a complex failure of the electric discharge machining (EDM) system located in hot area. Despite partial restoration, issues persist with the EDM’s secondary power supply, necessitating the exploration of alternative EDM or computer numerical control (CNC) machining approaches to facilitate tensile specimen preparation. Currently, low speed saw cutting is in progress to prepare specimens for hydrogen measurements and general microstructure analysis. The near-term goals include completing tensile specimen machining for advanced mechanical testing and characterizing fracture mechanisms, and stress-corrosion cracking testing, ultimately aiming to identify and mitigate the ambient-condition intergranular cracking through targeted post-irradiation annealing strategies.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

PIE-Enabled Study of Aqueous Corrosion & Zr Hydriding in Cr-Coated Cladding: M3GV-23PN0101132

Chromium-coated zirconium alloy cladding is under investigation as an accident-tolerant fuel (ATF) concept to extend performance to higher burnups via improved oxidation resistance and reduced hydrogen pickup. However, hydrothermal corrosion behavior and hydrogen transport mechanisms governing in-service hydriding of these coatings remain poorly understood. In a collaborative study between Pacific Northwest National Laboratory (PNNL), Idaho National Laboratory (INL), and the University of Huddersfield, cold spray (CS) and physical vapor deposition (PVD) Cr-coated Optimized ZIRLO™ cladding samples were characterized after exposure to PWR-simulated water chemistry under both in-core (neutron irradiation) and out-of-core (aqueous-only) conditions at the MIT Research Reactor. Multi-scale characterization was performed independently at PNNL and INL to evaluate coating integrity, microstructural evolution, Cr/Zr interface chemistry, and hydride formation. Both laboratories observed a consistent divergence in hydriding behavior: out-of-core CS Cr-coated cladding exhibited elevated hydride concentrations exceeding those of uncoated cladding, whereas in-core CS samples showed substantially suppressed hydriding. In-core specimens also displayed irradiation-specific features, including nanoscale voids within the Cr coating and radiation-induced segregation clusters in the Zr substrate. CS coatings retained an interdiffusion-free Cr/Zr bond with no intermetallic layer, whereas PVD coatings exhibited inferior quality with visible cracks and pores. An automated image-based hydride quantification method systematically overestimated bulk hydrogen content relative to inert gas fusion measurements, underscoring the need for standardized sample preparation and reference standards. Interpretation of hydride nucleation and growth is complicated by the open inner-diameter of the specimens, which provides additional hydrogen pathways. The results motivate further foundational studies to support predictive models of hydrothermal corrosion and hydrogen transport in unirradiated and irradiated Cr-coated cladding.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Long-Length SiC/SiC Tube Irradiation Experiment at LWR Temperatures Inserted into HFIR

This report describes the pre-irradiation characterization and successful assembly of a High Flux Isotope Reactor (HFIR) irradiation experiment designed to assess radiation-induced lateral bowing of silicon carbide fiber–reinforced, silicon carbide matrix composite (SiC/SiC) components under a radial fast neutron flux gradient at representative light-water reactor (LWR) temperatures of approximately 300 °C. The over-arching goal of this work is to validate thermomechanical models of SiC/SiC composite fuel assembly distortion used to predict the extent of lateral bowing and dose at which bowing reaches a maximum value. Ultimately, these models are used to ensure that fuel assembly distortion does not block coolant channels, interfere with control rod/blade movements, and/or impact local reactivity. The experiment contains six cladding tube specimens that are approximately 558.5 mm long with representative pressurized water reactor (PWR) outer and inner diameters of 9.5 mm and 7.5 mm, respectively. The specimens were thoroughly characterized prior to irradiation using traditional dimensional inspection and surface profilometry so that these measurements can later be compared with similar measurements that will be made post-irradiation to determine radiation-induced deformations. Furthermore, fine engraving markers were inscribed along the outer surfaces of the specimen and mapped using analytic photography and a three-dimensional stage. This technique allowed for accurate measurements of the marker spacings, which can be compared with similar measurements that will be made post-irradiation to provide local radiation-induced strain mapping. A light curtain optical micrometer was also employed to measure the pre-irradiation bowing of each specimen. The experiment was successfully assembled and inserted into position VXF-19 for HFIR cycle 516, which began May 12, 2026. The experiment will be removed after cycle 519, which is currently scheduled to end in December 2026.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2026

