Nuclear Energy – Knowledge Base for Advanced Modeling and Simulation (NE-KAMS) Code Verification and Validation Data Standards and Requirements: Fluid Dynamics Version 1.0

The Nuclear Energy - Knowledge base for Advanced Modeling and Simulation (NE-KAMS) is being developed at the Idaho National Laboratory in conjunction with Bettis Laboratory, Sandia National Laboratories, Argonne National Laboratory, Oak Ridge National Laboratory, Utah State University and others. The objective of this consortium is to establish a comprehensive knowledge base to provide Verification and Validation (V&V) and Uncertainty Quantification (UQ) and other resources for advanced modeling and simulation (M&S) in nuclear reactor design and analysis. NE-KAMS will become a valuable resource for the nuclear industry, the national laboratories, the U.S. NRC and the public to help ensure the safe operation of existing and future nuclear reactors. A survey and evaluation of the state-of-the-art of existing V&V and M&S databases, including the Department of Energy and commercial databases, has been performed to ensure that the NE-KAMS effort will not be duplicating existing resources and capabilities and to assess the scope of the effort required to develop and implement NE-KAMS. The survey and evaluation have indeed highlighted the unique set of value-added functionality and services that NE-KAMS will provide to its users. Additionally, the survey has helped develop a better understanding of the architecture and functionality of these data and knowledge bases that can be used to leverage the development of NE-KAMS.

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NRC test results and operations experience provide insights for a new gate valve stem force correlation

One of America’s best kept secrets is the success of its nuclear electric power industry. This paper presents data which support the construction and operating successes enjoyed by energy companies that operate nuclear power plants in the US. The result—the US nuclear industry is alive and well. Perhaps it’s time to start anew the building of nuclear power plants. Let’s take the wraps off the major successes achieved in the nuclear power industry. Over 20% of the electricity generated in the United States comes from nuclear power plants. An adequate, reliable supply of reasonably priced electric energy is not a consequence of an expanding economy and gross national product; it is an absolute necessity before such expansion can occur. It is hard to imagine any aspect of our business or personal lives not, in some way, dependent upon electricity. All over the world (in over 30 countries) nuclear power is a low-cost, secure, safe, dependable, and environmentally friendly form of electric power generation. Nuclear plants in these countries are built in six to eight years using technology developed in the US, with good performance and safety records. This treatise addresses the success experienced by the US nuclear industry over the last 40 years, and makes the case that this reliable, cost-competitive source of electric power can help support the economic engine of the country and help prevent experiences like the recent crises in California and the Northeast. Traditionally, the evaluation of electric power generation facility performance has focused on the ability of plants to produce at design capacity for high percentages of the time. Successful operation of nuclear facilities is determined by examining capacity or load factors. Load factor is the percentage of design generating capacity that a power plant actually produces over the course of a year’s operation. This paper makes the case that these operating performance indicators warrant renewed consideration of the nuclear option. Usage of electricity in the US now approaches total generating capacity. The Nuclear Regulatory Commission has pre-approved construction and operating licenses for several nuclear plant designs. State public service commissions are beginning to understand that dramatic reform is required. The economy is recovering and inflation is minimal. It’s time, once more, to turn to the safe, reliable, environmentally friendly nuclear power alternative.

Calendar- and cycle-life studies of advanced technology development program generation 1 lithium-ion batteries

For nearly fifty years, the US national laboratories, under the direction of the Department of Energy, have maintained a tradition of outstanding scientific research and innovative technological development. With the end of the Cold War, their roles have undergone profound changes. Although many of their original priorities remain--stewardship of the nation`s nuclear stockpile, for example--pressing budget constraints and new federal mandates have altered their focus. Promotion of energy efficiency, environmental restoration, human health, and technology partnerships with the goal of enhancing US economic and technological competitiveness are key new priorities. The multiprogram national laboratories offer unparalleled expertise in meeting the challenge of changing priorities. This volume aims to demonstrate each laboratory`s uniqueness in applying this expertise. It describes the laboratories` activities in eleven broad areas of research that most or all share in common. Each section of this volume is devoted to a single laboratory. Those included are: Argonne National Laboratory; Brookhaven National Laboratory; Idaho National Engineering Laboratory; Lawrence Berkeley Laboratory; Lawrence Livermore National Laboratory; Los Alamos National Laboratory; National Renewable Energy Laboratory; Oak Ridge National Laboratory; Pacific Northwest Laboratory; and Sandia National Laboratories. The information in this volume was provided by the multiprogram national laboratories and compiled at Lawrence Berkeley Laboratory.

