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At least 1,711 records · Page 95
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Computer Based Procedures for Field Workers - FY16 Research Activities

The Computer-Based Procedure (CBP) research effort is a part of the Light-Water Reactor Sustainability (LWRS) Program, which provides the technical foundations for licensing and managing the long-term, safe, and economical operation of current nuclear power plants. One of the primary missions of the LWRS program is to help the U.S. nuclear industry adopt new technologies and engineering solutions that facilitate the continued safe operation of the plants and extension of the current operating licenses. One area that could yield tremendous savings in increased efficiency and safety is in improving procedure use. A CBP provides the opportunity to incorporate context-driven job aids, such as drawings, photos, and just-in-time training. The presentation of information in CBPs can be much more flexible and tailored to the task, actual plant condition, and operation mode. The dynamic presentation of the procedure will guide the user down the path of relevant steps, thus minimizing time spent by the field worker to evaluate plant conditions and decisions related to the applicability of each step. This dynamic presentation of the procedure also minimizes the risk of conducting steps out of order and/or incorrectly assessed applicability of steps. This report provides a summary of the main research activities conducted in the Computer-Based Procedures for Field Workers effort since 2012. The main focus of the report is on the research activities conducted in fiscal year 2016. The activities discussed are the Nuclear Electronic Work Packages – Enterprise Requirements initiative, the development of a design guidance for CBPs (which compiles all insights gained through the years of CBP research), the facilitation of vendor studies at the Idaho National Laboratory (INL) Advanced Test Reactor (ATR), a pilot study for how to enhance the plant design modification work process, the collection of feedback from a field evaluation study at Plant Vogtle, and path forward to commercialize INL’s CBP system.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Example Evaluation of a Representative Heat Pipe Test Article Design for Structural Acceptability using ASME Design Rules

This report describes a design analysis of a test article mimicking a core block of a heat-pipe mi-croreactor using the ASME Boiler & Pressure Vessel Code Section III, Division 5, Subsection HB, Subpart B rules. These rules cover the design and construction of elevated temperature nu-clear reactor structural components, like the core block of a heat-pipe microreactor emulated by the test article. The purposes of the report are to: (1) provide a step-by-step example of applying the ASME rules to a reasonable core block geometry and corresponding loading conditions, (2) to provide design feedback on the test article, and (3) to evaluate the ASME rules for use in the design of complex heat-pipe microreactor components. The test article design has a relatively short ASME design life, limited by high thermal stresses. The current ASME rules are adequate for evaluating the core block component designs, though they could be optimized to increase the design efficiency and reduce the amount of designer/analyst time required to execute the design process. The report covers both the current, 2019 edition ASME rules and modified, currently unapproved, rules designed to improve the design process for components with complex geometries and high secondary stresses, like the core block test article.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2014

Evolution of pores in the fuel cladding of the electrogenerating channel

The results of reactor tests of carbonitride fuel in a monocrystalline cladding from a molybdenum-based alloy that were carried out earlier on an experimental setup Ya-82 for 8 to 300 h at a temperature of ∼1500°C can be used in order to prove the operational reliability of fuel elements in the design of a megawatt nuclear power plant for a spacecraft. A raster image of the surface of a sample of cladding shows that the interfaces between the layers are decorated by pores. This result is explained in this work by the theory of coalescence. The mechanisms responsible for the evolution of pores taking place at the parameters of a Ya-82 setup are considered. The effect of decoration by pores of the interfaces between the layers of a sample of cladding of the electrogenerating channel by carrying out reactor tests is explained. The dependence of the average radius of pores on the duration of the experiment is obtained. An evaluation of the average sizes of pores arising under the conditions of the experiment gives a value of ∼2 μm, which is in agreement with the experimental data. A computational study of swelling of the cladding material in the process of irradiation is performed. Predictive estimates of the behavior of the porous system and swelling of the cladding material for a megawatt class nuclear power plant are made.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NASA NTRS1978

Lithium and potassium heat pipes for thermionic converters

A prototypic heat pipe system for an out-of-core thermionic reactor has been built and tested. The emitter of the concentric thermionic converter consists of the condenser of a tungsten heat pipe utilizing a lithium working fluid. The evaporator section of the emitter heat pipe is radiation heated to simulate the thermal input from the nuclear reactor. The emitter heat pipe thermal transport is matched to the thermionic converter input requirement. The collector heat pipe of niobium, 1%-zirconium alloy uses potassium as the working fluid. The thermionic collector is coupled to the heat pipe by a tapered conical joint designed to minimize the temperature drop. The area ratio of the evaporator to condenser is 16:1, which increases the radiation area. The composite wick structure consists of seven arteries and cylindrical wraps. The collector heat flux matches the design requirements of the thermionic converter.

