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Qualification Plus is an accelerated test protocol and quality management system that gives higher confidence in field performance of PV modules compared with conventional qualification testing. The test sequences are being developed as consensus standards, but the early publication of these tests enables the community to begin benefiting from them sooner.
This report summarizes a set of test methods that are in the midst of being incorporated into IEC 61215 for certification of a module design or other tests that go beyond certification to establish bankability.
As the PV Industry has matured, there is increasing interest in quantifying higher quality so that customers may select products that meet their needs, and investors are demanding high confidence before committing millions of dollars. The community is developing the knowledge and standard methods needed to address these. This talk will give an overview of new developments in the last year including descriptions and status of Qualification Plus (a test protocol that goes beyond the standard qualification tests), a PV-specific version of ISO 9001 (for increased confidence in manufacturing consistency), and the Regional Test Centers (where new products can be tested to quantify performance and durability in multiple climate zones.) Development of PV system-level certification will also be described.
Each year the world invests tens of billions of dollars or euros in PV systems with the expectation that these systems will last approximately 25 years. Although the disciplines of reliability, quality, and service life prediction have been well established for numerous products, a full understanding of these is currently challenging for PV modules because the desired service lifetimes are decades, preventing direct verification of lifetime predictions. A number of excellent reviews can be found in the literature summarizing the types of failures that are commonly observed for PV modules. This chapter discusses key failure/degradation mechanisms selected to highlight how the kinetics of failure rates can and cannot be confidently predicted. For EVA-encapsulated modules, corrosion is observed to follow delamination, which then allows water droplets to directly contact the metallization. Extended test protocols such as Qualification Plus were created to address the known problems while standards groups update standard tests through the consensus process.
The rate and type of in-service PV performance degradation vary for different module designs, vintages, and operating climates. The current IEC 61215 module qualification testing standard can be used to identify design and manufacturing flaws that may lead to early life failures, but it was not intended to accurately predict degradation rates in the first 5 to 10 years of a PV plant’s lifetime, nor long-term performance, under varied operating conditions. Correlations between accelerated aging data and real-world degradation data are needed to better predict wear-out degradation in response to environmental stressors, such as elevated temperature, temperature cycling, moisture, wind load, and solar radiation. The PREDICTS2 project aged 36 pristine-condition ‘spare’ modules from a commercial PV power plant in the U.S. Southwest using the IEC 61215 and Qualification Plus (Qual Plus) testing standards. The study is unique in its ability to compare performance degradation data for fielded modules operating outdoors for 6 years in a commercial plant environmental with degradation data for modules of the same make, model and vintage that have been aged artificially in an indoor laboratory. Changes in individual module current-voltage (I-V) curve data were measured before and after the artificial aging process and compared to I-V curve data for the fielded modules. This project advanced the industry’s understanding of how environmental stressors affect performance degradation and established relationships between degradation mechanisms and changes to module electrical parameters and power loss. Improved certainty in degradation rates is expected to be the greatest source of LCOE reduction resulting from this project. Reducing the spread in the degradation rate probability distribution lowers project risk, which translates to lower project financing costs.
The mechanical behavior of a cast form of an advanced austenitic stainless steel, CF8C-Plus, is compared with that of its wrought equivalent in terms of both tensile and creep-rupture properties and estimated allowable stress values for pressurized service at temperatures up to about 850°C. A traditional Larson-Miller parametric model is used to analyze the creep-rupture data and to predict long-term lifetimes for comparison of the two alloy types. The cast CF8C-Plus exhibited lower yield and tensile strengths, but higher creep strength compared to its wrought counterpart. Two welding methods, shielded-metal-arc welding (SMAW) and gas-metal-arc welding, met the weld qualification acceptance criteria in ASME BPVC Section IX for the cast CF8C-Plus. However, for the wrought CF8C-Plus, while SMAW and gas-tungsten-arc welding passed the tensile acceptance criteria, they failed the side bend tests due to lack of fusion or weld metal discontinuities.
The objective of this white paper is to provide an overview of the state of knowledge regarding issues associated with fuel qualification for a liquid-fueled molten salt reactor. The overview consists of fuel qualification expectations for heterogeneous reactors—conventional light water reactors, plus advanced reactors such as sodium fast reactors and modular high-temperature gas-cooled reactors—which consider the mechanical behavior of the fuel itself, interaction between fuel and its cladding, behavior of the cladding, and structural behavior of fuel assemblies. Fuel qualification expectations for molten salt reactors are viewed in contrast to the fuel qualification for heterogeneous reactors in that the fuel is liquid (molten) and there is no accumulation of mechanical effects such as stress, creep, or swelling. However, the liquid fuel salts do accumulate fission products, corrosion byproducts, and buildup of actinides which impact the chemistry of the fuel salt and thermal-physical properties that may impact the ability to remove heat from the core and retain radioactive materials in the fuel salt. Finally, the role of the US Department of Energy in research on tri-structural-isotropic and metal fuel for advanced reactors for fuel qualification purposes is briefly discussed as a potential model for the Department’s research to gather quality chemical and thermal-physical property data necessary to qualify MSR fuel performance under normal and accident conditions.