JOM surveyed the leaders of the U.S. Department of Energy’s (DOE) national laboratories for perspectives on some of the U.S. science and engineering community’s most pressing issues regarding RD Dan Arvizu, National Renewable Energy Laboratory (NREL); John J. Grossenbacher, Idaho National Laboratory (INL); Thomas O. Hunter, Sandia National Laboratories; Robert Rosner, Argonne National Laboratory; and Jeffrey Wadsworth, Oak Ridge National Laboratory (ORNL).

Research studies by the Electric Power Research Institute (EPRI) established the technical and operational requirements necessary to enable the onsite cask-to-cask dry transfer of spent nuclear fuel. Use of the dry transfer system has the potential to permit shutdown reactor sites to decommission pools and provide the capability of transferring assemblies from storage casks or small transportation casks to sealed transportable canisters. Following an evaluation by the Department of Energy (DOE) and the National Academy of Sciences, a cooperative program was established between DOE and EPRI, which led to the cost-shared design of a dry transfer system (DTS). EPRI used Transnuclear, Inc., of Hawthorne, New York, to design the DTS in accordance with the technical and quality assurance requirements of the code of Federal Regulations, Title 10, Part 72 (10CFR72). EPRI delivered the final design report to DOE in 1995 and the DTS topical safety analysis report (TSAR) in 1996. DOE submitted the TSAR to the United States Nuclear Regulatory Commission (NRC) for review under 10CFR72 and requested that the NRC staff evaluate the TSAR and issue a Safety Evaluation Report (SER) that could be used and referenced by an applicant seeking a site-specific license for the construction and operation of a DTS. DOE also initiated a cold demonstration of major subsystem prototypes in 1996. After careful assessment, the NRC agreed that the DTS concept has merit. However, because the TSAR was not site-specific and was lacking some detailed information required for a complete review, the NRC decided to issue an Assessment Report (AR) rather than a SER. This was issued in November 2000. Additional information that must be included in a future site-specific Safety Analysis Report for the DTS is identified in the AR. The DTS consists of three major sections: a Preparation Area, a Lower Access Area, and a Transfer Confinement Area. The Preparation Area is a sheet metal building where casks are prepared for loading, unloading, or shipment. The Preparation Area adjoins the Lower Access Area and is separated from the Lower Access Area by a large shielded door. The Lower Access Area and Transfer Confinement Area are contained within concrete walls approximately three feet thick. These are the areas where the casks are located and where the fuel is moved during transfer operations. A floor containing two portals separates the Lower Access Area and the Transfer Confinement Area. The casks are located below the floor, and the fuel transfer operation occurs above the floor. The cold demonstration of the DTS was successfully conducted at the Idaho National Engineering and Environmental Laboratory (INEEL) as a cooperative effort between the DOE and EPRI. The cold demonstration was limited to the fuel handling equipment, the cask lid handling equipment, and the cask interface system. The demonstration included recovery operations associated with loss of power or off-normal events. The demonstration did not include cask receiving and lid handling; cask transport and lifting; vacuum/inerting/leak test; canister welding; decontamination; heating, ventilation, and air conditioning; and radiation monitoring. The demonstration test was designed to deliberately challenge the system and determine whether any specific system operation could adversely impact or jeopardize the operation or safety of any other function or system. All known interlocks were challenged. As in all new systems, there were lessons learned during the operation of the system and a few minor modifications made to ease operations. System modifications were subsequently demonstrated. The demonstration showed that the system operated as expected and provided times for normal fuel transfer operations. The demonstration also showed that recovery could be made from off-normal events.

The potential release of radioactive material during a plant incident, referred to as the source term, is a vital design metric and will be a major focus of advanced reactor licensing. The U.S. Nuclear Regulatory Commission has stated an expectation for advanced reactor vendors to present a mechanistic assessment of the potential source term in their license applications. The mechanistic source term presents an opportunity for vendors to realistically assess the radiological consequences of an incident, and may allow reduced emergency planning zones and smaller plant sites. However, the development of a mechanistic source term for advanced reactors is not without challenges, as there are often numerous phenomena impacting the transportation and retention of radionuclides. This project sought to evaluate U.S. capabilities regarding the mechanistic assessment of radionuclide release from core damage incidents at metal fueled, pool-type sodium fast reactors (SFRs). The purpose of the analysis was to identify, and prioritize, any gaps regarding computational tools or data necessary for the modeling of radionuclide transport and retention phenomena. To accomplish this task, a parallel-path analysis approach was utilized. One path, led by Argonne and Sandia National Laboratories, sought to perform a mechanistic source term assessment using available codes, data, and models, with the goal to identify gaps in the current knowledge base. The second path, performed by an independent contractor, performed sensitivity analyses to determine the importance of particular radionuclides and transport phenomena in regards to offsite consequences. The results of the two pathways were combined to prioritize gaps in current capabilities.

