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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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All sourcesNASA NTRSDOE OSTI
DOE OSTI2013

Fukushima Daiichi Information Repository FY13 Status

The accident at the Fukushima Daiichi nuclear power station in Japan is one of the most serious in commercial nuclear power plant operating history. Much will be learned that may be applicable to the U.S. reactor fleet, nuclear fuel cycle facilities, and supporting systems, and the international reactor fleet. For example, lessons from Fukushima Daiichi may be applied to emergency response planning, reactor operator training, accident scenario modeling, human factors engineering, radiation protection, and accident mitigation; as well as influence U.S. policies towards the nuclear fuel cycle including power generation, and spent fuel storage, reprocessing, and disposal. This document describes the database used to establish a centralized information repository to store and manage the Fukushima data that has been gathered. The data is stored in a secured (password protected and encrypted) repository that is searchable and available to researchers at diverse locations.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2014

AGC-3 Experiment Irradiation Monitoring Data Qualification (Final Report)

The Graphite Technology Development Program will run a series of six experiments to quantify the effects of irradiation on nuclear grade graphite. The third experiment, Advanced Graphite Creep 3 (AGC 3), began with Advanced Test Reactor (ATR) Cycle 152B on November 27, 2012, and ended with ATR Cycle 155B on April 23, 2014. This report documents qualification of AGC 3 experiment irradiation monitoring data for use by the Very High Temperature Reactor (VHTR) Technology Development Office (TDO) Program for research and development activities required to design and license the first VHTR nuclear plant. Qualified data meet the requirements for data collection and use as described in the experiment planning and quality assurance documents. Failed data do not meet the requirements. Trend data may not meet the requirements, but may still provide some useable information. All thermocouples (TCs) functioned throughout the AGC 3 experiment. There was one interval between December 18, 2012, and December 20, 2012, where 10 NULL values were reported for various TCs. These NULL values were deleted from the Nuclear Data Management and Analysis System database. All temperature data are Qualified for use by the VHTR TDO Program. Argon, helium, and total gas flow data were within expected ranges and are Qualified for use by the VHTR TDO Program. Total gas flow was approximately 50 sccm through the AGC 3 experiment capsule. Helium gas flow was briefly increased to 100 sccm during ATR shutdowns. At the start of the AGC 3 experiment, moisture in the outflow gas line was stuck at a constant value of 335.6174 ppmv for the first cycle (Cycle 152B). When the AGC 3 experiment capsule was reinstalled in ATR for Cycle 154B, a new moisture filter was installed. Moisture data from Cycle 152B are Failed. All moisture data from the final three cycles (Cycles 154B, 155A, and 155B) are Qualified for use by the VHTR TDO Program.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Phenomenon Identification and Ranking Table Analysis for Thermal Energy Storage Technologies Integration with Advanced Nuclear Reactors

This report provides an overview of the Phenomena Identification and Ranking Table (PIRT) analysis of thermal energy storage (TES) systems for possible integration with various types of advanced nuclear power plants (NPPs). Advanced NPPs will potentially need to operate in environments where power generation flexibility is more highly valued than the stability or baseload generation capability for conventional demand curves. TES systems would enable NPPs to respond nimbly to market variability and could also position advanced NPPs to participate differently in restructured markets, thus further enhancing their economic competitiveness. TES systems could also benefit the electric grid by eliminating the need for peaking plants, as well as by improving the economic performance of baseload NPPs. While TES technologies afford a unique opportunity to address many of these challenges, the applicability of these systems is also complicated by the fact that various advanced NPPs are designed differently, each with its own temperature range, size, operating fluids, and operating conditions. Hence, TES systems face significant barriers to investment, as more information on their compatibility and performance metrics is needed to quantify the advantages provided by each, as well as the challenges these technologies might face if coupled with a particular type of advanced NPP. This report explores the possibility of integrating a wide variety of TES technologies with various categories of advanced NPPs, based on their operating characteristics. To help users and developers decide which TES technology is best suited to a particular category of advanced NPPs, this research developed a PIRT of 10 TES systems that could potentially be coupled with advanced NPPs, which themselves are divided into nine categories based on their operating conditions. Each advanced NPP category is evaluated for compatibility with the 10 TES systems by assembling and discussing a database of information concerning 10 engineering questions (defined herein in as figures of merit [FOMs]), such as: technology readiness level (TRL), temperature compatibility, energy density, size, cycle frequency, ramp time, realignment frequency, geographic needs, environmental impact, and interventions. By assembling a database of information concerning the TES technologies’ compatibility with various advanced NPP systems, this study can help developers acquaint themselves with a particular TES technology before choosing to build a new integrated installation.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

