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NASA NTRS1993

Diploid yeast cells yield homozygous spontaneous mutations

A leucine-requiring hybrid of Saccharomyces cerevisiae, homoallelic at the LEU1 locus (leu1-12/leu1-12) and heterozygous for three chromosome-VII genetic markers distal to the LEU1 locus, was employed to inquire: (1) whether spontaneous gene mutation and mitotic segregation of heterozygous markers occur in positive nonrandom association and (2) whether homozygous LEU1/LEU1 mutant diploids are generated. The results demonstrate that gene mutation of leu1-12 to LEU1 and mitotic segregation of heterozygous chromosome-VII markers occur in strong positive nonrandom association, suggesting that the stimulatory DNA lesion is both mutagenic and recombinogenic. In addition, genetic analysis of diploid Leu+ revertants revealed that approximately 3% of mutations of leu1-12 to LEU1 result in LEU1/LEU1 homozygotes. Red-white sectored Leu+ colonies exhibit genotypes that implicate post-replicational chromatid breakage and exchange near the site of leu1-12 reversion, chromosome loss, and subsequent restitution of diploidy, in the sequence of events leading to mutational homozygosis. By analogy, diploid cell populations can yield variants homozygous for novel recessive gene mutations at biologically significant rates. Mutational homozygosis may be relevant to both carcinogenesis and the evolution of asexual diploid organisms.

Non-NASA Center↗

NASA NTRS2018

Exergy Efficiency of Interplanetary Transfer Vehicles

In order to optimize systems, systems engineers require some sort of measure with which to compare vastly different system components. One such measure is system exergy, or the usable system work. Exergy balance analysis models provide a comparison of different system configurations, allowing systems engineers to compare different systems configuration options. This paper presents the exergy efficiency of several Mars transportation system configurations, using data on the interplanetary trajectory, engine performance, and vehicle mass. The importance of the starting and final parking orbits is addressed in the analysis, as well as intermediate hyperbolic escape and entry orbits within Earth and Mars' spheres of influence (SOIs). Propulsion systems analyzed include low-enriched uranium (LEU) nuclear thermal propulsion (NTP), high-enriched uranium (HEU) NTP, LEU methane (CH4) NTP, and liquid oxygen (LOX)/liquid hydrogen (LH2) chemical propulsion.

Owen, Sean T.↗

NASA NTRS2018

Exergy Efficiency of Interplanetary Transfer Vehicles

Owen, Sean T.↗

NASA NTRS2018

Exergy Efficiency of Interplanetary Transfer Vehicles

Owen, Sean T.↗

DOE OSTI2014

Developing A Laser Shockwave Model For Characterizing Diffusion Bonded Interfaces

12. Other advances in QNDE and related topics: Preferred Session Laser-ultrasonics Developing A Laser Shockwave Model For Characterizing Diffusion Bonded Interfaces 41st Annual Review of Progress in Quantitative Nondestructive Evaluation Conference QNDE Conference July 20-25, 2014 Boise Centre 850 West Front Street Boise, Idaho 83702 James A. Smith, Jeffrey M. Lacy, Barry H. Rabin, Idaho National Laboratory, Idaho Falls, ID ABSTRACT: The US National Nuclear Security Agency has a Global Threat Reduction Initiative (GTRI) which is assigned with reducing the worldwide use of high-enriched uranium (HEU). A salient component of that initiative is the conversion of research reactors from HEU to low enriched uranium (LEU) fuels. An innovative fuel is being developed to replace HEU. The new LEU fuel is based on a monolithic fuel made from a U-Mo alloy foil encapsulated in Al-6061 cladding. In order to complete the fuel qualification process, the laser shock technique is being developed to characterize the clad-clad and fuel-clad interface strengths in fresh and irradiated fuel plates. The Laser Shockwave Technique (LST) is being investigated to characterize interface strength in fuel plates. LST is a non-contact method that uses lasers for the generation and detection of large amplitude acoustic waves to characterize interfaces in nuclear fuel plates. However the deposition of laser energy into the containment layer on specimen’s surface is intractably complex. The shock wave energy is inferred from the velocity on the backside and the depth of the impression left on the surface from the high pressure plasma pulse created by the shock laser. To help quantify the stresses and strengths at the interface, a finite element model is being developed and validated by comparing numerical and experimental results for back face velocities and front face depressions with experimental results. This paper will report on initial efforts to develop a finite element model for laser shock.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2014

Nuclear Naval Propulsion: A Feasible Proliferation Pathway?

