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All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

AGR-5/6/7 Fuel Fabrication Report

The U.S. Department of Energy Office of Nuclear Energy (DOE NE) and the Idaho National Laboratory (INL) Advanced Reactor Technologies (ART) Advanced Gas Reactor (AGR) Fuel Development and Qualification program (referred to as AGR Fuel program hereafter) are pursuing qualification of tristructural isotropic (TRISO) coated particle fuel for use in high temperature gas cooled reactors (HTGRs). The AGR Fuel program was established to provide a fuel qualification data set in support of the licensing and operation of an HTGR. BWX Technologies Nuclear Operations Group (BWXT-NOG) was subcontracted to fabricate the fuel for the AGR program. Several investments and innovations were realized in preparation to fabricate fuel for the AGR-5/6/7 irradiation experiments that brought fuel fabrication fully out of the laboratory and into engineering-scale operations. These included: • Increased the kernel fabrication line capacity and uniformity • Upgraded ancillary support equipment and processes for the tristructural isotropic (TRISO) coating furnace • Demonstrated efficient production of the matrix precursor powder by dry jet milling of co mingled components • Demonstrated an engineering-scale method for quick and efficient overcoating TRISO particles with the matrix precursor • Demonstrated an automated, multi cavity compacting system with a volumetric feed system • Demonstrated a combined-cycle thermal treatment furnace These changes increased production rates of some of these processes by an order of magnitude or more while eliminating the use of flammable solvents, multiple grinding and sorting operations, and the weighing out of individual die charges. Three fuel kernel lots were fabricated for production of the fuel for AGR-5/6/7. The initial lot (J52R-16-39316) was certified to fuel specifications but was not used because the kernels had a high fraction of internal fissures that caused an unacceptable fraction of the kernels to fragment when charged to the coating furnace where the TRISO coating would be deposited. Fragmented kernels increased the dispersed uranium in the particles and produced a worrisome fraction of dimpled particles with an elevated probability of in-pile failure. After some efforts to identify the cause of the fissure formation, two additional lots were produced with much lower fissure fractions, J52R-16-69317 and 69318. The latter kernel lot was a backup to the first and was not needed. Multiple kernel batches were composited to form each of the lots so as to simulate a commercial-scale operation where kernel batches would also be composited. Multiple TRISO coating runs were performed and the product characterized so that several could be composited into a TRISO lot. TRISO lot J52R-16-98005 conformed to all fuel specifications except the mean outer pyrocarbon (OPyC) layer thickness was thinner than specified. Furthermore, it was determined that the TRISO lot had a dispersed uranium fraction (DUF) that might result in the compacts not meeting the DUF specification. A review of the role of the OPyC layer and consequences of the DUF by the Technical Coordination Team and INL concluded that the fuel was acceptable for use in the AGR-5/6/7 irradiation experiment. The TRISO particles were overcoated with the matrix precursor that had been produced in a jet-milling operation. The overcoating was performed in equipment originally designed to coat pharmaceuticals. The overcoating process performed well; producing highly spherical and uniform overcoats requiring little upgrading and no recycle or rework. TRISO particles were overcoated with the matrix precursor to achieve nominal volumetric packing fractions (PFs) of TRISO particles of 25% and 40% for the irradiation experiments. The 40% PF compacts occupy the first and fifth test capsule in the test train while the inner three capsules are loaded with 25% PF compacts. The resinated graphite matrix precursor powder was a derivative of the German A3-27 matrix formulation, which differs from previous AGR irradiation campaigns that used an A3-3 formulation. Jet milling of the matrix powder precursor produced a finer mean graphite particle size than the milling operations used for the A3-3 matrix powder precursor. Changes made in the matrix formula and equipment yielded compacts with significantly higher matrix density than was attained in previous AGR irradiation campaigns. The changes in the matrix formulation and the means of milling the powders also complicated resolution of the three fuel compact defect fractions, DUF, exposed kernel fraction (EKF), and the silicon carbide defect fraction (SDF). Characterization data from BWXT-NOG had some anomalous results, so samples of the fuel compacts and overcoated TRISO particles were also analyzed by Oak Ridge National Laboratory (ORNL) to ensure that the defect fractions were accurately characterized.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Texture analysis of AGR program matrix materials