Refining a Novel Process Monitoring Method to Safeguard Continuously Cycling Designs Using Isotopic Ratios

In advanced reactor (AR) designs, a common feature is continuous chemical processing and circulation of the nuclear material. This work bridges a significant measurement gap in safeguarding reactors with circulating fuel or continuous refueling by leveraging and building on the isotope ratio method first developed by our team under an FY21 Advanced Reactors International Safeguards Engagement (ARISE) project (Uribe et al. 2021). In circulating fuel designs, the radioisotope inventory changes from traditional effects (e.g., radioactive decay, fission) but also includes material transport due to pressure and temperature gradients. Such designs may also require regular or continuous additions or removals during operation, which significantly increases the rate of inventory change compared to traditional pressurized water reactor (PWR)s. Thus, directly tracking the nuclear inventory is ineffective since the isotopes are continuously added and removed. The isotope ratio method instead focuses on detecting changes to the input and output flows of radioisotopes. Previous work showed that for well-chosen pairs of isotopes, the isotopic ratio provides a sensitive and lasting indicator of deviation from normal conditions (e.g., startup, shutdown, diversion). The isotope ratio method is a process monitoring method with potential for application in for forward-looking approaches to International Atomic Energy Agency (IAEA) safeguards. The original process monitoring method was developed for a specific case—the decay tank of a thorium-fueled molten salt breeder reactor. In this expanded work, we explored other types of reactors and processes with nonstationary (e.g., flowing) nuclear material, which are difficult to safeguard with traditional methods because of the transient nature of the systems. The goal of the present work is to generalize the isotope ratio method for use in processes with continuously flowing nuclear material. All continuous processes have an average time for isotopes to be replaced in the system. The isotope ratio method works by choosing radioisotopes with half-lives both above and below the processing time. The present work seeks to explore which isotopes are suitable for the method by simulating the nuclear inventory, radioisotope emissions, and detector responses for several classes of advanced reactors. While the method can in principle be applied to other processes (e.g., enrichment or reprocessing facilities), the present work limits scope to ARs with continuously flowing fuel. Section 2 details the mathematics supporting the isotope ratio method, and Section 3 introduces the representative ARs selected for this work. Section 4 discusses how each reactor was analyzed, and Section 5 showcases the results for each representative reactor. Finally, Section 6 provides concluding remarks and suggests pathways for further analysis.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2026

Non-Neutron Transmutation of Used Nuclear Fuel (Final Report)

