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All sourcesNASA NTRSDOE OSTI
DOE OSTI2013

Recent Improvement of Measurement Instrumentation to Supervise Nuclear Operations and to Contribute Input Data to 3D Simulation Code - 13289

The CEA has developed many compact characterization tools to follow sensitive operations in a nuclear environment. Usually, these devices are made to carry out radiological inventories, to prepare nuclear interventions or to supervise some special operations. These in situ measurement techniques mainly take place at different stages of clean-up operations and decommissioning projects, but they are also in use to supervise sensitive operations when the nuclear plant is still operating. In addition to this, such tools are often associated with robots to access very highly radioactive areas, and thus can be used in accident situations. Last but not least, the radiological data collected can be entered in 3D calculation codes used to simulate the doses absorbed by workers in real time during operations in a nuclear environment. Faced with these ever-greater needs, nuclear measurement instrumentation always has to involve on-going improvement processes. Firstly, this paper will describe the latest developments and results obtained in both gamma and alpha imaging techniques. The gamma camera has been used by the CEA since the 1990's and several changes have made this device more sensitive, more compact and more competitive for nuclear plant operations. It is used to quickly identify hot spots, locating irradiating sources from 50 keV to 1500 keV. Several examples from a wide field of applications will be presented, together with the very latest developments. The alpha camera is a new camera used to see invisible alpha contamination on several kinds of surfaces. The latest results obtained allow real time supervision of a glove box cleaning operation (for {sup 241}Am contamination). The detection principle as well as the main trials and results obtained will be presented. Secondly, this paper will focus on in situ gamma spectrometry methods developed by the CEA with compact gamma spectrometry probes (CdZnTe, LaBr{sub 3}, NaI, etc.). The radiological data collected is used to quantify the activity of hot spots and can also then be entered in 3D models of nuclear plants to simulate intervention scenarios. Recent developments and results will be presented regarding this. Finally, thanks to a large amount of feedback, the interest of using complementary measurements will be discussed. In fact, the recent use of 3D simulation codes requires very accurate knowledge of nuclear plant radiological data. The use of coupled devices such as imaging devices, (gamma and alpha cameras), gamma spectrometry, dose rate mapping, collimated / un-collimated measurements and many other physical values gives an approach to the radiological knowledge of a process or plant with the lowest possible uncertainty. In line with this, the paper will conclude with the future developments and trials that could be assessed in that field of application. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

An Information Building on Radioactivity and Nuclear Energy for the French CEA Cadarache Research Center - 13492

The CEA Cadarache research center is one of the 10 research centers of the French Alternative Energies and Atomic Energy Commission (CEA). Distributed throughout various research platforms, it focuses on nuclear fission, nuclear fusion, new energy technologies (hydrogen, solar, biomass) and fundamental research in the field of vegetal biology. It is the most important technological research and development centers for energy in Europe. Considering the sensitive nature of nuclear activities, the questions surrounding the issue of radioactive waste, the nuclear energy and the social, economic and environmental concerns for present and future generations, the French Government asked nuclear actors to open communication and to give all the information asked by the Local Information Commission (CLI) and the public [1]. In this context, the CEA Cadarache has decided to better show and explain its expertise and experience in the area of nuclear energy and nuclear power plant design, and to make it available to stakeholders and to the public. CEA Cadarache receives each year more than 9000 visitors. To complete technical visits of the research facilities and laboratories, a scientific cultural center has been built in 2011 to inform the public on CEA Cadarache research activities and to facilitate the acceptance of nuclear energy in a way suited to the level of knowledge of the visitors. A modern interactive exhibition of 150 m{sup 2} allows visitors to find out more about energy, CEA Cadarache research programs, radioactive waste management and radiological impact on the research center activities. It also offers an auditorium for group discussions and for school groups to discover science through enjoyment. This communication center has received several thousand visitors since its opening on October 2011; the initial results of this experience are now available. It's possible to explain the design of this exhibition, to give some statistics on the number of the visitors, their characteristics and their perception after their center visits. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Study of the influence of heat sources on the out-of-pile calibration curve of calorimetric cells used for nuclear energy deposition quantification

