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The prototype of an industrial reactor antineutrino detector iDREAM is constructed to demonstrate the capability of real-time remote monitoring of PWR operation by the neutrino method.
The Spectroscopic Planar Germanium (SPG) has been demonstrated at the IAEA Physical Inventory Verification (PIV) in South Korea, as well as at the Joint Fuel Cycle Study at Idaho National Laboratory. These field tests were successful and provided valuable feedback from KINAC, IAEA, and national laboratory staff. This feedback was used to direct efforts to improve the instrument this year.
In this study, an emissivity measurement system for a liquid molten salt was developed on the bases on the Stefan Boltzmann Law. A prototype device operating in ambient atmosphere was constructed an initial result compared well with available literature. The diffuse emissivity of NaCl-KCl system (NaCl, 51mol% NaCl – 49mol% KCl and KCl) was measured from 550 to 850 °C at a wavelength of 1.3 µm. The system appears to be relatively invariant with composition but a clear trend of decreasing emissivity values from ~0.9 at 550 °C to 0.75 at 850 °C was observed. A schematic for an enclosed system for inert atmosphere and actinide salts was drafted. Several of the components for fabrication of this device, improved laser system, furnace, nickel for capsule construction, have been purchased and are ready for assembly. The new laser system will have an improved measurement range of 275-1000 °C. The clam shell furnace has a maximum temperature of 1200 °C and will support the hermetically sealed inert encapsulation vessels. Nickel plate and tubing will create a relatively corrosion resistance capsule for testing that can withstand temperature up to 1000 °C. Emissivity measurements are important for a larger effort to understand radiative heat transfer within molten salts at elevated temperatures (600-1000 °C). Literature review of optical salt properties and radiative heat transfer models indicated that systems within elevated temperature regimes could see upwards of 20% or more of their total heat transferred through photonic mechanisms. Ultimately, future work will measure the total thermal conductivity and break down the molecular and radiative portions. Thermal conductivity, along with other thermal properties of molten salts, both coolant and fuel, are fundamental to modeling and simulations that support MSR design. The results presented here and ongoing work at PNNL will be useful in selecting optimal salt compositions, designing reactor systems, and optimizing heat transfer within a MSR power plant.
Nuclear Safety is a journal that covers significant issues in the field of nuclear safety. Its primary scope is safety in the design, construction, operation, and decommissioning of nuclear power reactors worldwide and the research and analysis activities that promote this goal, but it also encompasses the safety aspects of the entire nuclear fuel cycle, including fuel fabrication, spent-fuel processing and handling, and nuclear waste disposal, the handling of fissionable materials and radioisotopes, and the environmental effects of all these activities. Table of Contents for this issue follows. GENERAL SAFETY CONSIDERATIONS 181 Attitudes and Practices Regarding Disposal of Liquid Nuclear Waste at Clinton Laboratories in the Very Early Years: A Historical Analysis, S. H. Stow; 202 Off-Site Nuclear Emergency Management in Germany Under the Auspices of the Federal Structure, A. Bayer, S. Bittner, and H. Korn; DESIGN FEATURES: 211 Twenty-Fourth DOE/NRC Nuclear Air-Cleaning and Treatment Conference, R. R. Bellamy, J. J. Hayes, R. D. Porco, and M. W. First; OPERATING EXPERIENCES: 222 Explosion in the Tomsk-7 Reprocessing Plant on April 6, 1993, M. L. Hyder, W. G. Lussie, and F. E. Witmer; 234 Reactor Shutdown Experience, Compiled by J. W. Cletcher; U.S. NUCLEAR REGULATORY COMMISSION INFORMATION AND ANALYSES: 237 Assessment of Spent Fuel Cooling, J. G. Ibarra, W. R. Jones, G. F. Lanik, H. L. Ornstein, and S. V. Pullani; RECENT DEVELOPMENTS: 256 Reports, Standards, and Safety Guides, D. S. Queener; 260 Proposed Rule Changes as of June 30, 1996; ANNOUNCEMENTS: 210 Tennessee Industries Week; 221 21st International Symposium on The Scientific Basis for Nuclear Waste Management; 259 IAEA Programme of Scientific Meetings for 1997; 270 PSAM 4 (International Conference on Probabilistic Safety Assessment and Management); 265 The Authors; 268 Errata.