TIMS at SRNL • SRNL consolidated analytical laboratories and services into one area – loss of TIMS capability during relocation. • SRNL installed a new Thermo Scientific TRITON Plus TIMS in our Category II Nuclear Facility to re-establish the lab’s TIMS capabilities in Feb. 2024. • Allows for handling, preparation, purification, and analysis of high-level radiological materials within a single location • Pu and U isotopic composition and assay measurement qualification have been priority since installation. • Once the instrument has been qualified, this capability will support internal and external customers.
Reinvestment in the Los Alamos Canyon Bridge (LACB) deck, superstructure, and substructure, ensures continued access between Los Alamos National Laboratory (LANL) and the Los Alamos County (LAC) town site; as well as maintains emergency response qualifications to/from the Los Alamos Medical Center (LAMC). The operability of the LACB is critical to the institution and with the evolving oversight from the Federal Highway Administration (FHWA), Department of Energy (DOE) Headquarters, and National Nuclear Security Administration (NNSA); it is imperative safety of the infrastructure is sustained. The Los Alamos Canyon Bridge is 60 plus years old (typical bridge design life of 50 years), a fracture critical structure (no girder redundancies), has an average daily traffic in excess of 13,000 vehicles/day (primary route for LAC commuters), is deteriorating (decreasing code ratings, per inspection), was last rehabilitated in 1992, and maintains access for fire evacuations (Cerro Grande, Las Conchas). Reinvestment in the bridge is expected to extend the service life of the LACB, reduce average maintenance costs by $0.5 M to $2.8M, per year, and provide compliance with current code requirements. Pros, cons, and proposed funding schedule are provided.
The fabrication of the modules for the ITER Central Solenoid (CS) is in progress at General Atomics (GA) in Poway, California, USA. This purpose built facility has been established with the requisite tools and machines to fabricate the seven 110 tonne CS modules (six required plus one spare). GA’s project responsibility is completing the fabrication design, developing and qualifying the fabrication processes and tools, fabricating the seven modules, and testing the modules at 4.5 K with full current. The current schedule has the first module’s completing fabrication in 2018 followed by electrical and full current testing. Upon completion of testing, the modules will be shipped to the ITER site for assembly into the CS stack. GA has co mpleted the fabrication of a qualification coil, simulating all of the processes and exercising all of the tooling required to fabricate a production module. After completion, the qualification coil underwent a cool down cycle to 4.5K which served as the final commissioning step of the helium cooling system on the parallel flow paths of a module. In addition, other final test station systems and procedures were simulated to verify their functionality including simulated critical current sharing temperature (Tcs) measurements, DC switch operation to dump the energy from the coil, and a helium leak test on the coil at <10K. After warming the coil, Paschen tests were repeated prior to the coil being dissected to evaluate the quality of the vacuum pressure impregnation (VPI). Dissection is scheduled for early spring 2018 and the results showed full resin impregnation of all areas. In the production and testing of the qualification coil, many lessons were learned and the improvements incorporated into the module fabrication processes. GA currently has five production modules in fabrication, at different stages in the process. The first module is nearly ready for vacuum pressure impregnation. The second and third modules are currently in the insulation process and the 7 pancakes of the forth module are being joined together into a 110 tonne module. Module 5 is currently being wound. The qualification coil has been completed and dissected to evaluate the VPI. A description of the VPI process is provided. To verify the turn insulation of a module, unique impulse testing methods were developed to test the insulation between the 560 turns of a module. These methods are reported in this paper.
