scholarly journals A Methodology for the Integration of a Mechanistic Source Term Analysis in a Probabilistic Framework for Advanced Reactors

2016 ◽  
Author(s):  
Dave Grabaskas ◽  
Acacia J. Brunett ◽  
Matthew Bucknor
Keyword(s):  
Source Term ◽  
National Laboratory ◽  
Argonne National Laboratory ◽  
Primary System ◽  
Separate Analysis ◽  
Event Sequences ◽  
Safety And Reliability ◽  
Detailed Assessment ◽  
Water Reactors ◽  
High Level

GE Hitachi Nuclear Energy (GEH) and Argonne National Laboratory are currently engaged in a joint effort to modernize and develop probabilistic risk assessment (PRA) techniques for advanced non-light water reactors. At a high level, the primary outcome of this project will be the development of next-generation PRA methodologies that will enable risk-informed prioritization of safety- and reliability-focused research and development, while also identifying gaps that may be resolved through additional research. A subset of this effort is the development of PRA methodologies to conduct a mechanistic source term (MST) analysis for event sequences that could result in the release of radionuclides. The MST analysis seeks to realistically model and assess the transport, retention, and release of radionuclides from the reactor to the environment. The MST methods developed during this project seek to satisfy the requirements of the Mechanistic Source Term element of the ASME/ANS Non-LWR PRA standard. The MST methodology consists of separate analysis approaches for risk-significant and non-risk significant event sequences that may result in the release of radionuclides from the reactor. For risk-significant event sequences, the methodology focuses on a detailed assessment, using mechanistic models, of radionuclide release from the fuel, transport through and release from the primary system, transport in the containment, and finally release to the environment. The analysis approach for non-risk significant event sequences examines the possibility of large radionuclide releases due to events such as re-criticality or the complete loss of radionuclide barriers. This paper provides details on the MST methodology, including the interface between the MST analysis and other elements of the PRA, and provides a simplified example MST calculation for a sodium fast reactor.

scholarly journals A Methodology for the Development of a Reliability Database for an Advanced Reactor Probabilistic Risk Assessment

2016 ◽  
Author(s):  
Dave Grabaskas ◽  
Acacia J. Brunett ◽  
Matthew Bucknor
Keyword(s):  
Risk Assessment ◽  
Failure Modes ◽  
National Laboratory ◽  
Argonne National Laboratory ◽  
Probabilistic Risk Assessment ◽  
Quality Of Data ◽  
Reliability Data ◽  
Water Reactors ◽  
High Level ◽  
Probabilistic Risk

GE Hitachi Nuclear Energy (GEH) and Argonne National Laboratory are currently engaged in a joint effort to modernize and develop probabilistic risk assessment (PRA) techniques for advanced non-light water reactors. At a high level the primary outcome of this project will be the development of next-generation PRA methodologies that will enable risk-informed prioritization of safety- and reliability-focused research and development, while also identifying gaps that may be resolved through additional research. A subset of this effort is the development of a reliability database (RDB) methodology to determine applicable reliability data for inclusion in the quantification of the PRA. The RDB method developed during this project seeks to satisfy the requirements of the Data Analysis element of the ASME/ANS Non-LWR PRA standard. The RDB methodology utilizes a relevancy test to examine reliability data and determine whether it is appropriate to include as part of the reliability database for the PRA. The relevancy test compares three component properties to establish the level of similarity to components examined as part of the PRA. These properties include the component function, the component failure modes, and the environment/boundary conditions of the component. The relevancy test is used to gauge the quality of data found in a variety of sources, such as advanced reactor-specific databases, non-advanced reactor nuclear databases, and non-nuclear databases. The RDB also establishes the integration of expert judgment or separate reliability analysis with past reliability data. This paper provides details on the RDB methodology, and includes an example application of the RDB methodology for determining the reliability of the intermediate heat exchanger of a sodium fast reactor. The example explores a variety of reliability data sources, and assesses their applicability for the PRA of interest through the use of the relevancy test.