Miskolczy, G.↗
DOE OSTI2023

Monte Carlo Analysis of Coolant Stream Impurity Gamma Emissions in Gas-Cooled Fast Reactors

It is well established that a rapid increase in the concentration of fission products in the reactor coolant stream can serve as an early indication of fuel failure. We use Monte Carlo simulations to investigate the feasibility of using several gamma detectors as diagnostic equipment to monitor the presence of major fission product isotopes in high-temperature gas-cooled reactor (HTGR) coolant streams for early detection of fuel failure and therefore the prevention of fuel failure conditions. Herein we model the response of high-purity germanium (HPGe), CdZnTe, NaI(Tl), and LaBr 3 (Ce) detectors of typical commercial sizes to the gamma emissions from nuclides expected to be found within the coolant stream of the Versatile Test Reactor (VTR) under development by the U.S. Department of Energy. The results indicate that for the 233- and 250-keV gamma rays from 133 Xe and 135 Xe, respectively, the 3σ detection criterion is met in under 1 min using a single HPGe detector. Changes in other spectral lines associated with Xe nuclides are detected within 1 h regardless of the choice of detector.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

FUEL PERFORMANCE STUDIES AT IDAHO NATIONAL LABORATORY MAKING USE OF THE BYRON FUEL SHIPMENT

In December of 2023 a shipment of commercially irradiated fuel rods from the Byron Generating Station in Illinois was successfully shipped to the Materials and Fuels Complex (MFC) at Idaho National Laboratory (INL). The make-up of the rods includes a mix of cladding types from traditional ZirloTM, advanced zirconium alloys, and chrome coated ZirloTM. Burnups range from regular end of life values to over 70 GWd/MTU rod average. The R&D plan for the rods involves multiple projects from developing licensing data for new claddings to integral transient tests to support burnup extension efforts in the United States. The R&D began in early 2024 with the nondestructive examinations of the rods after which they will be sectioned for microscopy, mechanical testing, and analytical chemistry. Additionally, many rod segments will be refabricated into new test pins and inserted into a static water capsule for integral Reactivity Initiated Accident (RIA) and Loss of Coolant Accident (LOCA) testing at the TREAT reactor.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Resumption of water capsule reactivity-initiated accident testing at TREAT

A series of integral reactivity-initiated accident (RIA) commissioning experiments were completed in a new static water capsule in the Transient Reactor Test Facility (TREAT), marking the first such tests in the U.S. in more than 40 years. The test campaign included a verification test followed by five tests in the Static Environment Rodlet Transient Test Apparatus (SERTTA) capsule. The capsule initial conditions varied from room temperature and pressure up to 200°C and 2.5 MPa, with energy depositions varying between ~500–1100 J/gUO 2 . The series of tests allowed for a number of instrumentation qualifications and demonstrations, including cladding thermometry, rodlet plenum pressure, cladding elongation, and an electro-impedance boiling detector. Post transient examinations such as gamma emission spectroscopy, profilometry, and microscopy were performed to document the end state of the fuel rods. Furthermore, the results from the experiments show that the fuel rodlets behaved very similar to historical tests under similar energy depositions. This paper documents the design of the capsule and highlights some results from the commissioning tests and post-transient examination.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Advanced Test Reactor Core Modeling Update Project Annual Report for Fiscal Year 2013

Legacy computational reactor physics software tools and protocols currently used for support of Advanced Test Reactor (ATR) core fuel management and safety assurance, and to some extent, experiment management, are inconsistent with the state of modern nuclear engineering practice, and are difficult, if not impossible, to verify and validate (V&V) according to modern standards. Furthermore, the legacy staff knowledge required for effective application of these tools and protocols from the 1960s and 1970s is rapidly being lost due to staff turnover and retirements. In late 2009, the Idaho National Laboratory (INL) initiated a focused effort, the ATR Core Modeling Update Project, to address this situation through the introduction of modern high-fidelity computational software and protocols. This aggressive computational and experimental campaign will have a broad strategic impact on the operation of the ATR, both in terms of improved computational efficiency and accuracy for support of ongoing DOE programs as well as in terms of national and international recognition of the ATR National Scientific User Facility (NSUF).

97 MATHEMATICS AND COMPUTING↗
NASA NTRS1980

LVDT gage for fracture-toughness tests in liquid hydrogen

Linear-variable differential transformer replaces conventional resistance strain gages to measure crack-mouth-opening displacement. LVDT is superior in tests under liquid hydrogen, where boiling of hydrogen on resistive is suited to broad temperature range and hostile environments such as nuclear reactors.