At present the Jules Horowitz Reactor is under construction in Cadarache research center of CEA 'French Alternative Energies and Atomic Energy Commission' center located in the south-east of France. This new Material Testing Reactor (MTR) will be operational in late 2019 and will allow the generation of a new experimental potential (up to 20 irradiation devices simultaneously) and new harsh conditions such as higher neutron fluxes (5.10{sup 14} n.cm{sup -2}.s{sup -1} for E≥1 MeV), faster material ageing and higher nuclear heating (up to 20 W/g for nominal capacity of 100 MW). In nuclear research field, the control and the measurement of the nuclear heating (energy deposition rate per mass unit induced by the interactions of radiations with matter) is crucial to carry out accurate studies on ageing of materials and on the behavior of nuclear fuels under irradiation. Several experiments need to know precisely this key parameter in order to establish dedicated thermal conditions. The measurement of the nuclear heating inside MTRs is realized by three kinds of sensors: single-cell calorimeter, differential calorimeter and gamma thermometer. One scientific objective of the IN-CORE program, between CEA and Aix-Marseille University in 2009, is to improve the nuclear heating measurement. In this context a new multi-sensor device, called CARMEN, was made. This device contains in particular a differential calorimeter which was designed to measure the nuclear heating in the periphery of OSIRIS reactor (a MTR located at Saclay, France) up to 2 W/g and tested during two irradiation campaigns. Results obtained during these campaigns showed that temperatures reached inside the calorimeter are higher than ones obtained during the preliminary out-of-pile calibration experiments. For instance for 1.74 W/g, the in-pile temperature of the calorimeter rod is equal to 305 deg. C against 225 deg. C in laboratory conditions by simulating the nuclear heating by Joule Effect inside the calorimeter cell head. This discrepancy is higher than in previous experiments because the calorimeter owns a high sensitivity. Consequently, a new prototype was created and instrumented by other heat sources in order to impose an energy deposition on the calorimetric cell structure (in particular in the base) and to improve the calibration step in out-of-pile conditions. In this paper, on the first part a detailed description of the new calorimetric sensor will be given. On the second part, the experimental response of the sensor obtained for several internal heating conditions will be shown. The influence of these conditions on the calibration curve will be discussed. Then the response of this prototype will be also presented for different external cooling fluid conditions (in particular flow temperature). In this part, the comparison between the in-pile and out-of-pile experimental results will be performed. On the last part, these out-of-pile experiments will be completed by 2D axisymmetrical thermal simulations with the CEA code CAST3M using Finite Elements Method. After a comparison between experimental and numerical works, improvements of the sensor prototype will be studied (new heat sources). (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Testing and Versatility of the SAVY 4000 Nuclear Material Container - 15634

The SAVY 4000 container was designed by Nuclear Filter Technology, Inc. to meet DOE M441.1-1 requirements for safe storage of nuclear material, and it is currently being used at Los Alamos National Laboratory, Livermore National Laboratory, Idaho National Laboratory and Nevada National Security Site. The 316L stainless steel container can be manufactured in a range of sizes from 1 quart to 10 gallon capacity, it can be opened and closed easily by hand, it is sealed with a Viton{sup R} O-ring, and the filter retains particulates and resists liquid water entry while allowing hydrogen gas to escape. This innovative design has proven to provide benefits and uses beyond its original intent. For example, the container was tested in a simulated facility design basis accident involving high temperatures, fuel fire, and subsequent drop, and the damage ratio was measured. This has allowed the LANL plutonium facility to take credit for a material-at-risk reduction of approximately two orders of magnitude for material stored in the SAVY 4000 container. In addition, the water resistant filter allows the nuclear facility to credit the container as a safety feature to mitigate the risk of a criticality accident in the event of facility sprinkler system activation. The container has also been approved to transfer small quantities of solid nuclear material by hand carry between Los Alamos facilities, and it has been approved and used as a waste container for transuranic waste destined for the Waste Isolation Pilot Plant. The 1 quart capacity container and the future 2 quart design both fit inside the 9977 shipping container, and this has facilitated the transport of nuclear material from Idaho National Laboratory to Nevada Nuclear Security Site, obviated the need for repackaging the material for storage upon arrival. Finally, Los Alamos has recently contracted with Nuclear Filter Technology, Inc. to perform testing to certify the SAVY 4000 container as a DOT Type A liquid transport container. This paper will provide an overview of the various testing regimes and describe the resulting versatility of the container for nuclear material containment. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Characterization of Fresh Nuclear Fuel using Time-Of-Flight Neutrons