The primary goal of this study is to develop a national facility concept for transmuting long-lived fission products (LLFP) to substantially reduce the disposal impact by minimizing the need for a geologic-timescale repository. As a charter for this study, the national transmutation facility was required to reduce the radiotoxicity and decay heat of LLFP isotopes by at least 90% relative to their values at discharge from a commercial LWR, while consuming less than 10% of the reactor's energy. The identified LLFP isotopes are Se-79, Zr-93, Tc-99, I-129, Sn-126, and Cs-135, whose radiotoxicity is about 99% of the total radiotoxicity of all fission products at 1,000 years. Approximately ~72 kg of LLFPs is discharged every year from a 1,000 MWe commercial or advanced nuclear reactor. First, LLFP transmutation options with non-neutron beams (photons and protons) were explored. The study concluded that LLFP transmutation is feasible with high-energy, high-intensity photons or protons, but impractical on an engineering scale due to low transmutation rates and the high energy requirements to produce the desired photon or proton beams. As alternatives, LLFP transmutation options with neutrons from fission, fusion, and spallation reactions were additionally explored. The transmutation options using advanced critical reactors are attractive only for selective LLFP isotopes because the production rates of several LLFP isotopes (Zr-93, Sn-126, and Cs-135) from fission reactions are larger than the transmutation rates. The transmutation options with only spallation neutrons are favorable to transmute all LLFP isotopes, but as a tradeoff, the net transmutation rates are reduced. The national transmutation facility concept was developed following an exploration of transmutation options using various incident particles. The proposed national LLFP transmutation comprises a dedicated molten-salt reactor (MSR), a proton accelerator, and a spallation neutron-based transmuter. The MSR power was set at 300 MWt and 120 MWe, with the thermal power approximately 10% of that of a commercial 1,000 MWe PWR. The electricity generated by the MSR powers the accelerator and transmuter. The accelerator produces 1 GeV, 30 mA protons, which are introduced into the spallation neutron-based transmuter. The spallation neutron-based transmuter consists of a central spallation target and LLFP target pins merged in a heavy water tank. The six LLFP isotopes are separated into two groups. Tc-99, I-129, and Se-79, having larger neutron cross sections, belong to group A, while Zr-93, Sn126, and Cs-135, having smaller neutron cross sections, belong to group B. Then, for effective transmutation, LLFPs in groups A and B are transmuted in the dedicated MSR and in a spallation neutron-based transmuter, respectively. The estimated capital cost of the national transmutation facility is approximately $\$$3.1B, and its annual O&M cost is expected to be ~$\$$182M. Radiotoxicity and decay heat of LLFPs were calculated and compared with those of the original LLFPs. It was assumed that the targets were made with elementwise LLFP rather than isotopic LLFP, owing to the potentially high cost of isotopic separation from used nuclear fuels. The decay heat of LLFPs can be reduced by more than 90% using a single national transmutation facility. However, radiotoxicity decreases by 79–84%, which does not meet the transmutation performance requirement, primarily because Cs-135 is produced rather than depleted. Thus, to meet the design requirement, Cs-135 should be separated from other Cs isotopes and irradiated in a spallation neutron-based transmuter. Then, radiotoxicity decreases by ~92%.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

SEM/EBSD Characterization of Baffle-Former Bolts Designated for High Energy X-ray Microscopy Studies at the Advanced Photon Source

This report documents the microstructural and surface characterization of two high-fluence baffle-former bolts (BFBs) harvested from a commercial pressurized water reactor (PWR) after nearly four decades of service. These bolts experienced extreme conditions including high neutron fluence, elevated temperatures, corrosive primary water, and sustained mechanical loads, all of which contribute to degradation mechanisms such as irradiation-assisted stress corrosion cracking (IASCC). Using scanning electron microscopy (SEM) and electron backscatter diffraction (EBSD), the study assessed bulk and near-surface microstructures, grain boundary characteristics, and in-service oxide layers. The results revealed significant differences between bolts exposed to 41 dpa and 29 dpa, including differences in recrystallization extent, deformation structures, and oxidation morphology. These findings establish critical baseline data for integration with advanced synchrotron-based high-energy X-ray microscopy (HEXM) and computational crystal plasticity models. Ultimately, the work supports improved mechanistic understanding and predictive capability regarding the long-term performance of reactor core internal components.

Lach, Tim [Oak Ridge National Laboratory (ORNL), O↗
DOE OSTI2025

Nuclear Reactor Heat Extraction for Synthetic Fuel Plants

This report looks at the viability and best approach when producing synthetic fuels using heat and power supplied by advanced nuclear energy systems. The report looks at a low temperature integration pathway with four types of advanced reactors: a pressurized water reactor (PWR), an advanced light water reactor (A-LWR), a sodium fast reactor (SFR), and a high-temperature gas-cooled reactor (HTGR). A failure modes and effects analysis (FMEA) of the coupling system between the nuclear plant and the synthetic fuel systems is performed with the goal of identifying the reliability of such a thermal delivery system. Furthermore, a high temperature pathway is investigated for synfuel coupling to determine if this is more efficient and more cost effective as a coupling approach.

10 - SYNTHETIC FUELS↗
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↗
NASA NTRS1991

WTEC panel report on European nuclear instrumentation and controls

Control and instrumentation systems might be called the 'brain' and 'senses' of a nuclear power plant. As such they become the key elements in the integrated operation of these plants. Recent developments in digital equipment have allowed a dramatic change in the design of these instrument and control (I&C) systems. New designs are evolving with cathode ray tube (CRT)-based control rooms, more automation, and better logical information for the human operators. As these new advanced systems are developed, various decisions must be made about the degree of automation and the human-to-machine interface. Different stages of the development of control automation and of advanced digital systems can be found in various countries. The purpose of this technology assessment is to make a comparative evaluation of the control and instrumentation systems that are being used for commercial nuclear power plants in Europe and the United States. This study is limited to pressurized water reactors (PWR's). Part of the evaluation includes comparisons with a previous similar study assessing Japanese technology.