We report the fuel form for high-temperature gas-cooled reactors consists of tristructural isotropic (TRISO) particles embedded in a matrix of graphite flake and carbonized resin. The process of overcoating particles prior to compacting yields a circumferential orientation of the graphite flake surrounding the TRISO particles, which is modified to varied extents when overcoated particles are pressed into the final fuel form. As graphite is highly anisotropic, the texture may impact the properties and performance of the fuel. Ellipsometry was used to measure the texture of the matrix for fueled compacts and unfueled “matrix-only” samples. Results indicated local texture related to the spherical particles in compacts associated with overcoating versus a more linear layered structure in “matrix-only” samples.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Identifying Challenges in Safeguards for Metallic Fuel Fabrication Facilities

As new advanced reactors gain popularity, there is an increasing interest in metallic fuel fabrication for fast reactors. While metallic fuels themselves are not a new idea, as many of the first reactors employed metallic fuels, new designs, compositions, and fabrication methods are appearing throughout the nuclear community. As the interest grows and facilities are constructed, both domestic and international safeguards will need to be heavily involved to support safeguards-by-design (SBD) measures from the start. This work compiles a review of historical and modern fuel types and fabrication methods, fabrication processes, safeguards gaps, and potential safeguards solutions. Metallic nuclear fuel types have been around for many decades and were included in some of the first reactors including the Experimental Breeder Reactor (EBR)-I and -II, the Fermi 1 reactor, the Integral Fast Reactor (IFR), and the Dounreay Fast Reactor (DFR). These reactors used various compositions including pure uranium (U) metal, U-zirconium (Zr) alloys, plutonium (Pu)-aluminum (Al) alloys, U-fissium (Fs) alloys, U-Pu-Zr alloys, and U-molybdenum (Mo) alloys [1, 2, 3, 4, 5]. These small alloying additions are included to improve the material properties of the pure U metal. The alpha-phase U (stable below 661C) suffers elongation in one direction causing grain boundary cracking and increasing creep rate due to irradiation growth, thermal cycling, and preferential crystal orientation. It is ideal to utilize the gamma-phase U (typically stable above 769C) by adding small amounts of alloying elements such as Zr or Mo to stabilize this phase down to room temperature [3]. Additionally, some research has been focused on U with transuranic (TRU) elements present, typically coming from the used fuel recycling process. Including these elements in fast reactor fuel can aid in the reduction of nuclear waste by burning minor long-lived actinides. However, the additions of TRU elements can cause concerns to arise when trying to fabrication or safeguard metallic fuels. A typical metallic fuel element is shown in Figure 1. Sodium is added into the cladding to create a thermal bond between the fuel slug and cladding wall. The fuel slug is then inserted and the end plug is welded on to the top of the fuel element. A gas plenum is left to create a headspace for gaseous fission products to escape rather than continue to build in the fuel itself [1, 5]. Other fuel element geometries exist as well, such as the Lightbridge twisted cruciform geometry shown in Figure 2 [6]. This design allows for better cooling performance and provides room for fuel rod swelling without impacting the fuel rod diameter. There are many different fabrication methods for metallic fuels, which is one of the many benefits of these fuel types. Many of these fabrication methods are relatively easy and cost-efficient. The most popular fabrication method is injection casting, sometimes called vacuum induction melting (VIM), shown in Figure 3 [4, 8, 9, 7, 10]. This method was largely used for EBR-II fuel fabrication. The injection casting system is contained inside of a vessel consisting of a Y2O3-coated graphite crucible surrounded by an induction coil with ZrO2-coated quartz molds suspended above the crucible. The fuel feedstock is placed inside of the graphite crucible and melted using the induction furnace. The induction furnace utilizes a dual frequency with the high frequency melting the feedstock and the low frequency causing stirring of the melted feedstock to form a homogeneous mixture. The mixture is heated to approximately 1600C in an argon environment. The vessel is evacuated and then the quartz molds are lowered into the graphite crucible containing the molten metal and the vessel is repressurized to inject the metal fuel upwards into the molds. The molds are removed and then shattered to release the fuel slugs. This fabrication method was used to fabricate 39,000 metallic fuel pins for EBR-II. While injection casting has been the most common metallic fuel fabrication method throughout the decades, many other methods have been explored including low-pressure gravity casting, microwave casting, continuous casting, centrifugal casting, coextrusion, and many others [11, 12, 8, 13, 14, 15]. Some of these methods aim to mitigate challenges that arise with americium (Am) volatilization during the casting process for TRU-containing fuel feedstocks, an issue with injection casting. Coextrusion is one of the methods explored at the Idaho National Laboratory (INL) and has been utilized for the initial fabrication tests of Lightbridge's unique fuels, as well as other metallic fuels with cladding coextruded. In this process, large billets are formed and machined and then inserted into a molten salt bath for approximately 30 minutes. The billets are then loaded into the extrusion press and extruded. This process can be seen in Figure 4 [15].