Neutrons offer excellent characterization capabilities for nuclear fuels and can provide non-destructive insight on micro structure, chemistry and crystallographic phase. Measurements are possible at reactors or pulsed sources, but the inherent energy resolution implicit at pulsed sources offers unique opportunities. Recent advances in 2D neutron detection enabled new capabilities of material characterization using neutron imaging and tomography for nuclear materials. Moreover, time-of-flight (TOF) neutron measurements can enable isotope specific tomography by deriving isotope densities from transmission spectra exhibiting neutron resonances, which are recorded for each pixel. Tomographic algorithms allow 3D reconstructions of the spatial distribution of isotopes, density and features in typical fuel pellet samples. The ability to select radiographs for different energies allows characterizing samples which are opaque for thermal neutrons, such as enriched fuels or fuels containing plutonium, by selecting epi-thermal neutron energies and by knowledge of the total neutron cross sections of the constituents, isotopic enrichment levels can be determined. Nuclear cross sections are typically not calculated directly and databases such as the Evaluated Nuclear Data File (ENDF) database provide this information with cross sections determined by standardized experimental measurements, such as neutron transmission measurements for individual isotopes. Utilizing the information in the ENDF database, the 'inverse' problem can be solved by assuming cross section parameters to be known parameters and areal densities can be determined. This work presents recent advances at the Los Alamos Neutron Science Center (LANSCE) facility at Los Alamos National Laboratory (LANL), which provides the infrastructure to handle radioactive samples, in the development of fuel characterization with TOF neutrons. Examples include recent work on prototype accident tolerant nuclear fuels (ATF). The potential application of these techniques for nondestructive examination of irradiated fuel is under consideration. Non-destructive access to materials features for specimens embedded in containers, particularly for hazardous samples, such as a radioactive nuclear fuel, is indispensable to characterize and ultimately license new fuel forms. Energy-resolve neutron imaging is a new tool for this research field. The results reported here present a new imaging technique, setting a milestone in the capabilities of measuring non-destructively absolute quantities of physical properties. The enabling physics behind this new technique are nuclear resonances. We have demonstrated how absolute areal densities can be extracted from radiographs utilizing pulsed neutron sources. This technique provides information that is otherwise difficult and costly to obtain. In the case of nuclear fuels, the results from energy-dispersive neutron imaging may guide destructive examination of irradiated fuel. This summery does not describe recent analysis, which is ongoing. Particularly, the implementation of calibration standards has improved the determination of enrichment levels without additional calibration. The analysis of the samples was performed with full tomographic reconstructions for isotope specific data and presents a first approach in absolute isotope specific 3D reconstruction. This work was predominantly funded by and received substantial support from the Fuel Cycle R and D program at Los Alamos National Laboratory. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Advancing Nuclear Energy to Support Economy-wide Net-zero Solutions: Challenges and Opportunities

Governments and private industry around the world have established aggressive goals to achieve net-zero emissions for the power, industrial, and transportation sectors by 2050. These aggressive goals demand immediate action if we are to be successful, and they require us to think more holistically about our clean energy options. Programs within the U.S. Department of Energy (DOE) are addressing these holistic solutions. Traditionally, electricity generation and management and meeting energy demands for industry and transportation are considered independently. As we seek to achieve net-zero, we need to reassess energy demands. When we consider overall energy use, only one-third is in the form of electricity. Additional energy demands are in the form of heat or steam for industrial processes, as well as transportation. These sectors are much harder to abate, and electrification may not be the best option. Reducing environmental emissions at an affordable cost, while maintaining grid reliability and resilience, will require us to use all of the clean energy resources that we have available. That means coordinating the use of nuclear, renewables, and fossil fuels with carbon capture to meet growing demands for electricity, industrial applications, and mobility. The DOE Office of Nuclear Energy (DOE-NE) program on Integrated Energy Systems (IES) is led by researchers at Idaho National Laboratory (INL) and work is conducted in partnership with an array of other DOE laboratories, industry, and academia. The primary focus of IES research is to assess the technical and economic potential of IES to enhance the flexibility and utilization of nuclear reactors working alongside renewable generators to meet an array of energy demands—thereby maximizing the utilization of clean energy resources across all energy sectors. Various energy applications and product streams beyond electricity are being evaluated, ranging from generation of potable water to production of hydrogen, fertilizers, synthetic fuels, and chemicals. The DOE-NE program additionally partners with the Hydrogen and Fuel Cell Technologies Office under the DOE Office of Energy Efficiency and Renewable Energy to jointly fund the development of analysis tools, technologies, and nuclear-integrated hydrogen demonstration projects. This presentation highlights the wide array of R&D being conducted across multiple DOE-funded programs to develop and deploy nuclear-based IES that will be key to achieving our net-zero goals. By working with key collaborators in the nuclear industry, analytical studies are now becoming a reality in demonstration projects. The presentation specifically focuses on motivations for the paradigm shift in how nuclear energy is deployed and used, opportunities for nuclear energy to support an array of non-electric application, integration options, computational tools and experimental systems to support evaluation and to accelerate deployment, and industry partnerships.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Modeling, analysis, and optimization of complex nuclear processes and facilities via computational methods: The HALEU process case study

