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Presentation to NNL/INL staff on the status of the Halden BWR and implications of its shutdown for irradiation needs.
The degradation of photovoltaic (PV) balance of systems (BoS) components is not well-studied, but the consequences include: offline-modules, -strings, -inverters; system shutdown; arc-faults; and fires. A utility provider experienced a ~30% failure rate in their power transfer chain, attributed to branch connectors. Field-failed specimen assemblies were therefore examined, consisting of cable connector, branch connector and discrete fuse components. Unused field-vintage specimens are presently being examined using combined-accelerated stress testing (C-AST) to clarify the most influential environmental stressors as well as the effect of external mechanical perturbation. A benchtop prototype fixture was used to develop the perturbation capability for C-AST. The benchtop experiments were also used to develop the in-situ data acquisition of specimen: current, voltage, and temperature. A significant increase in operating temperature, ~100 deg C from ~40 deg C, and a different failure mode was observed immediately once periodic mechanical perturbation was applied. The current at failure was decreased from 35 A (with failure occurring in the fuses) to 15 A (failure at the male/female metal pin connection). After initial examination using X-ray computed tomography, the external plastic was machined away from failed specimens to allow failure analysis, including the extraction of the internal convolute springs for morphological examination (optical- and electron-microscopy). Chemical composition analysis included: energy-dispersive X-ray spectroscopy, differential scanning calorimetry, and Fourier transform infrared spectroscopy.
The degradation of photovoltaic (PV) balance of systems (BoS) components is not well-studied, but the consequences include: offline-modules, -strings, -inverters; system shutdown; arc-faults; and fires. A utility provider experienced a ~30% failure rate in their power transfer chain, originally attributed to branch connectors. Field-failed specimen assemblies were therefore examined, consisting of cable connector, branch connector, and discrete fuse components. Unused field-vintage specimens are presently being examined using combined-accelerated stress testing (C-AST) to clarify the most influential environmental stressors as well as the effect of external mechanical perturbation. A benchtop prototype fixture was used to develop the perturbation capability for the C-AST chamber. The benchtop experiments were also used to develop the in-situ data acquisition of specimen: current, voltage, and temperature. A significant increase in operating temperature, ~100 deg C from ~40 deg C, and a different failure mode was observed promptly once periodic mechanical perturbation was applied. The current at failure was decreased from 35 A (with failure in the fuses) to 15 A (failure at the male/female metal pin connection). After initial examination using X-ray computed tomography, the external plastic was machined away from failed specimens to allow failure analysis, including the extraction of the internal convolute springs for morphological examination (optical- and electron-microscopy).
The degradation of photovoltaic (PV) balance of systems (BoS) components is not well-studied, but the consequences include offline modules, strings, and inverters; system shutdown; arc faults; and fires. A utility provider experienced a ~30% failure rate in their power transfer chain, originally attributed to branch connectors. Field-failed specimen assemblies were therefore examined, consisting of cable connector, branch connector, and discrete fuse components. Previous papers and presentations have focused on the development of the test method using a benchtop prototype. This presentation covers the early results of C-AST aging of static and dynamic specimens.
This project is a collaborative research effort between PKMJ Technical Services LLC, Idaho National Laboratory, and Public Service Enterprise Group (PSEG) Nuclear, LLC. The collaboration, led by PKMJ Technical Services LLC, is part of the industry Funding Opportunity Announcement (FOA) award under Advanced Nuclear Technology Development FOA #DE-FOA-0001817. The pilot demonstration focuses on the Circulating Water System (CWS), an important non-safety-related system that impacts the power generation capability of the plant site. Achieving riskinformed condition-based Predictive Maintenance (PdM) on the CWS will result in significant economic benefits, and the developed methodologies can also be applied to other plant systems. This approach supports an industry goal of ensuring that nuclear power generation remains a viable, economically competitive option in the energy market. Operation and Maintenance (O&M) costs include labor-intensive Preventive Maintenance (PM) programs that involve manually performed inspection, calibration, testing, and maintenance of plant assets at periodic frequencies as well as time-based replacement of assets, irrespective of condition. This project offers an alternative by focusing on riskinformed condition-based maintenance to reduce O&M costs while still maintaining plant health and safety. This report summarizes the progress made toward achieving a risk-informed condition-based maintenance approach. The research and development (R&D) activities presented in this report are associated with development of a nuclear digital platform application, integration of fault signature models, and automated work management processes. The fault signatures and Machine Learning (ML) models are key components in predictive analytics and are heavily leveraged to improve the insights received by existing plant process data sources. Availability of the analysis results within a centralized digital platform enhances efficiency by enabling automation of activities otherwise performed manually. Personnel are presented with enhanced information that can be used to evaluate plant status and risks. Utilizing the enhancements to data analytics supports automated responses, (i.e. issuance of work orders) to address developing equipment faults and thus preventing forced, unplanned shutdowns of components or systems. The R&D activities described within this report lay the foundation for developing and demonstrating a digital automated platform to centralize the implementation of condition monitoring and response to equipment faults. The digital automated platform is cloud-based and designed to enable improved efficiency of plant processes. The digital platform includes content related to maintenance optimization, fault signature analysis, and plant records, which can all be used to support efficiencies when located within a centralized digital platform. These efficiencies could be further enhanced when deployed through industry-wide deployment of the technology to improve insights and processes based upon economies of scale.