FY21 Status Report on the Molten Salt Thermal Properties Database (MSTDB) Development

This report describes the thermodynamic assessments for the ZrF 4 -BeF 2 , AlCl 3 -NaCl, AlCl3-KCl and the PuCl 3 -CsCl pseudo-binary subsystems. Calculation results are compared to the inputs used to optimize the adjustable model parameters. The data came from experimental studies reported in the open literature, from the MSR research community, and/or computational results generated within the NEAMS program and from collaborators. The models from the thermodynamic assessment of these salt systems are integrated into the MSTDB-TC (Molten Salt Thermal Properties Database-Thermochemical). A brief description of the companion MSTDB-TP (Molten Salt Thermal Properties Database-Thermophysical) is given along with an overview of the approach for predicting both thermodynamic and thermophysical property behavior of multicomponent systems.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

PRO-X Research Reactor Database Status

The PRO-X Research Reactor Database is a functioning tool that can be used to quickly view attributes of past, current, and future research reactors. It contains all data found in the IAEA Research Reactor Database and entries generated from the M3 fuel conversion program. Additionally, it can be downloaded to any computer with the ability to run Microsoft ACCESS. Several research reactor attributes visible to the user have a limited amount of information, much of which is available and vetted for entry into the database. However, the necessary resources have not been allocated to add this information to the data tables. A change management system has also been developed to track changes to the database. This would need to be incorporated into the database program prior to data table modification in order to ensure proper management of the changes.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Mapping thermal energy storage technologies with advanced nuclear reactors

Advanced nuclear power plants (NPPs) will potentially need to operate in environments where power generation flexibility is more highly valued than the stability or baseload generation capability for conventional demand curves. Thermal energy storage (TES) systems would enable NPPs to respond nimbly to market variability and could also position advanced NPPs to participate differently in restructured markets, thus further enhancing their economic competitiveness. TES systems could also benefit the electric grid by eliminating the need for peaking plants, as well as by improving the economic performance of baseload NPPs. While TES technologies afford a unique opportunity to address many of these challenges, the applicability of these systems is also complicated by the fact that various advanced NPPs are designed differently, each with its own temperature range, size, operating fluids, and operating conditions. Hence, TES systems face significant barriers to investment, as more information on their compatibility and performance metrics is needed to quantify the advantages provided by each, as well as the challenges these technologies might face if coupled with a particular type of advanced NPP. This study explores the possibility of integrating a wide variety of TES technologies with various categories of advanced NPPs, based on their operating characteristics. To help decision makers, users and developers decide which TES technology is best suited to a particular category of advanced NPPs, this research present a Phenomena Identification and Ranking Table (PIRT) analysis of 10 TES systems that could potentially be coupled with advanced NPPs, which themselves are divided into nine categories based on their operating conditions. Then, each advanced NPP category is evaluated for compatibility with the 10 TES systems by assembling and discussing a database of information concerning 10 engineering questions, defined herein in as figures of merit (FOMs), such as: technology readiness level (TRL), temperature compatibility, energy density, size, cycle frequency, ramp time, realignment frequency, geographic needs, environmental impact, and interventions. By assembling a database of information concerning the TES technologies' compatibility with various advanced NPP systems, this study can help developers acquaint themselves with a particular TES technology before choosing to build a new integrated installation.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

M4FT-16LL080302052-Update to Thermodynamic Database Development and Sorption Database Integration

