scholarly journals A self-teaching curriculum for the NRC/SNL (Nuclear Regulatory Commission/Sandia National Laboratory) low-level waste performance assessment methodology

1991 ◽  
Author(s):  
M.S.Y. Chu ◽  
M.W. Kozak ◽  
J.E. Campbell ◽  
B.H. Thompson
Keyword(s):  
Performance Assessment ◽  
National Laboratory ◽  
Sandia National Laboratory ◽  
Nuclear Regulatory Commission ◽  
Assessment Methodology ◽  
Low Level ◽  
Teaching Curriculum ◽  
Regulatory Commission

Response of a Dry Storage Spent Fuel Cask to Hypothetical Loading

2005 ◽  
Author(s):  
Gustavo A. Aramayo ◽  
Douglas J. Ammerman ◽  
Jeffrey A. Smith
Keyword(s):  
Analytical Methods ◽  
National Laboratory ◽  
Sandia National Laboratory ◽  
Nuclear Regulatory Commission ◽  
Spent Fuel ◽  
Oak Ridge ◽  
Transportation Risk ◽  
Dry Storage ◽  
Regulatory Commission ◽  
The U.S

This paper addresses the analytical methods used to determine the response of a dry storage spent fuel cask to hypothetical loading. Because of the sensitive nature of the topic under discussion, the response of the cask is described in qualitative terms, and the paper is intentionally vague on the parameters and results. This research was sponsored by the U.S. Nuclear Regulatory Commission (NRC) Spent Fuel Project Office. The work was performed under contract from the Sandia National Laboratory (SNL), Transportation Risk and Packing organization. The analytical effort was performed at the Oak Ridge National Laboratory (ORNL) facilities with loading specified by SNL.

scholarly journals Review of Practice for Deeply Embedded/Buried NPP Structures Subject to Seismic Loadings

2004 ◽  
Author(s):  
J. Xu ◽  
C. Miller ◽  
C. Hofmayer ◽  
H. Graves
Keyword(s):  
National Laboratory ◽  
Earth Pressure ◽  
Nuclear Regulatory Commission ◽  
Brookhaven National Laboratory ◽  
Nuclear Containment ◽  
The Us ◽  
Reactor Building ◽  
Seismic Loadings ◽  
Knowledge And Practice ◽  
Regulatory Commission

Motivated by many design considerations, several conceptual designs for advanced reactors have proposed that the entire reactor building and a significant portion of the steam generator building will be either partially or completely embedded below grade. For the analysis of seismic events, the soil-structure interaction (SSI) effect and passive earth pressure for these types of deeply embedded structures will have a significant influence on the predicted seismic response. Sponsored by the US Nuclear Regulatory Commission (NRC), Brookhaven National Laboratory (BNL) is carrying out a research program to assess the significance of these proposed design features for advanced reactors, and to evaluate the existing analytical methods to determine their applicability and adequacy in capturing the seismic behavior of the proposed designs. This paper summarizes a literature review performed by BNL to determine the state of knowledge and practice for seismic analyses of deeply embedded and/or buried (DEB) nuclear containment type structures. Included in the paper is BNL’s review of the open literature of existing standards, tests, and practices that have been used in the design and analysis of DEB structures. The paper also provides BNL’s evaluation of available codes and guidelines with respect to seismic design practice of DEB structures. Based on BNL’s review, a discussion is provided to highlight the applicability of the existing technologies for seismic analyses of DEB structures and to identify gaps that may exist in knowledge and potential issues that may require better understanding and further research.

