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This webinar kicked off the launch of the DOE EERE's Clean Energy to Communities (C2C) technical assistance program. It includes EERE's goals, a history of the C2C program development and structure, and details on its in-depth partnerships and Energyshed.
Emerging HPC platforms are becoming extraordinarily difficult to program as a result of complex, deep and heterogeneous memory hierarchies, heterogeneous cores, and the need to divide work and data among them. New programming models and libraries are being developed to aid in porting efforts of application developers, but substantial code restructuring is still necessary to fully make use of these new technologies. To do this effectively, these developers need information about their source code characteristics, including static and dynamic (e.g. performance) information to direct their optimisation efforts and make key decisions. On the other hand, system administrators need to understand how users are using the software stack and resources on their systems to understand what software they need to provide to improve the productivity of the users on a platform.In this paper we describe a tool that combines compiler and profiler information to query performance characteristics within an application or across multiple applications on a given platform. Static and dynamic data about applications are collected and stored together in an SQL database that can be queried by either a developer or a system administrator. We will demonstrate the capabilities of this tool via a real-world example from application-driven case studies that aims at understanding the use of scientific libraries on a routine from the molecular simulation application CP2K.
This fact sheet provides an overview of the National Renewable Energy Laboratory (NREL) State, Local, and Tribal Program. The State, Local, and Tribal Program partners with communities of all sizes to support the development of efficient, affordable, equitable, and resilient energy systems.
In this interactive workshop, attendees will hear from Grantees and Subgrantees of the U.S. Department of Energy's Weatherization Assistance Program (WAP), that have implemented innovative methods, materials and approaches in their weatherization program. Attendees will have the opportunity to exchange ideas and share their own stories and perspectives in breakout groups. Attendees will also learn how their state and agency may be able to work with the National Renewable Energy Lab and Oak Ridge National Lab on WAP-funded pilot projects to scale innovative methods throughout the WAP network.
Slides discussing the DOE Advanced Gas Reactor Fuel Development and Qualification Program Overview to include focus on LEU UCO TRISO fuel in cylindrical compacts, with data for fuel qualification in support of reactor licensing and a path to establish a domestic commercial TRISO fuel fabrication capability. Timelines from 2004 through 2029. Major program highlights include: Developed fuel performance models based on past observed behavior, Study to understand differences between US and German (“good”) fuel, Developed and improved fuel fabrication and characterization methods, Fabricated lab-scale fuel (AGR-1) with very high quality; excellent in-pile and post-irradiation high-temperature fuel performance, Scale-up coating fabrication (AGR-2) with excellent in-pile performance, AGR-3/4 fuel fabrication and irradiation experiment to assess FP transport behavior, Scale-up matrix fabrication (AGR-5/6/7); & fuel performance evaluation in progress.
Multiple studies suggest that by 2030, millions of jobs may go unfilled because of a significant skills gap between the current manufacturing workforce and the workforce needed to support the shift toward advanced and sustainable manufacturing. According to US manufacturers, finding the right talent is more difficult now than in previous years. Developing a well-trained workforce will play a critical role in transforming the manufacturing sector into a net-zero contributor of greenhouse gas emissions by 2050. To help reduce energy, water, waste, and greenhouse gas emissions as well as build a strong, sustainable manufacturing workforce, the US Department of Energy created the voluntary Better Plants program in 2011. This study discusses the program’s various workforce development activities, including in-plant trainings, virtual in-plant trainings, and in-person bootcamps. The target audiences, training objectives, and topics designed to upskill workers for a low-carbon manufacturing future are described. Feedback that was collected (e.g., new training topics, certifications, hands-on activities) while working with more than 270 US manufacturers for the past 12 years on workforce development are discussed. Finally, some strategies are proposed to address the feedback.
Interior point methods (IPMs) are a common approach for solving linear programs (LPs) with strong theoretical guarantees and solid empirical performance. The time complexity of these methods is dominated by the cost of solving a linear system of equations at each iteration. In common applications of linear programming, particularly in machine learning and scientific computing, the size of this linear system can become prohibitively large, requiring the use of iterative solvers, which provide an approximate solution to the linear system. However, approximately solving the linear system at each iteration of an IPM invalidates the theoretical guarantees of common IPM analyses. To remedy this, we theoretically and empirically analyze (slightly modified) predictor-corrector IPMs when using approximate linear solvers: our approach guarantees that, when certain conditions are satisfied, the number of IPM iterations does not increase and that the final solution remains feasible. We also provide practical instantiations of approximate linear solvers that satisfy these conditions for special classes of constraint matrices using randomized linear algebra.