White, James D.↗
NASA NTRS2007

Development of the Functional Flow Block Diagram for the J-2X Rocket Engine System

The J-2X program calls for the upgrade of the Apollo-era Rocketdyne J-2 engine to higher power levels, using new materials and manufacturing techniques, and with more restrictive safety and reliability requirements than prior human-rated engines in NASA history. Such requirements demand a comprehensive systems engineering effort to ensure success. Pratt & Whitney Rocketdyne system engineers performed a functional analysis of the engine to establish the functional architecture. J-2X functions were captured in six major operational blocks. Each block was divided into sub-blocks or states. In each sub-block, functions necessary to perform each state were determined. A functional engine schematic consistent with the fidelity of the system model was defined for this analysis. The blocks, sub-blocks, and functions were sequentially numbered to differentiate the states in which the function were performed and to indicate the sequence of events. The Engine System was functionally partitioned, to provide separate and unique functional operators. Establishing unique functional operators as work output of the System Architecture process is novel in Liquid Propulsion Engine design. Each functional operator was described such that its unique functionality was identified. The decomposed functions were then allocated to the functional operators both of which were the inputs to the subsystem or component performance specifications. PWR also used a novel approach to identify and map the engine functional requirements to customer-specified functions. The final result was a comprehensive Functional Flow Block Diagram (FFBD) for the J-2X Engine System, decomposed to the component level and mapped to all functional requirements. This FFBD greatly facilitates component specification development, providing a well-defined trade space for functional trades at the subsystem and component level. It also provides a framework for function-based failure modes and effects analysis (FMEA), and a rigorous baseline for the functional architecture.

White, Thomas↗
NASA NTRS2008

Geologic Mapping of the V-36 Thetis Regio Quadrangle: 2008 Progress Report

As a result of mapping, eleven material stratigraphic units and three structural units have been identified and mapped. The material units include (from older to younger): tessera terrain material (tt), material of densely fractured plains (pdf), material of fractured and ridged plains (pfr), material of shield plains (psh), material of plains with wrinkle ridges (pwr), material of smooth plains of intermediate brightness (psi), material of radardark smooth plains (psd), material of lineated plains (pli) material of lobate plains (plo), material of craters having no radar-dark haloes (c1), and material of craters having clear dark haloes (c2). The morphologies and probably the nature of the material units in the study area are generally similar to those observed in other regions of Venus [2]. The youngest units are lobate plains (plo) which here typically look less lobate than in other areas of the planet. Close to them in age are smooth plains which are indeed smooth and represented by two varieties mentioned above. Lineated plains (pli) are densely fractured in a geometrically regular way. Plains with wrinkle ridges, being morphologically similar to those observed in other regions, here occupy unusually small areas. Shield (psh) plains here are also not abundant. Locally they show wrinkle ridging. Fractured and ridged plains (pfr), which form in other regions, the so called ridge belts, are observed as isolated areas of clusters of ridged plains surrounded by other units. Densely fractured plains (pdf) are present in relatively small areas in association with coronae and corona-like features. Tessera terrain (tt) is dissected by structures oriented in two or more directions. Structures are so densely packed that the morphology (and thus nature) of the precursor terrain is not known. Structural units include tessera transitional terrain (ttt), fracture belts (fb) and rifted terrain (rt). Tessera transitional terrain was first identified and mapped by [4] as areas of fractured and ridged plains (pfr) and densely fractured plains (pdf) deformed by transverse faults that made it formally resemble tessera terrain (tt). The obvious difference between units tt and ttt is the recognizable morphology of precursor terrain of unit ttt. Fracture belts are probably ancient rift zones [3]. Rifted terrain (rt), as in other regions of Venus, is so saturated with faults that according to the recommendation of [1, 5] it should be mapped as a structural unit.