This report presents the measured data and the analyses performed to date on the full-scale high-energy qualification and flow interruption gate valve testing to develop technical insights for the United States Nuclear Regulatory Commission (USNRC) effort regarding Generic Issue 87 (GI-87). The research was sponsored by the USNRC and conducted by researchers from the Idaho National Engineering Laboratory. We tested two 6-in., 900-lb class valve assemblies, which represent a significant percentage of the reactor water cleanup isolation valves installed in plant applications. These valves were modified before testing by adding a high temperature load cell in the valve stems, which allowed the direct measurement of valve stem thrust during both opening and closing valve cycles. Instrumentation installed in the flow loop and on the valve assemblies measured the important valve system test responses. Additionally, during the test program, all of the currently popular motor operated valve diagnostic test systems monitored the performance of the valves. Initially the valves were subjected to the hydraulic and leakage qualification tests defined in ANSI B 16.41 and then to flow interruption and reopening valve tests at boiling water reactor primary system water temperature and pressure conditions with downstream line break flows. For the two valves tested, results show that (a) the disc factor used in current industry motor operator sizing equations may require an additional term to account for nonlinear performance, (b) the thrusts required to close the valves were sensitive to the fluid temperature, and (c) the results of testing at lower pressures, temperatures, and flows cannot be extrapolated to design basis pressures, temperatures, and flows for valve designs that have not exhibited linear performance behavior during design basis prototypical testing. 2 refs., 30 figs., 4 tabs.

The Nuclear Regulatory Commission (NRC) has considered revision of 10-CFR-50.46C rule (Borchard and Johnson, 2013, “10 CFR 50.46c Rulemaking: Request to Defer Draft Guidance and Extension Request for Final Rule and Final Guidance,” U.S. Nuclear Regulatory Commission, Washington, DC.) to account for burn-up rate effects in future analysis of reactor accident scenarios so that safety margins may evolve as dynamic limits with reactor operation and reloading. To find these limiting conditions, both cladding oxidation and maximum temperature must be cast as functions of fuel exposure. To run a plant model through a long operational transient to fuel reload is computationally intensive, and this must be repeated for each reload until the time of the accident scenario. Moreover for probabilistic risk assessment (PRA), this must be done for many different fuel reload patterns. To perform such new analyses in a reasonable amount of computational time with good accuracy, Idaho National Laboratory (INL) has developed new multiphysics tools by combining existing codes and adding new capabilities. The parallel highly innovative simulation INL code system (PHISICS) toolkit (Rabiti et al., 2016, “New Simulation Schemes and Capabilities for the PHISICS/RELAP5-3D Coupled Suite,” Nucl. Sci. Eng., 182(1), pp. 104–118; Alfonsi et al., 2012, “PHISICS Toolkit: Multi-Reactor Transmutation Analysis Utility—MRTAU,” PHYSOR 2012 Advances in Reactor Physics Linking Research, Industry, and Education, Knoxville, TN, Apr. 15–20.) for neutronic and reactor physics is coupled with the reactor excursion and leak analysis program—three-dimensional (RELAP5-3D) (The RELAP5-3D© Code Development Team, 2014, “RELAP5-3D© Code Manual Volume I: Code Structure, System Models, and Solution Methods,” Rev. 4.2, Idaho National Laboratory, Idaho Falls, ID, Technical Report No. INEEL-EXT-98-00834.) for the loss of coolant accident (LOCA) analysis and reactor analysis and virtual-control environment (RAVEN) (Alfonsi et al., 2013, “RAVEN as a Tool for Dynamic Probabilistic Risk Assessment: Software Overview,” 2013 International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering, Sun Valley, ID, May 5–9, pp. 1247–1261.) for the probabilistic risk assessment (PRA) and margin characterization analysis. For RELAP5-3D to process a single sequence of cores in a continuous run required a sequence of restarting input decks, each with different neutronics or thermal-hydraulic (TH) flow region and culminating in an accident scenario. A new multideck input processing capability was developed and verified for this analysis. The combined RAVEN/PHISICS/RELAP5-3D tool is used to analyze a typical pressurized water reactor (PWR).