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

AGR TRISO Fuel Fission Product Release Data Summary

Knowledge of fission product retention in and release from TRISO fuel under normal and off-normal conditions is needed for reactor safety analyses. This is important for coated-particle fuel used in high-temperature reactors relying on the functional containment strategy. Data on the release and retention of key fission products (e.g., Ag-110m, Cs-134, Eu-154, and Sr-90) in AGR UCO TRISO fuels have been summarized in this report, and empirical relationships with respect to time and temperature were developed. This included fission product accumulation in the OPyC and compact graphitic matrix during irradiation, release from compacts during irradiation, and release during post-irradiation safety testing at temperatures from 1600-1800°C. The frequencies of SiC failure and TRISO coating failure from irradiation and post-irradiation safety testing were also summarized as they have bearing on the quantities of Cs release from the fuel. In a forthcoming publication, a framework for combining and using these empirical relationships as part of a source term analysis will be presented.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Assessment of SCALE and MELCOR for a generic pebble bed fluoride high-temperature reactor

This paper presents the development and application of SCALE and MELCOR models for a fluoride salt-cooled high-temperature reactor (FHR) based on publicly available specifications. SCALE version 6.3beta15 was used to generate power distributions and decay heat curves, and MELCOR version 2.2.18019 was used to calculate the thermal hydraulic response of an assumed FHR primary system. Here, an approach was developed for determining the equilibrium state of the core using depletion of a core slice model and blending of fuel compositions at different burnups to provide three-dimensional fuel composition in the core. Results compared between a core with entirely fresh fuel and one with an equilibrium fuel composition revealed that the equilibrium core led to lower steady-state temperatures but slower cooldown during a loss of flow accident (LOFA). A sensitivity study was then conducted to explore the transient response of our FHR system to variations in thermal hydraulic parameters of the system using the equilibrium core. The inlet temperature, graphite thermal conductivity, and SCRAM time all provided significant control over peak fuel and coolant temperatures. Uncertainties in radionuclide decay data in SCALE were used to perform a decay heat sensitivity study, and we found that uncertainties in decay heat led to negligible impact on peak temperatures during the course of the transient. In all cases evaluated, the observed peak fuel temperatures remained approximately 700 K below anticipated failure limits for the LOFA.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Two decades of DOE investment lays the foundation for TRISO-fueled reactors

Tristructural isotropic (TRISO) coated particle fuel is a robust, microencapsulated fuel form developed originally for use in high-temperature gas-cooled reactors (HTGRs). The particles consist of a spherical fissile kernel surrounded by several layers of pyrocarbon and a silicon carbide (SiC) layer (Figure 1). The particles are formed into cylindrical or spherical fuel forms using a resinated graphite matrix material for insertion into an HTGR. The kernel and coating layers together act to retain fission products within the particle during normal reactor operation and during postulated accidents; TRISO particles can maintain structural integrity at extremely high temperatures, reaching as high as approximately 1,600°C in limiting HTGR accidents. This limits the fission product activity circulating in the helium coolant and the activity released to the environment during accidents. Acceptable performance of TRISO particles is therefore essential for reactor safety.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

