Search NASA⌕ Search

SEARCH · Search NASA

Results for “REACTOR INSTRUMENTATION”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 235 records · Page 13
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Reactor operator-in-the-loop dynamic modelling - 122

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)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Evaluating situation awareness in the control room: the development and initial testing of the process awareness and situation understanding (PASU) measure - 267

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)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Development and application of FPGA-based logic controller - 357

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)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Development of a transient signal validation technique via a modified kernel regression model - 365

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)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Finite-Test-Case Development to Demonstrate Logical Integrity of Software in DRPS

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)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

A modular reactor design for in situ synchrotron x-ray investigation of atomic layer deposition processes

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.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
NASA NTRS1985

Tethered nuclear power for the Space Station

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.

Bents, D. J.↗
NASA NTRS1985

Tethered nuclear power for the space station

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.

Bents, D. J.↗
DOE OSTI2017

Test case selection procedure for simulation-assisted automation testing - 321

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)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

Component-Level Prognostics Health Management Framework for Passive Components - Advanced Reactor Technology Milestone: M2AT-15PN2301043

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.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

IEEE STD. 1012 and NEI 96-07, Appendix D - Strange bedfellows? - 149

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)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Human factors case study: An iterative approach to verifying and validating a new suite of reactor inspection and maintenance tooling - 177

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)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Safety-Related Instrumentation and Control Pilot Upgrade: Initial Scoping Phase Implementation and Lessons Learned

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.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

APR1400 In-Core Instrumentation Penetration Failure by Zirconia Melt

Light water reactors have many penetrations at the reactor vessel lower head which are called ICI (in-core instrumentation tubes), ICM-GT (in-core monitoring guide tubes), CRGT (control rod guide tubes) and drain tubes. For Korean PWRs (pressurized water reactors), there are dozens of bottom-up type ICI penetration tubes (45 for OPR1000 and 61 for APR1400). For the typical BWRs (boiling water reactors), hundreds of penetrations are positioned at the reactor vessel lower head due to the structural complexity at the reactor upper head. These penetrations are regarded as the most vulnerable structures during a severe accident as they could be failed by high temperature of melt (or debris) and its decay heat. That is, tube rupture and/or tube ejection could occur because there exist an empty hole inside the penetration tube which can be a melt flow path, and they are attached to the inside of the reactor vessel by a partial weld with a gap between the tubes and reactor vessel. Therefore, the research on the penetration failure is of importance, as it could lead to melt discharge into the containment and subsequent release of radioactive materials to the environment due to the containment failure. This paper focuses on the experimental investigation on the tube ejection failure for the APR1400 ICI penetration. The tube ejection experiments were performed as it was exposed to high temperature of zirconia melt and high pressure. The tube ejection failure experiments were performed for the APR1400 in-core instrumentation penetration nozzle. Zirconia melt was used as a simulant of molten corium and interacted with the penetration specimen. The specimen was pressurized up to 2.6 bar during the interaction with melt to induce the tube ejection. The penetration weld was heated up to its melting temperature, and the penetration tube and weld above the reactor vessel surface were eroded severely by the melt. However, even though it was estimated that the weld failure occurred and the reactor vessel with stainless steel cladding was ablated severely by the melt, the tube ejection did not take place, and consequently the integrity for the APR1400 ICI penetration was confirmed in the present experimental conditions. The goal of this research is to improve the penetrations failure models in MAAP5 codes based on the experimental data. To achieve this goal, it is necessary to increase the heat flux and pressure acting on the penetration tube to find the conditions for the tube ejection failure. Moreover, the penetration failure experiments will be performed by delivering melt into the penetration specimen to simulate the melt relocation process in a real severe accident situation. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Summary Describing Integration of ERM Methodology into Supervisory Control Framework with Software Package Documentation; Advanced Reactor Technology Milestone: M4AT-16PN2301052

This project involved the development of enhanced risk monitors (ERMs) for active components in Advanced Reactor (AdvRx) designs by integrating real-time information about equipment condition with risk monitors. Health monitoring techniques in combination with predictive estimates of component failure based on condition and risk monitors can serve to indicate the risk posed by continued operation in the presence of detected degradation. This combination of predictive health monitoring based on equipment condition assessment and risk monitors can also enable optimization of maintenance scheduling with respect to the economics of plant operation. This report summarizes PNNL’s multi-year project on the development and evaluation of an ERM concept for active components while highlighting FY2016 accomplishments. Specifically, this report provides a status summary of the integration and demonstration of the prototypic ERM framework with the plant supervisory control algorithms being developed at Oak Ridge National Laboratory (ORNL), and describes additional case studies conducted to assess sensitivity of the technology to different quantities. Supporting documentation on the software package to be provided to ONRL is incorporated in this report.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Reduced order methods for the improvement of control-oriented modelling of nuclear power plants - 185

In the development of the control system of a nuclear power plant, it is of primary importance to rely on simulation tools for its realization, testing and validation. These tools should provide an accurate description of the controlled response of the entire power plant in different scenarios. The classic control-oriented modelling based on 0D/1D approach is well suited for control purposes since it provides fast-running simulations and it is easy to linearize, but it may not be sufficient to deeply investigate the complexity of some systems, in particular when/where spatial phenomena have a remarkable impact on dynamics. Reduced Order Methods (ROMs) can offer the proper trade-off between computational cost and solution accuracy, combining the high-detail modelling (e.g., 3D modelling) usually adopted for design purposes with the requirements demanded for a control-oriented tool, firstly the computational efficiency. In this work, ROMs are used to improve a control-oriented plant simulator of a Lead-cooled Fast Reactor (LFR). In particular, the attention is focused on the reactor core neutronics, and on the thermal-hydraulics of the reactor pool as well. A test case regarding the design of a Kalman Filter is set up showing promising results for the application of ROM in the control context. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