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All sourcesNASA NTRSDOE OSTI
DOE OSTI1994

Nuclear Safety [Vol. 35, No. 2, July-December 1994]

Nuclear Safety is a journal that covers significant issues in the field of nuclear safety. Its primary scope is safety in the design, construction, operation, and decommissioning of nuclear power reactors worldwide and the research and analysis activities that promote this goal, but it also encompasses the safety aspects of the entire nuclear fuel cycle, including fuel fabrication, spent-fuel processing and handling, and nuclear waste disposal, the handling of fissionable materials and radioisotopes, and the environmental effects of all these activities. Table of Contents for this issue follows. GENERAL SAFETY CONSIDERATIONS: 179 Consideration of Postaccident Consequences in the Determination of Safety Objectives for Future Nuclear Power Plants in France, D. Queniart, A. Sugier, and J. Lochard; ACCIDENT ANALYSIS: 187 Nuclear Safety Research: The Phebus FP Severe Accident Experimental Program, P. von der Hardt, A. V. Jones, C. Lecomte, and A. Tattegrain; 205 Containment Performance Analysis of the Advanced Neutron Source Reactor at the Oak Ridge National Laboratory, S. H. Kim, R. P. Taleyarkhan, and V. Georgevich; 213 Assessment of Fission Product Deposits in the Reactor Coolant System: The DEVAP Program, G. Le Marois and M. Megnin; 222 Erratum to “A Review of the Available Information on the Triggering Stage of a Steam Explosion," Vol. 35, No. 1, CONTROL AND INSTRUMENTATION: 223 Effects of Normal Aging on Calibration and Response Time of Nuclear Plant Resistance Temperature Detectors and Pressure Sensors, H. M. Hashemian; DESIGN FEATURES: 235 Defense in Depth Against the Hydrogen Risk—A European Research Program, F. Fineschi; ENVIRONMENTAL EFFECTS: 246 Technical Note: A Preliminary Analysis of the Risks to Hong Kong Resulting from Potential Accidents of Daya Bay Nuclear Power Plant, Z. Shi and X. Wei; OPERATING EXPERIENCES: 253 Reactor Shutdown Experience, Compiled by J. W. Cletcher; SPECIAL SECTION ON TMI-2 VESSEL INVESTIGATION PROJECT: 256 Three Mile Island—New Findings 15 Years After the Accident A. M. Rubin and E. Beckjord; 269 Relocation of Molten Material to the TMI-2 Lower Head, J. R. Wolf, D. W. Akers, and L. A. Neimark; 280 Insight Into the TMI-2 Core Material Relocation Through Examination of Instrument Tube Nozzles, L. A. Neimark; 288 Physical and Radiochemical Examinations of Debris from the TMI-2 Lower Head, D. W. Akers and B. K. Schuetz; 301 Results of Metallographic Examinations and Mechanical Tests of Pressure Vessel Samples from the TMI-2 Lower Head, D. R. Diercks and G. E. Korth; 313 Margin-to-Failure Calculations for the TMI-2 Vessel, J. Rempe, L. Stickler, S. Chavez, G. Thinnes, R. Witt, and M. Corradini; U.S. NUCLEAR REGULATORY COMMISSION INFORMATION AND ANALYSES: 328 1993 Accident Sequence Precursor (ASP) Program Results, L. N. Vanden Heuvel, J. W. Cletcher, D. A. Copinger, J. W. Minarick, B. W. Dolan, and P. D. O’Reilley; RECENT DEVELOPMENTS: 339 Reports, Standards, and Safety Guides, D. S. Queener; 345 Proposed Rule Changes as of June 30, 1994; ANNOUNCEMENTS: 234 Thirty-First Annual Meeting of the National Council on Radiation Protection and Measurements; 356 1995 International Incineration Conference; 358 Fifth International Controls and Instrumentation Conference; 359 ANS International Topical Meeting on Safety of Operating Reactors; 359 Fifth International Conference on Nuclear Criticality Safety; 360 International Conference on Probabilistic Safety Assessment Methodology and Applications; 351 The Authors; 357 Reviewers of Nuclear Safety, Vol. 35.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Computational and Experimental Benchmarking for Transient Fuel Testing