SCALE modeling of foil irradiations at WSU’s TRIGA with sensitivity/uncertainty analysis

We report the Washington State University TRIGA reactor was modeled in the SCALE system’s KENO neutron transport code. The model includes 119 fuel rods, a central boral control rod, 4 control blades, 5 irradiation ports, and 20 graphite reflector blocks within a rectangular aluminum grid box. The model accurately simulated the irradiation of depleted uranium and highly enriched uranium foils and predicted reaction rate ratios for 238 U(n,2n) 237 U / 238 U(n,γ) 239 U and 99 Mo / fission, which compared favorably to empirical values previously measured at Los Alamos. Finally, a sensitivity/uncertainty analysis was performed using the continuous-energy TSUNAMI-3D module of SCALE.

237Np↗
DOE OSTI2020

Hard, transparent, sp3-containing 2D phase formed from few-layer graphene under compression

Despite several theoretically proposed two-dimensional (2D) diamond structures, experimental efforts to obtain such structures are in initial stage. Recent high-pressure experiments provided significant advancements in the field, however, expected properties of a 2D-like diamond such as sp 3 content, transparency and hardness, have not been observed together in a compressed graphene system. Here, we compress few-layer graphene samples on SiO 2 /Si substrate in water and provide experimental evidence for the formation of a quenchable hard, transparent, sp 3 -containing 2D phase. Our Raman spectroscopy data indicates phase transition and a surprisingly similar critical pressure for two-, five-layer graphene and graphite in the 4e6 GPa range, as evidenced by changes in several Raman features, combined with a lack of evidence of significant pressure gradients or local non-hydrostatic stress components of the pressure medium up to ≈ 8 GPa. The new phase is transparent and hard, as evidenced from indentation marks on the SiO 2 substrate, a material considerably harder than graphene systems. Furthermore, we report the lowest critical pressure (≈ 4 GPa) in graphite, which we attribute to the role of water in facilitating the phase transition. Overall, theoretical calculations and experimental data indicate a novel, surface to- bulk phase transition mechanism that gives hint of diamondene formation.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2023

Development of Hydrogen Permeation Barrier and Matrix Interaction Protection Coatings for TZM Enclosure of Hydride Moderators (Microreactor Program M3 Report)

It is important to develop a high-performance enclosure solution for hydride-based microreactor moderation components to effectively suppress hydride thermal decomposition and prevent hydrogen loss in harsh in-reactor environments. Currently, the DOE-NE Microreactor Program (MR program) is investigating the titanium-zirconium-molybdenum (TZM) canned yttrium hydride (YH 2-x ) solution, which provides acceptable hydrogen retainment performance up to ~700 °C. However, improvements in the enclosure are necessary for increasing long-term operation temperature beyond 700 °C and enhancing short-term survivability at even higher temperatures during power transients. One potential solution is the addition of a dedicated hydrogen permeation barrier, which can potentially improve the enclosure's performance. Furthermore, the hydride moderator components will be inserted into the microreactor matrix, which may consist of materials that are not chemically inert against TZM. Therefore, the TZM might need to be protected from the external environment through the application of another surface protection coating. To address these critical issues, two types of barrier coatings optimized for TZM-based moderator enclosure has been developed. A multilayer coating approach has been used to identify TZM-compatible barrier materials. Behavior of these coating designs against H 2 permeation and long-term high-temperature graphite matrix interactions has been examined in detail. Additionally, the developed coatings' behavior when exposed to both high-temperature thermal cycling and heavy-ion irradiation has also been investigated. Overall, the development of these barrier coatings is essential for achieving maturity for the metal hydride moderator technology. The results in this report will provide crucial insights into the behavior of the developed coatings and their potential for improving the long-term performance of TZM-based moderator enclosures.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

