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At least 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

A Qualitative Assessment of Current CCF Guidance Based on a Review of Safety System Digital Implementation Changes with Evolving Technology

The US Nuclear Regulatory Commission (NRC) is initiating a new rulemaking project to develop a digital system common-cause failure (CCF) rule. This rulemaking will review and modify or affirm the NRC's current digital system CCF policy as discussed in the Staff Requirements Memorandum to the Secretary of the Commission, Office of the NRC (SECY) 93-087, Policy, Technical, and Licensing Issues Pertaining to Evolutionary and Advanced Light Water Reactor (ALWR) Designs, and Branch Technical Position (BTP) 7-19, Guidance on Evaluation of Defense-in-Depth and Diversity in Digital Computer-Based Instrumentation and Control Systems, as well as Chapter 7, Instrumentation and Controls, in NRC Regulatory Guide (NUREG)-0800, Standard Review Plan for Review of Safety Analysis Reports for Nuclear Power Plants (ML033580677). The Oak Ridge National Laboratory (ORNL) is providing technical support to the NRC staff on the CCF rulemaking, and this report is one of several providing the technical basis to inform NRC staff members. For the task described in this report, ORNL examined instrumentation and controls (I&C) technology implementations in nuclear power plants in the light of current CCF guidance. The intent was to assess whether the current position on CCF is adequate given the evolutions in digital safety system implementations and, if gaps in the guidance were found, to provide recommendations as to how these gaps could be closed.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2026

Fermilab PIP-II machine protection system digitized data noise elimination scheme and its FPGA implementation

In Fermilab's PIP-II machine protection system, beam loss signals from various detectors are digitized at 125 MS/s. Noise from both high-frequency sources and low-frequency 60 Hz AC power equipment can contaminate the data. To suppress noise across these ranges—especially 60 Hz and its harmonics, which overlap with beam loss signal frequencies—advanced digital processing beyond standard filtering is required. Several real-time functional blocks were simulated and tested on an FPGA: (1) a dual time-constant discharging integrator filter, (2) a de-ripple baseline extraction and storage block, and (3) a fast-recovery discharging integrator. The nonlinear IIR integrator filter removes high-frequency noise and feeds into the baseline extractor. Upon detecting abrupt beam loss, it switches to a longer time constant to prevent baseline distortion. The de-ripple block calculates a valid baseline by averaging over multiple 60 Hz periods, storing results in a 4096-word FPGA RAM. This baseline is subtracted from raw data before integration by the fast-recovery block, which resets quickly after use. All blocks achieved expected performance and were successfully implemented on a low-cost FPGA.

Wu, Jinyuan [Fermilab]↗
DOE OSTI2017

NRC technical basis for evaluation of its position on protection against common cause failure in digital systems used in nuclear power plants - 378

Digital technology has advantages over analog systems, including automated monitoring and alerts for standby safety functions, and predictive algorithms to maintain critical safety systems. Additionally, digital technology generally has higher reliability and can be designed to reduce single point vulnerabilities. For these reasons many nuclear plants have applied digital technology to safety and non-safety related applications, including reactor protection system, feedwater and turbine controls, etc. with a corresponding significant improvement in trip reduction. Nonetheless, digital instrumentation and control (I and C) systems also present potential new vulnerabilities that need to be assessed, including potential failures due to increased complexity of digital systems, the introduction of unique failure modes due to software (including software common cause failure (CCF)), and limited operating history of digital systems in nuclear safety related applications compared to analog systems. The fact that software is intangible means that common methods, such as analysis or testing, used for detecting CCF may not be effective when applied to software. Consequently, digital technology is perceived to pose a potential risk from the introduction of undetected systematic faults that could result in CCF. Despite the I and C system upgrades and modifications performed to date, the U.S. Nuclear Regulatory Commission (NRC) and industry stakeholders have identified the need to modernize the regulatory infrastructure to efficiently address risks associated with the use of digital technology for nuclear safety applications and address regulatory uncertainties. The NRC's current position on CCF is guided by the staff requirements memorandum (SRM) on SECY 93-087. The SRM provides specific acceptance criteria for the evaluation of CCF, which the staff implemented in the Branch Technical Position (BTP) 7-19. However, industry stakeholders have proposed using methods to characterize the likelihood of software CCF and eliminate it from further consideration in a defense-in-depth and diversity analysis. The NRC's current position does not consider these alternatives, and thus corresponding acceptance criteria is not currently available. The work discussed in this paper assesses the underlying technical basis associated with CCF, provides technical support for updating the NRC position and considers proposed methods for addressing potential CCF in digital systems while enhancing efficiency, clarity, and confidence. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Graded approach for assessing digital system failure susceptibilities - 305