scholarly journals Shutdown and Closure of the Experimental Breeder Reactor–II

2002 ◽  
Author(s):  
John A. Michelbacher ◽  
Carl E. Baily ◽  
Daniel K. Baird ◽  
S. Paul Henslee ◽  
Collin J. Knight ◽  
...  
Keyword(s):  
Liquid Metal ◽  
National Laboratory ◽  
Argonne National Laboratory ◽  
Passive Layer ◽  
Department Of Energy ◽  
Primary System ◽  
Secondary System ◽  
Reactor Plant ◽  
Carbon Dioxide Gas ◽  
Safe Condition

The Department of Energy mandated the termination of the Integral Fast Reactor (IFR) Program, effective October 1, 1994. To comply with this decision, Argonne National Laboratory-West (ANL-W) prepared a plan providing detailed requirements to maintain the Experimental Breeder Reactor-II (EBR-II) in a radiologically and industrially safe condition, including removal of all irradiated fuel assemblies from the reactor plant, and removal and stabilization of the primary and secondary sodium, a liquid metal used to transfer heat within the reactor plant. The EBR-II is a pool-type reactor. The primary system contained approximately 325 m3 (86,000 gallons) of sodium and the secondary system contained 50 m3 (13,000 gallons). In order to properly dispose of the sodium in compliance with the Resource Conservation and Recovery Act (RCRA), a facility was built to react the sodium to a solid sodium hydroxide monolith for burial as a low level waste in a land disposal facility. Deactivation of a liquid metal fast breeder reactor (LMFBR) presents unique concerns. Residual amounts of sodium remaining in circuits and components must be passivated, inerted, or removed to preclude future concerns with sodium-air reactions that could generate potentially explosive mixtures of hydrogen and leave corrosive compounds. The passivation process being implemented utilizes a moist carbon dioxide gas that generates a passive layer of sodium carbonate/sodium bicarbonate over any quantities of residual sodium. Tests being conducted will determine the maximum depths of sodium that can be reacted using this method, defining the amount that must be dealt with later to achieve RCRA clean closure. Deactivation of the EBR-II complex is on schedule for a March, 2002, completion. Each system associated with EBR-II has an associated layup plan defining the system end state, as well as instructions for achieving the layup condition. A goal of system-by-system layup is to minimize surveillance and maintenance requirements during the interim period between deactivation and decommissioning. The plans also establish document archival of not only all the closure documents, but also the key plant documents (P&IDs, design bases, characterization data, etc.) in a convenient location to assure the appropriate knowledge base is available for decommissioning, which could occur decades in the future.

Implementation of Liquid Metal Properties in RELAP5 MOD3.2 for Safety Analysis of Sodium-Cooled Fast Reactors

2017 ◽  
Author(s):  
Jian Song ◽  
Limin Liu ◽  
Simiao Tang ◽  
Yingwei Wu ◽  
Wenxi Tian ◽  
...  
Keyword(s):  
Experimental Data ◽  
Liquid Metal ◽  
Fast Reactor ◽  
Natural Circulation ◽  
National Laboratory ◽  
Argonne National Laboratory ◽  
Working Fluid ◽  
Primary System ◽  
Steady State Condition ◽  
Design Values

Due to great deal of operation experience and technology accumulation, sodium cooled fast reactor (SFR) is the most promising among the six Generation IV reactors, which has advantages of breeding nuclear fuel, transmuting long-lived actinides and good safety characteristics. Thermal-hydraulic computer codes will have to be developed, verified, and validated to support the conceptual and final designs of new SFRs. However, work on developing thermal hydraulic analysis code for SFR is very limited in China, while the common software RELAP5 MOD3 is unable to analyze liquid metal systems. So the modified RELAP5 MOD3.2 is being considered as the thermal-hydraulic system code to support the development of the SFRs. The thermodynamic and transport properties of sodium liquid and vapor have been implemented into the RELAP5 MOD3.2 code, as well as the specific heat transfer correlations for liquid metal. The sodium liquid properties use polynomial equations based on data obtained from Argonne National Laboratory, and the vapor is assumed to be perfect gas. The property equations are acceptably accurate for analysis of SFR, especially for single-phase liquid. New files are added to the fluids directory to generate property tables for new working fluid, which are similar to the table interpolation subroutines for light and heavy water in the original file directory. The method of code modifications are universal for other working fluids and will not affect the code original performance. Some basic verification work for the modified code are carried out. The steam generator of CEFR is analyzed to verify the modified code. The calculated results show that all the water will boil off in the evaporator and the calculated results are in good agreement with the design values. By using modified RELAP5 to model the primary loop of EBR-II fast reactor, the SHRT-17 PLOF test was analyzed. The results show that the natural circulation can be established in the EBR-II primary system after main pumps off to remove the core decay residual heat effectively, and the peak temperature under the safety limits. Moreover, the results computed in this work compared well with the test experimental data for the steady state condition. During the transients, the changing trends of temperature and pressure are similar to experimental data. The discrepancies between calculation and experiment are considered acceptably which need to be improved in the future work. Our work could demonstrate the capability and reliability of the modified RELAP5 for the analysis of SFRs further.