CSEWG Meeting Minutes

There were many improvements and additions to ENDF/B-VIII.1. We tested four key moderator materials: Be, HF, FLiBe, and graphite. Be-metal moderated critical assemblies of HEU only see 26 pcm difference. Sd library adds new physics to the TSL which improves the agreement with the benchmark. Other changes besides the TSL impact the benchmark: 235U and 9Be. Free gas change affects library heavily, hundreds of pcm in wrong way – need to look at FG libraries. F and Cl TSLs added to 8.1. A 3k pcm disagreement was seen without new TSL. Adding 8.1 + TSL reduces pcm disagreement A LOT.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2020

Gamma Spectrometry Examination of the AGR-3/4 Irradiation

The results from gamma spectrometry examination of the different components from the combined third and fourth U.S. Advanced Gas Reactor (AGR) TRISO-coated particle fuel irradiation tests (AGR-3/4) have been analyzed. This experiment was designed to provide information about in-pile fission product migration. In each of the 12 capsules, a single stack of four compacts with designed-to-fail (DTF) particles surrounded by inner and outer graphite and/or graphitic matrix rings and a graphite sink ring were irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). Gamma spectrometry has been used to evaluate the gamma-emitting fission product inventory of compacts from the irradiation and evaluate the burnup of these compacts based on the activity of the radioactive cesium isotopes (i.e., Cs-134 and Cs-137) in the compacts. Burnup from gamma spectrometry compares well with predicted burnup from simulations. The inner and outer rings were also examined by gamma spectrometry to evaluate the total fission product inventory of the rings, and gamma emission computed tomography (GECT) was used to investigate the spatial distribution of gamma-emitting fission products within the rings. This report focuses on results obtained via the Precision Gamma Scanner (PGS) for the compacts and inner and outer rings. These non-destructive gamma measurements are currently being compared to recent destructive measurements of spatial fission product distributions in the rings and to predictions made via mathematical modeling. A separate report discusses the entire AGR-3/4 fission product mass balance for all irradiation capsule components including the inner and outer rings, graphite sink rings, capsule spacers, capsule foils, through tubes, and felts. This report is similar to the conference proceedings: Harp, J.M., Demkowicz, P.A., and Stempien, J.D., “Initial gamma spectrometry examination of the AGR-3/4 irradiation,” International Topical Meeting High Temperature Reactor Technology (HTR 2016), Las Vegas, NV, USA, November 2016, Paper HTR2016-18593, but it has been updated to reflect all the data that has been collected on AGR-3/4 using PGS. The Appendices contain extensive data from the collected gamma spectra.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Gamma Spectrometry Examination of the AGR-3/4 Irradiation

The results from gamma spectrometry examination of the different components from the combined third and fourth U.S. Advanced Gas Reactor (AGR) TRISO-coated particle fuel irradiation tests (AGR-3/4) have been analyzed. This experiment was designed to provide information about in-pile fission product migration. In each of the 12 capsules, a single stack of four compacts with designed-to-fail (DTF) particles surrounded by inner and outer graphite and/or graphitic matrix rings and a graphite sink ring were irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). Gamma spectrometry has been used to evaluate the gamma-emitting fission product inventory of compacts from the irradiation and evaluate the burnup of these compacts based on the activity of the radioactive cesium isotopes (i.e., Cs-134 and Cs-137) in the compacts. Burnup from gamma spectrometry compares well with predicted burnup from simulations. The inner and outer rings were also examined by gamma spectrometry to evaluate the total fission product inventory of the rings, and gamma emission computed tomography (GECT) was used to investigate the spatial distribution of gamma-emitting fission products within the rings. This report focuses on results obtained via the Precision Gamma Scanner (PGS) for the compacts and inner and outer rings. These non-destructive gamma measurements are currently being compared to recent destructive measurements of spatial fission product distributions in the rings and to predictions made via mathematical modeling. A separate report discusses the entire AGR-3/4 fission product mass balance for all irradiation capsule components including the inner and outer rings, graphite sink rings, capsule spacers, capsule foils, through tubes, and felts. This report is similar to the conference proceedings: Harp, J.M., Demkowicz, P.A., and Stempien, J.D., “Initial gamma spectrometry examination of the AGR-3/4 irradiation,” International Topical Meeting High Temperature Reactor Technology (HTR 2016), Las Vegas, NV, USA, November 2016, Paper HTR2016-18593, but it has been updated to reflect all the data that has been collected on AGR-3/4 using PGS. The Appendices contain extensive data from the collected gamma spectra.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