The nuclear industry and the U.S. Department of Energy are jointly developing accident tolerant fuel (ATF) which has the potential to greatly improve the safety and economic performance of the current fleet of NPPs. The combination of ATF with the optimal use of FLEX equipment (deployed as an industry response to the Fukushima Daiichi Nuclear Power Plant (NPP) accident in 2012), enhancements on plant components and systems, incorporation of augmented cooling systems or new passive cooling systems, and improvements in fuel cycle efficiency, have the potential to render the NPPs more efficient and resilient to events (both internally and externally initiated) that have the potential to challenge the integrity of the fuel or plant containment. In this report, combinations of these plant improvements are called Enhanced Resilient Plant Systems. In order to accelerate the development and deployment of the components that comprise Enhanced Resilient Plant Systems, a well-integrated toolkit that is fuel/clad-, fuel cycle-, systems-, and scenario-centric is required to provide a ready platform for the analysis of the advanced ATF fuel and cladding materials that are being developed, as well as plant systems modifications to achieve enhanced resiliency. Since the deployment of these enhancements will comprise a risk/benefit tradeoff of their impacts on plant safety and economics, the envisioned toolkit will need to be capable to characterizing the multidimensional aspects associated with these evaluations and provide decision-makers with information that can optimize the technological, economic, and safety benefits that are to be obtained. The Risk-Informed Systems Analysis (RISA) Pathway of the DOE Light Water Reactor Sustainability (LWRS) Program initiated several research tasks to develop analytical capabilities to support the industry in the transition to the ATF and Enhanced Resilient Plant Systems. The general idea behind the initiative is the development of an integrated evaluation approach which combines the plant probabilistic risk assessment (PRA) models with Multi-Physics Best Estimate Plus Uncertainty (MP-BEPU) analyses in a seamless fashion. The integrated evaluation framework aims at enabling plant system configuration variations to be studied with speed and precision, including detailed risk and benefit assessments of introducing ATF and Enhanced Resilient Plant Systems into current LWR plants to achieve both safety and operational performance enhancements. The focus of this report is to present an integrated research and development (R&D) roadmap to identify and perform high value evaluations of proposed ATF and Enhanced Resilient Plant Systems concepts to identify both the technical (e.g. benefits to risk, safety, and operational margins) and the economic (i.e. business and cost) elements associated with industry adoption of the technology. The integrated evaluation approach will be developed to support the development, testing, qualification, licensing, and deployment of ATF and Enhanced Resilient Plant Systems technology that is capable of achieving substantial safety and economic improvements as well as timely widespread adoption by the US nuclear industry.
This paper presents a case study in utilizing information from experiments, models, and verification and validation (V&V) to support a decision. It consists of a simple system with data and models provided, plus a safety requirement to assess. The goal is to pose a problem that is flexible enough to allow challengers to demonstrate a variety of approaches, but constrained enough to focus attention on a theme. This was accomplished by providing a good deal of background information in addition to the data, models, and code, but directing the participants' activities with specific deliverables. In this challenge, the theme is how to gather and present evidence about the quality of model predictions, in order to support a decision. This case study formed the basis of the 2014 Sandia V&V Challenge Workshop and this resulting special edition of the ASME Journal of Verification, Validation, and Uncertainty Quantification.
Post-irradiation examination (PIE) is critical for the performance assessment and qualification of nuclear fuels. Secondary ion mass spectrometry (SIMS) is a powerful materials characterization technique that allows for elemental and isotopic mapping with a depth resolution greater than EDS and EPMA. However, it has not yet been applied to PIE of metallic nuclear fuel. Here, in this work, we characterize an fast neutron spectrum irradiated U-10Zr fuel sample using a time-of-flight SIMS (ToF-SIMS) system connected to a FIB/SEM system, which allows for flexible sample analysis compared to a dedicated ToF-SIMS instrument. Analysis of the resulting hyperspectral micrograph data was aided by the development of an unsupervised machine learning (ML) algorithm that iterates on existing methods to segment the 3D micrographic datasets based on the similarity of mass spectra. The results showed that the FIB-ToF-SIMS instrument was potentially capable of spatially resolving closed fission gas bubbles in 3D by continued ion sputtering of the analyzed volume. Additionally, the ML algorithm proved useful in revealing the chemical segregation of light fission products (those with an atomic mass between approximately 85–105 amu, such as ruthenium and rhodium) plus matrix zirconium, heavy fission products (those with an atomic mass between approximately 135–150 amu, such as the lanthanides) and uranium. Future studies are planned to conduct FIB-ToF-SIMS analysis on more irradiated U-Zr samples to study the constituent redistribution.