A workshop was held at the Idaho National Engineering Laboratory, August 16--18, 1994 on the topic of risk assessment on medical devices that use radioactive isotopes. Its purpose was to review past efforts to develop a risk assessment methodology to evaluate these devices, and to develop a program plan and a scoping document for future methodology development. This report contains a summary of that workshop. Participants included experts in the fields of radiation oncology, medical physics, risk assessment, human-error analysis, and human factors. Staff from the US Nuclear Regulatory Commission (NRC) associated with the regulation of medical uses of radioactive materials and with research into risk-assessment methods participated in the workshop. The workshop participants concurred in NRC`s intended use of risk assessment as an important technology in the development of regulations for the medical use of radioactive material and encouraged the NRC to proceed rapidly with a pilot study. Specific recommendations are included in the executive summary and the body of this report. An appendix contains the 8 papers presented at the conference: NRC proposed policy statement on the use of probabilistic risk assessment methods in nuclear regulatory activities; NRC proposed agency-wide implementation plan for probabilistic risk assessment; Risk evaluation of high dose rate remote afterloading brachytherapy at a large research/teaching institution; The pros and cons of using human reliability analysis techniques to analyze misadministration events; Review of medical misadministration event summaries and comparison of human error modeling; Preliminary examples of the development of error influences and effects diagrams to analyze medical misadministration events; Brachytherapy risk assessment program plan; and Principles of brachytherapy quality assurance.

The MELCOR Accident Consequence Code System, Version 1.5, (MACCS 1.5) calculates potential consequences resulting from atmospheric releases of radioactive materials. Sandia National Laboratories developed the code for the US Nuclear Regulatory Commission on a VAX/VMS mini-computer. This report documents the Idaho National Engineering Laboratory conversion of MACCS 1.5 for compilation and execution on an 80386-based IBM or IBM-compatible personal computer (PC). The resulting PC version of the code is available through the National Energy Software Center, Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL, 60439.

The paper is dedicated to the legal regulation of risk management and safety in the nuclear power industry. Legal regulation of risk management and safety in the nuclear power industry can significantly increase the damage compensation. It is possible to involve nuclear power industry’s enterprises in the formation of measures for the safe use of nuclear energy by an effective legal system creation. The legal system should include national and international principles regulating the nuclear energy use. This will reduce the negative impact on the environment and ensure the nuclear industry’s energy safety. The main subjects involved in the nuclear energy management and regulation are principal, manufacturer, operating organization. The principal is the developer of normative legal acts which indicate the basic institutional requirements and provide for state regulation of safety when using of nuclear energy. This regulation is achieved by such methods as licensing and certification. The manufacturers are various subjects of productive activities in the frame of the nuclear energy objects’ life cycle, and entities for these objects design, construction, and operation. Safety in the area of nuclear energy use includes mandatory and discretionary methods for regulation of relations between different actors. Methods for legal regulation of the safety level in the nuclear industry come from the condition of ecological and economic optimality, selection of appropriate principles of the law in the nuclear energy area. A tool for improvement of the effectiveness related to legal instruments in the area of nuclear energy management and use is a system of norms, rules, and standards, regulating and providing the stages of energy resources’ extraction, production and consumption in the context of atomic power facilities safe operation and subject activities strict regulation.

The Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) Version 8 is a software application developed for performing a complete probabilistic risk assessment using a personal computer running the Microsoft Windows™ operating system. SAPHIRE 8 is funded by the U.S. Nuclear Regulatory Commission (NRC). The role of the Idaho National Laboratory (INL) in this project is that of software developer and tester. In older versions of SAPHIRE, the model creation and analysis functions were intermingled. However, in SAPHIRE 8, the act of creating a model has been separated from the analysis of that model in order to improve the quality of both the model (e.g., by avoiding inadvertent changes) and the analysis. Consequently, in SAPHIRE 8, the analysis of models is performed by using what are called Workspaces. Currently, there are Workspaces for three types of analyses: (1) the NRC’s Accident Sequence Precursor program, where the workspace is called “Events and Condition Assessment (ECA);” (2) the NRC’s Significance Determination Process (SDP); and (3) the General Analysis (GA) workspace. Workspaces for each type are created and saved separately from the base model which keeps the original database intact. Workspaces are independent of each other and modifications or calculations made within one workspace will not affect another. In addition, each workspace has a user interface and reports tailored for their intended uses.