A new multiparameter representative approach for the design of Advanced full MOX BWR core physics experimental programs is developed. The approach is based on sensitivity analysis of integral parameters to nuclear data, and correlations among different integral parameters. The representativeness method is here used to extract a quantitative relationship between a particular integral response of an experimental mock-up and the same response in a reference project to be designed. The study is applied to the design of the 100% MOX BASALA ABWR experimental program in the EOLE facility. The adopted scheme proposes an original approach to the problem, going from the initial 'microscopic' pin-cells integral parameters to the whole 'macroscopic' assembly integral parameters. This approach enables to collect complementary information necessary to optimize the initial design and to meet target accuracy on the integral parameters to be measured. The study has demonstrated the necessity of new fuel pins fabrication, fulfilling minimal costs requirements, to meet acceptable representativeness on local power distribution. (authors)
700 MWe Indian Pressurized Heavy Water Reactor (IPHWR) is horizontal channel type reactor with partial boiling at channel outlet. Due to boiling, it has a large volume of vapor present in the primary loops. It has two primary loops connected with the help of pressurizer surge line. The pressurizer has a large capacity and is partly filled by liquid and partly by vapor. Large vapor volume improves compressibility of the system. During turbine trip or load rejection, pressure builds up in Steam Generator (SG). This leads to pressurization of Primary Heat Transport System (PHTS). To control pressurization of SG and PHTS, around 70% of the steam generated in SG is dumped into the condenser by opening Condenser Steam Dump Valves (CSDVs) and rest of the steam is released to the atmosphere by opening Atmospheric Steam Discharge Valves (ASDVs) immediately after sensing the event. This is accomplished by adding anticipatory signal to the output of SG pressure controller. Anticipatory signal is proportional to the thermal power of reactor and the proportionality constant is set so that SG pressure controller's output jacks up to ASDV opening range when operating at 100% FP. To simulate this behavior for 700 MWe IPHWR, Primary and secondary heat transport system is modeled. SG pressure control and other process control program have also been modeled to capture overall plant dynamics. Analysis has been carried out with 3-D neutron kinetics coupled thermal hydraulic computer code ATMIKA.T to evaluate the effect of the anticipatory signal on PHT pressure and over all plant dynamics during turbine trip in 700 MWe IPHWR. This paper brings out the results of the analysis with and without considering anticipatory signal in SG pressure control program during turbine trip. (authors)
The Formerly Utilized Sites Remedial Action Program (FUSRAP) has addressed sites across the nation for almost 4 decades. Multiple stake holder pressures, multiple regulations, and process changes occur over such long time periods. These result in many challenges to the FUSRAP project teams. Initial FUSRAP work was not performed under Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) Records of Decision (ROD). The ROD identifies the remedy decision and ultimately the criteria to be used to release a site. Early FUSRAP projects used DOE Orders or the Uranium Mill Tailings Radiation Control Act (UMTRCA) standards. Under current RODs, regulations may differ, resulting in different cleanup criteria than that used in prior Vicinity Property (VP) remediation. The USACE, in preparation for closeout of Sites, conducts reviews to evaluate whether prior actions were sufficient to meet the cleanup criteria specified in the current ROD. On the basis of these reviews, USACE has conducted additional sampling, determined that prior actions were sufficient, or conducted additional remediation consistent with the selected remedy in the ROD. As the public pressures, regulations, and processes that the FUSRAP encounters continue to change, the program itself continues to evolve. Assessment of VPs at FUSRAP sites is a necessary step in the life cycle of our site management. (authors)
The NREL Student Training in Applied Research (STAR) internship is a year-long program that provides undergraduate students from minority-serving institutions the opportunity to contribute to cutting-edge research while progressing in their studies at their home universities. The program connects students and faculty with NREL's team to foster a future generation of industry leaders and sustained research engagement.