Nuclear power plant cables were originally qualified for 40 year life and generally have not required specific test verification to assure service availability through the initial plant qualification period. However, license renewals to 60 and 80 years of operation require a cable aging management program that depends on some form of test and verification to assure fitness for service. Environmental stress (temperature, radiation, chemicals, water, and mechanical) varies dramatically within a nuclear power plant and, in some cases, cables have degraded and required repair or replacement before their qualified end-of-life period. In other cases, cable conditions have been mild and dependable cable performance confirmed to extend well beyond the initial qualified life. Most offline performance-based testing requires cables to be de-coupled and de-energized for specially trained technicians to perform testing. These offline tests constitute an expensive operational burden that limits the economic viability of nuclear power plants. Although initial investment may be higher, new online test practices are emerging as options or complements to offline testing that avoid or minimize the regularly scheduled offline test burden. These online methods include electrical and fiber-optic partial discharge measurement, spread spectrum time or frequency domain reflectometry, distributed temperature profile measurements, and local interdigital capacitance measurement of insulation characteristics. Introduction of these methods must be supported by research to confirm efficacy plus either publicly financed or market driven investment to support the start-up expense of cost-effective instrumentation to monitor cable condition and assure reliable operation. This work summarizes various online cable assessment technologies plus introduces a new cable motor test bed to assess some of these technologies in a controlled test environment.
This comprehensive investigation examines the structure–property relationships in two nuclear alloy systems—Alloy 709 (A709) austenitic stainless steel and Grade 92 (G-92) ferritic/martensitic (F/M) steel—manufactured via directed energy deposition (DED) for sodium-cooled fast reactor applications. This study establishes the fundamental mechanisms for controlling microstructures for optimizing the mechanical performance of additively manufactured nuclear materials through systematic heat treatment optimization and multiscale characterization. As-deposited A709 steel develops a complex multiscale strengthening architecture consisting of a fine cellular solidification structure with diameter of 2-3 µm within10–50 µm grains, elevated dislocation densities from rapid thermal cycling, and grain boundary precipitates that activate concurrent Hall–Petch, dislocation, and precipitation hardening mechanisms to achieve exceptional properties [yield strength (YS): 603 MPa, ultimate tensile strength (UTS): 844 MPa, Vickers hardness: 220 HV] that achieve a 44% superior strength compared to that of the wrought material. Heat treatments produce different results. Solution annealing (SA) dissolves the cellular structure and reduces the hardness to 190 HV. Precipitation treatment (PT) keeps the cellular structure but adds carbides, allowing the hardness to reach 205 HV. The best approach combines both treatments (SA+PT) and creates uniform precipitate distributions with M 23 C 6 carbides at the grain boundaries and MX carbonitrides in the matrix, achieving a hardness of 195 HV. However, directional differences persist, with a 12%–15% strength variation between orientations due to the inherited layered microstructural architecture that survives aggressive heat treatment. While tensile testing at 550°C demonstrates 40%–50% thermal softening with dynamic strain aging, DED A709 steel still maintains a 71% higher YS than that of the wrought material. Ion irradiation studies (100–400 dpa) of DED A709 steel reveal progressive radiation damage with increasing void density and radiation-induced segregation causing nickel enrichment and chromium depletion, which will ultimately compromise mechanical properties. As-deposited G-92 exhibits exceptional strength (UTS: 1650–1700 MPa, 430 HV) through a complex microstructure containing both ferrite and martensite phases, a high geometrically necessary dislocation (GND) density (17.04×10 14 /m 2 ), and fine carbides. Heat treatments create distinct changes. Normalizing produces fresh martensite with the highest hardness (460 HV) and an increased GND density (20.23×10 14 /m 2 ). Tempering develops dual precipitation systems and reduces the hardness to 290 HV. The optimal approach uses sequential normalizing plus tempering, achieving balanced properties with the lowest hardness (250 HV) and a reduced GND density (11.01×10 14 /m 2 ). A processing-dependent anisotropy is observed: horizontal specimens achieve superior ductile behavior, while vertical specimens exhibit brittle failure. A tempering heat treatment successfully mitigates this anisotropic behavior by transforming the hard martensitic as-deposited structure into tempered martensite enabling both horizontal and vertical specimens to exhibit similar stress–strain characteristics with visible necking behavior. Remarkably, testing at 550°C reveals a reversal in the anisotropy, where as-deposited specimens achieve near isotropy with superior thermal stability (a 15%–20% strength reduction), while tempered specimens develop an orientation dependence with a 25%–30% strength reduction. Both alloy systems demonstrate that DED processing creates specimens with a superior strength through refined microstructural features, though with distinct strengthening mechanisms—austenitic through cellular structures and precipitates versus F/M through phase transformations and precipitates. Heat treatment optimization requires alloy-specific approaches, with A709 benefiting from controlled precipitation while G-92 requires careful phase transformation control. The results show that DED manufacturing can produce nuclear materials with exceptional performance, but directional effects and temperature-dependent behavior must be carefully considered for reactor component design and qualification.