Basilevsky, A. T.↗
NASA NTRS2008

Structural Maps of the V-17 Beta Regio Quadrangle, Venus

These represent slices of the geologic map into 7 time-stratigraphic levels whose descriptions are found in [3-6]. From older to younger they are: 1) Tessera material unit (t), 2) Densely fractured plains material unit (pdf), 3) Fractured and ridged plains material unit (pfr), 4) Tessera transitional terrain structural unit (tt), 5) Fracture belts structural unit (fb), 6) Shield plains (psh) and plains with wrinkle ridges (pwr) material units combined, and 7) Lobate (pl) and smooth (ps) plains material units combined and, approximately contemporaneous with them, the structural unit of rifted terrain (rt). Each slice shows the generalized pattern of structures typical of these units. Figures 1-7 show the seven maps and Figure 8 shows the combined map illustrating what is shown in the seven maps. To visualize the Beta Regio uplift outlines, the major structure of this area, we show the +0.5 km and +2.5 km contour lines, corresponding respectively to the base and the mid-height of the uplift. It is seen in Figures 1-2 and 4 the trends of t, pdf and tt occupy relatively small areas and their structures seen in these small windows appear rather variable and with almost no orientation heritage with time. Figure 3 shows that swarms of ridge belts trend mostly NW and go through the Beta structure with no alignment with it, suggesting that this structure did not yet exist at this time. Figure 5 shows that fracture belts align along the northern base of the Beta uplift suggesting onset of the formation of this structure. Figure 6 shows that wrinkle ridges do not show alignment with the Beta uplift suggesting that this already forming structure was not high enough to exert topographic stress in its vicinity. Figure 7 shows that the Beta uplift has Devana Chasma as an axial rift zone, suggesting a genetic link between the uplift and rifting. Figure 8 shows that structural trends in this area significantly changed with time.

Basilevsky, A. t.↗
NASA NTRS2008

Ares I Crew Launch Vehicle Upper Stage/Upper Stage Engine Element Overview

The Ares I upper stage is an integral part of the Constellation Program transportation system. The upper stage provides guidance, navigation and control (GN and C) for the second stage of ascent flight for the Ares I vehicle. The Saturn-derived J-2X upper stage engine will provide thrust and propulsive impulse for the second stage of ascent flight for the Ares I launch vehicle. Additionally, the upper stage is responsible for the avionics system of the the entire Ares I. This brief presentation highlights the requirements, design, progress and production of the upper stage. Additionally, test facilities to support J-2X development are discussed and an overview of the operational and manufacturing flows are provided. Building on the heritage of the Apollo and Space Shuttle Programs, the Ares I Us and USE teams are utilizing extensive lessons learned to place NASA and the US into another era of space exploration. The NASA, Boeing and PWR teams are integrated and working together to make progress designing and building the Ares I upper stage to minimize cost, technical and schedule risks.

McArthur, J. Craig↗
NASA NTRS2010

Testing for the J-2X Upper Stage Engine

NASA selected the J-2X Upper Stage Engine in 2006 to power the upper stages of the Ares I crew launch vehicle and the Ares V cargo launch vehicle. Based on the proven Saturn J-2 engine, this new engine will provide 294,000 pounds of thrust and a specific impulse of 448 seconds, making it the most efficient gas generator cycle engine in history. The engine's guiding philosophy emerged from the Exploration Systems Architecture Study (ESAS) in 2005. Goals established then called for vehicles and components based, where feasible, on proven hardware from the Space Shuttle, commercial, and other programs, to perform the mission and provide an order of magnitude greater safety. Since that time, the team has made unprecedented progress. Ahead of the other elements of the Constellation Program architecture, the team has progressed through System Requirements Review (SRR), System Design Review (SDR), Preliminary Design Review (PDR), and Critical Design Review (CDR). As of February 2010, more than 100,000 development engine parts have been ordered and more than 18,000 delivered. Approximately 1,300 of more than 1,600 engine drawings were released for manufacturing. A major factor in the J-2X development approach to this point is testing operations of heritage J-2 engine hardware and new J-2X components to understand heritage performance, validate computer modeling of development components, mitigate risk early in development, and inform design trades. This testing has been performed both by NASA and its J-2X prime contractor, Pratt & Whitney Rocketdyne (PWR). This body of work increases the likelihood of success as the team prepares for testing the J-2X powerpack and first development engine in calendar 2011. This paper will provide highlights of J-2X testing operations, engine test facilities, development hardware, and plans.