Combining RAVEN, RELAP5-3D, and PHISICS for Fuel Cycle and Core Design Analysis for New Cladding Criteria

2017 ◽  
Vol 3 (2) ◽  
Author(s):  
Andrea Alfonsi ◽  
George L. Mesina ◽  
Angelo Zoino ◽  
Nolan Anderson ◽  
Cristian Rabiti
Keyword(s):  
Risk Assessment ◽  
National Laboratory ◽  
Nuclear Regulatory Commission ◽  
Probabilistic Risk Assessment ◽  
Reactor Physics ◽  
Idaho National Laboratory ◽  
Accident Scenario ◽  
Regulatory Commission ◽  
Fuel Reload ◽  
Probabilistic Risk

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).

scholarly journals An Assessment of Simplified vs. Detailed Methodologies for SSI Analyses of Deeply Embedded Structures

2004 ◽  
Author(s):  
J. Xu ◽  
C. Miller ◽  
C. Hofmayer ◽  
H. Graves
Keyword(s):  
Performance Assessment ◽  
Nuclear Power ◽  
National Laboratory ◽  
Response Spectra ◽  
Nuclear Regulatory Commission ◽  
Brookhaven National Laboratory ◽  
Beam Model ◽  
Simplified Approach ◽  
Embedded Structures ◽  
Different Depths

Sponsored by the US Nuclear Regulatory Commission (NRC), Brookhaven National Laboratory (BNL) is carrying out a research program to develop a technical basis to support the safety evaluation of deeply embedded and/or buried (DEB) structures as proposed for advanced reactor designs. In this program, the methods and computer programs established for the assessment of soil-structure interaction (SSI) effects for the current generation of light water reactors are evaluated to determine their applicability and adequacy in capturing the seismic behavior of DEB structures. This paper presents an assessment of the simplified vs. detailed methodologies for seismic analyses of DEB structures. In this assessment, a lump-mass beam model is used for the simplified approach and a finite element representation is employed for the detailed method. A typical containment structure embedded in a soil profile representative of a typical nuclear power plant site was utilized, considering various embedment depths from shallow to full burial. BNL used the CARES program for the simplified model and the SASSI2000 program for the detailed analyses. The calculated response spectra at the key locations of the DEB structure are used for the performance assessment of the applied methods for different depths of burial. Included in the paper are: 1) the description of both the simplified and detailed models for the SSI analyses of the DEB structure, 2) the comparison of the analysis results for the different depths of burial between the two methods, and 3) the performance assessment of the analysis methodologies for SSI analyses of DEB structures. The resulting assessment from this study has indicated that simplified methods may be capable of capturing the seismic response for much deeper embedded structures than would be normally allowed by the standard practice.

Proceeding Toward a License Application for U.S. Nuclear Waste Repository: Total System Performance Assessment Approach

2003 ◽  
Author(s):  
Jerry McNeish ◽  
Peter Swift ◽  
Rob Howard ◽  
David Sevougian ◽  
Donald Kalinich ◽  
...  
Keyword(s):  
Performance Assessment ◽  
System Performance ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Yucca Mountain ◽  
Nuclear Waste Repository ◽  
Total System ◽  
License Application ◽  
Regulatory Commission ◽  
The U.S

The development of a deep geologic repository system in the United States has progressed to the preparation of an application for a license from the U.S. Nuclear Regulatory Commission. The project received site recommendation approval from the U.S. President in early 2002. The next phase of the project involves development of the license application (LA) utilizing the vast body of information accumulated in study of the site at Yucca Mountain, Nevada. Development of the license application involves analyses of the total system performance assessment (TSPA) of the repository, the TSPA-LA. The TSPA includes the available relevant information and model analyses from the various components of the system (e.g., unsaturated geologic zone, engineered system (waste packaging and drift design), and saturated geologic zone) (see Fig. 1 for nominal condition components), and unites that information into a single computer model used for evaluating the potential future performance or degradation of the repository system. The primary regulatory guidance for the repository system is found in 10 CFR 63, which indicates the acceptable risk to future populations from the repository system. The performance analysis must be traceable and transparent, with a defensible basis. The TSPA-LA is being developed utilizing state-of-the-art modeling software and visualization techniques, building on a decade of experience with such analyses. The documentation of the model and the analyses will be developed with transparency and traceability concepts to provide an integrated package for reviewers. The analysis relies on 1000’s of pages of supporting information, and multiple software and process model analyses. The computational environment represents the significant advances in the last 10 years in computer workstations. The overall approach will provide a thorough, transparent compliance analysis for consideration by the U.S. Nuclear Regulatory Commission in evaluating the Yucca Mountain repository.