Collecting radiation measurements can be a hazardous task, especially in the presence of highly active sources. Various scenarios necessitate source search, classification, and quantification, often with limited or no a priori information. Activities such as disaster mitigation, emergency response, environmental monitoring, and site remediation may involve dangerous radioactive sources. In these situations, there is a pressing need for remote monitoring capabilities that protect human operators from potential harm and enhance adherence to the principle of “As Low as Reasonably Achievable” (ALARA) for radiation doses. The first step toward achieving remote radiation measurement capabilities is the remote operation of a radiation sensor. Once this milestone is reached, the next challenge is to integrate this remote sensing capability into suitable actuation agents, collectively referred to as uncrewed systems (UxS). These systems include familiar platforms such as robotic quadrupeds, aerial multirotor vehicles, and ground vehicles, any of which may be teleoperated, act autonomously, or utilize a combination of both. A critical factor in achieving remote radiation sensing is the availability of data from the appropriate sensor. Many commercially available radiation sensors have closed-source documentation for their communication protocols. Typical end-user products are often self-contained, handheld devices designed for manual measurement scenarios. While there are commercial off-the-shelf (COTS) integrations of radiation sensors with UxS available for purchase, these solutions are typically tailored for specific use cases and may not meet the requirements of different applications. This paper discusses efforts to remotely acquire radiation measurements from a small form-factor CZT gamma spectrometer. Sandia National Laboratories has successfully demonstrated the initial capability to integrate low size, weight, and power (SWaP) gamma spectroscopy into various UxS, alongside co-located GPS data logging and sensor calibration and qualification. With remote gamma spectroscopy achieved, the stage is set for UxS integration of this capability.
In the UK, nuclear skills and qualifications have been the subject of several studies in higher levels of Government, including one by UK Cabinet Office. The newly formed UK Nuclear Skills Strategy Group (NSSG) has been tasked with addressing the current skills and people gaps for the UK nuclear industry. The NSSG is working with several key organisations in the UK: The National Skills Academy for Nuclear, (NSAN), which covers skills interventions, training and workforce development and experience, and the newly forming National College for Nuclear (NCfN) which will become a nuclear qualifications awarding body, regulating educational qualifications and curriculum in the nuclear sector. The NSSG has published its own UK Nuclear Skills Strategy in December 2016, aligned to the UK Cabinet Office study. It is the intent that, in the near future, nuclear qualifications and experience will be directly transferable across different organisations and nuclear sites in the UK, enabling the maintenance of high standards of behaviours and improved mobility throughout the sector. The UK needs to maintain and grow a diverse workforce and nuclear Suitably Qualified and Experienced Personnel (SQEP) people. The 2015 UK Nuclear Labour Market Intelligence (LMI) report, produced by Cogent for the Nuclear Energy Skills Alliance (NESA), concluded that the total nuclear workforce demand for the UK Nuclear Civil and Defence sector needed to grow by 33,000 people in the next five years, which includes replacement and expansion demand. This involved all disciplines and grades, with for example, higher level skilled workers needing to increase by 2200 FTEs every year. The main discipline pinch points were in, for example, Mechanical; Electrical; Project Management; Safety Case Specialists; Commissioning Engineers. Of the overall workforce, around 10% is over 60 years old and around 38% is over 50 years old. Jacobs in the UK is widely involved in many aspects of skills and competence development, with active involvement in local and national training programmes, and driving new methodologies to accelerate our staff's journey to nuclear competence. As well as looking at the immediate future, there is a risk in the UK that, over the 120-year plus clean-up mission, there won't be enough people with the right skills. Jacobs is developing sustainable solutions to this in our skills and development strategy for the nuclear sector. Jacobs has developed a variety of approaches to help us maintain and grow our diverse workforce and nuclear 'SQEP' people. This includes: - Identifying and developing our employees' technical and personal skills and capabilities, particularly the requirements of our millennials in both technical and social aspects of their working lives. - Identifying and planning for our future skills needs to meet our projected business requirements. - Our programme to transfer skilled employees from other sectors into the nuclear industry. - Our extensive school student placement scheme, aligned to the UK's 'Industrial Cadet' scheme, which has helped to bring school-aged students into Jacobs on workplace experience. - Our collaboration with our customers to develop UK national level Trailblazer Apprentice programmes. - Our involvement at a UK national level in the work of the National Skills Academy for Nuclear (NSAN), the Nuclear Energy Skills Alliance (NESA) High Level Skills Committee and the newly forming UK National College for Nuclear (NCfN). - Our work to ensure diversity is valued in our workplace, for example in our involvement in 'Women in Nuclear'. (authors)