Improving and adapting industrial systems to timely meet changing programmatic and market demands is an important goal to achieve, including when operating and maintaining complex nuclear processes and facilities. However, changes to these complex systems are costly, particularly when they are already in place and bounded to stringent requirements and constraints such as when handling radioactive material and contaminated equipment. These conditions often exist when treating spent nuclear fuel remotely within shielded nuclear radiation chambers, commonly referred as hot cells, to condition nuclear material and/or fabricate products for utilization in other nuclear enterprises such as in the manufacture of advanced nuclear fuel. The illustrative case considered here is the production of high assay low enriched uranium (HALEU) products supporting the deployment of advanced nuclear reactors. For the HALEU program, resources invested were and are being systematically analyzed so that these investments are maximized in a facility that is nearly 60 years old. A methodology that has effectively enabled optimized and improvements in the Spent Fuel Treatment (SFT) program, and consequently the HALEU program, involves discrete event simulation as addressed in this article. Here, the quantification of multiple productivity metrics, including material processing rates, cycle times, bottlenecks, number of material transfers as well as equipment, workstation, and material handling utilization, has resulted in a myriad of diverse discoveries and data-informed decisions regarding process layout and constituent, labor levels and schedules, selection of new process units, storage needs, and other critical process configurations. This article describes such a computational capability being applied for decision-making, illustrates its application to an actual process and program, provides illustrative results, and argues how computational methods for the modeling, analysis, and optimization of complex processes and facilities does lead to informed decisions derived from data and not only from intuition.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI1989

Nuclear Safety [Vol. 30, No. 1, January-March 1989]

Nuclear Safety is a review journal that covers significant developments in the field of nuclear safety. Its scope includes the analysis and control of hazards associated with nuclear energy, operations involving fissionable materials, and the products of nuclear fission and their effects on the environment. Primary emphasis is on safety in reactor design, construction, and operation; however, the safety aspects of the entire fuel cycle, including fuel fabrication, spent-fuel processing, nuclear waste disposal, handling of radioisotopes, and environmental effects of these operations, are also treated. Table of Contents for this issue follows. GENERAL SAFETY CONSIDERATIONS: 1 International Conference on Thermal Reactor Safety: NUCSAFE 88 by A. P. Malinauskas and J. G. Pruett; ACCIDENT ANALYSIS: 18 Chernobyl Today: State of Research by A. Yu. Gagarinskii; CONTROL AND INSTRUMENTATION: 23 Problems Associated with Routine In-Plant Radioactive Effluent Monitoring Systems at U.S. Light-Water Reactors by Barry G. Wahlig, David M. Walker, Mahmoud R. Ghavi, and John M. Palms; DESIGN FEATURES: 34 Twentieth DOE/NRC Nuclear Air-Cleaning Conference by R. R. Bellamy, D. W. Moeller, and M. W. First, 46 Design Options for the Core Support Structure for Liquid-Metal-Cooled Reactor Plants by J. P. Burelbach, W. J. Kann, J. G. Saiveau, and R. W. Seidensticker; ENVIRONMENTAL EFFECTS: 53 The Radiological Consequences of the Chernobyl Accident by L. V. Konstantinov and A. J. González; WASTE AND SPENT FUEL MANAGEMENT: 70 Activities Related to Waste Management Compiled by E. G. Silver; OPERATING EXPERIENCES: 80 Fuel Performance Annual Report for 1986 by W. J. Bailey, 103 Safety Considerations Related to 12-Hour Shift Schedules by S. Gould, 105 Reactor Shutdown Experience Compiled by J. W. Cletcher, 108 Operating U.S. Power Reactors Compiled by E. G. Silver; RECENT DEVELOPMENTS: 130 General Administrative Activities Compiled by E. G. Silver, 140 Reports, Standards, and Safety Guides by D. S. Queener, 145 Status of Power-Reactor Licensing Activities Compiled by E. G. Silver, 153 Proposed Rule Changes as of Sept. 30, 1988; ANNOUNCEMENTS: 69 Northwestern University Short Course on Radiation Safety, 79 International Seminar on Fission Product Transport Processes in Reactor Accidents, 129 Harvard School of Public Health Offers Short Courses, 152 MIT Offers Summer Short Courses on Reactor Safety and Nodal Analysis, 152 Seminar on Use of PCs in Probabilistic Safety Analysis, 157 The Authors, 160 Indexes to Nuclear Safety, Volume 29.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI1993

Nuclear Safety [Vol. 34, No. 1, January-March 1993]