SiC as a core-edge integrated wall solution in DIII-D

Silicon carbide (SiC) is a promising material for use in a fusion reactor due to its low hydrogenic diffusivity, high temperature strength and resilience under neutron irradiation. To assess SiC as a main wall material in DIII-D, simulations with TRIM.SP and DIVIMP are performed on a well-diagnosed L-mode discharge. The effective charge, Z eff , across the separatrix is used as a figure of merit in comparing SiC to the current graphite walls. It is found that SiC is expected to reduce Z eff , potentially by as much as ~50%. It is discussed how SiC may be expected to "self-condition" and create wall conditions similar to siliconization, further lowering Z eff due to efficient oxygen gettering. Furthermore, the potential benefits are reviewed and a path towards SiC walls in DIII-D is presented.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

A review on the state of thermal hydraulics research on air ingress scenarios in High-Temperature Gas-cooled Reactors following a D-LOFC

With the expectation of near-immediate carbon neutrality, widespread implementation of proven High-Temperature Gas-cooled Reactors (HTGRs) embodies a viable solution pathway given their inherent, passive safety features and high thermal efficiency. This study provides an overview of the current state of research involving the thermal hydraulics associated with air ingress from a depressurized loss of forced cooling (D-LOFC) in HTGRs. Accurately characterizing and predicting the physical phenomena underlying air ingress is of paramount concern, as the integrity of the fuel and core graphite support structures are threatened by the presence of oxygen. Broadly speaking, the air ingress scenario can be delineated into three main stages: (1) Depressurization, (2) Density-Driven Flow, and (3) Natural Convection. In tandem with the underlying fundamental theory, this review collates and synthesizes the existing body of contemporary research concerning the air ingress scenario following a D-LOFC. As evinced by this review, our current understanding and predictive abilities have benefited from extensive research, predominantly concentrated on the rate of air ingestion into the core. Here, additional research is necessary to holistically capture the phenomenology of an air ingress scenario following a D-LOFC by considering an additional variable: the oxygen content of the ingressing air. The latter variable requires investigation into the complex interactions of the fully integrated system. Additionally, while numerical tools are evolving domestically through the Nuclear Energy Advanced Modeling and Simulation program, a sufficiently validated code remains absent.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Remote isotope detection and quantification using femtosecond filament-laser ablation molecular isotopic spectrometry

We demonstrate the detection and quantification of isotopes from solid samples remotely at a distance of 36 m. This was accomplished through the combination of femtosecond filaments and laser ablation molecular isotopic spectrometry (F 2 -LAMIS). Isotopically enriched graphite samples with varying 13 C concentration were used and the F 2 -LAMIS emission was measured remotely using a portable spectrometer system. A standardless quantification approach was used to determine the corresponding isotopic ratios. The use of F 2 -LAMIS for the detection and quantification of isotopes at atmospheric pressure without the use of calibration standards opens the possibility to implement this technology for remote sensing of isotopes in field applications

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2020

SCALE capabilities for high temperature gas-cooled reactor analysis

The SCALE code system’s ability to address stochastic distributions of fuel particles within a graphite matrix has been revisited in both multigroup (MG) features and continuous-energy (CE) Monte Carlo methods. Furthermore, this paper describes current and emergent SCALE capabilities within the CSAS sequence to address double-heterogeneous systems and presents verification and validation studies of these methods and data. Good agreement was obtained for a high temperature gas-cooled reactor (HTGR) fuel pebble model between CSAS MG eigenvalue calculations and corresponding CE reference solutions. Code-to-code comparisons for this HTGR pebble model showed good agreement of CSAS-KENO and CSAS-Shift CE calculations and the Serpent and MCNP codes in terms of eigenvalues and reaction rate ratios. Validation studies based on two HTGR experiments resulted in good agreement between MG and CE results, as well as between experiment and calculation, although the level of agreement was significantly influenced by the applied ENDF/B nuclear data library.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2026

