Development of an In-Situ Method for Cable Condition Monitoring in Nuclear Power Plants
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21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
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Optimization of Facility Maintenance Program through Application of Streamlined Reliability - Centered Maintenance (RCM)
Abstract not provided
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
IMPLEMENTING PRIMARY CREEP CALCULATIONS DURING STRESS CHANGES AND REVERSALS IN THE FUEL PERFORMANCE CODE FRAPCON-3
Recently, the creep models in FRAPCON-3.4 were used to model the creep tests performed in IFA-699 up through cycle 4. IFA-699 is an instrumented test rig in the Halden Boiling Water Reactor that is being used to perform in-reactor creep experiments under changing stress conditions. The cladding creep tests in IFA-699 were on four fresh fuel segments with Zircaloy-4, E110, M-MDA, and M5™ cladding. These fuel segments were loaded in a test rig and placed in a PWR loop and operated under PWR conditions. Each of the test fuel rods were connected to gas lines in order to apply tensile and compressive stresses to the cladding. Diameters of the cladding were measured by diameter gages at several intervals to measure the creep behavior of the cladding as a function of time. The creep tests were performed at a temperature of around 360°C and fast neutron flux of 1.8 to 2.4×1017 n/m²-s (depends on specimen location). During the four cycles of operation, hoop stresses of -75, -50, +30, and +110 MPa were applied to the specimens. In addition to providing assessment of the FRAPCON-3.4 creep modeling, the creep model comparisons were also used to provide insight into the nature of primary creep following stress changes, whether creep behavior is symmetric under tensile and compressive loading, and the effect of fast fluence on creep rate. It was determined that creep strain should be calculated as a function of effective stress rather than hoop stress. This accounts for the observed difference between creep under compressive and tensile stress for the same hoop stress. It was also determined that the primary creep changes in proportion to the change in stress following stress change or stress reversal. Finally, it was observed that the irradiation creep rate decreases somewhat with increasing fluence. The creep model in FRAPCON-3.4 has already been developed to be a function of effective stress and includes a term that decreases the irradiation creep rate with increasing fluence. A new mathematical formulation has been developed that is suitable for a fuel performance code to calculate the primary creep strain under conditions of changing stress and stress reversal. The creep model in FRAPCON-3.4 and the formulation for calculating primary creep under changing stress conditions will be described in this paper. Comparisons to data will also be provided.
Geelhood, Kenneth J.↗
Fuel Refabrication Prototype - Presentation for 2021 ASI Annual Webinar
This power point presentation is a summary of the work produced in FY2021 under the LWR Fuel Reinstrumentation project funded by DOE's Advanced Sensor Initiative, and the outlook for FY22. The objective of this portion of the ASI program is to "Capture critical technology created by the Institute For Energy (formerly Halden Reactor Project) to reinstrument irradiated fuel rodlets, and further this technology to enable incorporation of advanced instrumentation: fiberoptics, LVDTs, ultrasonic based sensors". The presentation explains that two of the three equipment modules were received and INL and a number of practice runs were conducted with them. In FY22 the third equipment module will arrive and using the three equipment modules measurement sciences personnel will create a complete prototypical instrumented fuel rodlet (using ceramic surrogates in the place of UO2).
