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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.

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NASA NTRS1974

Compatibility of refractory materials for nuclear reactor poison control systems

Metal-clad poison rods have been considered for the control system of an advanced space power reactor concept studied at the NASA Lewis Research Center. Such control rods may be required to operate at temperatures of about 140O C. Selected poison materials (including boron carbide and the diborides of zirconium, hafnium, and tantalum) were subjected to 1000-hour screening tests in contact with candidate refractory metal cladding materials (including tungsten and alloys of tantalum, niobium, and molybdenum) to assess the compatibility of these materials combinations at the temperatures of interest. Zirconium and hafnium diborides were compatible with refractory metals at 1400 C, but boron carbide and tantalum diboride reacted with the refractory metals at this temperature. Zirconium diboride also showed promise as a reaction barrier between boron carbide and tungsten.

Sinclair, J. H.↗

NASA NTRS1996

Development of a Software Safety Process and a Case Study of Its Use

Research in the year covered by this reporting period has been primarily directed toward: continued development of mock-ups of computer screens for operator of a digital reactor control system; development of a reactor simulation to permit testing of various elements of the control system; formal specification of user interfaces; fault-tree analysis including software; evaluation of formal verification techniques; and continued development of a software documentation system. Technical results relating to this grant and the remainder of the principal investigator's research program are contained in various reports and papers.

Knight, J. C.↗

DOE OSTI2019

Transformational Challenge Reactor Autonomous Control System Framework and Key Enabling Technologies

Autonomous operation of nuclear plants has the potential for significantly improving the economics of existing plants while maintaining operation within the safety envelope of the reactor. Autonomous operation technology, if developed and deployed, may enable deployment of small reactors at remote sites while requiring minimal operator support. In addition to reducing costs, the technology can add additional layers of safety through rapid decision making and actuation in off-normal situations. Within the context of the Transformational Challenge Reactor (TCR), autonomous control along with the other unique characteristics of the reactor (advanced manufacturing) is expected to enable rapid deployment. On the other hand, technology for autonomy can add risk to the system if not properly developed. There is therefore a need for systematic development and demonstration of autonomy technology for TCR. This document describes a framework for autonomy for the TCR. It defines autonomous operation and examines whether autonomy is possible for nuclear systems. The report also describes enabling sensing technologies for autonomy and describes a framework to facilitate autonomy.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2019

FY19 Transient Prescriptions in TREAT Full-Slotted Core

The transient reactor test facility (TREAT) is a graphite-based transient test reactor at the Idaho National Laboratory. TREAT started operation in 1959 and performed thousands of transient irradiations for the primary purpose of nuclear fuel safety research. TREAT operations were suspended in 1994 and then, several years later, resumed reactor operations in 2017 to support development of advanced nuclear fuels. During TREAT’s first era of operation the transient rod drive system and eventual Automatic Reactor Control System (ARCS) were upgraded and improved to create a transient reactor capable of unparalleled power shaping capabilities. TREAT’s first year of modern experiment work was performed in 2018 and saw the demonstration of new kinds of transients focused on modern experiment needs; especially focused on light water reactor (LWR) category transients including reactivity-initiated accidents (RIA) and loss of coolant accidents (LOCA). These transients were performed using a non-fueled neutronic mockup of a historical sodium loop which was not a perfect representation of future core configurations, but was adequate for demonstrating capabilities, successfully troubleshooting transient programming parameters, and calibrating nuclear models of the core. These physics tests were termed “transient prescription tests”. Later in 2018 the core was reconfigured to a modern fueled experiment arrangement and TREAT began performing irradiations on LWR rodlets where crucial fuel specimen nuclear heating data were obtained, and several supporting systems were commissioned. This successful effort gave way to additional fueled experiments which were performed throughout 2019.

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DOE OSTI2025

Poster: Hardware-Based Demonstration of Temperature Control Functions for Reactor Systems

Autonomous thermal regulation in nuclear reactors remains crucial for maintaining stable operation and ensuring the integrity of fuel. To alleviate public skepticism of the safety of nuclear reactors, demonstrating control over this key factor is pivotal. Utilizing electric heat pads to simulate the heat released in a reactor core, thermocouples for temperature monitoring, and an Arduino micro programmable logic controller (PLC) with an embedded proportional-integral-derivative (PID) algorithm for control, a hardware-based demonstration of a reactor heating system will validate the efficacy of reactor control over this key parameter.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗

DOE OSTI2025

Project Presentation: Hardware-Based Demonstration of Temperature Control Functions for Reactor Systems

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

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↗

NASA NTRS1969

A second-order system model for pneumatic instrumentation lines.