In 2018 the Transient Reactor Test (TREAT) Facility was restarted. This now-operational facility brings back world-class transient fuel testing capabilities to the U.S. Among a number of parallel efforts prior to the TREAT Facility’s restart a Department of Energy (DoE) Integrated Research Project (IRP) was funded for the purpose of benchmarking the TREAT with several goals in mind: To produce a comprehensive evaluation of existing TREAT Facility neutronics data using the next generation reactor core neutronics codes in accordance with established guidelines per the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP). To perform a benchmark using ready-use computational fluid dynamics tools against that of a historically collected sodium loop experiment collected in the TREAT. To design, develop and test a new experimental thermal hydraulic system of a representative TREAT Facility water flow loop. And benchmark the data collected against that produced via RELAP5-3D and TRACE. To develop a comprehensive instrumentation plan for the TREAT Facility by benchmarking performing of advanced in-pile instruments in the Massachusetts Institute of Technology Research Reactor (MITR) and the TREAT. The integrated team brought together collaborating members from three universities including Oregon State University (OSU), the University of Michigan (UM), and Massachusetts Institute of Technology (MIT); three national laboratories, Idaho National Laboratory, Argonne National Laboratory (ANL), and Oak Ridge National Laboratory (ORNL); and two industry partners, TerraPower (TP) and Harris Thermal Transfer Products (HTTP). A brief summary of accomplishments resulting from this IRP includes two filed patents, more than 40 conference papers, two published benchmark reports through IRPhEP, a new representative thermal hydraulic test facility alike the in-pile PWR water flow loop for TREAT and the first advanced instruments tested within the TREAT core upon its restart. The papers presented herein were developed and provide a summary of the significant technical accomplishments which align with the four project’s objectives. The collaborators has significant support from countless parties including the operational and technical staff at TREAT to ensure that their outcomes most impactfully aligned with program needs. Lastly the collaborators would like to acknowledge the fiscal support of the DoE’s Office of Nuclear Energy under work package identification number NU-15-OR-OSU- 0701-01. This report contains selected papers from the Computational and Experimental Benchmarking for Transient Fuel Testing Integrated Research Project.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

High-resolution phonon energy shift measurements with the inelastic neutron spin echo technique

The energy resolution of the conventional way of measuring a small change in a phonon dispersion curve using neutron scattering is restricted by the relatively coarse intrinsic resolution ellipsoid of the neutron triple-axis spectrometer (TAS). By implementing inelastic neutron spin echo on the host TAS using the Larmor precession of the neutron spin, the energy resolution of such measurements can be further improved without reducing the resolution ellipsoid. Measurements of the temperature-dependent phonon energy change are demonstrated using superconducting magnetic Wollaston prisms at the HB-1 instrument of the High-Flux Isotope Reactor at Oak Ridge National Laboratory, and the achievable resolution is <10 µeV.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗
NASA NTRS1981

Low cost solar aray project: Experimental process system development unit for producing semiconductor-grade silicon using the silane-to-silicon process

This phase consists of the engineering design, fabrication, assembly, operation, economic analysis, and process support R&D for an Experimental Process System Development Unit (EPSDU). The mechanical bid package was issued and the bid responses are under evaluation. Similarly, the electrical bid package was issued, however, responses are not yet due. The majority of all equipment is on order or has been received at the EPSDU site. The pyrolysis/consolidation process design package was issued. Preparation of process and instrumentation diagram for the free-space reactor was started. In the area of melting/consolidation, Kayex successfully melted chunk silicon and have produced silicon shot. The free-space reactor powder was successfully transported pneumatically from a storage bin to the auger feeder twenty-five feet up and was melted. The fluid-bed PDU has successfully operated at silane feed concentrations up to 21%. The writing of the operating manual has started. Overall, the design phase is nearing completion.

