The Official Web Site of the U.S. Department of Energy's Nuclear Criticality Safety Program
Abstract not provided
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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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Comparison of Hand-Held and Pager Radionuclide Identification Systems for Inspections
Abstract not provided
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Validation of Full-Domain Massively Parallel Transport Sweep Algorithms
Abstract not provided
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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.
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THE OFFICIAL WEBSITE OF THE U.S. DEPARTMENT OF ENERGY'S NUCLEAR CRITICALITY SAFETY PROGRAM
Abstract not provided
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A Review of the Upstream Corner Balance Spatial Discretization
Abstract not provided
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A Review of the Upstream Corner Balance Spatial Discretization
Abstract not provided
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Sectioning, Metallography and Burnup Determination of MFF Fuel Pins
This test plan describes the requirements for sample preparation only. Metallography is also covered. Burnup and density exams will be performed in the Idaho National Laboratory (INL) Analytical Laboratory and planning will be done using the required paperwork of that facility; the retained fission gas work will be covered by a separate test plan.
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ECAR-1944 AGC-1 AS RUN THERMAL RESULTS
Purpose of this ECAR is to provide documentation of as-run analysis results for each specimen temperature in the AGC-1 test train at specific time points during the irradiation.
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U.S. Nuclear Industry Processes for Probabilistic Risk Assessment
Slides for an invited seminar for a safety/risk class at MIT.
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Design and Fabrication of a Thermodynamic Reference Electrode Using Chemical Vapor Deposition
Explore the source record for details and available documents.
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ALEPH2 - A general purpose Monte Carlo depletion code
The Monte-Carlo burn-up code ALEPH is being developed at SCK-CEN since 2004. A previous version of the code implemented the coupling between the Monte Carlo transport (any version of MCNP or MCNPX) and the ' deterministic' depletion code ORIGEN-2.2 but had important deficiencies in nuclear data treatment and limitations inherent to ORIGEN-2.2. A new version of the code, ALEPH2, has several unique features making it outstanding among other depletion codes. The most important feature is full data consistency between steady-state Monte Carlo and time-dependent depletion calculations. The last generation general-purpose nuclear data libraries (JEFF-3.1.1, ENDF/B-VII and JENDL-4) are fully implemented, including special purpose activation, spontaneous fission, fission product yield and radioactive decay data. The built-in depletion algorithm allows to eliminate the uncertainties associated with obtaining the time-dependent nuclide concentrations. A predictor-corrector mechanism, calculation of nuclear heating, calculation of decay heat, decay neutron sources are available as well. The validation of the code on the results of REBUS experimental program has been performed. The ALEPH2 has shown better agreement with measured data than other depletion codes. (authors)
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Monte Carlo depletion calculations using VESTA 2.1 new features and perspectives
VESTA is a Monte Carlo depletion interface code that is currently under development at IRSN. With VESTA, the emphasis lies on both accuracy and performance, so that the code will be capable of providing accurate and complete answers in an acceptable amount of time compared to other Monte Carlo depletion codes. From its inception, VESTA is intended to be a generic interface code so that it will ultimately be capable of using any Monte-Carlo code or depletion module and that can be tailored to the users needs. A new version of the code (version 2.1.x) will be released in 2012. The most important additions to the code are a burn up dependent isomeric branching ratio treatment to improve the prediction of metastable nuclides such as {sup 242m}Am and the integration of the PHOENIX point depletion module (also developed at IRSN) to overcome some of the limitations of the ORIGEN 2.2 module. The task of extracting and visualising the basic results and also the calculation of physical quantities or other data that can be derived from the basic output provided by VESTA will be the task of the AURORA depletion analysis tool which will be released at the same time as VESTA 2.1.x. The experimental validation database was also extended for this new version and it now contains a total of 35 samples with chemical assay data and 34 assembly decay heat measurements. (authors)
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Development of a statistical sampling method for uncertainty analysis with SCALE
A new statistical sampling sequence called Sampler has been developed for the SCALE code system. Random values for the input multigroup cross sections are determined by using the XSUSA program to sample uncertainty data provided in the SCALE covariance library. Using these samples, Sampler computes perturbed self-shielded cross sections and propagates the perturbed nuclear data through any specified SCALE analysis sequence, including those for criticality safety, lattice physics with depletion, and shielding calculations. Statistical analysis of the output distributions provides uncertainties and correlations in the desired responses. (authors)
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Second order perturbation theory for nonlinear time-dependent problems: Application to a simplified coupled transient
In this paper a second-order perturbation technique for nonlinear time-dependent problems is presented and applied to a simplified multi-physics model. This method is developed by using the properties of the adjoint problem which allows calculating the set of first and second order coefficients by solving a number of linear systems. As an illustrative example the adjoint procedure is applied to a reference transient problem, represented by a coupled point-kinetic/lumped-parameters model, and used to calculate the sensitivity coefficients of a safety related response with respect to a set of input parameters. The results obtained are compared with the values given by a direct sampling of the forward nonlinear problem. A way to reduce the number of calculations required for the application of second order adjoint techniques is also discussed. Our first results show that the procedure provides good estimations in presence of higher order perturbation components, being able to reconstruct the responses of interest even in presence of non-Gaussian probability density functions. Furthermore, the use of reduced second order information decreases the computational requirements of the method, making it appealing for possible large scale applications. (authors)
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Application of wavelet scaling function expansion continuous-energy resonance calculation method to MOX fuel problem
More and more MOX fuels are used in all over the world in the past several decades. Compared with UO{sub 2} fuel, it contains some new features. For example, the neutron spectrum is harder and more resonance interference effects within the resonance energy range are introduced because of more resonant nuclides contained in the MOX fuel. In this paper, the wavelets scaling function expansion method is applied to study the resonance behavior of plutonium isotopes within MOX fuel. Wavelets scaling function expansion continuous-energy self-shielding method is developed recently. It has been validated and verified by comparison to Monte Carlo calculations. In this method, the continuous-energy cross-sections are utilized within resonance energy, which means that it's capable to solve problems with serious resonance interference effects without iteration calculations. Therefore, this method adapts to treat the MOX fuel resonance calculation problem natively. Furthermore, plutonium isotopes have fierce oscillations of total cross-section within thermal energy range, especially for {sup 240}Pu and {sup 242}Pu. To take thermal resonance effect of plutonium isotopes into consideration the wavelet scaling function expansion continuous-energy resonance calculation code WAVERESON is enhanced by applying the free gas scattering kernel to obtain the continuous-energy scattering source within thermal energy range (2.1 eV to 4.0 eV) contrasting against the resonance energy range in which the elastic scattering kernel is utilized. Finally, all of the calculation results of WAVERESON are compared with MCNP calculation. (authors)
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Development and validation of instantaneous risk model in nuclear power plant's risk monitor
The instantaneous risk model is the fundament of calculation and analysis in a risk monitor. This study focused on the development and validation of an instantaneous risk model. Therefore the principles converting from the baseline risk model to the instantaneous risk model were studied and separated trains' failure modes modeling method was developed. The development and validation process in an operating nuclear power plant's risk monitor were also introduced. Correctness of instantaneous risk model and rationality of converting method were demonstrated by comparison with the result of baseline risk model. (authors)
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