Pneumatic instrumentation line transfer function approximated by system model for simplifying reactor control system analysis

Samuelson, R. D.↗

DOE OSTI2017

Test based reliability assessment method for a safety critical software in reactor protection system - 284

Digitalization of the instrumentation and control systems in nuclear power plant entails some new features which do not exist in analog systems. When the new features are applied in safety critical systems, the risk stemming from them should be quantified properly to assure the reliability of the entire power plant. Among the new features, software is considered as the most important factor because it can cause common cause failure to many redundant systems. Therefore, in this work, a test based reliability assessment method for a safety-critical software is developed. An output of a software is determined by not only inputs but also the internal state of the software at the time. In the case of safety critical software developed for usage in nuclear power plant, it is possible to directly investigate probable internal states thanks to detailed design specification and programing features, and the probable inputs also can be identified based on the physical linearity of each process parameter and hardware characteristics related to data acquisition. In this work, a development method for practical exhaustive test case consisting of investigation of internal state and probable input sets to a specific internal state are suggested. In addition to this, software logic simulator for execution of software logic test and the reliability quantification method based on the test result are developed. The feasibility of the suggested method is demonstrated via a case study. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2017

Control system design for a small pressurized water reactor - 141

Small Pressurized Water Reactors (SPWR) have unique designs from the commercial large Pressurized Water Reactors (PWRs). Integral design is adopted. There are no hot legs and cold legs between the reactor core and the steam generators like in the PWR. Strong interactions exist among different loops. The pressure vessel is large in volume and the coolant inventory is also in a large amount. The inertia of the coolant is large and it takes more time for the primary system to respond to disturbances. The steam generator adopts the once-through design and the water inventory is small. The steam generator is very sensitive to disturbances. These unique characteristics challenge the control system design of SPWR. Relap5 is used to model the SPWR. Steady-state simulation is performed to verify the correctness of the model. In the reactor power control system, both the reactor fission power and the coolant average temperature are regulated by the control rod reactivity. In the feedwater flow control system, the coordination between the reactor and the turbine is considered and coolant average temperature is adopted as one measurable disturbance for the control system. The coolant pressure is adjusted based on the heaters and spray in the pressurizer. The water level in the pressurizer is controlled by the charging flow and the letdown flow is compensated to calculate the charging flow needed. Transient simulations are carried out to evaluate the control system performance. When the reactor is perturbed, the reactor can be stabilized under the control system. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2017

Online monitoring of rod control systems in combustion engineering pressurized water reactors - 195

Current designs of Pressurized Water Reactors (PWRs) are equipped with control and shutdown rods that are inserted and withdrawn from the reactor core to control reactivity. In Combustion Engineering (CE) plants, these rods are moved using the Control Element Drive Mechanism (CEDM) and controlled by the CEDM Control System (CEDMCS). In the event of system failure, troubleshooting activities can have a substantial impact on plant resources. Furthermore, CEDM component failures have occurred without warning, which has led to increased costs for the plants. These issues, combined with the inherently complex design of the CEDMCS, create a need for an online diagnostics system that can improve system health monitoring, increase diagnostic capabilities, and decrease troubleshooting time. This paper describes an advanced analysis and diagnostics system that monitors the CEDMCS during plant operation and provides critical information to onsite personnel to help detect and prevent problems before they impact normal plant operation. This system uses existing CEDMCS test points to acquire data and needs no modifications to the CEDMCS for implementation. The online monitoring capabilities automatically collect data during normal operation and diagnose failure events, reduce troubleshooting time and provide plant personnel with valuable information which is not currently available. Additionally, the system has built-in features used to streamline routine surveillance testing, collect stationary and movement data on multiple rods simultaneously, automate analysis, and generate reports. This system was developed with input from industry CEDMCS engineers at existing commercial facilities and has been deployed in multiple operating nuclear plants. The system has demonstrated itself to be a valuable tool for CEDMCS health monitoring, not only in reducing required man-power for routine system testing, but also in detected impending coil faults and allowing plant engineers to mitigate system component failures to prevent impending plant shut downs. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

NASA NTRS1967

Design of linear and nonlinear control systems via state variable feedback, with applications in nuclear reactor control

Linear and nonlinear control systems via state variable feedback with applications in nuclear reactor control

LINEAR SYSTEM↗

NASA NTRS1993

SP-100 control drive assembly development

The SP-100 is an electrical generating nuclear power system for space operation. This paper describes the nuclear reactor control systems and the methods used to assure reliable performance for the 10-year design life. Reliable performance is achieved by redundancy and by selecting highly reliable components and design features. Reliability is quantified by analysis using established reliability data. Areas lacking reliability data are identified for development testing. A specific development test description is provided as an example to demonstrate how this process is meeting the system reliability goals.