This progress report (Level 4 Milestone Number M4FT-16LL080302052) summarizes research conducted at Lawrence Livermore National Laboratory (LLNL) within the Argillite Disposal R&D Work Package Number FT-16LL08030205. The focus of this research is the thermodynamic modeling of Engineered Barrier System (EBS) materials and properties and development of thermodynamic databases and models to evaluate the stability of EBS materials and their interactions with fluids at various physico-chemical conditions relevant to subsurface repository environments. The development and implementation of equilibrium thermodynamic models are intended to describe chemical and physical processes such as solubility, sorption, and diffusion.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Consolidating Data on Spent Nuclear Fuel into a Unified Database

This paper describes the comprehensive commercial spent nuclear fuel (SNF) database, the Unified Database (UDB), and discusses the UDB's (1) content and design with limited data examples and potential applications, (2) external interfaces, and (3) planned enhancements to streamline national SNF data collection and preservation. The UDB is the nucleus of the Used Nuclear Fuel Storage, Transportation and Disposal Analysis Resource and Data System (UNF-ST and DARDS). The UDB is being developed for the US Department of Energy Office of Nuclear Energy (DOE-NE) to provide a credible, controlled data source for key information for: various analytical purposes, including assessing issues and uncertainties related to the extended storage and transportability of SNF after extended storage; supporting research and development (R and D) prioritization; and preserving SNF related information for future generations. The UDB will serve as a national resource for use by DOE-NE or a future management and disposal organization in planning for future storage, transport, and disposal of SNF. The UDB is also being designed with the flexibility to provide information on greater-than-class C (GTCC) waste, government-owned SNF, and high-level radioactive waste (HLW). (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Evaluation of Used Fuel Disposition in Clay-Bearing Rock

Deep geological disposal of nuclear waste in clay/shale/argillaceous rock formations has received much consideration given its desirable attributes such as isolation properties (low permeability), geochemically reduced conditions, slow diffusion, sorbtive mineralogy, and geologically widespread (Jové Colón et al., 2014). There is a wealth of gained scientific expertise on the behavior of clay/shale/ argillaceous rock given its focus in international nuclear waste repository programs that includes underground research laboratories (URLs) in Switzerland, France, Belgium, and Japan. Jové Colón et al. (2014) have described some of these investigative efforts in clay rock ranging from site characterization to research on the engineered barrier system (EBS). Evaluations of disposal options that include nuclear waste disposition in clay/shale/argillaceous rock have determined that this host media can accommodate a wide range of waste types. R&D work within the Used Fuel Disposition Campaign (UFDC) assessing thermal effects and fluid-mineral interactions for the disposition of heat-generating waste have so far demonstrated the feasibility for the EBS and clay host rock to withstand high thermal loads. This report represents the continuation of disposal R&D efforts on the advancement and refinement of coupled Thermal-Hydrological-Mechanical-Chemical (THMC), hydrothermal experiments on clay interactions, used fuel degradation (source term), and thermodynamic modeling and database development. The development and implementation of a clay/shale/argillite reference case described in Jové Colón et al. (2014) for FY15 will be documented in another report (Mariner et al. 2015) – only a brief description will be given here. This clay reference case implementation is the result of integration efforts between the GDSA PA and disposal in argillite work packages. The assessment of sacrificial zones in the EBS is being addressed through experimental work along with 1D reactive-transport and reaction path modeling. The focus of these investigations into the nature of sacrificial zones is to evaluate the chemical effects of heterogeneous chemical reactions at EBS interfaces. The difference in barrier material types and the extent of chemical reactions within these interfacial domains generates changes in mineral abundances. These mineralogical alterations also result in volume changes that, although small, could affect the interface bulk porosity. As in previous deliverables, this report is structured according to various national laboratory contributions describing R&D activities applicable to clay/shale/argillite media.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Advanced Sensors and Instrumentation Issue 10