scholarly journals Development of a ceramic waste form for high-level waste disposal

1999 ◽  
Vol 556 ◽  
Author(s):  
D. W. Esh ◽  
K. M. Goff ◽  
K. T. Hirsche ◽  
T. J. Battisti ◽  
M. F. Simpson ◽  
...  
Keyword(s):  
Spent Nuclear Fuel ◽  
National Laboratory ◽  
Argonne National Laboratory ◽  
Fission Products ◽  
Waste Form ◽  
High Level Waste ◽  
Durability Test ◽  
Ceramic Waste ◽  
High Level ◽  
The Impact

AbstractA ceramic waste form is being developed by Argonne National Laboratory* (ANL) as part of the demonstration of the electrometallurgical treatment of spent nuclear fuel [1]. The halide, alkaline earth, alkali, transuranic, and rare earth fission products are stabilized in zeolite which is combined with glass and processed in a hot isostatic press (HIP) to form a ceramic composite. The mineral sodalite is formed in the HIP from the zeolite precursor. The process, from starting materials to final product, is relatively simple. An overview of the processing operations is given. The metrics that have been developed to measure the success or completion of processing operations are developed and discussed. The impact of variability in processing metrics on the durability of the final product is presented. The process is demonstrated to be robust for the type and range of operation metrics considered and the performance metric (PCT durability test) against which the operation metrics are evaluated.

A Methodology for the Integration of Passive System Reliability With Success Criteria in a Probabilistic Framework for Advanced Reactors

2016 ◽  
Author(s):  
Acacia J. Brunett ◽  
Dave Grabaskas ◽  
Matthew Bucknor ◽  
Stefano Passerini
Keyword(s):  
System Reliability ◽  
National Laboratory ◽  
Argonne National Laboratory ◽  
Passive System ◽  
Passive Systems ◽  
Success Criteria ◽  
Mission Success ◽  
Reliability Method ◽  
High Level ◽  
The Impact

GE Hitachi Nuclear Energy (GEH) and Argonne National Laboratory are currently engaged in a joint effort to modernize and develop probabilistic risk assessment (PRA) techniques for advanced non-light water reactors. At a high level, the primary outcome of this project will be the development of next-generation PRA methodologies that will enable risk-informed prioritization of safety- and reliability-focused research and development, while also identifying gaps that may be resolved through additional research. A subset of this effort is the development of PRA methodologies that can be used for the determination of passive system reliability while integrating quantitative success criteria into the risk analysis framework. An updated passive system reliability approach has been developed for utilization in the PRISM PRA that systematically characterizes the impact of passive safety systems on key success criteria. This methodology is derived from the Reliability Method for Passive Systems (RMPS), but is refined to explicitly include consideration of overall mission success through satisfaction of success criteria, rather than only focusing on the passive system itself. This paper provides details on the integrated methodology, focusing on the interface between passive system reliability and success criteria. Specific examples for the passive systems/features of interest, RVACS and inherent reactivity feedback, are included. Additionally, aspects of the integrated passive system and success criteria methodology as they relate to the ASME/ANS Non-LWR PRA Standard are identified and discussed.