The description of human operator dynamic characteristics in mathematical terms compatible with control engineering practice is an essential prerequisite to the analytical treatment of manual reactor control systems. Safe reactor operation requires effective operator control through interaction with plant dynamics, manipulators and displays. Traditional static analysis methods consider only specific situations; they fail to adequately explain the mutual interactions between the operator and the reactor plant characteristics. In this paper we investigate the theory for describing operator-reactor characteristics based on the methods of conventional control engineering techniques. The primary purpose of the experiments reported is the validation of the quasi-linear operator model. (authors)
What should we ask the control room crew about when evaluating situation awareness? Currently available freeze probe techniques concentrate on trends of process parameters. Analysis of control room work shows that there are far more involved in control room crews' Situation Awareness (SA). The Process Awareness and Situation Understanding (PASU) measure has been developed from an analysis of control room work and includes probing the crew about the status of safety functions and main plant systems such as coolant supply, ordinary and emergency power supply, main reactor parameters, and main plant parameters. Results from a nuclear control room evaluation showed that the crew's overview of safety functions developed differently across scenario phases compared to the crew's overview of main process parameters. SA in the nuclear control room seems to be a multifaceted phenomenon and valid measurement needs to cover its aspects. Initial empirical testing showed that PASU was sensitive to scenario phases and related to other performance indicators. PASU, or measures based on similar principles, can be a valid and useful probe based technique for use in the evaluation of control room design and control room work. The measure will fill a gap in the research literature and will serve as an example of SA measurement with utility for control room evaluation. (authors)
The control systems of nuclear power plants contain many components which are becoming obsolete at an increasing rate. Nowadays maintenance for analog-based protection systems may be difficult as analog parts become obsolete or hard to obtain. Various studies have been conducted to address this control system hardware obsolescence issue. Those control systems of nuclear power plants have been replaced with modern digital control systems such as the programmable logic controller (PLC) and the distributed control system (DCS). The central processing unit (CPU)- based platforms such as the PLC and the DCS require an operating system (OS) and application software, and thus they might result in a common-cause failure when a problem occurs in the operating system or application software. The field programmable gate array (FPGA) is highlighted as an alternative to the conventional protection or control systems. This paper presents the development of an FPGA-based logic controller (FLC) and a diverse protection system (DPS) with the FLC platform. Protection functions for the DPS were implemented in FPGAs without any CPU or OS. (authors)
The demand for robust and resilient performance has led to the use of online-monitoring and signal validation techniques to monitor the process parameters and equipment conditions using several empirical models. One of these models is kernel regression (KR). Although KR is being used and performed excellently when applied to steady state, it has limitation in time-varying data that has several repetition of the same data point of the signals, especially if those signals are used to infer the other signals. In addition, many situations are related to variation and fluctuation during normal states, such as transient - start-up and shutdown mode. However, accurate estimation of the process signal can lead to the proper understanding of the equipment behaviours as well as enhance the online monitoring applications. In the sight of foregoing, this paper proposed a modified KR model, especially for transient signal validation to resolve the setback of conventional KR by improving the measure of similarities in the conventional KR. The performance of the proposed model is verified using both experimental data and the plant simulation data, and its results are compared with that of conventional KR. The results show that the proposed model has a greater robustness in particular situations compared to conventional KR, and it can be used to improve process and equipment monitoring applications. (authors)
Reactor protection system (RPS) in nuclear power plant (NPP) is in the process of digitalization in response to advanced features of digital systems. In this context, software integrity in the RPS becomes an important issue since it may cause common-cause failure. Therefore, to guarantee the safety of NPPs, the integrity of the software need to be properly demonstrated or assessed. There are three typical approaches for software reliability quantification: software reliability growth model (SRGM), Bayesian belief network (BBN), and test-based method. Basically, SRGM is difficult to apply to the safety-critical software because not only sufficient failure sets are unavailable but consequence of system failure is also unacceptable. Then the result obtained through the BBN model may have large uncertainty, so it produces difficulties in integrity demonstration of safety-critical software. Thus, the BBN result need to be complemented or verified by test-based method. Currently, for the integrity demonstration of safety-critical software, black-box test which feeds inputs and examines just outputs whether they succeeded or failed is considered based on trajectory-form-of-input (a series of successive values for the input variables that occur during the operation of the software over time). Usually these trajectory-inputs are randomly sampled from operational profile. In this approach, however, there are some practical difficulties. Therefore, we proposed a new test method for safety-critical software in digitalized reactor protection system (DRPS). The main contents described in this paper are about basic concepts under development of finite-test-cases by which the integrity of the software logic can be verified. It should be noted that the influence of the environments, such as operating system and other hardware, are not considered since the method focuses on software logic only. (authors)