46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
Characterization of cartridge filters from the IEA-R1 Nuclear Reactor
The management of radioactive waste ensures safety to human health and the environment nowadays and for the future, without overwhelming the upcoming generations. The primary characterization of radioactive waste is one of the main steps in the management of radioactive waste. This step permits to choose the best treatment for the radioactive waste before forwarding it to its final disposal. The aim of the present work is the primary characterization of cartridge filters from the IEA-R1 nuclear reactor utilizing gamma-ray spectrometry, and the method of Monte Carlo for calibration. The IEA-R1 is located in the Nuclear and Energy Research Institute (IPEN - CNEN) in the city of Sao Paulo, Brazil. Cartridge filters are used for purification of the cooling water that is pumped through the core of the pool type nuclear research reactors. Once worn out, these filters are replaced and then become radioactive waste. Determination of the radioactive inventory is of paramount importance in the management of such radioactive waste, and one of the main methods for doing so is the gamma-ray spectrometry, which can identify and quantify high energy photon emitters. The technique chosen for the characterization of radioactive waste in the present work is the gamma-ray spectrometry with High purity Germanium (HPGe) detectors. From the energy identified in the experimental spectrum, three radioisotopes were identified in the cartridge filter: {sup 108m}Ag, {sup 110m}Ag, {sup 60}Co. For the estimated activity of the filter, the calibration in efficiency was made utilizing the MCNP4C code of the Monte Carlo method. Such method was chosen because there is no standard source available in the same geometry of the cartridge filter, therefore a simulation had to be developed in order to reach a calibration equation, necessary to estimate the activity of the radioactive waste. The results presented an activity value in the order of MBq for all radioisotopes. (authors)
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
A human automation interaction concept for a small modular reactor control room - 101
Many advanced nuclear power plant (NPP) designs incorporate higher degrees of automation than the existing fleet of NPPs. Automation is being introduced or proposed in NPPs through a wide variety of systems and technologies, such as advanced displays, computer-based procedures, advanced alarm systems, and computerized operator support systems. Research consistently finds that there is a fundamental tradeoff between system performance with increased automation and reduced human performance. Idaho National Laboratory (INL) researchers and NuScale Power human factors and operations staff are working on a collaborative project to address the human performance challenges of increased automation and to determine the principles that lead to optimal performance in highly automated systems. This paper will describe this concept in detail and will describe an experimental test of the concept. The benefits and challenges of the approach will be discussed. (authors)
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Model-free adaptive control for MHTGR-based nuclear steam supply systems - 144
Modular high temperature gas-cooled reactor (MHTGR) uses helium as coolant and graphite as both moderator and structural materials. Due to its inherent safety feature, MHTGR has been regarded as an important candidate for next generation of nuclear plants. By coupling with the once-through steam generator (OTSG), MHTGR can be utilized to form the nuclear steam supply system (NSSS) for producing superheated steam, which can be adopted for nuclear plants with high efficiency. Power-level control of is important for the stable and efficient operation of the MHTGR-based NSSS, and it is necessary to develop the power-level control law that is robust to physical and thermal parameters and can guarantee the closed-loop stability in normal and even abnormal conditions. In this paper, a model-free adaptive power-level control law is developed for the MHTGR-based NSSS, which is free from physical and thermal-hydraulic parameters, takes the simple forms of proportional-integral (PI) or proportional-differential (PD) feedback law, and can provide globally asymptotic closed-loop stability. Then, it is applied to the NSSS control of the two modular high temperature gas-cooled nuclear plant HTR-PM. Numerical simulation results in normal power maneuver and some typical abnormal cases show that this new control law can provide satisfactory closed-loop stability as well as transient performance. (authors)
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
The overview in safety review of human factors engineering and control room design in chinese AP1000 nuclear power plant - 174
The first AP1000 nuclear power plant is constructed in Sanmen county in Zhejiang Province in China, which has creative and distinctive design characteristics. Human factors engineering disciplines are applied to the design of the AP1000. In addition to the elements of the program review model(including human factors engineering program management, operating experience review, functional requirements analysis and allocation, task analysis, staffing, integration of human reliability analysis with human factors engineering, human system interface design, procedure development, training program development, human system interface verification and validation, design implementation, human performance monitoring),the minimum inventory of controls, displays, and alarms present in the main control room and at the remote shutdown workstation. These contents mentioned above are reviewed according to standards, rules and regulations. This article introduces review process, and several important issues during the reviewing are described in detail. These issues include many elements, such as operating experience review, task analysis, human system interface design, verification and validation,and so on. This paper emphasizes on the main control room design(including environment and layout). Background noise in the actual main control room may exceed too much to design value. Another important issue is about main control room habitability systems(VES) changes to satisfy post-actuation performance requirements in AP1000 design change proposal. All the wall panel displays will be closed in this design change proposal. Review process and proposal are described in the paper. Since the AP1000 unit in China is the first constructed all over the world, verification and validation are especially important and necessary. The plan of verification and validation, the results summary report of verification and validation, the report of human engineering discrepancies are focus attention. Verification and validation in human factors engineering include HSI task support verification, HFE design verification and integrated system validation. This paper introduces the review process of verification and validation, the issues found in verification and validation. (authors)