Nuclear Safety is a review journal that covers significant developments in the field of nuclear safety. Its scope includes the analysis and control of hazards associated with nuclear energy, operations involving fissionable materials, and the products of nuclear fission and their effects on the environment. Primary emphasis is on safety in reactor design, construction, and operation; however, the safety aspects of the entire fuel cycle, including fuel fabrication, spent-fuel processing, nuclear waste disposal, handling of radioisotopes, and environmental effects of these operations, are also treated. Table of Contents for this issue follows. GENERAL SAFETY CONSIDERATIONS: 1 An Integrated Regional Approach To Risk Management of Industrial Systems, S. Chakraborty and R. Stratton; 9 Annual Technical Meeting of the NRC Cooperative Severe Accident Research Program, E. G. Silver; ACCIDENT ANALYSIS: 13 Deterministic Severe Accident Criteria as Severe Accident Design Criteria and Policy for the New Production Reactor-Heavy Water Reactor, P. T. Rhoads; 20 Proposed Deterministic Severe Accident Criteria for the Heavy Water Reactor-New Production Reactor Containment, K. D. Bergeron, S. E. Slezak, and C. E. Leach; 33 Quasi-Static Core Liquid Level Depression and Long-Term Core Uncovery During a PWR LOCA, Y. Kukita, R. R. Schultz, H. Nakamura, and J. Katayama; 49 LIRA: An Advanced Containment System to Minimize the Accidental Radioactivity Releases, A. Turricchia; 63 Errata to “A Review of Hydrogen Production During Melt/Water Interaction in LWRs,” Vol. 33, No. 4; CONTROL AND INSTRUMENTATION: 64 The Nuplex 80+™ Advanced Control Complex from ABB Combustion Engineering, F. Ridolfo, D. Harmon, and K. Scarola; DESIGN FEATURES: 76 A Scheme for Passive Isolation of the Containment of a Reactor, A. K. Ghosh, V. V. Raj, and K. Kakodkar; 84 Utility Requirements for Safety in the Passive Advanced Light-Water Reactor, T. U. Marston, W. H. Layman, and G. Bockhold, Jr.; WASTE AND SPENT FUEL MANAGEMENT: 91 Activities Related to Waste and Spent Fuel Management, Compiled by M. D. Muhlheim and E. G. Silver; OPERATING EXPERIENCES: 103 Testing Deficiencies in Auxiliary Systems Feedwater Systems, J. D. Kueck; 110 Reactor Shutdown Experience, Compiled by J. W. Cletcher; 113 Selected Safety-Related Events, Compiled by G. A. Murphy; 115 Operating U.S. Power Reactors, Compiled by M. D. Muhlheim and E. G. Silver; RECENT DEVELOPMENTS: 133 General Administrative Activities, Compiled by M. D. Muhlheim and E. G. Silver; 145 Reports, Standards, and Safety Guides, D. S. Queener; 150 Proposed Rule Changes as of Sept. 30, 1992; ANNOUNCEMENTS: 75 Harvard Short Course on Risk: Science, Assessment, and Management; 102 International Conference on Reactor Physics and Reactor Computations; 102 The Fourth International Conference on Nuclear Waste Reprocessing and Waste Management [RECOD ’94]; 132 International Symposium on the Radiological Exposure of the Population of the European Community from Radioactivity in the Mediterranean Sea [MARINA-MED]; 162 International High-Level Radioactive Waste Conference ‘Technology Integration—Above and Below”; 154 The Authors; 158 Indexes to Nuclear Safety, Volume 33.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Lysine 271 but not lysine 210 of XRCC4 is required for the nuclear localization of XRCC4 and DNA ligase IV

XRCC4 and DNA Ligase IV (LIG4) cooperate to join two DNA ends at the final step of DNA double-strand break (DSB) repair through non-homologous end-joining (NHEJ). However, it is not fully understood how these proteins are localized to the nucleus. Here we created XRCC4{sup K271R} mutant, as Lys271 lies within the putative nuclear localization signal (NLS), and XRCC4{sup K210R} mutant, as Lys210 was reported to undergo SUMOylation, implicated in the nuclear localization of XRCC4. Wild-type and mutated XRCC4 with EGFP tag were introduced into HeLa cell, in which endogenous XRCC4 had been knocked down using siRNA directed to 3′-untranslated region, and tested for the nuclear localization function by fluorescence microscopy. XRCC4{sup K271R} was defective in the nuclear localization of itself and LIG4, whereas XRCC4{sup K210R} was competent for the nuclear localization with LIG4. To examine DSB repair function, wild-type and mutated XRCC4 were introduced into XRCC4-deficient M10. M10-XRCC4{sup K271R}, but not M10-XRCC4{sup K210R}, showed significantly reduced surviving fraction after 2 Gy γ-ray irradiation as compared to M10-XRCC4{sup WT}. The number of γ-H2AX foci remaining 2 h after 2 Gy γ-ray irradiation was significantly greater in M10-XRCC4{sup K271R} than in M10-XRCC4{sup WT}, whereas it was only marginally increased in M10-XRCC4{sup K210R} as compared to M10-XRCC4{sup WT}. The present results collectively indicated that Lys271, but not Lys210, of XRCC4 is required for the nuclear localization of XRCC4 and LIG4 and that the nuclear localizing ability is essential for DSB repair function of XRCC4. - Highlights: • XRCC4{sup K271R} is defective in the nuclear localization of itself and LIG4. • XRCC4{sup K210R} is competent for the nuclear localization of itself and LIG4. • XRCC4{sup K271R} is deficient in DSB repair function. • XRCC4{sup K210R} is mostly normal in DSB repair function.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2018