Pierce, W. S.↗
NASA NTRS1994

SP-100 Dynamic Power and Lithium-Propellant MPD Nuclear Electric Propulsion Technology Requirements

The objective of this study was to evaluate the requirements (including system integration, design, test requirements, and schedule) for the propulsion and power conversion systems of a nuclear electric propulsion (NEP) vehicle using an SP-100 reactor with a dynamic power conversion system, Li-propellant magnetoplasmadynamic (MPD) thrusters, Li-propellant storage and feed systems, and the power conditioning electronics required to convert the power output from the power system to the form (voltage, current) needed by the thrusters. Potassium-Rankine power conversion systems have the potential for the greatest mission benefit in terms of minimum mass and volume (as compared to Brayton or Stirling power conversion systems), but they require the most development.

Murray, Kathy H.↗
DOE OSTI2017

Flux in a 1D-Reactor Cell by Response Matrix

Legacy computational reactor physics software tools and protocols currently used for support of the Advanced Test Reactor (ATR) core fuel management and safety assurance, and to some extent, experiment management, are inconsistent with the state of modern nuclear engineering practice, and are difficult, if not impossible, to verify and validate (V&V) according to modern standards. Furthermore, the legacy staff knowledge required for effective application of these tools and protocols from the 1960s and 1970s is rapidly being lost due to staff turnover and retirements. Thus in late 2009, the Idaho National Laboratory (INL) initiated a focused effort, the ATR Core Modeling Update Project, to address this situation through the introduction of modern high-fidelity transport theory based computational software and models to replace the legacy methods, with appropriate (V&V) protocols according to applicable national standards.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Production of Xenon-135 from isotopically enriched Xenon-134 and Xenon-136 Targets

Atmospheric radioxenon analyses conducted under the Comprehensive Test Ban Treaty Organization (CTBTO) necessitates the collection and analysis of air samples to quantify 135Xe, an indicator for nuclear explosions and reactor operations. Analytical instrument calibration for these analyses requires high quality, monoisotopic reference standards that are typically short lived, and therefore must be produced and shipped promptly to monitoring stations worldwide. This work describes the preparation of 135Xe by neutron capture (n, ?) on monoisotopic 134Xe, neutron irradiation (n,2n) on monoisotopic 136Xe, and high energy photon irradiation (?, n) on monoisotopic 136Xe. The targets are prepared by electromagnetic isotope separation of natural xenon. Results from each process are compared for 135Xe yield and purity.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2022

Advanced Graphite Creep Uncertainty Analysis

Radiation damage estimation is an important component of the post irradiation analysis of the Advanced Graphite Creep (AGC) experiment. It depends primarily on the fast fluence, which is determined using well established methods of spectral adjustment. These are based on best estimates from models such as Monte Carlo N-Particle (MCNP), input cross-sections, and measured activities from flux wires in the experiment. Each of these parameters can propagate uncertainties which will affect the uncertainty in the calculated dose levels for AGC, or any experiment irradiated within a reactor. While the methods of propagating uncertainty are well-established, the final uncertainty estimates they provide are only as good as the estimates of uncertainty in the inputs on which they are based. The purpose of this work is to outline some deficiencies in the ways these input uncertainties are presently estimated, and to outline a methodology by which they can be improved. The fast fluence and radiation damage received by graphite specimens irradiated in the Advanced Graphite Creep (AGC) experiments is presently estimated using spectral adjustment methods that are based on both flux wire activity measurements, and MCNP model predictions. This work describes an ongoing effort to quantify and propagate uncertainties in inputs to the spectral adjustment process, and thereby quantify the resultant error in radiation damage (dpa) estimates. The effort is multi-faceted, and we consider the impacts of both the set of flux wires selected, and the counting process. An expanded set of flux wires is identified that provides a more comprehensive data set on the fast spectrum. To address the counting process itself, a series of round-robin measurements in several reactor metrology laboratories across the Department of Energy (DOE) complex and nuclear industry are being undertaken to refine the American Society for Testing and Materials (ASTM) standards for flux wire measurements. To address the contribution of uncertainty in the MCNP model predictions, an uncertainty quantification (UQ) tool has been developed that statistically samples the model input parameters, runs a series of cases, and assimilates the results to provide an overall uncertainty. The impact of the MCNP UQ tool results is demonstrated by re-analyzing previous AGC flux wire and irradiation data. While the expanded flux wire set obviously cannot be added to these experiments retroactively, plans for future graphite irradiations are outlined.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Fact Sheet - Long-length Scintillating Fibers for Radiation Detection

As the world looks to underground geological repositories for storing nuclear waste, technologies to safeguard the material are needed. An example of a project that developed a new technology is Tripwire. Tripwire proposes a multi-sensor system approach for geological spent nuclear fuel repositories that relies of radiation, vibration, and electromagnetic detection. Tripwire was developed by the Applied Radiation Measurements and Systems (ARMS) group at INL and sponsored by the National Nuclear Security Administration (NNSA). The ARMS group performs research and development, testing and evaluation, operational support, and training focused on applied ionizing radiation detection and measurement. Application areas include nuclear engineering for advanced reactors and fuel cycle operations, nuclear nonproliferation, nuclear counterproliferation, nuclear forensics, and arms control and disarmament.