The High Luminosity LHC (HL-LHC) is expected to deliver an integrated luminosity of 3000 - 4000 fb -1 in its lifetime with peak instantaneous luminosities reaching about 5 - 7.5×10 34 cm -2 s -1 . During the Long Shutdown 3 period, several components of the CMS detector will undergo major changes, called Phase-2 upgrades, to be able to operate in the challenging environment of the HL-LHC, including the replacement of the entire silicon tracker. The Phase-2 Outer Tracker (OT) will have high radiation tolerance, higher granularity, and the capability to handle higher data rates. Another key feature of the OT will be to provide tracking information to the Level-1 trigger, allowing trigger rates to be kept at a sustainable level without sacrificing physics potential. For this, the OT will be made of modules that have two closely spaced sensors read out by the same electronics, which can correlate hits in the two sensors creating short track segments called ”stubs”. The stubs will be used for tracking at 40 MHz. In this contribution, the design of the CMS Phase-2 OT, the technological choices, and highlights of research and development activities are reported.
The United States Department of Energy (DOE) has long term goals to develop solutions for managing the nation’s spent nuclear fuel (SNF) and high-level waste (HLW) inventory. The Integrated Waste Management (IWM) program under the DOE office of Nuclear Energy (DOE-NE) is employing system-level engineering and analysis principles to inform potential future waste management system architectures. Managing the spent nuclear waste requires the use of system-level analysis software that takes various aspects of the fuel cycle into account like waste generation, on-site/centralized storage, transportation infrastructure, and long-term disposal. The Next Generation System Analysis Model (NGSAM) is an agent-based model that was developed to simulate the transportation and storage of SNF and HLW. As an agent-based model, NGSAM has the capability to detail the interaction and movement of individual components and groups, such as rail cars and casks. The SNF inventory from commercial nuclear reactors is currently in temporary storage at multiple locations spread across the United States. Shipping of SNF from these locations relies on one of three transportation modes: rail, heavy-haul truck, or barge. Out of the three modes identified, rail is generally the most preferred due to the size of the canisters and casks the SNF would be shipped in. However, under some scenarios, a direct rail route might not be readily available to a reactor site or improving the rail infrastructure at shutdown sites might be too cost-prohibitive for utilities to opt for a direct rail transfer. Under such scenarios, using a barge or heavy haul truck to de-inventory the site and transfer the SNF to a nearby intermodal transfer site with adequate rail infrastructure where the payload could be transferred to a rail car might prove to be an attractive option. This work initially presents the various intermodal transportation options that could be used to transfer SNF from reactor sites to rail cars. This is followed by exploring the operational steps in each of these modes to move the SNF from a reactor site and transfer it to a rail car. This work also presents the procedure of implementing the intermodal transfer methodology in NGSAM using various Java methods. Finally, the process times for accomplishing each of the individual steps are furnished. The implementation ideology, assumptions, and future steps are presented in this work.
Ion Chamber and Photomultiplier Tube (PMT) are both used for beam loss monitor in the Machine Protection System (MPS) at Jefferson Lab. The requirements of signal processing of these detectors are different, so two VME-based signal processing boards, Beam Loss Monitor (BLM) board and Ion-Chamber board, were developed. The BLM board has fast response (< 1us) and 5 decades dynamic range from 10nA to 1 mA, while the Ion-Chamber board has 8 decades dynamic range from 100 pA to 10 mA and slower response. Both of boards provide functions of machine protection and beam diagnostics, and features of fast shutdown (FSD) interface, beam sync interface, built-in-self-test, remotely controlled bias signals, and on-board memory buffer.