Gleason, Thomas↗

NASA NTRS2005

STS-114: Discovery Day 13 Mission Status Briefing

LeRoy Cain, STS-114 Ascent/Entry Flight Director, takes a solo stand with the Press in this briefing. He reports that the vehicle is in good shape, consumable status is excellent, and the shuttle crew is in high spirits and preparing for de-orbit and landing. LeRoy and his team have completed the entry system check up, flight control check up, reactor control system check up, and noted that all are at nominal performance; weather forecast is very good, the Entry team is ready and looking forward to de-orbit and landing at the Kennedy Space Center on Monday, August 8th. Re-entry, personal feelings, Columbia accident, data gathering, consumable situation, back up sites, weather, communication block out, night and day landing, and Commander Collin's piloting skills during night flight are some of the topics covered with the News media.

Source record↗

DOE OSTI2017

Nuclear power plant instrumentation and control cyber security common vector access leading to relational common cause failures - 340

Nuclear power plants and facilities have been implementing digital system upgrades into their previously analog systems for well over twenty years. New nuclear facilities' control, security, and emergency preparedness systems are almost exclusively built on digital architectures with a high degree of communication between the various systems that are often integrated together into a central control station to aid in operation or security of the facility. As digital systems become more widespread in nuclear facility control system architectures, cyber security related issues have become a significant concern to operators, regulators, governments, and other groups. Among the many concerns related to digital systems and cyber security is the area of common cause and common mode failures. This paper introduces, defines, and discusses some sources of common cause failure from a cyber security perspective: common vector access. This refers to specific access points that an adversary can exploit through a single attack sequence that have the potential to provide relational failures through common cause on multiple components, subsystems, systems, or plants. This paper will further discuss interconnected processes where these access points may exist, the importance of limiting or controlling these pinch points, and some methods of protecting common vector access points. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2016

Design of a Commercial Scale Accelerator Driven Subcritical Aqueous Assembly

SHINE Medical Technologies is working towards producing a reliable, domestic supply of medical isotopes, including molybdenum-99 (Mo-99), xenon-133, and iodine 131, without using highly enriched uranium. Approximately 88% of Mo-99 decays to technetium-99m, which is the most widely used medical isotope in the world, accounting for approximately 80% of all nuclear medicine procedures. A novel aqueous, subcritical nuclear assembly has been designed by SHINE Medical Technologies. The assembly can produce medical radioisotopes in significant quantities for U.S. and foreign needs. The assembly produces isotopes without using a reactor, instead using a fusion-based neutron source coupled to a fission-based aqueous assembly. The subcritical nature of the aqueous assembly has safety and licensing benefits, including the elimination of the reactor control system and elimination of high level waste. SHINE plans to construct eight subcritical assemblies with a plant capacity of greater than 3000 6-day Curies, which is more than one-quarter of the world's Mo-99 requirements. From a nuclear design standpoint, the aqueous subcritical assembly needs to be able to generate sufficient fission power to meet the production capacity requirements while remaining subcritical at all times with an acceptable margin. To address these conflicting design goals, a simple nuclear design and reactivity control philosophy has been developed that allows for high power operation with a high degree of safety. A fusion-driven subcritical fission assembly has been designed by SHINE Medical Technologies to provide a reliable, domestic supply of medical isotopes. Nuclear models have been created and show that the assembly can be started up, operated, and shutdown in a safe manner without active reactivity control systems. A unique reactivity control strategy allows SHINE to operate the aqueous subcritical assembly at much higher power levels than would otherwise be practical, while still retaining the benefits of subcritical operation and avoiding the need for complex subcritical measurement techniques. Based on the current design, the subcritical target is capable of generating significant fission power (on the order of 100 kW) when being driven by a simple, low-cost deuterium-tritium accelerator, enabling the SHINE facility to supply over one-fourth of the worldwide demand of Mo- 99. SHINE has recently been issued a construction permit by the U.S. Nuclear Regulatory Commission to build the commercial facility in Janesville, WI, with eight subcritical assemblies and the associated production equipment. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI1994

Nuclear Safety [Vol. 35, No. 1, January-June 1994]