Owner/operators of nuclear plants must be able to identify and assess susceptibilities to digital system failures and unintended behaviors that could lead to plant system malfunctions, including common-cause failures (CCFs) of multiple controlled components that may impact overall plant safety. Nuclear plant designers and regulators often assess and manage potential failure modes by assuming the failure occurs and showing by analysis that the results are acceptable. In 2016 EPRI published a guideline on Methods for Assuring Safety and Dependability when Applying Digital Instrumentation and Control Systems (EPRI 3002005326), which takes a more holistic approach that considers digital system failure modes from the perspective of their impact on plant risk and includes a graded approach based on safety significance. This paper describes that graded approach. The methodology in EPRI 3002005326 systematically identifies potential I and C vulnerabilities that could lead to significant malfunctions of controlled components and systems, including common-cause failures (CCF), and discusses in detail methods to protect against them. It considers both preventive measures that reduce the likelihood of failures, and plant systems and features that mitigate the effects of component failures and misbehaviors. Coping analysis is then performed as appropriate to provide additional assurance of protection. The safety significance based graded approach is an important feature in that it can help users focus attention on overall plant safety, including potentially risk-significant scenarios that might not be considered in traditional safety analyses. The purpose is to help the user ensure that modifications to the I and C that could potentially be safety significant are treated appropriately and at the same time not waste valuable resources on excessive protection against changes that have little or no impact on safety. The graded approach allows the design engineer to tailor the rigor of the preventive, limiting, and mitigative measures commensurately with the effects the I and C modification has on overall plant safety. The graded approach focuses on safety significance impact, which is effectively a qualitative measure of the potential change in safety (or risk) caused by a proposed I and C modification, as compared to the I and C that is being replaced. Safety significance impact should not be confused with importance to safety or risk significance. It is possible that an upgrade to safety-related and/or risk-significant I and C systems may have little or no safety significant impact and vice versa. The approach considers three factors that influence the impact on safety significance of a proposed I and C modification: context, likelihood of failure, and consequences of failure. The most significant of the three factors is context - what the I and C is connected to, both directly and indirectly, including its potential effects on plant systems that respond to transients or accidents. The graded approach is particularly helpful in assessing potential CCFs resulting from digital I and C failures, and shows why the most problematic CCFs will be those that can affect multiple plant systems. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2025

Microreactor Automated Control System - Digital Twin Models and Advanced Control Systems Updates

Automation of control systems is expected to be important in the economic and safe operation of microreactors. Therefore, there is a need to develop and demonstrate automated control for microreactors, along with the development of testbeds for this purpose. This report provides updates on the status of a nonnuclear microreactor automated control system (MACS)—a real-time, hardware-in-the-loop testbed for non-nuclear testing of microreactor control system automation. A real-time hardware-in-the-loop testbed incorporates the realistic dynamics of physical systems into control system development and testing. The collaborative effort between Oak Ridge National Laboratory (ORNL) and Idaho National Laboratory (INL) resulted in the development of a prototypic microreactor plant-level digital twin that includes the reactor and a balance of plant system. Advanced control strategies were incorporated to demonstrate testing of control automation solutions. The gRPC communication protocol, which was implemented in the hardware-in-the-loop testbed by INL, was coupled to a digital twin model developed using the TRANsient Simulation Framework of Reconfigurable Models (TRANSFORM) library in Modelica. This digital twin simulation was tested with the ViBRANT hardware for realistic feedback and visual representation of control action in real time. A modular Python client structure was developed to manage functional mock-up unit-based simulation and real-time gRPC communication. Hardware-in-the-loop testing indicated that the modeled reactor—a natural-convection, molten-salt coolant loop configuration—responds well to control of drum positioning for modulation of reactor core power, as well as system-level control and downstream demand changes. Ongoing research is focused on integrating additional control algorithms that utilize data from newly included sensors within the MACS hardware testbed, as well as demonstrating and assessing the performance of the different automated control algorithms on multiple additional operational scenarios.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