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Implementing a control application on an FPGA Platform - 356
Today, many I and C system suppliers propose FPGA-based I and C platforms. In that context, EDF has set up a project to assess the use of such platforms for safety-related systems and to enhance EDF's ability to manage this technology for the upgrade of existing I and C systems or for new constructions. One aim of the project is to assess the adequacy of these platforms for the development of control functions. This paper presents the development of a case study on a Radiy FPGA-based platform in a Hardware-In-the-Loop configuration, and the comparison with a Dymola Model-In-the-Loop and a Siemens T2000 S7 Hardware-In-the-Loop approach. The case study is a part of the level control of steam generators of French PWR NPP. The control flow diagram uses complex analog elementary blocks coded with fixed point format for the analog values. The use of IP cores of the FPGA supplier was avoided to enable a complete mastership of the VHDL code, to provide evidence of compliance with safety-related requirements, and to allow porting this code to another FPGA platform. The case study has highlighted the specificity of using a FPGA, and has shown that a complex control application in a FPGA-based PLC presents no major differences in precision and response time with a processor-based PLC. (authors)
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Introduction of Diverse Manual Actuation Design for APR1400
APR1400 is an Advanced Light Water Reactor (ALWR) design with enhanced plant safety and advanced design features. One unit is in commercial operation and 11 are currently under construction, 4 abroad and 7 in Korea. APR1400 adopts full digitalized Man-Machine Interface System (MMIS) to perform the function of control, indication and alarming for safety and non-safety systems APR1400 MMIS is designed to operate the plant from the Main Control Room (MCR) under all anticipated operating conditions including design basis accidents. This advanced MMIS is designed to meet the latest codes, standards and regulatory requirements including the defense-in-depth and diversity requirements, which is intended to mitigate the consequence of a total failure of the digital safety system caused by a software Common Cause Failure (CCF) coupled with anyone of the Design Basis Events. The redundancy of the safety system could be defeated during this CCF condition and the protective function could be rendered inoperable.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Sensitivity Analyses of Natural Convection in the AP1000 Passive Containment Cooling System Following LBLOCA Using CFD
The AP1000 power plant has implemented multiple layers of passive safety systems to enhance reactor safety and ensure system integrity during postulated scenarios, such as loss of coolant accident (LOCA) and main steam line break (MSLB). Among these, the passive containment cooling system (PCCS) is a safety-related system designed to prevent the containment from exceeding the design limits of both pressure and temperature following a postulated design basis accident, by transferring heat from the steel containment vessel to the atmosphere. During LOCA, natural air convection plays a vital role in removing heat from the steel containment vessel to the atmosphere. This is integrated with the condensate film from the Passive Containment Cooling Water Storage Tank (PCCWST). Air natural convection is also the only heat removal mechanism after the dry-out of the PCCWST and during the loss of shutdown decay heat removal (LOSDHR) event. Therefore, it is essential to investigate the thermal hydraulics behavior of the containment under transient conditions to ensure its safety and integrity. This paper presents a simplified CFD/ANSYS FLUENT model developed to analyze the impact of different parameters, such as air relative humidity, air temperature, and steel containment temperature, on the natural convection capability to remove the decay heat following LOCA. The model aims to examine the thermal behavior of the containment and provides a comprehensive understanding of the system's natural convection capability during postulated accidents. The results obtained from the model can be used to improve the safety and integrity of the containment system and provide valuable insights for future research in the field.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
FIDES-II/P2M Simulation Exercise on AN3 and AN10 bump tests: main results
This paper presents simulations of bump tests conducted as part of the P2M (Power to Melt and Maneuverability) project within the OECD/NEA FIDES-II framework (Framework for Irradiation Experiments). To prepare fuel performance codes for analyzing the P2M experiments, an international Simulation Exercise (SE) was initiated under the P2M project. The objective of this exercise was to model the AN3 and AN10 bump tests carried out in the 1980s at the DR3 reactor (3rd RISØ Fission Gas Project), during which instrumented fuel rodlets (equipped with thermocouples and pressure sensors) were subjected to prolonged power transients and power dips. The ongoing simulation exercise involves seven organizations from industry, research, and regulatory bodies, each using different fuel performance codes. This paper provides detailed descriptions of the test cases and simulation results from six fuel performance codes (ALCYONE, BISON, FAST, FEMAXI, FRAPCON, and TRANSURANUS), with a particular focus on fission gas release kinetics.