Buzzell, James C.↗
NASA NTRS2010

Three Years of on Orbit ISS Oxygen Generation System Operation 2007-2010

The International Space Station (ISS) United States Orbital Segment (USOS) Oxygen Generation System (OGS) has accumulated 240 days of continuous operation at varied oxygen production rates within the US Laboratory Module (LAB) since it was first activated in July 2007. OGS relocated from the ISS LAB to Node 3 during 20A Flight (February 2010). The OGS rack delivery was accelerated for on-orbit checkout in the LAB, and it was launched to ISS in July of 2006. During the on-orbit checkout interval within the LAB from July 2007 to October 2008, OGS operational times were limited by the quantity of feedwater in a Payload Water Reservoir (PWR) bag. Longer runtimes are now achievable due to the continuous feedwater availability after ULF2 delivery and activation of the USOS Water Recovery System (WRS) racks. OGS is considered a critical function to maintaining six crew capability. There have been a number of failures which interrupted or threatened to interrupt oxygen production. Filters in the recirculation loop have clogged and have been replaced, Hydrogen sensors have fallen out of specifications, a pump delta pressure sensor failed, a pump failed to start, and the voltage on the cell stack increased out of tolerance. This paper will discuss the operating experience and characteristics of the OGS, as well as operational issues and their resolution.

Diderich, Greg S.↗
NASA NTRS2011

The J-2X Fuel Turbopump - Design, Development, and Test

Pratt and Whitney Rocketdyne (PWR), a NASA subcontractor, is executing the design, development, test, and evaluation (DDT&E) of a liquid oxygen, liquid hydrogen two hundred ninety four thousand pound thrust rocket engine initially intended for the Upper Stage (US) and Earth Departure Stage (EDS) of the Constellation Program Ares-I Crew Launch Vehicle (CLV). A key element of the design approach was to base the new J-2X engine on the heritage J-2S engine with the intent of uprating the engine and incorporating SSME and RS-68 lessons learned. The J-2S engine was a design upgrade of the flight proven J-2 configuration used to put American astronauts on the moon. The J-2S Fuel Turbopump (FTP) was the first Rocketdyne-designed liquid hydrogen centrifugal pump and provided many of the early lessons learned for the Space Shuttle Main Engine High Pressure Fuel Turbopumps. This paper will discuss the design trades and analyses performed for the current J-2X FTP to increase turbine life; increase structural margins, facilitate component fabrication; expedite turbopump assembly; and increase rotordynamic stability margins. Risk mitigation tests including inducer water tests, whirligig turbine blade tests, turbine air rig tests, and workhorse gas generator tests characterized operating environments, drove design modifications, or identified performance impact. Engineering design, fabrication, analysis, and assembly activities support FTP readiness for the first J-2X engine test scheduled for July 2011.

Tellier, James G.↗
NASA NTRS2011

The J-2X Oxidizer Turbopump - Design, Development, and Test

Pratt and Whitney Rocketdyne (PWR), a NASA subcontractor, is executing the Design, Development, Test, and Evaluation (DDT&E) of a liquid oxygen, liquid hydrogen two hundred ninety-four thousand pound thrust rocket engine initially intended for the Upper Stage (US) and Earth Departure Stage (EDS) of the Constellation Program Ares-I Crew Launch Vehicle (CLV). A key element of the design approach was to base the new J-2X engine on the heritage J-2S engine which was a design upgrade of the flight proven J-2 engine used to put American astronauts on the moon. This paper will discuss the design trades and analyses performed to achieve the required uprated Oxidizer Turbopump performance; structural margins and rotordynamic margins; incorporate updated materials and fabrication capability; and reflect lessons learned from legacy and existing Liquid Rocket Propulsion Engine turbomachinery. These engineering design, analysis, fabrication and assembly activities support the Oxidizer Turbopump readiness for J-2X engine test in 2011.

Brozowski, Laura A.↗