The Advanced Sensors and Instrumentation (ASI) program, on its seventh fiscal year, continues to address key technology gaps for future nuclear energy systems. ASI’s goal is to enable the advanced instrumentation and control (I&C) technologies essential to the Office of Nuclear Energy’s research and development (R&D) efforts needed to realize its mission goals. Information on our projects can be found in our ASI annual awards summaries, annual review webinars, and newsletters posted in the Office of Nuclear Energy website (energy.gov/ne). This year, we have initiated three newly awarded projects in the area of data analytics. We also initiated a project to develop and demonstrate a prototype passive wireless sensor network. This sensor will use advanced manufacturing and printing, and be tested and demonstrated at a U.S. nuclear power plant facility. In addition, we are developing a sensor database to capture advanced sensor and instrumentation technology that is applicable to nuclear power plants. Under the In-Pile Instrumentation program, we continue to develop unique in-pile sensors and instrumentation to provide real-time, accurate, and spatially resolved information regarding the performance of fuels and materials. Due to the closure of the Halden reactor, we are focusing our near-term research in providing equivalent capability for the instrumentation of irradiation experiments in the U.S. that are needed by the nuclear industry. As the ASI program continues to award projects through competitive solicitations, we encourage the I&C community to visit our website for current funding opportunity announcements, especially for the industry-focused solicitation entitled, “U.S. Industry Opportunities for Advanced Nuclear Technology Development.”

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2015

High Fidelity Forward Model Development for Nuclear Reactor Spent Fuel Technical Nuclear Forensics

The fidelity of the forward model within a spent fuel forensic analysis system was improved by using two unique methodologies. The first consisted of developing a system to create accurate one-group neutron cross-section libraries for any user specified reactor system. In such, a detailed model is developed using the depletion code MONTEBURNS. During MONTEBURNS execution, cross-section libraries are generated at every user specified burnup step in time. These libraries could be developed for many reactor systems, then housed in a database and used for analyzing unknown fuel samples. The forensic analysis system for spent fuel resulted in higher accuracy at predicting the initial uranium isotopic compositions and burnup from spent fuel samples. Using this method, the error in results was reduced from the order of 1-6% down to less than 1% when recovering a fuel sample's burnup and initial uranium isotopic composition. The second method consisted of implementing 2D/3D reactor depletion codes as the forward model within the system's framework. This method would allow the usage of potentially recoverable geometric information from an unknown sample. No predetermined cross-section library is required for the system using this method, therefore potentially reducing model error associated with the neutron flux spectrum. The accuracy of the recovered initial uranium isotopic compositions and burnup from spent fuel samples was also improved using this method, even more so than the first. For MTR reactors, the error using this method was significantly reduced and was driven to below 0.5%. However, additional research may be required to determine the ideal fission yield and recoverable energy per fission for cases where significant amounts of 239 PU are bred and burned throughout the life of the fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Novel Platform for Digitization of Radiation Measurements, Decommissioning and Dismantling and Nuclear Waste Management - 18229

An innovative Platform for a seamless digitization of the information produced during the characterization of radioactive waste, contaminated objects, and in general radioprotection measurements has ben developed by CAEN SyS. The Platform is based on a spectroscopic handheld instrument integrating RFID technology, barcode reader, camera, voice recorder, and geo-localization, a selection of radiation tolerant RFID tags tested above 300 Gy integrated dose, ruggedized portable RFID readers and a secure database framework capable to collect data also from external sources such as gamma scanners and imaging systems. The Platform allows for real-time measurement of dose rate, radiation spectroscopy, tagging, tracking and inventorying of items. All data produced by the Platform are securely stored into the non-volatile memory of the RFID tags and in the database for both onsite and remote access to the information by operators. This paper presents the current development of the Platform and shows the advantages of such approach to achieve an easy yet reliable waste traceability, waste bagging optimization, and a structured methodology for radiation measurements, meeting an excellent quality control as increasingly required by Safety Authorities. Applications here envisaged and explored are dismantling and decontamination operations, radioprotection measurements and patrols, nuclear material management, safeguards and verification inspections. More in general, the Platform can address the needs of all plants, research laboratories and facilities that are looking to improve their Safety Standards, decrease operational costs, have a precise knowledge of the contaminated environment and objects as well as optimization for personnel dose minimization. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Research and Test Reactor Fuels