Source record↗
DOE OSTI2012

CATHARE calculations of Phenix ultimate natural convection test

The Phenix Sodium cooled Fast Reactor (SFR) started operation in 1973 and it was stopped in 2009. Before the reactor was definitively shutdown, in order to collect experimental data for code assessments in the frame of Generation IV an intensive program of ultimate tests was set up. Among these ultimate experiments, two thermal hydraulic tests were performed: an asymmetrical test consisting in a trip on one secondary pump and a natural convection test in the primary circuit. The natural convection test has been used for an international benchmark on system codes in the frame of the IAEA. The CATHARE code - initially developed for water cooled reactors and now extended for safety analyses for other kinds of reactors, including Sodium Fast Reactor - was used by CEA for this benchmark. The paper reminds briefly the Phenix reactor with the main physical parameters and the instrumentation used during the natural convection test. Main test results are also briefly reminded including the evolution of the core and the heat exchangers inlet and outlet temperatures, and some local temperature measurements. The main developments to perform CATHARE SFR computations and the strategy of system code assessment are presented. Then the CATHARE modelling of Phenix reactor is depicted and the various assumptions are pointed out. CATHARE encountered no problem to predict the initial nominal state. Afterwards, the whole transient scenario is calculated and CATHARE calculations are compared to the Phenix measurements. The global trend is rather well predicted by the CATHARE code. Nevertheless, due to complex flow phenomena occurring in large plena and components, the system code encountered physical limitations, leading to remaining discrepancies between code prediction and plant data. Various sensitivity calculations are presented and they bring partial answers. Additional analyses are in progress to understand more deeply the complex 3D phenomena involved during the different phases of the natural convection test. Additional work for coupling CATHARE system code and TRIO-U CFD code is in progress and will bring useful information to better understand the physical phenomena involved during the natural convection test and to improve system modeling for future SFR safety analysis. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Identifying and bounding uncertainties in nuclear reactor thermal power calculations

Determination of the thermal power generated in the reactor core of a nuclear power plant is a critical element in the safe and economic operation of the plant. Direct measurement of the reactor core thermal power is made using neutron flux instrumentation; however, this instrumentation requires frequent calibration due to changes in the measured flux caused by fuel burn-up, flux pattern changes, and instrumentation drift. To calibrate the nuclear instruments, steam plant calorimetry, a process of performing a heat balance around the nuclear steam supply system, is used. There are four basic elements involved in the calculation of thermal power based on steam plant calorimetry: The mass flow of the feedwater from the power conversion system, the specific enthalpy of that feedwater, the specific enthalpy of the steam delivered to the power conversion system, and other cycle gains and losses. Of these elements, the accuracy of the feedwater mass flow and the feedwater enthalpy, as determined from its temperature and pressure, are typically the largest contributors to the calorimetric calculation uncertainty. Historically, plants have been required to include a margin of 2% in the calculation of the reactor thermal power for the licensed maximum plant output to account for instrumentation uncertainty. The margin is intended to ensure a cushion between operating power and the power for which safety analyses are performed. Use of approved chordal ultrasonic transit-time technology to make the feedwater flow and temperature measurements (in place of traditional differential-pressure- based instruments and resistance temperature detectors [RTDs]) allows for nuclear plant thermal power calculations accurate to 0.3%-0.4% of plant rated power. This improvement in measurement accuracy has allowed many plant operators in the U.S. and around the world to increase plant power output through Measurement Uncertainty Recapture (MUR) up-rates of up to 1.7% of rated power, while also decreasing the probability of significant over-power events. This paper will examine the basic elements involved in calculation of thermal power using ultrasonic transit-time technology and will discuss the criteria for bounding uncertainties associated with each element in order to achieve reactor thermal power calculations to within 0.3% to 0.4%. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

FY21 Progress Report for Advanced Re fabrication/Re-instrumentation Capability Development