11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Nuclear heating measurements by in-pile calorimetry: prospective works for a microsensor design
Since 2009 works have been performed in the framework of joint research programs between CEA and Aix-Marseille University. The main aim of these programs is to design and develop in-pile instrumentations, advanced calibration procedure and accurate measurement methods in particular for the new Material Testing Reactor (MTR) under construction in the South of France: Jules Horowitz Reactor (JHR). One major sensor is a specific radiometric calorimeter, which was studied out-of-pile from a thermal point of view and in-pile during irradiation campaigns. This sensor type is dedicated to measurements of nuclear heating (energy deposition rate per mass unit induced by interactions between nuclear rays and matter) inside experimental channels of MTRs. This kind of in-pile calorimeter corresponds to heat flux calorimeter exchanging with the external cooling fluid. This thermal running mode allows the establishment of steady thermal conditions inside the sensor to carry out online continuous measurements inside the reactor (core or reflector). Two main types of calorimeters exist. The first type consists of a single cell calorimeter. It is divided into a sample of material to be tested and a jacket instrumented with two thermocouples or a single thermocouple (Gamma Thermometer). The second, called a differential calorimeter, is composed of two superposed twin cells (a measurement cell containing a sample of material, and a reference cell to remove the heating of the cell body) instrumented with four thermocouples and two electrical heaters. Contrary to a single-cell calorimeter, a differential calorimeter allows the compensation of the parasite nuclear heating of the sensor body or jacket. Moreover, it possesses interesting advantages: thanks to the heaters embedded in the cells, three different measurement methods can be applied during irradiations to quantify nuclear heating. The first one is based on the use of out-of-pile calibration curves obtained by generating a heat source by the Joule Effect inside each calorimetric cell. The second one is a zero method consisting in cancelling the difference in cell responses with an additional energy into the reference cell. The last measurement method is based on current additions in the two calorimetric cells. However, one drawback of the existing differential calorimeter is the size of the sensor: a great length equal to 220 mm and a diameter equal to 18 mm. This current size leads to measurement limitations. This paper will begin with a presentation of these measurement limitations from a bibliographic state. Each limitation will be detailed and in particular in the case of a high nuclear heating level expected, for instance, inside the JHR's core at its highest nominal power. The second part of the paper will develop the scientific skills of each partner in heat sciences, micro technology and nuclear physics necessary to design a new calorimetric micro-system: the advantages of studied microelements such as micro-thermocouples, micro- fluxmeters and micro-heaters will be presented. The last part will discuss preliminary designs. (authors)
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Technical Challenges and Gaps in Integration of Advanced Sensors, Instrumentation, and Communication Technologies with Digital Twins for Nuclear Application
This paper explores integrating advanced sensor, instrumentation, and communication technologies with digital twin technologies for nuclear energy application. Digital twins and digital-twin-enabling technologies are expected to integrate with future nuclear reactor designs and have the potential to impact currently operating nuclear power plants. Greater digital integration, advanced instrumentation and control systems, and advanced operations and maintenance practices are all associated with digital-twin-enabling technologies. Advanced sensors, instrumentation, and communication technology are expected to comprise important elements of the infrastructure required to develop and operate a nuclear digital twin system. This paper identifies and discusses challenges and gaps in developing and implementing advanced sensors and instrumentation and communication technology to be integrated with a digital twin in current and advanced reactor applications. It is important to address some challenge and gap to enable a successful near-term deploying advanced sensors, instrumentation, and communication technologies integrated with digital twins.