PRO-RR is the research reactor focused program element of the broader Proliferation Resistance Optimization program (PRO-X) under the National Nuclear Safety Administration (NNSA) in the U.S. Department of Energy (DOE). PRO-X provides a framework for integrating proliferation resistance in nuclear system designs to minimize weapons usable nuclear materials (WUNM) production and diversion pathways while optimizing systems performance for peaceful use missions. PRO-RR applies the PRO-X mission objectives to research reactor system design. This document serves as one of the foundational documents for the PRO-RR-Fuel System Design technical team by documenting current research reactor fuels usage. The PRO-RR-Fuel System Design technical team consists of subject matter experts from Argonne National Laboratory (Argonne) and Savannah River National Laboratory (SRNL). In order to determine the preferred fuel of use in upcoming research and test reactors to optimize proliferation resistance, performance, and safety, it is useful to assess the fuels that have been used in the past, or are currently in use. This report reviews the historical and current fuels used in research and test reactors to inform future fuel selection. Chapter 2 discusses the low-enriched uranium (LEU) fuels currently in use in terms of thermal power level and utilization of the reactor. Chapter 3 summarizes the fabrication processes for common fuel types. Chapter 4 discusses in detail the fuel types in use in research and test reactors. A review of the cladding types in use is presented in Chapter 5, and a historical review of research and test reactor fuel fabricators is presented in Chapter 6. The data collection strategy used the International Atomic Energy Agency (IAEA) research reactor database [1] as a starting point. Information on the fuel used was gathered on research reactors (other than critical assemblies) that were listed as operational, planned, or in temporary shutdown in the IAEA database. Data on the fuel type, geometry, enrichment, uranium loading, cladding type, and fabricator were collected for each of the reactors available in the public domain. Sources of data included conference papers, journal articles, and facility and fabricator websites. Data on research reactors operating on LEU fuels are presented in Appendix A, while Appendix B presents data collected on all reactors at the time of publication of this report. Appendix C presents data collected on reactors that were part of the M3 research and test reactor conversion program.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2014

Nuclear Science References Database

The Nuclear Science References (NSR) database together with its associated Web interface, is the world's only comprehensive source of easily accessible low- and intermediate-energy nuclear physics bibliographic information for more than 210,000 articles since the beginning of nuclear science. The weekly-updated NSR database provides essential support for nuclear data evaluation, compilation and research activities. The principles of the database and Web application development and maintenance are described. Examples of nuclear structure, reaction and decay applications are specifically included. The complete NSR database is freely available at the websites of the National Nuclear Data Center (http://www.nndc.bnl.gov/nsr) and the International Atomic Energy Agency (http://www-nds.iaea.org/nsr)

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2015

SU-E-T-241: Design and Implement of An Information Management System for Radiation Workers in a Hospital

Purpose: To meet the special needs of Medical Administration Division, an information management system for radiation workers in a hospital(IMSRWH) has been developed. Methods: There are about 200 radiation workers in 20 departments, such as department of radiology, department of radiation oncology, department of nuclear medicine, and so on. An IMSRWH server was used to run a database and web service of Apache+PHP+MySQL. The IMSRWH web service could be accessed with Internet Explorer(IE) to input, search, count, and print the radiation workers’ data. Every worker’s data information includes the following tables: registration, personal dose monitoring records, personnel certificate, training certificate, personnel license, medical examination, etc, and each table includes more optional detail information. Results: In the past year, the IMSRWH has been successfully developed and implemented in the hospital. According to the needs, the user rights are divided into: the system administrator, office manager, department manager and radiation worker. The different users have the different interfaces, and the appropriate authority permissions mean corresponding operations. The administrator and office manager could manage all personnel information, department manager could manage the staff information of his department, and the workers could view or modify their own information. Because information updates, involving hundreds of each entry and modification are complicated, each table information of all workers can be batch import from or export to a Microsoft Excel CSV format file, to update all together. However, it has only Chinese version now. Conclusion: It was demonstrated to be user-friendly and was proven to significantly improve the efficiency of the hospital management. Since it is an in-house developed system, more functions can be added or modified to further enhance its potentials in research and clinical practice.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2020