In support of performing follow on irradiation experiments with previously irradiated materials, the Halden Reactor Project developed unique and state of the art capabilities to refabricate and re-instrument previously irradiated materials. Such materials were used in in-pile tests at the Halden reactor, and out-of-pile tests for example using furnaces as a heat source. The decision to close the Halden Reactor Project results in the loss of this refabrication and reinstrumentation capability. As a result, the United States Department of Energy has determined to develop refabrication and re-instrumentation capability at the unique shielded facilities at Idaho National Laboratory. The development of refabrication capability has been completed and demonstrated. This report focuses on the complementary aspects of re-instrumentation and the progress to date. Halden spent nearly 30 years developing both refabrication and reinstrumentation. Collaboration with Halden is allowing INL to develop this capability much more rapidly. The results include development of the capability to drill annular center holes in ceramic UO2 fuel pellets, development of fuel rod end caps with feedthroughs for centerline instrumentation inside the rodlet, The procurement of both fuel drilling and welding demonstration equipment from Halden, evaluation of surface thermocouple attachments to support better understanding of temperature measurement uncertainties, and finally, the conceptual design of a new shielded enclosure where advanced refabrication and re-instrumentation equipment can be housed. The authors would like to thank the numerous colleagues at INL and Halden who provided support in accomplishing this work. Their support both material and intellectual is invaluable in advancing the state of the art and establishing the capabilities for refabrication and re-instrumentation at INL.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

An Integrated Framework for Risk Assessment of High Safety Significant Safety-related Digital Instrumentation and Control Systems in Nuclear Power Plants: Methodology and Demonstration

This report documents the activities performed by Idaho National Laboratory (INL) during Fiscal Year (FY) 2022 for the U.S. Department of Energy (DOE) Light Water Reactor Sustainability (LWRS) Program, Risk Informed Systems Analysis (RISA) Pathway, digital instrumentation and control (DI&C) risk assessment project. In FY 2019, the RISA Pathway initiated a project to develop a risk assessment strategy for delivering a technical basis to support effective and secure DI&C technologies for digital upgrades/designs. A framework was proposed for this strategy, which aims to (1) provide a best-estimate, risk-informed capability to quantitatively and accurately estimate the risk impact of plant modernization, considering the introduction of high safety-significant safety-related (HSSSR) DI&C systems, (2) support and supplement existing risk-informed DI&C design guides by providing quantitative risk information and evidence, (3) offer a capability of design architecture evaluation of various DI&C systems, (4) assure the long-term safety and reliability of HSSSR DI&C systems, and (5) reduce uncertainty in costs and support integration of DI&C systems in the plant. To achieve these technical goals, the framework provides a means to address relevant technical issues by: (1) defining a risk-informed analysis process for DI&C upgrade, that integrates hazard analysis, reliability analysis, and consequence analysis, (2) applying risk-informed tools to address common cause failures (CCFs) and quantify corresponding failure probabilities for DI&C technologies, particularly software CCFs, (3) evaluating the impact of digital failures at the component level, system level, and plant level, and (4) providing insights and suggestions on designs to manage the risks, thus to support the development and deployment of advanced DI&C technologies on nuclear power plant (NPPs).

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Experimental design and analysis for irradiation of SiC/SiC composite tubes under a prototypic high heat flux

The purpose of this work is to design an irradiation vehicle for testing silicon carbide (SiC) fiber-reinforced SiC matrix composite cladding materials under conditions representative of a light water reactor in order to validate thermo-mechanical models of stress states in these materials due to irradiation swelling and differential thermal expansion. The design allows for a constant tube outer surface temperature in the range of 300–350 °C under a representative high heat flux (~0.66 MW/m 2 ) during one cycle of irradiation in an un-instrumented “rabbit” capsule in the High Flux Isotope Reactor. An engineered aluminum foil was developed to absorb the expansion of the cladding tubes, due to irradiation swelling, without changing the thermal resistance of the gap between the cladding and irradiation capsule. Finite-element analyses of the capsule were performed, and the models used to calculate thermal contact resistance were validated by out-of-pile testing and post-irradiation examination of the foils and passive SiC thermometry. Six irradiated cladding tubes (both monoliths and composites) were irradiated and subsequently disassembled in a hot cell. The calculated temperatures of passive SiC thermometry inside the capsules showed good agreement with temperatures measured post-irradiation, with two calculated temperatures falling within 10 °C of experimental measurements. Furthermore, the success of this design could lead to new opportunities for irradiation applications with materials that suffer from irradiation swelling, creep, or other dimensional changes that can affect the specimen temperature during irradiation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Manufactured chemistry: Rethinking unit operation design in the age of additive manufacturing