Nuclear import of transcriptional corepressor BCOR occurs through interaction with karyopherin α expressed in human periodontal ligament

Highlights: • Human periodontal ligament cells express karyopherin α (KPNA) of three subclasses. • BCL6 corepressor (BCOR) is translocated to the nucleus by KPNA2, 4, and 6. • BCOR interacts with KPNA through nuclear localization signal 1. • BCOR mutation causing OFCD syndrome downregulates the SMAD4-SHH-NFIC pathways. Mutations in the gene encoding BCL-6 corepressor (BCOR) are responsible for oculofaciocardiodental (OFCD) syndrome, which is a rare X-linked dominant disorder characterized by radiculomegaly of permanent teeth as the most typical symptom. To function as a transcriptional corepressor, BCOR needs to enter the nucleus; however, the molecular pathway for its nuclear translocation during dental root formation remains unclear. The purpose of this study was to determine the mechanism underlying BCOR transport into the nucleus. Our results showed that human periodontal ligament (PDL) cells expressed karyopherin α (KPNA)2, KPNA4, and KPNA6 belonging to a family of nuclear import proteins, which interacted with BCOR in the immunoprecipitation assay. Site-directed mutagenesis targeting the two nuclear localization signals (NLSs) within BCOR reduced its nuclear translocation; however, co-expression of KPNA2, KPNA4, or KPNA6 with BCOR carrying a previously described mutation which eliminated one of the two NLSs significantly increased nuclear accumulation of the mutant BCOR, indicating participation of KPNA in BCOR nuclear translocation. Comparative expression profiling of PDL cells isolated from normal and OFCD patients revealed significant downregulation of SMAD4, GLI1, and nuclear factor 1-C (NFIC) mRNA expression, suggesting that BCOR mutations cause hyperactive root formation in OFCD syndrome by inhibiting SMAD4-Hedgehog-NFIC signaling implicated in dental root development. Our study contributes to understanding of the mechanisms providing nuclear import of BCOR during root formation.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2016

Elevated nuclear sphingoid base-1-phosphates and decreased histone deacetylase activity after fumonisin B1 treatment in mouse embryonic fibroblasts

Fumonisin B1 (FB1) is a mycotoxin produced by a common fungal contaminant of corn. Administration of FB1 to pregnant LM/Bc mice induces exencephaly in embryos, and ingestion of FB1-contaminated food during early pregnancy is associated with increased risk for neural tube defects (NTDs) in humans. FB1 inhibits ceramide synthase enzymes in sphingolipid biosynthesis, causing sphinganine (Sa) and bioactive sphinganine-1-phosphate (Sa1P) accumulation in blood, cells, and tissues. Sphingosine kinases (Sphk) phosphorylate Sa to form Sa1P. Upon activation, Sphk1 associates primarily with the plasma membrane, while Sphk2 is found predominantly in the nucleus. In cells over-expressing Sphk2, accumulation of Sa1P in the nuclear compartment inhibits histone deacetylase (HDAC) activity, causing increased acetylation of histone lysine residues. In this study, FB1 treatment in LM/Bc mouse embryonic fibroblasts (MEFs) resulted in significant accumulation of Sa1P in nuclear extracts relative to cytoplasmic extracts. Elevated nuclear Sa1P corresponded to decreased histone deacetylase (HDAC) activity and increased histone acetylation at H2BK12, H3K9, H3K18, and H3K23. Treatment of LM/Bc MEFs with a selective Sphk1 inhibitor, PF-543, or with ABC294640, a selective Sphk2 inhibitor, significantly reduced nuclear Sa1P accumulation after FB1, although Sa1P levels remained significantly increased relative to basal levels. Concurrent treatment with both PF-543 and ABC294640 prevented nuclear accumulation of Sa1P in response to FB1. Other HDAC inhibitors are known to cause NTDs, so these results suggest that FB1-induced disruption of sphingolipid metabolism leading to nuclear Sa1P accumulation, HDAC inhibition, and histone hyperacetylation is a potential mechanism for FB1-induced NTDs. - Highlights: • FB1 treatment results in accumulation of Sa1P primarily in the nucleus of MEFs. • FB1 treatment and elevated nuclear Sa1P are associated with HDAC inhibition. • Sphk2 inhibition alone significantly decreases nuclear Sa1P in response to FB1. • Sphk1 and Sphk2 inhibitors prevent nuclear Sa1P accumulation in response to FB1.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2017