MECHANICAL SHOCK AND VIBRATION ANALYSIS OF SPENT NUCLEAR FUEL CARRIED BY THE ATLAS RAILCAR

Researchers at Pacific Northwest National Laboratory have completed a structural-dynamic analysis of spent nuclear fuel subjected to the mechanical shock and vibration environment that is anticipated during normal conditions of transport in casks carried by the Atlas railcar. The Atlas railcar is a new railcar design that is being developed specifically for the purpose of carrying spent nuclear fuel casks. The analysis used best-estimate railcar dynamics models of the Atlas railcar and considered 17 different spent nuclear fuel transportation cask systems, representing the current fleet of cask options. This work used NUCARS, a specialized railcar dynamics explicit finite element code to calculate railcar dynamic response to prescribed speeds and track configurations. The railcar dynamics models provided cask transient motion for a wide range of speeds and track conditions, generating a relatively large database of potential cask motion. All of the cask motion transients were then applied as loading conditions to LS-DYNA structural-dynamic models of a single fuel rod. The analyses predict that the Equipos Nucleares S.A./U.S. Department of Energy (ENSA/DOE) multimodal transportation test of 2017 provided a relatively stronger vibration environment than is expected from the Atlas railcar. This paper describes the analysis methods, the analysis results, and compares the results of the Atlas transportation analysis to the test results and analyses of the ENSA/DOE multimodal transportation test of 2017.

Klymyshyn, Nicholas A.↗
DOE OSTI2020

M4SF-20LL010301042: Updated Thermodynamic Database for Use in Generic Disposal System Assessment

This progress report (Level 4 Milestone Number M4SF-20LL010301042) summarizes research conducted at Lawrence Livermore National Laboratory (LLNL) within the Argillite Activity Number SF-20LL010301041. LLNL is leading efforts in the development of thermodynamic databases in support of the Spent Fuel and Waste Science Technology (SFWST) program. Thermodynamic models provide the basis for understanding the stability of solid phases and speciation of aqueous species and modeling the evolution of repository conditions. The LLNL effort is being performed in coordination with other US database development efforts. The effort includes a review of available thermochemical databases and a path forward for database integration. International coordination with the NEA-TDB is supported through crystalline international work package.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Standardized Procedure Content And Data Structure Based On Human Factors Requirements For Computer-Based Procedures

Most activities that involve human interaction with systems in a nuclear power plant are guided by procedures. Traditionally, the use of procedures has been a paper-based process that supports safe operation of the nuclear power industry. However, the nuclear industry is constantly trying to find ways to decrease the human error rate, especially the human errors associated with procedure use. Advances in digital technology make computer-based procedures (CBPs) a valid option that provides further enhancement of safety by improving human performance related to procedure use. The transition from paper-based procedures (PBPs) to CBPs creates a need for a computer-based procedure system (CBPS). A CBPS needs to have the ability to perform logical operations in order to adjust to the inputs received from either users or real time data from plant status databases. Without the ability for logical operations the procedure is just an electronic copy of the paper-based procedure. In order to provide the CBPS with the information it needs to display the procedure steps to the user, special care is needed in the format used to deliver all data and instructions to create the steps. The procedure should be broken down into basic elements and formatted in a standard method for the CBPS. One way to build the underlying data architecture is to use an Extensible Markup Language (XML) schema, which utilizes basic elements to build each step in the smart procedure. The attributes of each step will determine the type of functionality that the system will generate for that step. The CBPS will provide the context for the step to deliver referential information, request a decision, or accept input from the user. The XML schema needs to provide all data necessary for the system to accurately perform each step without the need for the procedure writer to reprogram the CBPS. The research team at the Idaho National Laboratory has developed a prototype CBPS for field workers as well as the underlying data structure for such CBPS. The objective of the research effort is to develop guidance on how to design both the user interface and the underlying schema. This paper will describe the result and insights gained from the research activities conducted to date.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