There is no better time than now to close the loophole in Article IV of the Nuclear Non-proliferation Treaty (NPT) that excludes military uses of fissile material from nuclear safeguards. Several countries have declared their intention to pursue and develop naval reactor technology, including Argentina, Brazil, Iran, and Pakistan, while other countries such as China, India, Russia, and the United States are expanding their capabilities. With only a minority of countries using low enriched uranium (LEU) fuel in their naval reactors, it is possible that a state could produce highly enriched uranium (HEU) under the guise of a nuclear navy while actually stockpiling the material for a nuclear weapon program. This paper examines the likelihood that non-nuclear weapon states exploit the loophole to break out from the NPT and also the regional ramifications of deterrence and regional stability of expanding naval forces. Possible solutions to close the loophole are discussed, including expanding the scope of the Fissile Material Cut-off Treaty, employing LEU fuel instead of HEU fuel in naval reactors, amending the NPT, creating an export control regime for naval nuclear reactors, and forming individual naval reactor safeguards agreements.

export controls↗

DOE OSTI2015

Analysis of LOCA Scenarios in the NIST Research Reactor Before and After Fuel Conversion

An analysis has been done of hypothetical loss-of-coolant-accidents (LOCAs) in the research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The purpose of the analysis is to determine if the peak clad temperature remains below the Safety Limit, which is the blister temperature for the fuel. The configuration of the NBSR considered in the analysis is that projected for the future when changes will be made so that shutdown pumps do not operate when a LOCA signal is detected. The analysis was done for the present core with high-enriched uranium (HEU) fuel and with the proposed low-enriched uranium (LEU) fuel that would be used when the NBSR is converted from one to the other. The analysis consists of two parts. The first examines how the water would drain from the primary system following a break and the possibility for the loss of coolant from within the fuel element flow channels. This work is performed using the TRACE system thermal-hydraulic code. The second looks at the fuel clad temperature as a function of time given that the water may have drained from many of the flow channels and the water in the vessel is in a quasi-equilibrium state. The temperature behavior is investigated using the three-dimensional heat conduction code HEATING7.3. The results in all scenarios considered for both HEU and LEU fuel show that the peak clad temperature remains below the blister temperature.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2017

Heat Source Characterization in a TREAT Fuel Particle Using Coupled Neutronics-BCMC Calculations

This work presents a multi-physics, multi-scale approach to modeling the Transient Test Reactor (TREAT) currently prepared for restart at the Idaho National Laboratory. TREAT fuel is made up of microscopic fuel grains (r ˜ 20µm) dispersed in a graphite matrix. The novelty of this work is in coupling a binary collision Monte-Carlo (BCMC) model to the Finite Element based code Moose for solving a microsopic heat-conduction problem whose driving source is provided by the BCMC model tracking fission fragment energy deposition. This microscopic model is driven by a transient, engineering scale neutronics model coupled to an adiabatic heating model. The macroscopic model provides local power densities and neutron energy spectra to the microscpic model. Currently, no feedback from the microscopic to the macroscopic model is considered. TREAT transient 15 is used to exemplify the capabilities of the multi-physics, multi-scale model, and it is found that the average fuel grain temperature differs from the average graphite temperature by 80 K despite the low-power transient. The large temperature difference has strong implications on the Doppler feedback a potential LEU TREAT core would see, and it underpins the need for multi-physics, multi-scale modeling of a TREAT LEU core.

36 MATERIALS SCIENCE↗

DOE OSTI2019

Assessment of Low Enriched Uranium Fueled Core Configurations for the Versatile Test Reactor

The Versatile Test Reactor (VTR) Program was initiated in FY-2017. This multi-laboratory effort supported by the U.S. Department of Energy has the long-term objective of completing a Versatile Test Reactor concept with associated cost and schedule estimate that enables a fully informed DOE decision within 3 years, to have a fully operating VTR by 2026 if acquisition is authorized, and to enable accelerated testing of advanced reactor fuels and materials required for advanced reactor technologies. The VTR concept uses sodium as coolant, features a core power of 300 MWth, and employs reactor-grade plutonium as fissile material in a U-Pu-10Zr fuel alloy. The reference core layout provides 10 assembly locations for high-flux irradiation testing with peak fast flux in excess of 4x1015 n/cm2-s. In order to establish a comparison point with other types of fuel, alternative core configurations were investigated using Low Enriched Uranium (LEU) fuel. The objective of this study was to modify the reference VTR core such as to use LEU fuel and assess the performance tradeoff, while retaining the same core envelope.