Advanced manufacturing enables new approaches to chemical reactor design due to the ability to build complex geometries and topologies unaccessible to traditional manufacturing technologies. These benefits have been demonstrated in the design and manufacture of novel heat exchangers and fluidic devices. However, due to prevailing heuristics in chemical reactor design, many of the benefits of advanced manufacturing have yet to be realized by the chemical engineer. As such, in order to reap these benefits, a formalized design methodology such as topology optimization should be employed for the development of high-efficiency small scale modular reactors capable of being manufactured at large volumes. In this way we may be able to start to descend fast manufacturing learning curves and unleash a new wave of innovation in the chemical engineering sciences.

47 OTHER INSTRUMENTATION↗
DOE OSTI2014

Effects of Levels of Automation for Advanced Small Modular Reactors: Impacts on Performance, Workload, and Situation Awareness

The Human-Automation Collaboration (HAC) research effort is a part of the Department of Energy (DOE) sponsored Advanced Small Modular Reactor (AdvSMR) program conducted at Idaho National Laboratory (INL). The DOE AdvSMR program focuses on plant design and management, reduction of capital costs as well as plant operations and maintenance costs (O&M), and factory production costs benefits.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Neutron irradiation control in the neutron transmutation doping process in HANARO using SPND

The neutron irradiation control method by using self-powered neutron detector (SPND) is developed for the neutron transmutation doping (NTD) application in HANARO. An SPND is installed at a fixed position of the upper part of the sleeve in HANARO NTD hole for real-time monitoring of the neutron irradiation. It is confirmed that the SPND is significantly affected by the in-core condition and surroundings of the facility. Furthermore, the SPND signal changes about 15% throughout a whole cycle according to the change of the control rod position. But, it is also confirmed that the variation of the neutron flux on the silicon ingots inside the irradiation can is not so big while moving of the control rod. Accordingly, the relationship between the ratio of the neutron flux to the SPND signal output and the control rod position is established. In this procedure, the neutron flux measurement by using zirconium foil is utilized. The real NTD irradiation experiments are performed using the established relationship. The irradiated neutron fluence can be controlled within ±1.3% of the target one. The mean value of the irradiation/target ratio of the fluence is 0.9992, and the standard deviation is 0.0071. Thus, it is confirmed that the extremely accurate irradiation would be accomplished. This procedure can be useful for the SPND application installed at the fixed position to the field requiring the extremely high accuracy. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Quantitative void fraction detection with an eddy current flowmeter for generation IV Sodium cooled Fast Reactor

This study was carried out to understand the response of an eddy current type flowmeter in two phase liquid-metal flow. We use the technique of ellipse fit and correlate the fluctuations in the angle of inclination of this ellipse with the void fraction. The effects of physical parameters such as coil excitation frequency and flow velocity have been studied. The results show the possibility of using an eddy current flowmeter as a gas detector for large void fractions. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Feasibility study of Self Powered Neutron Detectors in Fast Reactors for detecting local change in neutron flux distribution

Neutron flux monitoring system forms an integral part of the design of a Generation IV sodium cooled fast reactor system. Diverse possibilities of detector systems installation have to be investigated with respect to practicality and feasibility according to the detection parameters. In this paper, we demonstrate the feasibility of using self powered neutron detectors as in-core detectors in fast reactors for detecting local change in neutron flux distribution. We show that the gamma contribution from fission products decay in the fuel and activation of structural materials is very small compared to the fission gammas. Thus, it is possible for the in-core SPND signal to follow changes in local neutron flux as they are proportional to each other. This implies that the signal from an in-core SPND can provide dynamic information on the neutron flux perturbations occurring inside the reactor core. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