The Transient REActor Test (TREAT) facility was restarted in 2017 to perform reactor fuels testing under transient conditions, following shutdown in 1994 and 23 years in non-operational standby. TREAT is a transient reactor that, like most pulsing reactors, can create natural shaped power pulses. However, TREAT provides the ability to shape the power pulse and control the energy deposition rate on the test specimens by automatically controlling the amount of reactivity insertion over time. The original automatic reactor control system (ARCS) was created in the 1980s using INTEL 8086 CPU boards and software written in assembly code. The need to maintain the system for the next 40+ years of TREAT life led to the decision to replace ARCS with modern hardware while maintaining the original functionality. After a prototyping effort, the replacement was successfully completed in the first quarter of 2020. The process and results of that effort are reported here.
CEBAF (Continuous Electron Beam Accelerator Facility) at Jefferson Lab was initially designed to deliver up to a 4 GeV electron beam simultaneously to three experimental halls. The maximum deliverable energy was raised to 6 GeV and CEBAF ran there until May of 2012 after which CEBAF entered an upgrade shutdown. After the upgrade, scientists plan to use CEBAF to deliver 11 GeV beam to the three original experimental halls and 12 GeV to new hall ? Hall D. Delivering beams to multiple halls at the new energy requires increased RF power to drive the separation cavities and adding a 750 MHz beam repetition rate to deliver beam to all halls simultaneously. Doubling the beam energy means doubling field strength in the cavities by quadrupling delivered power. 750 MHz beam repetition rate requires new 750 MHz capable cavities and a frequency change to one of the existing IOT amplifiers powering these cavities, and new 499 MHz Solid State Amplifiers to power four downstream cavities. Topics covered will include: verifying the Cavity?s power capabilities, modifications to the 750 MHz system and new cavities, and the decision process in selecting an amplifier type to power four downstream cavities.
The CEBAF Element Database (CED) was first developed in 2010 as a resource to support model-driven configuration of the Jefferson Lab Continuous Electron Beam Accelerator (CEBAF). Since that time, its uniquely flexible schema design, robust programming interface, and support for multiple concurrent versions has permitted it to evolve into a more broadly useful operational and control system tool. The CED played a critical role before and during the 2013 startup and commissioning of CEBAF following its 18-month long shutdown and upgrade. Information in the CED about hardware components and their relations to one-another facilitated a thorough Hot Checkout process involving more than 18,000 system checks. New software relies on the CED to generate EDM screens for operators on-demand thereby ensuring that the information on those screens is correct and up-to-date. The CED also continues to fulfill its original mission of supporting model-driven accelerator setup. Using the new ced2elegant and eDT (elegant Download Tool), accelerator physicists have proven able to compute and apply energy-dependant set points with greater efficiency than ever before.
This paper describes a new VME based machine protection Beam Loss Monitor (BLM) signal processing board designed at Jefferson Lab which features high sensitivity, high resolution, and low cost per channel. This double-wide VME board provides eight-channel signal processing with each channel having linear, logarithmic, and integrating amplifiers that simultaneously provide the optimal signal processing for both machine protection and beam diagnostics. Amplified signals are independently sampled by 16-bit, fast ADCs and then further processed through a Field Programmable Gate Array (FPGA). It also has features of fast shutdown interface (FSD), beam sync interface, built-in-self-test, remotely controlled bias signals, and on-board memory buffer. The initial experimental test shows that the BLM board has fast response (<< 1us) for integrating and wide dynamic range from 10 nA to 1 mA (>50 dB) for logarithmic signals. The board is suitable for signal processing from photomultiplier (PMT) and ion chamber detectors. Notice: Authored by Jefferson Science Associates, LLC under U.S. DOE Contract No. DE-AC05-06OR23177. The U.S. Government retains a non-exclusive, paid-up, irrevocable, world-wide license to publish or reproduce this manuscript for U.S. Government purposes.
For decades, the Halden Boiling Water Reactor (HBWR) in Norway was an international resource for assessing nuclear fuels and materials behavior and its unexpected shutdown in 2018 represented a significant loss in experimental capability for prototypical irradiation testing. In the aftermath of the closure, a study was performed to assess capability gaps related to the Accident Tolerant Fuels (ATF) program. It was concluded that the primary capability gaps left by the closure of the HBWR were the loss of prototypic in-pile Light Water Reactor (LWR) loops that provide operational transient testing and in-pile Loss-of-Coolant Accident (LOCA) testing capabilities. The study also concluded that the Advanced Test Reactor (ATR) and Transient Reactor Test Facility (TREAT) likely have the necessary key capabilities to absorb the breadth of the HBWR mission gaps related to the ATF program, but additional investments in experimental infrastructure were needed. One such investment is the design and installation of additional pressurized water loops in the medium-I positions of the ATR, hereafter referred to as I-Loops, to support power ramp testing and testing in Boiling Water Reactor (BWR) conditions. This paper gives an overview of the planned ATR I-Loops and assesses the feasibility of performing power ramp testing inside such a loop.