Synchrotron characterization techniques provide some of the most powerful tools for the study of film structure and chemistry. The brilliance and tunability of the Advanced Photon Source allow access to scattering and spectroscopic techniques unavailable with in-house laboratory setups and provide the opportunity to probe various atomic layer deposition (ALD) processes in situ starting at the very first deposition cycle. Here, we present the design and implementation of a portable ALD instrument which possesses a modular reactor scheme that enables simple experimental switchover between various beamlines and characterization techniques. As first examples, we present in situ results for (1) X-ray surface scattering and reflectivity measurements of epitaxial ZnO ALD on sapphire, (2) grazing-incidence small angle scattering of MnO nucleation on silicon, and (3) grazing-incidence X-ray absorption spectroscopy of nucleation-regime Er{sub 2}O{sub 3} ALD on amorphous ALD alumina and single crystalline sapphire.
A nuclear space power system the SP-100 is being developed for future missions where large amounts of electrical power will be required. Although it is primarily intended for unmanned spacecraft, it can be adapted to a manned space platform by tethering it above the station through an electrical transmission line which isolates the reactor far away from the inhabited platform and conveys its power back to where it is needed. The transmission line, used in conjunction with an instrument rate shield, attenuates reactor radiation in the vicinity of the space station to less than one-one hundredth of the natural background which is already there. This combination of shielding and distance attenuation is less than one-tenth the mass of boom-mounted or onboard man-rated shields that are required when the reactor is mounted nearby. This paper describes how connection is made to the platform (configuration, operational requirements) and introduces a new element the coaxial transmission tube which enables efficient transmission of electrical power through long tethers in space. Design methodology for transmission tubes and tube arrays is discussed. An example conceptual design is presented that shows SP-100 at three power levels 100 kWe, 300 kWe, and 1000 kWe connected to space station via a 2 km HVDC transmission line/tether. Power system performance, mass, and radiation hazard are estimated with impacts on space station architecture and operation.
A nuclear space power system the SP-100 is being developed for future missions where large amounts of electrical power will be required. Although it is primarily intended for unmanned spacecraft, it can be adapted to a manned space platform by tethering it above the station through an electrical transmission line which isolates the reactor far away from the inhabited platform and conveys its power back to where it is needed. The transmission line, used in conjunction with an instrument rate shield, attenuates reactor radiation in the vicinity of the space station to less than one-one hundredth of the natural background which is already there. This combination of shielding and distance attenuation is less than one-tenth the mass of boom-mounted or onboard man-rated shields that are required when the reactor is mounted nearby. This paper describes how connection is made to the platform (configuration, operational requirements) and introduces a new element the coaxial transmission tube which enables efficient transmission of electrical power through long tethers in space. Design methodology for transmission tubes and tube arrays is discussed. An example conceptual design is presented that shows SP-100 at three power levels 100 kWe, 300 kWe, and 1000 kWe connected to space station via a 2 km HVDC transmission line/tether. Power system performance, mass, and radiation hazard are estimated with impacts on space station architecture and operation.
Simulation-assisted automation testing is an efficient tool for validating the automation systems for nuclear power plants. This paper presented a systematic procedure for selecting test cases for simulation-assisted automation testing. The procedure is relies on the plant's licensing framework and allows associating the test cases to the functional requirements in order to produce traceable test set for validation. The procedure is demonstrated with a case study concerning the automation tests of Loviisa NPP's automation renewal project. (authors)
Modification, major overhaul, repair, and instrument calibration of equipment used in cryogenic irradiation facility at Plum Brook reactor installation
This report describes research results to date in support of the integration and demonstration of diagnostics technologies for prototypical advanced reactor passive components (to establish condition indices for monitoring) with model-based prognostics methods. Achieving this objective will necessitate addressing several of the research gaps and technical needs described in previous technical reports in this series.