The human factors verification at plant startup for the AP1000 Plant - 173

The Human Factors (HF) verification at plant startup is the final HF assessment for the AP1000{sup R} HF program and is performed using the as-built plant design. This HF verification has been implemented for the China AP1000 Standard Plant. The scope and objectives of the HF verification at plant startup met the requirements for Verification and Validation (V and V) as described in NUREG-0711, 'Human Factors Engineering Program Review Model'. The majority of the verifications have been completed. Overall, the results confirmed the adequacy of the China AP1000 Standard Plant operating procedures and operator training. A number of design issues were identified and were categorized as safety issues, operability and maintainability issues or minor issues. These issues are being addressed and tracked as site test deficiencies. Based on the current results, it can be concluded that the China AP1000 Standard Plant is safe, operable and maintainable, pending completion of the outstanding verifications and resolution of the identified issues. This paper will describe the overall strategy and discuss the lessons learned in respect to providing HF support to the startup activities for the AP1000 plant. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Qualified display system architecture - 246

There is a need in the nuclear industry for a Qualified Display System (QDS) architecture as described in this paper. The function of the QDS is to process and display Regulatory Guide 1.97 post accident monitoring parameters. The QDS is implemented on the Lockheed Martin Field-Programmable Gate Array (FPGA) based Nuclear Protection and Control (NuPAC) digital platform, coupled with a Qualified Display Controller (QDC) and a Safety Display Unit (SDU). The QDS is designed to meet the requirements of IEEE 603-1991 (endorsed by Regulatory Guide 1.153, Rev. 1), IEEE 497-2002 (endorsed by Regulatory Guide 1.97, Rev. 4), and IEEE 7-4.3.2- 2003 (endorsed by Regulatory Guide 1.152, Rev. 3). The QDS is designed with flexibility such that it can be implemented in an advanced reactor design, Small Modular Reactor (SMR) with any cooling medium, or digital upgrade project for an operating nuclear plant. The QDS has four independent and redundant divisions. Each division has five layers in the architecture: Process variable input signals; Process Protection Logic (PPL); Algorithm Logic (AL), Interface Panel (IP); and Safety Display Unit (SDU). The QDS meets the following design criteria: redundancy, independence, predictability and repeatability, and diversity capability. The QDS architecture also meets the simplicity design principle. The QDS architecture design provides flexibility in the implementation in a plant. The QDS can be implemented in the design of a new advanced plant or as a digital upgrade in an operating plant. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Feature selection based on PCA and the transient identification in nuclear power plants - 289

For the sake of enhancing safety and achieving more economic benefits, it is very important to timely and correctly identify the transients during the operation of nuclear power plants. There are thousands of monitoring signals in a nuclear power plant. It is difficult to complete the calculations timely if all the related signals are chosen for an algorithm of transient identification. Moreover, the irrelevant state variables may also do harm to the correctness of the identification process. Thus, one of the key tasks for transient identification is to figure out effective methods to extract the features from state variables. In this paper, we will address an efficient method for feature selection and transient identification based on principal component analysis (PCA). The effectiveness of the method are illustrated by the tests based on the data from the simulator designed for the High Temperature gas-cooled Reactor Pebble-bed Module (HTR-PM), which is under construction in China and the construction will be completed around 2017. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Integration and assessment of component health prognostics in supervisory control systems - 291

Enhanced risk monitors (ERMs) for active components in advanced reactor concepts use predictive estimates of component failure to update, in real time, predictive safety and economic risk metrics. These metrics have been shown to be capable of use in optimizing maintenance scheduling and managing plant maintenance costs. Integrating this information with plant supervisory control systems increases the potential for making control decisions that utilize realtime information on component conditions. Such decision making would limit the possibility of plant operations that increase the likelihood of degrading the functionality of one or more components while maintaining the overall functionality of the plant. ERM uses sensor data for providing real-time information about equipment condition for deriving risk monitors. This information is used to estimate the remaining useful life and probability of failure of these components. By combining this information with plant probabilistic risk assessment models, predictive estimates of risk posed by continued plant operation in the presence of detected degradation may be estimated. In this paper, we describe this methodology in greater detail, and discuss its integration with a prototypic software-based plant supervisory control platform. In order to integrate these two technologies and evaluate the integrated system, software to simulate the sensor data was developed, prognostic models for feedwater valves were developed, and several use cases defined. The full paper will describe these use cases, and the results of the initial evaluation. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