A qualitative assessment of current CCF guidance based on a review of safety system digital implementation changes with evolving technology - 148

This paper is a summary of a study performed in support of the US Nuclear Regulatory Commission (NRC) to evaluate current policy on software common-cause failure (CCF). The study reported in this paper is one of several that contributed to the technical basis to inform NRC staff. The study first reviewed policies and assessment guidance as discussed in the Staff Requirements Memorandum to the Secretary of the Commission, Office of the NRC (SECY) 93-087, Policy, Technical, and Licensing Issues Pertaining to Evolutionary and Advanced Light Water Reactor (ALWR) Designs, and Branch Technical Position (BTP) 7-19, Guidance on Evaluation of Defense-in-Depth and Diversity in Digital Computer-Based Instrumentation and Control Systems, as well as Chapter 7, 'Instrumentation and Controls', in NUREG-0800, Standard Review Plan for Review of Safety Analysis Reports for Nuclear Power Plants. The study then examined instrumentation and controls (I and C) technology implementations in nuclear power plants in the light of current CCF guidance. The intent was to assess whether the current position on CCF remains appropriate given the evolutions in digital safety system implementations and, if gaps in the guidance were found, to provide recommendations as to how these gaps could be closed. The methodology adopted was to review the vendors' technology and software implementation processes for digital safety systems as technology evolved. The following three representative safety systems were selected to provide illustrative examples: - Eagle 21 was selected to represent vintage microprocessor-based technology used from the 1980's to mid-1990's, - TELEPERM XS (TXS) was selected to represent second generation microprocessor-based technology used from the mid-1990's to the 2000's, and - The advanced logic system (ALS) was selected to represent the latest trend of using technology based on field programmable gate arrays (FPGAs). Technology implementations were reviewed in light of the basic premise of the NRC in BTP 7-19 (Revision 6), that software-based or software-logic-based digital system development errors are a credible source of CCF and therefore are susceptible to CCF because identical copies of the software-based logic and architecture are present in redundant divisions of safety-related systems. BTP 7-19 categorizes firmware and logic developed from software-based development systems all under software. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

Signal digitizing system and method based on amplitude-to-time optical mapping

A signal digitizing system and method based on analog-to-time optical mapping, optically maps amplitude information of an analog signal of interest first into wavelength information using an amplitude tunable filter (ATF) to impress spectral changes induced by the amplitude of the analog signal onto a carrier signal, i.e. a train of optical pulses, and next from wavelength information to temporal information using a dispersive element so that temporal information representing the amplitude information is encoded in the time domain in the carrier signal. Optical-to-electrical conversion of the optical pulses into voltage waveforms and subsequently digitizing the voltage waveforms into a digital image enables the temporal information to be resolved and quantized in the time domain. The digital image may them be digital signal processed to digitally reconstruct the analog signal based on the temporal information with high fidelity.

47 OTHER INSTRUMENTATION↗
DOE OSTI2017

A readout digitization system for the ATLAS muon spectrometer phase-I tRPC trigger upgrade

The front-end digitization electronics for thin-Resistive Plate Chambers to be installed as part of the ATLAS Muon Spectrometer Phase-I upgrade project is described. Here, the design is based on the CERN HPTDC and GOL chips. In this paper, the design of the digitization system is presented, including power, connection interfaces to front-end amplifier-shaper-discriminator board and to the low level trigger processor board. System performance tests were conducted and validated with a prototype which includes one HPTDC/GOL pair.