In Situ Radiation Damage Studies of Ca3Zr2FeAlSiO12 and Ca3Hf2FeAlSiO12

2008 ◽  
Vol 1124 ◽  
Author(s):  
Karl R Whittle ◽  
Mark Blackford ◽  
Gregory R Lumpkin ◽  
Katherine L Smith ◽  
Nestor J Zaluzec
Keyword(s):  
Nuclear Waste ◽  
National Laboratory ◽  
Argonne National Laboratory ◽  
Voltage Electron ◽  
Significant Difference ◽  
User Facility ◽  
High Level Nuclear Waste ◽  
Nuclear Waste Immobilization ◽  
High Level

AbstractGarnets, A3B2C3O12, are considered to be potential host phases for the immobilization of high-level nuclear waste as they can accommodate a number of elements of interest, including Zr, Ti and Fe. The naturally occurring garnet, kimzeyite, Ca3(Zr,Ti)2(Si,Al,Fe)3O12, can contain ˜30wt% Zr. An understanding of the radiation tolerance of these materials is crucial to their potential use in nuclear waste immobilization. In this study two synthetic analogues of kimzeyite of composition Ca3Zr2FeAlSiO12 and Ca3Hf2FeAlSiO12 were monitored in situ during irradiation with 1.0 MeV Kr ions using the intermediate voltage electron microscope-Tandem User Facility (IVEM) at Argonne National Laboratory. The structure of these materials was previously determined by neutron diffraction and 57Fe Mössbauer spectroscopy. Ca3Zr2FeAlSiO12 and Ca3Hf2FeAlSiO12 have very similar structural properties with cubic Ia3d symmetry, the only significant difference being the presence of Zr and Hf, respectively, on the 6 coordinated B sites.

scholarly journals VERIFICATION OF THE DIFFUSION AND TRANSPORT SOLVERS WITHIN DIF3D FOR 3D HEXAGONAL GEOMETRIES

2021 ◽  
Vol 247 ◽  
pp. 10030
Author(s):  
A. G. Nelson ◽  
M. A. Smith ◽  
F. Heidet
Keyword(s):  
National Laboratory ◽  
Argonne National Laboratory ◽  
Three Dimensional ◽  
Analytic Solutions ◽  
Specific Level ◽  
Structured Grid ◽  
Design Activities ◽  
Transport And Diffusion ◽  
High Level ◽  
And Diffusion

The DIF3D code (DIFfusion 3D) has been a workhorse of fast reactor analysis work at Argonne National Laboratory for over 40 years. DIF3D was primarily built in the late 1970s as a three-dimensional multigroup diffusion equation solver operating on semi-structured grid geometries. In the mid-1990s, transport capabilities needed for high-leakage reactor configurations were added to DIF3D with the variational anisotropic nodal transport approach. Recent reactor design activities at Argonne are requiring that a thorough verification of the Argonne Reactor Computation (ARC) codes be performed. With DIF3D being central to the entire ARC system, the verification efforts are focused on the 3D Cartesian, 3D triangular, and 3D hexagonal core geometry options of DIF3D. Validation activities, while needed for the ongoing design activities at Argonne, are handled at a project-specific level. This paper summarizes the verification work so far on the forward and adjoint forms of the fixed source, inhomogeneous fixed source, and k-eigenvalue steady state transport and diffusion equations as implemented specifically for 3D triangular and hexagonal geometries in DIF3D. Since analytic solutions of the neutron diffusion and transport equations are either limited in scope or not possible, this verification required multiple tiers of problems unique to each solver and geometry type, each testing features independent and complementary arguments for why this separate testing of functionalities is acceptable. This separate testing was also supplemented with a high-level integral check of each the diffusion and transport capabilities and applicable geometries.

scholarly journals HVEM-Tandem and Eels Study of Radiation Damage in Zirconolite

1997 ◽  
Vol 3 (S2) ◽  
pp. 773-774
Author(s):  
Katherine L. Smith ◽  
Nestor J. Zaluzec ◽  
Gregory R. Lumpkin
Keyword(s):  
Radiation Damage ◽  
Electron Spectroscopy ◽  
National Laboratory ◽  
Argonne National Laboratory ◽  
Predictive Modelling ◽  
Transmission Electron ◽  
Ion Accelerator ◽  
Diffraction Patterns ◽  
Individual Grains ◽  
High Level