The French Planning Act of 28 June 2006 prescribed that a reversible repository in a deep geological formation be chosen as the reference solution for the long-term management of high-level and intermediate-level long-lived radioactive waste. It also entrusted the responsibility of further studies and design of the repository (named Cigeo) upon the French Radioactive Waste Management Agency (Andra), in order for the review of the creation-license application to start in 2015 and, subject to its approval, the commissioning of the repository to take place in 2025. Andra is responsible for siting, designing, implementing, operating the future geological repository, including operational and long term safety and waste acceptance. Nuclear operators (Electricite de France (EDF), AREVA NC, and the French Commission in charge of Atomic Energy and Alternative Energies (CEA) are technically and financially responsible for the waste they generate, with no limit in time. They provide Andra, on one hand, with waste packages related input data, and on the other hand with their long term industrial experiences of high and intermediate-level long-lived radwaste management and nuclear operation. Andra, EDF, AREVA and CEA established a cooperation agreement for strengthening their collaborations in these fields. Within this agreement Andra and the nuclear operators have defined an industrial program for waste management. This program includes the waste inventory to be taken into account for the design of the Cigeo project and the structural hypothesis underlying its phased development. It schedules the delivery of the different categories of waste and defines associated flows. (authors)
The Port Hope Initiative (PHAI) involves the cleanup of historic low-level radioactive waste in various locations throughout the communities of Port Hope and Clarington, Ontario, as well as the construction of two engineered aboveground mounds for safe long-term management. The PHAI is comprised of two major projects - the Port Hope Project and the Port Granby Project. An Environmental Assessment (EA) was undertaken for each project and as a result EA Follow-up Programs were developed and are being implemented addressing both biophysical and socioeconomic aspects. This paper provides insight on elements of the EA Follow-up Program development, and its implementation. (authors)
The Net Zero World initiative, a U.S. flagship program launched at COP26, is designed to support countries in accelerating transitions to clean energy systems. Combining expertise from 10 U.S. Department of Energy laboratories and 9 U.S. government agencies, Net Zero World collaborates with governments in Argentina, Chile, Egypt, Indonesia, Nigeria, Singapore, Thailand, and Ukraine to fortify net-zero pathways, advance implementation of priority actions, and mobilize significant investment in clean energy projects and infrastructure, targeting $10 billion by 2025. This fact sheet highlights program impacts to be presented at COP28.
The U.S. Department of Energy (DOE) methods and protocols allow evaluation of remediation and final site conditions to determine if remediated sites remain protective. Two case studies are presented that involve the Niagara Falls Storage Site (NFSS) and associated vicinity properties (VPs), which are being remediated under the Formerly Utilized Sites Remedial Action Program (FUSRAP). These properties are a part of the former Lake Ontario Ordnance Works (LOOW). In response to stakeholders concerns about whether certain remediated NFSS VPs were putting them at risk, DOE met with stakeholders and agreed to evaluate protectiveness. Documentation in the DOE records collection adequately described assessed and final radiological conditions at the completed VPs. All FUSRAP wastes at the completed sites were cleaned up to meet DOE guidelines for unrestricted use. DOE compiled the results of the investigation in a report that was released for public comment. In conducting the review of site conditions, DOE found that stakeholders were also concerned about waste from the Separations Process Research Unit (SPRU) at the Knolls Atomic Power Laboratory (KAPL) that was handled at LOOW. DOE agreed to determine if SPRU waste remained at that needed to be remediated. DOE reviewed records of waste characterization, historical handling locations and methods, and assessment and remediation data. DOE concluded that the SPRU waste was remediated on the LOOW to levels that pose no unacceptable risk and allow unrestricted use and unlimited exposure. This work confirms the following points as tenets of an effective long-term surveillance and maintenance (LTS and M) program: - Stakeholder interaction must be open and transparent, and DOE must respond promptly to stakeholder concerns. - DOE, as the long-term custodian, must collect and preserve site records in order to demonstrate that remediated sites pose no unacceptable risk. - DOE must continue to maintain constructive relationships with the U.S. Army Corps of Engineers and state and federal regulators. After review of historical site documentation, DOE reports, and USACE radiological data, DOE concluded the following: - DOE had access to adequate documentation to evaluate site conditions at the former LOOW. This is important to confirm now, while institutional knowledge of early FUSRAP work remains available. - DOE remediated the completed VPs to conditions that are protective for unrestricted residential use. Sample and walkover gamma scan results indicate that no wastes remain that exceed cleanup criteria. - Process knowledge