Stauff, Nicolas E.↗

DOE OSTI2017

Hybrid Gamma Emission Tomography for the Verification of Unirradiated Fuel: A Viability Study

Current International Atomic Energy Agency (IAEA) methodologies for the verification of fresh low-enriched uranium (LEU) and mixed oxide (MOX) fuel assemblies are volume-averaging methods that lack sensitivity to individual pins. Further, as unirradiated fuel assemblies become more and more complex (e.g., heavy gadolinium loading, high degrees of axial and radial variation in fissile concentration), the accuracy of current IAEA instruments degrades and measurement time increases. Particularly in light of the fact that no special tooling is required to remove individual pins from modern fuel assemblies, new capabilities for the verification of unirradiated (i.e., fresh LEU and MOX) assemblies are needed to ensure that fissile material has not been diverted. Passive gamma emission tomography has demonstrated potential to provide pin-level verification of spent fuel, but gamma-ray emission rates from unirradiated fuel emissions are significantly lower, precluding purely passive tomography methods. The work presented here introduces the concept of Hybrid Gamma Emission Tomography (HGET) for verification of unirradiated fuels, in which a neutron source is used to actively interrogate the fuel assembly and the resulting gamma-ray emissions are imaged using tomographic methods to provide pin-level verification of fissile material concentration. This paper describes the status of a viability study on the HGET concept, including: envisioned use-case scenarios and corresponding definitions of fuel assemblies; modeling framework based on Monte Carlo and deterministic transport methods, and its validation; quantitative assessment of candidate HGET signatures with a focus on prompt fission gamma rays and delayed fission gamma rays; a nominal HGETv1 instrument design; candidate HGET-specific tomographic reconstruction methods that fully incorporate declared information; and examples of simulation-based predictions of HGET performance.

Miller, Erin A.↗

DOE OSTI2021

U-10Mo Monolithic Fuel Qualification Plan

The Material Management and Minimization (M3) Program’s primary objective within the U.S. Department of Energy/National Nuclear Security Administration is to achieve permanent threat reduction by minimizing and when possible, eliminating weapon-usable nuclear material around the world. Under M3’s Office of Conversion, the U.S. High Performance Research Reactor (USHPRR) Project, is pursuing a fuel qualification and licensing effort focused on converting high-performance research reactors in the United States from using highly enriched uranium (HEU) fuel to using low-enriched uranium (LEU) fuel. The body of this document is focused on defining the activities required for generic qualification of high density, LEU fuel that then allows reactor conversions to proceed for the four high-performance research reactors and the critical assembly that will operate using the U-10Mo monolithic fuel design (see Subsection 1.3): MITR, MURR, NBSR, ATR, and ATRC. Fuel qualification in this report means generic fuel acceptance of the U-10Mo monolithic fuel form by the NRC for use in reactor conversions in the United States, which allows the fuel to be used in subsequent reactor specific licensing requests. The U 10Mo Monolithic Fuel Research, Development, and Qualification Plan utilizes the functions and requirements of the USHPRR Project that was established in the Functions and Requirements Document (F&RD) [1] and expands on these requirements to ensure that planned tests have traceable results that will ensure the requirement has been met. The methods by which data will be collected to show that these requirements have been met are described in this document.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

UPDATES FROM THE INVOLUTE WORKING GROUP

The HFIR, RHF, and FRM II reactors represent a particular class of Research and Test Reactors that provide some of the most intense and continuous neutron fluxes for science, industry, and medical applications. These high-performance reactors have achieved compact cores by operating with Highly Enriched Uranium fuel (HEU, 235U/U ≥ 20 wt. %) and utilizing fuel plates curved as an involute. Due to the proliferation risks, the international community aims to reduce or eliminate, when possible, the use of HEU fuel in civilian facilities by converting them to a Low-Enriched Uranium fuel (LEU, 235U/U < 20 wt. %). Conversion of these reactors without significantly compromising their performance or safety is a challenging endeavor that can tremendously benefit from advanced computational tools and thus, eliminate unnecessary conservatism to ensure sufficient thermal margins. Therefore, models are being developed using modern Computational Fluid Dynamics (CFD) and Computational Structural Mechanics (CSM) software to evaluate the steady-state safety margins of various LEU designs instead of being reliant on the more traditional, conservative methods. To gain the confidence and acceptance of high-fidelity modeling by the nuclear regulators, Argonne National Laboratory (ANL) and the involute reactors have formed an informal scientific group, the Involute Working Group (IWG). The IWG facilitates inter-organizational collaboration on experimental benchmarking, code-to-code comparisons, and Verification and Validation (V&V). This paper describes some of the recent IWG efforts in validating software against the existing experimental data, as well as code-to-code comparisons of different software used by the IWG members.