Zirconolite (CaZrTi2O7) is the major host phase for actinides in Synroc, a promising waste form for the immobilisation of high-level radioactive waste. The effect of radiation damage on the structure and durability of zirconolite are important to predictive modelling of zirconolite's behaviour in the repository environment and risk assessment.In this study, radiation damage effects in zirconolite were investigated by irradiating samples with 1.5 MeV Kr+ ions using the HVEM-Tandem at Argonne National Laboratory (ANL) and energy loss electron spectroscopy (EELS). The HVEM-Tandem consists of a modified AEI high votage transmission electron microscope interfaced to to a 2 MV tandem ion accelerator. EELS spectra were collected using a Philips 420 TEM, operated at 120 kV, fitted with a Gatan Model 607 Serial EELS. EELS data were recorded at resolutions of ˜1.0 eV and at a dispersion of about ˜0.25 eV.Selected area diffraction patterns (SADs) of individual grains of various zirconolites were monitored as a function of dose to establish the critical dose for aniorphisation (Dc).

Electron and ion irradiation-induced dissolution of mechanical twins in the intermetalic U3Si: An in situ HVEM study

1989 ◽  
Vol 47 ◽  
pp. 652-653
Author(s):  
Charles W. Allen ◽  
Robert C. Birtcher
Keyword(s):  
Elevated Temperatures ◽  
Ion Irradiation ◽  
Ion Beam ◽  
National Laboratory ◽  
Argonne National Laboratory ◽  
Heavy Ion ◽  
High Energy ◽  
Mechanical Twinning ◽  
In Situ Experiments

The uranium silicides, including U3Si, are under study as candidate low enrichment nuclear fuels. Ion beam simulations of the in-reactor behavior of such materials are performed because a similar damage structure can be produced in hours by energetic heavy ions which requires years in actual reactor tests. This contribution treats one aspect of the microstructural behavior of U3Si under high energy electron irradiation and low dose energetic heavy ion irradiation and is based on in situ experiments, performed at the HVEM-Tandem User Facility at Argonne National Laboratory. This Facility interfaces a 2 MV Tandem ion accelerator and a 0.6 MV ion implanter to a 1.2 MeV AEI high voltage electron microscope, which allows a wide variety of in situ ion beam experiments to be performed with simultaneous irradiation and electron microscopy or diffraction.At elevated temperatures, U3Si exhibits the ordered AuCu3 structure. On cooling below 1058 K, the intermetallic transforms, evidently martensitically, to a body-centered tetragonal structure (alternatively, the structure may be described as face-centered tetragonal, which would be fcc except for a 1 pet tetragonal distortion). Mechanical twinning accompanies the transformation; however, diferences between electron diffraction patterns from twinned and non-twinned martensite plates could not be distinguished.

Environmental cell studies of deformation and fracture

1990 ◽  
Vol 48 (4) ◽  
pp. 516-517
Author(s):  
H. K. Birnbaum ◽  
I. M. Robertson
Keyword(s):  
National Laboratory ◽  
Argonne National Laboratory ◽  
Damage Threshold ◽  
Chromatic Aberration ◽  
Good Choice ◽  
Point Of View ◽  
Deformation And Fracture ◽  
Environmental Cell ◽  
Gas Molecules ◽  
The University

Studies of the effects of hydrogen environments on the deformation and fracture of fcc, bcc and hep metals and alloys have been carried out in a TEM environmental cell. The initial experiments were performed in the environmental cell of the HVEM facility at Argonne National Laboratory. More recently, a dedicated environmental cell facility has been constructed at the University of Illinois using a JEOL 4000EX and has been used for these studies. In the present paper we will describe the general design features of the JEOL environmental cell and some of the observations we have made on hydrogen effects on deformation and fracture.The JEOL environmental cell is designed to operate at 400 keV and below; in part because of the available accelerating voltage of the microscope and in part because the damage threshold of most materials is below 400 keV. The gas pressure at which chromatic aberration due to electron scattering from the gas molecules becomes excessive does not increase rapidly with with accelerating voltage making 400 keV a good choice from that point of view as well. A series of apertures were placed above and below the cell to control the pressures in various parts of the column.

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