and field observations establish that Cs-137 is the predominant radionuclide in the KAPL waste stream. Cs-137, a strong gamma emitter, was used as an indicator for remediation of KAPL waste. Other radionuclides were present in much lower relative concentrations and were likely also removed during remediation of the VPs. - KAPL contaminants were removed during remedial activities at the former LOOW as either co-located or co-mingled with other radionuclides. - For the active VPs (VP-E, VP-E', and VP-G), results of DOE's cleanup of the accessible portions of these properties indicate that KAPL waste does not remain at concentrations greater than the DOE cleanup limit: - Inaccessible areas were not associated with historic KAPL waste handling. Therefore, it is unlikely that KAPL waste remains on the active VPs. - Because gamma activity was used by DOE during remediation/verification activities for excavation control, additional USACE cleanup of FUSRAP wastes on these properties will likely result in the remediation of any co-located residual KAPL wastes to acceptable levels or identification of KAPL waste that is not co-located. - Although USACE has not established a cleanup level for Cs-137 on the active NFSS VPs, DOE assessment and remediation data indicate that assessed Cs-137 was remediated and significant Cs-137 is unlikely to remain. Because of the low likelihood of encountering significant KAPL waste on the active NFSS VPs, additional remediation is not anticipated at these properties. - USACE assessment soil sampling results on the NFSS proper indicate that KAPL waste does not exceed the DOE cleanup level for Cs-137. USACE has not established a cleanup level for Cs-137 on NFSS proper. The USACE cleanup of FUSRAP wastes on the NFSS proper will likely result in the remediation of any co-located residual KAPL wastes or identification of KAPL waste that is not co-located. DOE is drafting a report of the investigation of KAPL waste at LOOW. The report will be released to the public for comment when the draft is complete. DOE responses to stakeholder inquiries resulted in a common understanding of site conditions and site risk. DOE expects additional interaction with stakeholders at the former LOOW as USACE completes remediation of the active VPs and the NFSS proper, and these relationships will hopefully have built trust between DOE and the stakeholders that DOE will perform its duties in an open and transparent manner that includes stakeholders as stewards for remediated FUSRAP sites. (authors)
Spain has to manage about 3700 tons of irradiated graphite from the reactor Vandellos I as radioactive waste. 2700 tons are the stack of the reactor and are still in the reactor core waiting for retrieval. The rest of the quantities, 1000 tons, are the graphite sleeves which have been already retrieved from the reactor. During operation the graphite sleeves were stored in a silo and during the dismantling stage a retrieval process was carried out separating the wires from the graphite, which were crushed and introduced into 220 cubic containers of 6 m{sup 3} each and placed in interim storage. The graphite is an intermediate level radioactive waste but it contains long lived radionuclides like {sup 14}C which disqualifies disposal at the low level waste repository of El Cabril. Therefore, a new project has been started in order to investigate two new options for the management of this waste type. The first one is based on a selective decontamination of {sup 14}C by thermal methods. This method is based on results obtained at the Research Centre Juelich (FZJ) in the Frame of the EC programs 'Raphael' and 'Carbowaste'. The process developed at FZJ is based on a preferential oxidation of {sup 14}C in comparison to the bulk {sup 12}C. Explanations for this effect are the inhomogeneous distribution and a weaker bounding of {sup 14}C which is not incorporated in the graphite lattice. However these investigations have only been performed with graphite from the high temperature reactor Arbeitsgemeinschaft Versuchsreaktor Juelich AVR which has been operated in a non-oxidising condition or research reactor graphite operated at room temperature. The reactor Vandellos I has been operated with CO{sub 2} as coolant and significant amounts of graphite have been already oxidised. The aim of the project is to validate whether a {sup 14}C decontamination can also been achieved with graphite from Vandellos I. A second possibility under investigation is the encapsulation of the graphite in a long term stable glass matrix. The principal applicability has been already proved by FNAG. Crushed graphite mixed with a suitable glass powder has been pressed at elevated temperature under vacuum. The vacuum is required to avoid gas enclosures in the obtained product. The obtained products, named IGM for 'Impermeable Graphite Matrix', have densities above 99% of theoretical density. The amount of glass has been chosen with respect to the pore volume of the former graphite parts. The method allows the production of encapsulated graphite without increasing the disposal volume. This paper will give a short overview of characterisation results of different irradiated graphite materials obtained at CIEMAT and in the Carbowaste project as well as the proposed methods and the actual status of the program including first results about leaching of non-radioactive IGM samples and hopefully first tendencies concerning the C-14 separation from graphite of Vandellos I by thermal treatment. Both processes, the thermal treatment as well as the IGM, have the potential to solve problems related to the management of irradiated graphite in Spain. However the methods have only been tested with different types of i-graphite and virgin graphite, respectively. Only investigations with real i-graphite from Spain will reveal whether the described methods are applicable to graphite from Vandellos I. However all partners are convinced that one of these new methods or a combination of them will lead to a feasible option to manage i-graphite in Spain on an industrial scale. (authors)