Bergeron, Aurelien↗

DOE OSTI2022

Sensitivity/Uncertainty Comparison Study Involving IRSN, LANL, and ORNL Tools to Support Validation

Under a DOE Nuclear Criticality Safety Program (NCSP) task involving Analytical Methods, three Laboratories collaborated in a comparison of results obtained from Sensitivity/Uncertainty (S/U) packages relevant to validation of transport codes. The task involves Institut de Radioprotection et de Sûreté Nucléaire (IRSN), Los Alamos National Laboratory (LANL), and Oak Ridge National Laboratory (ORNL) comparing results of MORET 5/MACSENS V3.0, MCNP6.2/Whisper-1.1, and SCALE 6.2.3/TSUNAMI/USLSTATS respectively. All Monte Carlo transport code results utilize nuclear data from ENDF/B-VII.1 evaluation. This study examines five cases from the International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP Handbook) selected as application models: IEU-MET- FAST-002-001, LEU-COMP-THERM-001-001, LEU-SOL-THERM-004-001, MIX-COMP- THERM-001-001, and U233-SOL-THERM-001-001. This is a continuation of a previous study to examine Pu and HEU cases: HEU-MET-FAST-013-001, HEU-SOL-THERM-001-008, PU-MET- FAST-022-001, and PU-SOL-THERM-001-001. Ultimately, comparison is made between Upper Subcritical Limits (USLs) obtained using each code package for each application case. Since differences exist in whether packages take into account margin of subcriticality (MOS), the USL is computed using only bias and bias uncertainty, also known as the calculational margin (CM) in ANSI/ANS-8.24. Results comparison appears to show that benchmark selection has a greater influence on the USL than the method used for calculation of bias and bias uncertainty.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2022

Overview of Base Model in Parametric Studies Specific to Performance of U-Mo Plates

This paper provides an overview of the base model specifically developed to perform parametric sensitivity studies on the U-10Mo monolithic fuel system. U-Mo monolithic fuels are being considered for the conversion of test reactors into high-performance research reactors that operate using proliferation-resistant, low-enriched uranium (LEU) fuels. These plate-type fuels contain a high-density, low-enrichment fuel sandwiched between zirconium diffusion barriers and encapsulated in aluminum claddings. All U.S. high-performance research reactors have released the designs of their LEU monolithic fuel reactor cores. These designs include nearly 50 distinct fuel plate geometries with different operational parameters. Consequently, a single generic plate geometry representing all the extreme points in this design matrix is unrealistic. To evaluate the performance for various parameters, a set of sensitivity studies was performed. These studies considered various input parameters (i.e., geometric, operational, and material property-related). The results revealed valuable information about plate performance and the sensitivity of this performance to various modeling inputs. To establish a reference state for comparing these result, base model featuring representative irradiation conditions was developed. To capture in-reactor behavior accurately, incorporation of representative constitutive models capable of evolving properties with respect to temperature, irradiation time, and burnup was needed. The behavioral models considered burnup-dependent properties, swelling, creep, and degradation. This paper introduces the base model created for the parametric sensitivity studies. The detailed description of the procedure includes the model geometry, model discretization, thermo-mechanical coupling, material properties and behavioral models. This paper also provides selected results and assesses the performance of the base model.

42 ENGINEERING↗

DOE OSTI2023

MCNP Neutronic Design Optimization of an Accelerator-Driven Subcritical Assembly for Mo-99 Production

Molybdenum-99 (99Mo) has been in short supply in the recent decades. The most common productionmechanism, irradiation in a reactor fueled by highly enriched uranium (HEU), raisesproliferation concerns. Novel efforts are being pursued to meet the demand for 99Mo withoutHEU irradiation, relying on the processing of low-enriched uranium or natural uranium. Onesuch concept is to use an electron linear accelerator to generate bremsstrahlung that, through aphotonuclear reaction, can help multiply neutrons in a subcritical assembly. The goal of this workwas to neutronically optimize the subcritical assembly for keff and 99Mo production. This optimizationis unique for The design was optimized in terms of reflector type, pitch, and LEU andNU mass. Results of the paper show the effects of fuel mass, pitch, and reflector material on keffand the NU fission fraction. keff was found to be mostly driven by LEU mass, and the NU fissionfraction was dependent on several parameters, primarily pitch and mass.