One-dimensional neutron diffraction from layered graphite: Reciprocal space structure and grating behavior

In this work we report observations of one-dimensional neutron diffraction from highly oriented pyrolytic graphite (HOPG), where the scattering angle varies continuously with incident angle following classical grating-like behavior. The 2D polycrystalline structure of HOPG—with highly aligned layers along the 𝑐 axis but random in-plane rotations—creates planes of scattering intensity in reciprocal space at 𝑄 𝑐 =𝑛⁢(2⁢𝜋/𝑑) where 𝑑=3.35Å is the interlayer spacing. As the Ewald sphere sweeps through reciprocal space during sample rotation, it continuously intersects these planes, producing the observed angular dispersion. We observe both first-order (𝑛=1) and second-order (𝑛=2) diffraction at conventional scattering angles (25°–70°), with peak positions that remain temperature-independent between 10 K and 294 K and follow quantitative agreement with momentum conservation 𝑄 𝑐 =𝑘⁢[sin⁡𝜓−sin⁡𝜓 𝑓 ]=𝑛⁢(2⁢𝜋/𝑑). X-ray diffraction under similar conditions shows no comparable behavior, confirming that sharp nuclear-vacuum contrast is essential. While diffraction intensities are weak (∼10 −6 of Bragg peaks), the observations demonstrate how the interplay of atomic-scale periodicity, nuclear contrast, and structural disorder enables observation of continuous diffraction curves at thermal neutron wavelengths, illustrating how HOPG's unique microstructure determines its scattering properties beyond conventional Bragg diffraction.

2-dimensional systems↗
NASA NTRS1995

An Accelerated Development, Reduced Cost Approach to Lunar/Mars Exploration Using a Modular NTR-Based Space Transportation System

The results of integrated systems and mission studies are presented which quantify the benefits and rationale for developing a common, modular lunar/Mars space transportation system (STS) based on nuclear thermal rocket (NTR) technology. At present NASA's Exploration Program Office (ExPO) is considering chemical propulsion for an 'early return to the Moon' and NTR propulsion for the more demanding Mars missions to follow. The time and cost to develop these multiple systems are expected to be significant. The Nuclear Propulsion Office (NPO) has examined a variety of lunar and Mars missions and heavy lift launch vehicle (HLLV) options in an effort to determine a 'standardized' set of engine and stage components capable of satisfying a wide range of Space Exploration Initiative (SEI) missions. By using these components in a 'building block' fashion, a variety of single and multi-engine lunar and Mars vehicles can be configured. For NASA's 'First Lunar Outpost' (FLO) mission, an expendable NTR stage powered by two 50 klbf engines can deliver approximately 96 metric tons (t) to translunar injection (TLI) conditions for an initial mass in low earth orbit (IMLEO) of approximately 198 t compared to 250 t for a cryogenic chemical TLI stage. The NTR stage liquid hydrogen (LH2) tank has a 10 m diameter, 14.5 m length, and 66 t LH2 capacity. The NTR utilizes a UC-ZrC-graphite 'composite' fuel with a specific impulse (Isp) capability of approximately 900 s and an engine thrust-to-weight ratio of approximately 4.3. By extending the size and LH2 capacity of the lunar NTR stage to approximately 20 m and 96 t, respectively, a single launch Mars cargo vehicle capable of delivering approximately 50 t of surface payload is possible. Three 50 klbf NTR engines and the two standardized LH2 tank sizes developed for lunar and Mars cargo vehicle applications would be used to configure the Mars piloted vehicle for a mission as early as 2010. The paper describes the features of the 'common' NTR-based moon/Mars STS, examines performance sensitivities resulting from different 'mission mode' assumptions, and quantifies potential schedule and cost benefits resulting from this modular moon/Mars NTR vehicle approach.

Borowski, S.↗