Yadav, Vaibhav↗
Laboratory Experiments and Modeling for Interpreting Field Studies of Secondary Organic Aerosol Formation Using an Oxidation Flow Reactor
This grant was originally funded for deployment of a suite of aerosol instrumentation by our group in collaboration with other research groups and DOE/ARM to the Ganges Valley in India (GVAX) to study aerosols sources and processing. Much of the first year of this grant was focused on preparations for GVAX. That campaign was cancelled due to political reasons and with the consultation with our program manager, the research of this grant was refocused to study the applications of oxidation flow reactors (OFRs) for investigating secondary organic aerosol (SOA) formation and organic aerosol (OA) processing in the field and laboratory through a series of laboratory and modeling studies. We developed a gas-phase photochemical model of an OFR which was used to 1) explore the sensitivities of key output variables (e.g., OH exposure, O 3 , HO 2 /OH) to controlling factors (e.g., water vapor, external reactivity, UV irradiation), 2) develop simplified OH exposure estimation equations, 3) investigate under what conditions non-OH chemistry may be important, and 4) help guide design of future experiments to avoid conditions with undesired chemistry for a wide range of conditions applicable to the ambient, laboratory, and source studies. Uncertainties in the model were quantified and modeled OH exposure was compared to tracer decay measurements of OH exposure in the lab and field. Laboratory studies using OFRs were conducted to explore aerosol yields and composition from anthropogenic and biogenic VOC as well as crude oil evaporates. Various aspects of the modeling and laboratory results and tools were applied to interpretation of ambient and source measurements using OFR. Additionally, novel measurement methods were used to study gas/particle partitioning. The research conducted was highly successful and details of the key results are summarized in this report through narrative text, figures, and a complete list of publications acknowledging this grant.
54 ENVIRONMENTAL SCIENCES↗
An automatic multi-precursor flow-type atomic layer deposition system
Designs for two automated atomic layer deposition (ALD) flow reactors are presented, and their capabilities for coating additively manufactured (AM) metal prints are described. One instrument allows the coating of several AM parts in batches, while the other is useful for single part experiments. To demonstrate reactor capabilities, alumina (Al 2 O 3 ) was deposited onto AM 316L stainless steel by dosing with water (H 2 O) vapor and trimethylaluminum (TMA) and purging with nitrogen gas (N 2 ). Both instruments are controlled by custom-programmed LabVIEW software that enables in situ logging of temperature, total pressure, and film thickness using a quartz crystal microbalance. An initial result shows that 150 ALD cycles led to a film thickness of ~55 nm, which was verified with Rutherford backscattering spectroscopy. This indicates that the reactors were indeed depositing single atomic layers of Al 2 O 3 per ALD cycle, as intended.
47 OTHER INSTRUMENTATION↗
Ultrasonic imaging in liquid sodium: topological energy for damages detection
Under-sodium imaging at high temperature is an important requirement in sodium cooled fast reactors during structural inspection. It aims at checking the integrity of immersed structures and assessing component degradation. The work presented in this paper focuses on designing an advanced ultrasound methodology for detecting damages such as deep crack defects. For that purpose, a topological imaging approach was implemented. This method takes advantage of all available prior knowledge about the environment through the integration of differential imaging and time reversal techniques. The quality of the topological energy distribution (the image) is further enhanced by applying a time gating related to each reconstructed pixel. Numerical and experimental results are presented using this method in order to confirm its reliability. These images are compared to a B-scan to emphasize the localization performances of this method. (authors)
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
An intelligent scanning algorithm for lost fuel rod recovery in the MYRRHA lower plenum
This paper presents a system for visualizing the interior structure of a vessel filled with a liquid metal. Due to the opaque and conductive nature of the medium, regular visualization systems based on cameras, electromagnetic or laser based techniques can't be used. The operation, strengths and weaknesses of an existing set-up are investigated by creating a scenario to find and extract a rectangular object inside the vessel in the shortest possible time. The main focus of this paper is not on designing a new system but to investigate the bottlenecks that are faced and need to be kept in mind in future designs. (authors)
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