Prolonged exposure to particulate chromate inhibits RAD51 nuclear import mediator proteins

Particulate hexavalent chromium (Cr(VI)) is a human lung carcinogen and a human health concern. The induction of structural chromosome instability is considered to be a driving mechanism of Cr(VI)-induced carcinogenesis. Homologous recombination repair protects against Cr(VI)-induced chromosome damage, due to its highly accurate repair of Cr(VI)-induced DNA double strand breaks. However, recent studies demonstrate Cr(VI) inhibits homologous recombination repair through the misregulation of RAD51. RAD51 is an essential protein in HR repair that facilitates the search for a homologous sequence. Recent studies show prolonged Cr(VI) exposure prevents proper RAD51 subcellular localization, causing it to accumulate in the cytoplasm. Since nuclear import of RAD51 is crucial to its function, this study investigated the effect of Cr(VI) on the RAD51 nuclear import mediators, RAD51C and BRCA2. We show acute (24 h) Cr(VI) exposure induces the proper localization of RAD51C and BRCA2. In contrast, prolonged (120 h) exposure increased the cytoplasmic localization of both proteins, although RAD51C localization was more severely impaired. These results correlate temporally with the previously reported Cr(VI)-induced RAD51 cytoplasmic accumulation. In addition, we found Cr(VI) does not inhibit interaction between RAD51 and its nuclear import mediators. Altogether, our results suggest prolonged Cr(VI) exposure inhibits the nuclear import of RAD51C, and to a lesser extent, BRCA2, which results in the cytoplasmic accumulation of RAD51. Cr(VI)-induced inhibition of nuclear import may play a key role in its carcinogenic mechanism since the nuclear import of many tumor suppressor proteins and DNA repair proteins is crucial to their function. - Highlights: • RAD51C nuclear localization is reduced after prolonged Cr(VI) exposure. • Prolonged Cr(VI) inhibits nuclear localization of BRCA2 less severely than RAD51C. • RAD51C deficiency induces the cytoplasmic accumulation of RAD51. • Cr(VI) does not inhibit interactions between RAD51 and its nuclear import partners.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2013

Antibody-based analysis reveals “filamentous vs. non-filamentous” and “cytoplasmic vs. nuclear” crosstalk of cytoskeletal proteins

To uncover the molecular composition and dynamics of the functional scaffold for the nucleus, three fractions of biochemically-stable nuclear protein complexes were extracted and used as immunogens to produce a variety of monoclonal antibodies. Many helix-based cytoskeletal proteins were identified as antigens, suggesting their dynamic contribution to nuclear architecture and function. Interestingly, sets of antibodies distinguished distinct subcellular localization of a single isoform of certain cytoskeletal proteins; distinct molecular forms of keratin and actinin were found in the nucleus. Their nuclear shuttling properties were verified by the apparent nuclear accumulations under inhibition of CRM1-dependent nuclear export. Nuclear keratins do not take an obvious filamentous structure, as was revealed by non-filamentous cytoplasmic keratin-specific monoclonal antibody. These results suggest the distinct roles of the helix-based cytoskeletal proteins in the nucleus. - Highlights: • A set of monoclonal antibodies were raised against nuclear scaffold proteins. • Helix-based cytoskeletal proteins were involved in nuclear scaffold. • Many cytoskeletal components shuttle into the nucleus in a CRM1-dependent manner. • Sets of antibodies distinguished distinct subcellular localization of a single isoform. • Nuclear keratin is soluble and does not form an obvious filamentous structure.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2016

Distinct hydrophobic “patches” in the N- and C-tails of beta-catenin contribute to nuclear transport