In the framework of the PROMETRA program, clad hardening constitutive laws (stress-strain curves) have been already implemented in the SCANAIR code. These laws were derived from uniaxial hoop tensile tests performed on narrow machined ring samples. However, they do not reproduce the conditions of strain multi-axiality acting on cladding during a RIA (Reactivity-Initiated Accident) in a PWR. As a consequence, a new specimen called Plane Strain Tensile (PST) similar to the annular sample developed by Pennsylvania State University has been developed at CEA on a wider notched ring. It aims to accurately predict the cladding failure in more prototypical conditions, close to plane strain hoop tension in the middle of the gage section. This paper details the PST tests implementation and results in the framework of the PROMETRA program on fresh and irradiated claddings. The development of this test has led also to the development of a technique to measure local deformations in the gage section of the PST specimen using 2-D Digital Image Correlation (DIC). Tests performed on fresh Zircaloy-4 and on irradiated Zircaloy-4, M5 and ZIRLO specimens are analyzed relying of post-test examinations and finite element simulations. In the case of tests performed on irradiated Zry-4 and ZIRLO at 280 and 350 Celsius degrees, the presence of a hydride rim favors the crack nucleation on the outer surface. In this case, rupture nucleation occurs preferentially in the vicinity of notches. Such tests are not representative of the rupture of a cladding subjected to plane strain loading. Nevertheless, they give a good representation of the initiation of rupture observed during RIA on irradiated cladding, resulting in a hydride damage. Zry-4 specimens tested at 480 Celsius degrees and M5TM at 280 and 350 C. degrees result in a ductile fracture nucleation in the center of the specimen. This rupture is prototypical of a ductile rupture of an irradiated specimen subjected to a plane strain loading. However, the nucleation on inner surface by shear at the mandrel contact is not prototypical of a clad rupture during the PCMI (Pellet Clad Mechanical Interaction) phase of a RIA (membrane stresses in the clad thickness in RIA). An important lesson learned from this experimental approach is that the PST test is a particularly complex test, both in its implementation and interpretation.
On May 2012, Best Medical Belgium (BMB) S.A., a manufacturer of pharmaceutical products located on the nuclear site of Fleurus, was declared bankrupt. According to the Belgian law, ONDRAF/NIRAS (Belgian Radioactive Waste Management Agency) was entrusted with the clean-up and the decommissioning operations of the former BMB facilities, nowadays known as 'ONDRAF/NIRAS Site Fleurus' (ONSF). In June 2015, ONSF launched a Decontamination and Decommissioning (D and D) program, for which Tractebel supports the definition of the clean-up and decommissioning strategies; the drafting of the Final Decommissioning Plan (FDP) and License Application for decommissioning; and the follow-up of the clean-up and upcoming D and D works on site. Four nuclear buildings are part of the clean-up and D and D activities. One of them, Building 14 (B14), represents the most challenging task with the clean-up of highly contaminated areas and their further decommissioning up to greenfield. Beside its two cyclotrons successively used for the production of radioisotopes, this building features hot cells and glove boxes used for the purification of radioisotopes and the conditioning of the final pharmaceutical product. The present paper aims at presenting the development of the D and D program in terms of modifications of existing facilities to allow carrying out the clean-up and D and D activities and the management of dismantled radioactive materials in a safe, efficient and cost-effective way as well as in terms of strategic options to be developed for the FDP. To support the clean-up and D and D activities, several workshops have been designed in order to treat radioactive materials removed from the building rooms. Those workshops feature an input buffer storage area, a common output buffer storage area and installations for the treatment of radioactive materials and waste (handling devices, cutting means, decontamination means, free release measurements means, etc.). Some modifications to the existing auxiliary systems will be required in order to dispose of appropriate systems to operate those workshops in a safe way. Activated concrete structures will constitute the most important part of radioactive materials that will arise from the decommissioning activities. To limit the total decommissioning costs, and especially the costs associated to radioactive waste management, two strategies have been developed in order to manage a part of activated concrete as candidate for free release with a certain decay period or as candidate for conventional landfill. Both strategies will have to be further detailed and discussed with the competent Authorities to ensure they can actually be implemented. (authors)