The final draft of a revision to IEEE Std. 1012-2012, 'IEEE Standard for System, Software and Hardware Verification and Validation', is expected to be approved for publication in June 2017. A draft of NEI 96-07, Appendix D, 'Supplemental Guidance for Application of 10 CFR 50.59 to Digital Modifications', has been submitted to the NRC staff for review and endorsement. What, if any, relationship exists between these two documents? How can the nuclear industry utilize the first one to help ensure success when applying the second one? NRC Regulatory Guides only endorse digital system and software related industry standards for application to safety-related systems and components, but the vast majority of digital modifications are made to non-safety related systems and components, so are these industry standards of little or no use when performing digital upgrades to non-safety related equipment? This paper answers these and other related questions and shows how application of a graded approach to verification and validation, utilizing the IEEE Std. 1012 concept of 'integrity level', can increase the likelihood of a 10 CFR 50.59 evaluation outcome that does not require prior NRC review and approval for non-safety related digital modifications. (authors)
An iterative approach was taken to perform the HF Verification and Validation (V and V) activities for a complex set of reactor inspection and maintenance tooling equipment known as the Bruce Reactor Inspection and Maintenance System (BRIMS). This paper is divided into the verification activities (of Human System Information Task Support and Design verifications) and the validation activities. The validation method used a graded approach by breaking validation activities into early usability testing and high fidelity validations. Early usability testing, allowed for issues and design concepts to be identified and explored. High fidelity validation activities tested that the final design requirements and operational goals of the system were met in the production equipment. The validation method has been described in detail by depicting the following: task selection criteria, tools, performance measures and data collection. Validation results demonstrate the benefits of the graded approach as issues were to be identified early in the design process, improvements were made during detailed design and additional testing could confirm that issues were resolved prior to commissioning. To capture and share the success of this iterative V and V approach lessons learned have been created for use in future V and V projects. (authors)
In May 2016, the U.S. Nuclear Regulatory Commission (NRC) staff provided a digital instrumentation and control (I&C) regulatory infrastructure integrated action plan to the NRC for approval. One of the objectives of that plan was to establish a clear regulatory structure with reduced regulatory uncertainty to enable the expanded safe use of digital I&C in commercial nuclear reactors while continuing to ensure safety and security. To achieve this end, the NRC, with collaboration from industry, developed a streamlined License Amendment Request Alternate Review (AR) process for safety-related (SR) digital I&C upgrades. In spite of this effort, the industry has remained reluctant to perform such I&C upgrades because of perceived regulatory and financial risks associated with being the first or an early adopter of the AR process for SR I&C upgrades. The U.S. Department of Energy Light Water Reactor Sustainability Program at the Idaho National Laboratory performed Initial Scoping Phase research to help break this impasse by supporting a SR I&C Pilot Upgrade, working with MPR Associates, ScottMadden Inc., and Exelon Generation. Exelon’s Limerick Generating Station (LGS) was selected as the target for this research. This paper summarizes the Initial Scoping Phase engineering and operations, licensing, and project management activities necessary to bound the scope, schedule, and estimated cost of the project sufficiently to enable utility management authorization of Conceptual Design Phase activities. These efforts and associated products are intended to provide a template to support larger industry efforts to perform similar upgrades as a foundation stone for a digital transformation that will improve plant safety, reliability, and operational performance while lowering plant total cost of ownership. As a result of the combined effort of Exelon Generation and research participants, Conceptual Design Phase activities for the subject upgrade at LGS were approved by Exelon. Further, the U.S. Department of Energy also awarded a $50 million cost share award to Exelon in order to pave the way for SR I&C modernization and associated control room upgrades across the U.S. nuclear fleet. Additional research reports are planned for the Conceptual Design Phase, Detailed Design Phase, and the Implementation Phase of the LGS project to document the process followed and promulgate lessons learned to industry.