46 - INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AN↗
DOE OSTI2017

Computer-based human-machine interfaces for emergency operation - 233

The Halden Reactor Project (HRP) has many years of experience in developing and evaluating principles for computer-based human-machine interfaces (HMIs). We have synthesized the lessons learned in an integrated design concept that is implemented on a boiling water reactor (BWR) simulator in the Halden Man Machine Laboratory (HAMMLAB). The design concept is intended to alert and support the operators in monitoring and controlling the plant during normal operation, emergency situations and outages. The computer-based HMIs include an alarm system, operating displays, operating procedures and large screen displays. The design concept has been tested in a simulator study in HAMMLAB. This paper describes the design and initial testing of the computer-based emergency operating procedures (EOPs) and other displays to support the operators when using the EOPs. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Development of Performance Validation Tool for the NPCS for APR1400

Availability of nuclear power plant is highly dependent on the quality of major control systems including NSSS Control Systems. It has been experienced that a small flaw hidden among the control functions resulted in large operational transient or even reactor trip. The proactive quality control of these systems is not easy because of the dynamical characteristics of control algorithms. In a stand-alone or open loop state, it is not possible to perform integral system tests because they need control loads to form an appropriate closed control loop. In the construction plant, Power Ascension Test (PAT) is the first chance to witness the overall performance of the control systems. Because the PAT is the last activity of plant construction schedule, the cost of corrective actions is very high if any functional defect is found in this stage. To minimize risk caused by late detection of control system defects, we developed a validation tool as a proactive test device for the NSSS Process Control System (NPCS) for the SKN Unit 3 which is the first APR1400 plant in Korea. In this paper, the features of the dynamic performance test tool, test methodology, and the test results are described.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Benefits of digitalizing and employing simulation to increase plant system performance and ensure compliance with technical specifications - 109

The essential services water system in a BWR nuclear power plant was affected by corrosion, soiling and blockages due to ageing. Those episodes affected safety margins hindering the Technical Specifications compliance during periodic system testing. Tecnatom designed, implemented and tested a design modification for the on-line monitoring of this system by integrating it in the existing Digital Control System (DCS). Key system parameters were acquired in real time to be displayed in the Human System Interface, used in performing calculations and stored their historical evolution. The objective was to optimize monitoring and surveillance of the essential service water system. Furthermore, Tecnatom developed an engineering simulator ('What if' simulator) consisting of a hydraulic model of the system. This simulator takes as inputs each heat exchanger performance parameters from the plant DCS, allows the user to change the essential services water system configuration (valves position, UHS level...) and calculates theoretical process values predicting the system real behavior. In a second stage, more system instruments were wired to the DCS for Technical Specifications compliance automatic surveillance regarding opening and closing times of system valves. Operating experience shows important benefits after the digitalization. It enhances on-line information of system operation and automates calculation of load loss factor of each heat exchanger. It eases the evaluation of blockages and soiling in the system allowing optimization of cleaning proceedings. System Surveillance Requirements and Test Requirements reports have been automated optimizing the operation workload. The simulator use enables to identify optimal system configuration increasing safety margins of the system and ensuring Technical Specifications fulfillment. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Digital engineering implementation in nuclear demonstration and nonproliferation projects at Idaho National Laboratory

Digital engineering and digital twins are increasingly being used in nuclear energy projects with important impacts. At Idaho National Laboratory, these approaches have been applied in a variety of nuclear energy research, development, and demonstration projects, with key lessons and evolutions occurring for each. In this paper, we describe the use of digital engineering and digital twins in the Versatile Test Reactor design, National Reactor Innovation Center test beds, and nonproliferation analysis of the AGN-201 reactor design. We share key lessons learned for these projects related to tool selection, adoption and training, and working with existing assets versus beginning at the design phase. We also share highlights of future potential uses of digital twins and digital engineering, including using artificial intelligence to perform repetitive design tasks and digital twins to move towards semiautonomous nuclear power plant operations.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