β-catenin is a key mediator of Wnt signaling and its deregulated nuclear accumulation can drive cancer progression. While the central armadillo (Arm) repeats of β-catenin stimulate nuclear entry, the N- and C-terminal “tail” sequences are thought to regulate turnover and transactivation. We show here that the N- and C-tails are also potent transport sequences. The unstructured tails of β-catenin, when individually fused to a GFP-reporter, could enter and exit the nucleus rapidly in live cells. Proximity ligation assays and pull-down assays identified a weak interaction between the tail sequences and the FG-repeats of nucleoporins, consistent with a possible direct translocation of β-catenin through the nuclear pore complex. Extensive alanine mutagenesis of the tail sequences revealed that nuclear translocation of β-catenin was dependent on specific uniformly distributed patches of hydrophobic residues, whereas the mutagenesis of acidic amino acids had no effect. Moreover, the mutation of hydrophobic patches within the N-tail and C-tail of full length β-catenin reduced nuclear transport rate and diminished its ability to activate transcription. We propose that the tail sequences can contribute to β-catenin transport and suggest a possible similar role for hydrophobic unstructured regions in other proteins. - Highlights: • We show that the N- and C-tails of beta-catenin possess nuclear transport activity. • Nuclear transport of the N- or C-tails requires specific hydrophobic amino acids. • Mutagenesis of the N-terminus diminished nuclear entry of full-length beta-catenin. • We propose the N-tail contributes to beta-catenin nuclear entry and transactivation.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2022

Techno-economic performances and life cycle greenhouse gas emissions of various ammonia production pathways including conventional, carbon-capturing, nuclear-powered, and renewable production

Ammonia (NH 3 ) is conventionally produced using fossil natural gas (NG) for hydrogen production through steam reformation and synthesis in the Haber–Bosch (HB) process. The global conventional ammonia production contributes more than 420 million tons of CO 2 emissions annually. Here, in this work, we investigated the techno-economics and well-to-plant-gate (WTG) greenhouse gas (GHG) emissions of conventional NG-based, carbon-capturing, nuclear-powered, and renewable ammonia production by developing an engineering process model for each. Carbon-capturing ammonia production refers to the NG-based ammonia production while capturing CO 2 and transporting it via pipelines for storage or utilization. Nuclear-powered and renewable ammonia production represents an alternative ammonia production through water electrolysis, air separation, and the HB process using carbon-free energy sources. Nuclear and renewable energy sources are assumed to be used for nuclear-powered and renewable ammonia production, respectively. Sensitivity analyses are performed for CO 2 pipeline transport distances, potential carbon capture tax credits, and clean H 2 production cost. Carbon-capturing ammonia production reduces WTG GHG emissions by 55–70% compared to conventional NG-based ammonia production methods. Nuclear-powered and renewable ammonia production almost eliminate GHG emissions since energy supply is either zero carbon or near-zero carbon. However, when $4.16–4.83 per kg is the cost assumed for clean H 2 production using state-of-the-art electrolysis technologies, the levelized costs of nuclear-powered and renewable ammonia are calculated to be $0.92–1.06 per kg NH 3 , which are approximately four times higher than the conventional NG-based ammonia production cost. The cost of CO 2 emission avoidance is estimated in the range of $266–318 per metric ton of CO 2 for nuclear-powered and renewable ammonia production. The clean H 2 production cost is the major contributor to the levelized costs of nuclear-powered and renewable ammonia production. A lower production cost, near $1 per kg H 2 , for clean H 2 is required for nuclear-powered and renewable ammonia production to be cost-competitive with conventional NG-based ammonia production.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2015

Hydrodynamical simulations of nuclear rings in barred galaxies

Dust lanes, nuclear rings, and nuclear spirals are typical gas structures in the inner region of barred galaxies. Their shapes and properties are linked to the physical parameters of the host galaxy. We use high-resolution hydrodynamical simulations to study 2D gas flows in simple barred galaxy models. The nuclear rings formed in our simulations can be divided into two groups: one group is nearly round and the other is highly elongated. We find that roundish rings may not form when the bar pattern speed is too high or the bulge central density is too low. We also study the periodic orbits in our galaxy models, and find that the concept of inner Lindblad resonance (ILR) may be generalized by the extent of x{sub 2} orbits. All roundish nuclear rings in our simulations settle in the range of x{sub 2} orbits (or ILRs). However, knowing the resonances is insufficient to pin down the exact location of these nuclear rings. We suggest that the backbone of round nuclear rings is the x{sub 2} orbital family, i.e., round nuclear rings are allowed only in the radial range of x{sub 2} orbits. A round nuclear ring forms exactly at the radius where the residual angular momentum of infalling gas balances the centrifugal force, which can be described by a parameter f{sub ring} measured from the rotation curve. The gravitational torque on gas in high pattern speed models is larger, leading to a smaller ring size than in the low pattern speed models. Our result may have important implications for using nuclear rings to measure the parameters of real barred galaxies with 2D gas kinematics.

79 ASTRONOMY AND ASTROPHYSICS↗
DOE OSTI2024

Space Nuclear Power Autonomous Control Algorithm and Control Element Test Bed

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

Wilson, Brandon↗