Thorium is a fertile material and most of the past research has focused on breeding thorium to fissile material. In this paper, the focus is on using thorium to improve the thermal margins by homogeneously distributing thorium in the fuel pellets. A proposed uranium-thorium-based fuel assembly is simulated for the Swedish Ringhals-3 PWR core in a realistic demonstration. All the key safety parameters, such as isothermal temperature coefficient of reactivity, Doppler temperature of reactivity, boron worth, shutdown margins and fraction of delayed neutrons are studied in this paper, and are within safety limits for the new core design using the uranium-thorium-based fuel assemblies. The calculations were performed by the two-dimensional transport code CASMO-4E and the two group steady-state three dimensional nodal code SIMULATE-3 from Studsvik Scandpower. The results showed that the uranium-thorium-based fuel assembly improves the thermal margins, both in the pin peak power and the local power (Fq). The improved thermal margins would allow more flexible core designs with less neutron leakage or could be used in power uprates to offer efficient safety margins. (authors)
This paper was focused on core design, neutronics evaluation and fuel cycle analysis for Thorium-Uranium Breeding Recycle in current PWRs, without any major change to the fuel lattice and the core internals, but substituting the UOX pellet with Thorium-based pellet. The fuel cycle analysis indicates that Thorium-Uranium Breeding Recycle is technically feasible in current PWRs. A 4-loop, 193-assembly PWR core utilizing 17 x 17 fuel assemblies (FAs) was taken as the model core. Two mixed cores were investigated respectively loaded with mixed reactor grade Plutonium-Thorium (PuThOX) FAs and mixed reactor grade {sup 233}U-Thorium (U{sub 3}ThOX) FAs on the basis of reference full Uranium oxide (UOX) equilibrium-cycle core. The UOX/PuThOX mixed core consists of 121 UOX FAs and 72 PuThOX FAs. The reactor grade {sup 233}U extracted from burnt PuThOX fuel was used to fabrication of U{sub 3}ThOX for starting Thorium-. Uranium breeding recycle. In UOX/U{sub 3}ThOX mixed core, the well designed U{sub 3}ThOX FAs with 1.94 w/o fissile uranium (mainly {sup 233}U) were located on the periphery of core as a blanket region. U{sub 3}ThOX FAs remained in-core for 6 cycles with the discharged burnup achieving 28 GWD/tHM. Compared with initially loading, the fissile material inventory in U{sub 3}ThOX fuel has increased by 7% via 1-year cooling after discharge. 157 UOX fuel assemblies were located in the inner of UOX/U{sub 3}ThOX mixed core refueling with 64 FAs at each cycle. The designed UOX/PuThOX and UOX/U{sub 3}ThOX mixed core satisfied related nuclear design criteria. The full core performance analyses have shown that mixed core with PuThOX loading has similar impacts as MOX on several neutronic characteristic parameters, such as reduced differential boron worth, higher critical boron concentration, more negative moderator temperature coefficient, reduced control rod worth, reduced shutdown margin, etc.; while mixed core with U{sub 3}ThOX loading on the periphery of core has no visible impacts on neutronic characteristics compared with reference full UOX core. The fuel cycle analysis has shown that {sup 233}U mono-recycling with U{sub 3}ThOX fuel could save 13% of natural uranium resource compared with UOX once through fuel cycle, slightly more than that of Plutonium single-recycling with MOX fuel. If {sup 233}U multi-recycling with U{sub 3}ThOX fuel is implemented, more natural uranium resource would be saved. (authors)
Thorium is a fertile material and most of the past research has focused on breeding thorium to fissile material. In this paper, the focus is on using thorium to improve the thermal margins by homogeneously distributing thorium in the fuel pellets. A proposed uranium-thorium-based fuel assembly is simulated for the Swedish Ringhals-3 PWR core in a realistic demonstration. All the key safety parameters, such as isothermal temperature coefficient of reactivity, Doppler temperature of reactivity, boron worth, shutdown margins and fraction of delayed neutrons are studied in this paper, and are within safety limits for the new core design using the uranium-thorium-based fuel assemblies. The calculations were performed by the two-dimensional transport code CASMO-4E and the two group steady-state three dimensional nodal code SIMULATE-3 from Studsvik Scandpower. The results showed that the uranium-thorium-based fuel assembly improves the thermal margins, both in the pin peak power and the local power (Fq). The improved thermal margins would allow more flexible core designs with less neutron leakage or could be used in power uprates to offer efficient safety margins. (authors)