Digital electronic circuits↗
DOE OSTI2024

Development of Whole System Digital Twins for Advanced Reactors: Leveraging Graph Neural Networks and SAM Simulations

Here, in this work, we introduce a novel method to develop whole system digital twins (DTs) for advanced nuclear reactors. This method treats a complex reactor system as a heterogeneous graph: with the system components as different types of graph nodes and their physical interconnections as edges. Based on the heterogeneous graph, a graph neural network combining graph convolution and temporal node attention is developed as the DT, facilitating a comprehensive understanding of the system's dynamic behavior. By utilizing the System Analysis Module (SAM) code for simulating various operational transients, we develop a graph-based database that trains the DT. This DT is characterized by two primary functions: It can infer the entire system's status using sparse node information, and it can predict the progress of transients based on current and historical system information. Our approach is validated through case studies on the Experimental Breeder Reactor II (EBR-II) system and a generic Fluoride-salt-cooled High-temperature Reactor (gFHR), demonstrating the DT's accuracy in forecasting operational transients. The DT's rapid computation capabilities enhance its potential for supporting advanced reactor operations, offering benefits in intelligent simulation, autonomous control, and anomaly detection, paving the way for improved safety analysis and intelligent component health management for advanced reactor systems and reducing their operations and maintenance cost.

EBR-II↗
DOE OSTI2026

Los Alamos Portable Pulser (LAPP) Marx Bank and Digitizer System Characterization

Over the duration of June 22 nd to July 15 th , I interned at Los Alamos National Laboratory within ISR-2 on the Los Alamos Portable Pulser (LAPP) Program. This program concerns a radar dish used for characterizing signal propagation through earth’s ionosphere. The bulk of my internship consisted of characterizing the delay of a Marx Bank Generator and writing hardware-interface C/C++ code for the pulse-recording Digitizer System.

47 OTHER INSTRUMENTATION↗
DOE OSTI2025

Fermilab PIP II Machine Protection System Digitized Data Noise Elimination Scheme and Its FPGA Implementation

In Fermilab's PIP-II machine protection system, beam loss signals from various detectors are digitized at 125 MS/s. Noise from both high-frequency sources and low-frequency 60 Hz AC power equipment can con-taminate the data. To suppress noise across these ranges especially 60 Hz and its harmonics, which overlap with beam loss signal frequencies advanced digital processing beyond standard filtering is re-quired. Several real-time functional blocks were simu-lated and tested on an FPGA: (1) a dual time-constant discharging integrator filter, (2) a de-ripple baseline extraction and storage block, and (3) a fast-recovery discharging integrator. The nonlinear IIR integrator filter removes high-frequency noise and feeds into the baseline extractor. Upon detecting abrupt beam loss, it switches to a longer time constant to prevent baseline distortion. The de-ripple block calculates a valid base-line by averaging over multiple 60 Hz periods, storing results in a 4096-word FPGA RAM. This baseline is subtracted from raw data before integration by the fast-recovery block, which resets quickly after use. All blocks achieved expected performance.

Wu, Jinyuan [Fermilab] (ORCID:0000000344329521)↗
DOE OSTI2026

Fermilab PIP II machine protection system digitized data noise elimination scheme and its FPGA implementation

In Fermilab's PIP-II machine protection system, beam loss signals from various detectors are digitized at 125 MS/s. Noise from both high-frequency sources and low-frequency 60 Hz AC power equipment can contaminate the data. To suppress noise across these ranges especially 60 Hz and its harmonics, which overlap with beam loss signal frequencies advanced digital processing beyond standard filtering is required. Several real-time functional blocks were simulated and tested on an FPGA: (1) a dual time-constant discharging integrator filter, (2) a de-ripple baseline extraction and storage block, and (3) a fast-recovery discharging integrator. The nonlinear IIR integrator filter removes high-frequency noise and feeds into the baseline extractor. Upon detecting abrupt beam loss, it switches to a longer time constant to prevent baseline distortion. The de-ripple block calculates a valid baseline by averaging over multiple 60 Hz periods, storing results in a 4096-word FPGA RAM. This baseline is subtracted from raw data before integration by the fast-recovery block, which resets quickly after use. All blocks achieved expected performance.

Wu, J. [Fermilab] (ORCID:0000000344329521)↗
DOE OSTI2025

Modeling and Design of a Sub-Nyquist PDV Digitization System

Photon-Doppler Velocimetry (PDV) uses interferometry of a transmitted optical signal reflected from a fast-moving device under test (DUT) to generate a product signal at a doppler shifted frequency which is proportional to the velocity of the object being measured. The technique has applications in shock physics, where it is used to measure fast-moving objects over a short lifespan of travel, such as a bullet, shrapnel, or other shock accelerated object.

42 ENGINEERING↗