Conant, Andrew↗

DOE OSTI2016

Role of Core Exit Axial Reflectors for NTP Full Submersion Criticality Accident Mitigation

One of the only design basis accidents currently codified into law for space nuclear systems is the full-submersion criticality accident. This accident scenario prescribes that the reactor core must remain sub-critical in the event of a full water submersion with the control system in the shut-down position. This accident scenario is of particular interest and relevance for nuclear thermal propulsion (NTP) reactors because of their unique core periphery and internal design considerations. In previous work it has been shown that this accident is particularly sever for low-enriched uranium (LEU) fueled NTP cores due to their moderated spectrum and the minimization of parasitic neutron absorption. These then add to the existing problems associated with the numerous voids within the core and the extensive neutron leakage due to its compact size. In this abstract we examine the role of a core exit axial reflector in minimizing the submersion worth of a graphite composite LEU-NTP core. The neutronics simulations were done using the ENDF/B-VII cross-section library in conjunction with MCNP6.1. The MCNP calculations were done using 20000 particles per cycle with 50 inactive and 300 active cycles. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2016

Monolithic U-10Mo Fuel Development for U.S. High Performance Research Reactors

To support the U.S Department of Energy (DOE), National Nuclear Security Administration (NNSA) mission to reduce the threat posed by the civilian use of highly enriched uranium (HEU) worldwide, the US High Performance Research Reactor (US HPRR) fuel development program is pursuing development and qualification of a new high-density monolithic U-Mo fuel to facilitate conversion of five higher power research reactors located in the U.S. (ATR, HFIR, NBSR, MIT and MURR) and one critical facility (ATR-C) from HEU to low enriched uranium (LEU). These reactors require higher uranium fuel density than is achievable with dispersion fuel systems used to convert lower power research reactors. The down selected fuel system consists of U-10Mo alloy fuel foils having a thin Zr diffusion barrier interlayer, clad in 6061 Al alloy by hot isostatic pressing. In order to support fabrication development, fuel performance evaluations, generic fuel qualification through the Nuclear Regulatory Commission (NRC), and eventual reactor licensing, a series of irradiation experiments and testing campaigns are being planned. The tests are designed to provide data needed for fabrication process down selection, to establish acceptable fuel performance under normal operating conditions and anticipated transients, and to demonstrate successful scale-up to prototypic reactor-specific fuel element designs. A comprehensive testing program is being executed as part of the US HPRR fuel development program to develop and qualify a new monolithic UMo fuel system to allow conversion of five high power research reactors and one critical facility within the US in support of the NNSA mission to convert research reactors worldwide from HEU to LEU fuels. A significant outcome of this program will be the qualification and eventual licensing of a new plate-type fuel system for use in research reactors through the NRC for the first time in nearly 3 decades. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

Identification and Quantification of Carbon Phases in Conversion Fuel for the Transient Reactor Test Facility

As part of an overall effort to convert US research reactors to low-enriched uranium (LEU) fuel, a LEU conversion fuel is being designed for the Transient Reactor Test Facility (TREAT) at the Idaho National Laboratory. TREAT fuel compacts are comprised of UO{sub 2} fuel particles in a graphitic matrix material. In order to refine heat transfer modeling, as well as determine other physical and nuclear characteristics of the fuel, the amount and type of graphite and non-graphite phases within the fuel matrix must be known. In this study, we performed a series of complementary analyses, designed to allow detailed characterization of the graphite and phenolic resin based fuel matrix. Methods included Scanning Electron Microscopy, Raman spectroscopy, and X-ray Diffraction. Our results indicate that no single characterization technique will yield all of the desired information; however, through the use of statistical and empirical data analysis, such as curve fitting, partial least squares regression, volume extrapolation and spectra peak ratios, a degree of certainty for the quantity of each phase can potentially be obtained. Although no one type of analysis has provided a complete determination of the quantities and identities of all carbon phases present in the fuel matrix, additional information has been gathered and is being used to refine future analysis methods. From the three types of quantification performed, results have been within 1% (92.8%, 93.02% and 93.6% for BSE image analysis, XRD and Raman spectroscopy, respectively). More work remains to identify any trace crystalline carbonaceous phases that may be present, quantify any traces that are identified, and develop and use further complementary bulk phase quantification techniques. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