The French sodium cooled fast reactor Phenix was shut down in 2009 after 35 years of operation. Before decommissioning, a final set of tests were performed by the CEA during 9 months. Several topics were involved such as thermal hydraulics, core physics and fuel behaviour. Among these ultimate experiments, two thermal hydraulic tests were performed: an asymmetrical test consisting in a trip of one secondary pump and a natural convection test in the primary circuit. Recognizing the unique opportunity offered by these Phenix ultimate tests, IAEA decided in 2007 to launch a Coordinated Research Project (CRP) devoted to benchmarking analyses with system codes on the Phenix natural convection test. One objective of the natural convection test in Phenix reactor is the assessment of the CATHARE system code for safety studies on future and advanced sodium cooled fast reactors. The aim of this paper is to describe this test, which was performed on June 22-23, 2009, and the associated benchmark specifications for the CRP work. The paper reminds briefly the Phenix reactor with the main physical parameters and the instrumentation used during the natural convection test. After that, the test scenario is described: - initial state at a power of 120 MWth, - test beginning resulting from a manual dry out of the two steam generators, - manual scram, - manual trip on the three primary pumps without back-up by pony motors, - setting and development of natural convection in the primary circuit, in a first phase without significant heat sink in the secondary circuits and in a second phase with significant heat sink in the secondary circuits, by opening the casing of steam generators to create an efficient heat sink, by air natural circulation in the steam generators casing. The benchmark case ends after this second phase, which corresponds to the experimental test duration of nearly 7 hours. The paper presents also the benchmark specifications data supplied by the CEA to all participants of this CRP in order to perform calculations (core, primary circuit, primary pumps, IHX, shutdown system, operating parameters, test scenario and real test conditions). Finally, main test results and analyses are presented including the evolution of the core and of the heat exchangers inlet and outlet temperatures, and some local temperature measurements. The natural convection has been easily set up in the pool type reactor Phenix with different boundary conditions at the secondary side, with or without heat sink. The data obtained during this unique test represent some very useful and precious results for the development of SFR in a wide range of thematic such as numerical methods dedicated to thermal-hydraulics safety analyses (system codes, CFD codes and coupling system and CFD codes) and instrumentation. (authors)
The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor (iPWR), in which all of the components typically associated with the nuclear steam supply system (NSSS) of a nuclear power plant are incorporated within a single reactor pressure vessel. This paper is the third in a series of four papers, which describe the design and functionality of the Westinghouse SMR. It focuses in particular upon the plant building layout and modular design of the Westinghouse SMR. In the development of small modular reactors, the building layout is an area where the safety of the plant can be improved by applying new design approaches. This paper will present an overview of the Westinghouse SMR building layout and indicate how the design features improve the safety and robustness of the plant. The Westinghouse SMR is designed with no shared systems between individual reactor units. The main buildings inside the security fence are the nuclear island, the rad-waste building, the annex building, and the turbine building. All safety related equipment is located in the nuclear island, which is a seismic class 1 building. To further enhance the safety and robustness of the design, the reactor, containment, and most of the safety related equipment are located below grade on the nuclear island. This reduces the possibility of severe damage from external threats or natural disasters. Two safety related ultimate heat sink (UHS) water tanks that are used for decay heat removal are located above grade, but are redundant and physically separated as far as possible for improved safety. The reactor and containment vessel are located below grade in the center of the nuclear island. The rad-waste and other radioactive systems are located on the bottom floors to limit the radiation exposure to personnel. The Westinghouse SMR safety trains are completely separated into four unconnected quadrants of the building, with access between quadrants only allowed above grade. This is an improvement to conventional reactor design since it prevents failures of multiple trains during floods or fires and other external events. The main control room is located below grade, with a remote shutdown room in a different quadrant. All defense in depth systems are placed on the nuclear island, primarily above grade, while the safety systems are located on lower floors. The economics of the Westinghouse SMR challenges the established approach of large Light Water Reactors (LWR) that utilized the economies of scale to reach economic competitiveness. To serve the market expectation of smaller capital investment and cost competitive energy, a modular design approach is implemented within the Westinghouse SMR. The Westinghouse SMR building layout integrates the three basic design constraints of modularization; transportation, handling and module-joining technology. (authors)