Nuclear Safety is a review journal that covers significant developments in the field of nuclear safety. Its scope includes the analysis and control of hazards associated with nuclear energy, operations involving fissionable materials, and the products of nuclear fission and their effects on the environment. Primary emphasis is on safety in reactor design, construction, and operation; however, the safety aspects of the entire fuel cycle, including fuel fabrication, spent-fuel processing, nuclear waste disposal, handling of radioisotopes, and environmental effects of these operations, are also treated. Table of Contents for this issue follows. GENERAL SAFETY CONSIDERATIONS: THE CHERNOBYL ACCIDENT: 1 Chernobyl Accident Management Actions, A. R Sich 25 The IAEA-ASSET Approach to Avoiding Accidents is to Recognize the Precursors to Prevent Incidents, F. Reisch; ACCIDENT ANALYSIS: 36 A Review of the Available Information on the Triggering Stage of a Steam Explosion, D. F. Fletcher; 58 Analysis and Modeling of Flow-Blockage-Induced Steam Explosion Events in the High-Flux Isotope Reactor, R. P. Taleyarkhan, V. Georgevich, C. W. Nestor, U. Gat, B. L. Lepard, D. H. Cook, J. Freels, S. J. Chang, C. Luttrell, R. C. Gwaltney, and J. Kirkpatrick; 74 An Analysis of Disassembling the Radial Reflector of a Thermionic Space Nuclear Reactor Power System, M. S. El-Genk and D. V. Paramonov; CONTROL AND INSTRUMENTATION: 86 Standards for High-Integrity Software, D. R. Wallace, D. R. Kuhn, L M. Ippolito, and L. Beltracchi; DESIGN FEATURES: 98 Adoption of New Design Features for the Next Generation Nuclear Power Reactors, L. S. Tong; 114 Review of Nuclear Piping Seismic Design Requirements, G. C. Slagis and S. E. Moore; ENVIRONMENTAL EFFECTS: 128 PC-Based Probabilistic Safety Assessment Study for a Geological Waste Repository Placed in a Bedded Salt Formation, S. A. Khan; OPERATING EXPERIENCES: 142 Managing Aging in Nuclear Power Plants: Insights from NRC’s Maintenance Team Inspection Reports, A. Fresco and M. Subudhi; 150 Reactor Shutdown Experience, Compiled by J. W. Cletcher; 153 Selected Safety-Related Events, Compiled by G. A. Murphy; RECENT DEVELOPMENTS: 158 Reports, Standards, and Safety Guides, D. S. Queener; 168 Proposed Rule Changes as of Dec. 31, 1993; ANNOUNCEMENTS: 177 Symposium on Radioactive and Mixed Waste—Risk as a Basis for Waste Classification; 177 1995 Incineration Conference 178 Ninth Power Plant Dynamics Control and Testing Symposium; 174 The Authors

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI1994

Nuclear Safety [Vol. 35, No. 1, January-June 1994]

Nuclear Safety is a review journal that covers significant developments in the field of nuclear safety. Its scope includes the analysis and control of hazards associated with nuclear energy, operations involving fissionable materials, and the products of nuclear fission and their effects on the environment. Primary emphasis is on safety in reactor design, construction, and operation; however, the safety aspects of the entire fuel cycle, including fuel fabrication, spent-fuel processing, nuclear waste disposal, handling of radioisotopes, and environmental effects of these operations, are also treated. Table of Contents for this issue follows. THE CHERNOBYL ACCIDENT: 1 Chernobyl Accident Management Actions, A. R Sich; GENERAL SAFETY CONSIDERATIONS: 25 The IAEA-ASSET Approach to Avoiding Accidents is to Recognize the Precursors to Prevent Incidents, F. Reisch; ACCIDENT ANALYSIS: 36 A Review of the Available Information on the Triggering Stage of a Steam Explosion, D. F. Fletcher; 58 Analysis and Modeling of Flow-Blockage-Induced Steam Explosion Events in the High-Flux Isotope Reactor, R. P. Taleyarkhan, V. Georgevich, C. W. Nestor, U. Gat, B. L. Lepard, D. H. Cook, J. Freels, S. J. Chang, C. Luttrell, R. C. Gwaltney, and J. Kirkpatrick; 74 An Analysis of Disassembling the Radial Reflector of a Thermionic Space Nuclear Reactor Power System, M. S. El-Genk and D. V. Paramonov; CONTROL AND INSTRUMENTATION: 86 Standards for High-Integrity Software, D. R. Wallace, D. R. Kuhn, L M. Ippolito, and L. Beltracchi; DESIGN FEATURES: 98 Adoption of New Design Features for the Next Generation Nuclear Power Reactors, L. S. Tong; 114 Review of Nuclear Piping Seismic Design Requirements, G. C. Slagis and S. E. Moore; ENVIRONMENTAL EFFECTS: 128 PC-Based Probabilistic Safety Assessment Study for a Geological Waste Repository Placed in a Bedded Salt Formation, S. A. Khan; OPERATING EXPERIENCES: 142 Managing Aging in Nuclear Power Plants: Insights from NRC’s Maintenance Team Inspection Reports, A. Fresco and M. Subudhi; 150 Reactor Shutdown Experience, Compiled by J. W. Cletcher; 153 Selected Safety-Related Events, Compiled by G. A. Murphy; RECENT DEVELOPMENTS: 158 Reports, Standards, and Safety Guides, D. S. Queener; 168 Proposed Rule Changes as of Dec. 31, 1993; ANNOUNCEMENTS: 177 Symposium on Radioactive and Mixed Waste—Risk as a Basis for Waste Classification; 177 1995 Incineration Conference 178 Ninth Power Plant Dynamics Control and Testing Symposium; 174 The Authors

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