Status of the ANS Low Power and Shutdown PRA Standard

2004 ◽  
Author(s):  
Kenneth L. Kiper
Keyword(s):  
Low Power ◽  
National Laboratory ◽  
Nuclear Regulatory Commission ◽  
Current Status ◽  
Public Comment ◽  
Writing Group ◽  
Draft Standard ◽  
American Nuclear Society ◽  
Regulatory Commission

The Low Power & Shutdown (LPSD) PRA Standard (ANS-58.22) is currently being drafted by a Writing Group under the auspices of the Risk Informed Standards Committee of the American Nuclear Society. The Writing Group includes representatives from nuclear utilities, US Nuclear Regulatory Commission, national laboratory, university, and consultants with substantial experience producing LPSD PRAs. This draft standard is scheduled to be released for public comment in the second quarter of 2004, with publication by the end of 2004. This paper presents the current status of this standard in preparation for its public release.

Benchmarking of a Thermal Finite Element Approximation Scheme for Externally Cooled Spent Fuel Storage Casks

2007 ◽  
Author(s):  
Christopher S. Bajwa ◽  
Ian F. Spivack
Keyword(s):  
Pacific Northwest ◽  
National Laboratory ◽  
Pacific Northwest National Laboratory ◽  
Finite Element Approximation ◽  
Nuclear Regulatory Commission ◽  
Element Approximation ◽  
Spent Fuel ◽  
Spent Fuel Storage ◽  
Fuel Storage ◽  
Regulatory Commission

The US Nuclear Regulatory Commission (NRC) is responsible for licensing spent fuel storage casks under Title 10 of the Code of Federal Regulations Part 72 (10 CFR Part 72). Under these regulations, storage casks must be evaluated to verify that they meet various criteria, including acceptable thermal performance requirements. The purpose of the evaluation described in this paper is to establish the effectiveness of a medium-effort modeling approach and associated simplifying assumptions in closely approximating spent fuel cask component temperature distributions. This predictive evaluation is performed with the ANSYS® code, and is applicable to externally cooled cask designs. The results are compared against experimental measurements and predictions of the COBRA-SFS finite-difference code developed at Pacific Northwest National Laboratory.

Structural Analysis Approach for the Defense Programs Package 3 (DPP-3)

2020 ◽  
Author(s):  
Peter J. Sakalaukus ◽  
Nathan P. Barrett ◽  
Brian J. Koeppel
Keyword(s):  
Pacific Northwest ◽  
National Laboratory ◽  
Pacific Northwest National Laboratory ◽  
Analytical Techniques ◽  
Nuclear Regulatory Commission ◽  
Department Of Energy ◽  
Design Criteria ◽  
Element Analysis ◽  
Regulatory Testing ◽  
Regulatory Commission

Abstract The Pacific Northwest National Laboratory (PNNL) is the design authority for a new Type B hazardous materials transportation package designated as the Defense Programs Package 3 (DPP-3) for the U.S. Department of Energy (DOE) National Nuclear Security Administration (NNSA). The DPP-3 has been developed using similar materials and fabrication methods employed in previous U.S. Nuclear Regulatory Commission (NRC), DOE, and NNSA certified packages. The DPP-3 design criteria are derived from the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC), NNSA guidance and NRC regulatory guides in order to safely and securely transport a variety of payloads. Final regulatory approval by the NNSA will require regulatory testing to demonstrate that the containment vessel (CV) remains leaktight after enduring the entire regulatory testing sequence prescribed in Title 10 of the Code of Federal Regulations Part 71 (10 CFR 71). In order to gain confidence that the DPP-3 will remain leaktight after testing, the DPP-3 has been structurally analyzed using the Finite Element Analysis (FEA) software LS-DYNA. The FEA analyses serve two general purposes: first, they aid in design and development of the package, and second, they advise as to which drop orientations are expected to cause the most damage during regulatory testing. This paper will discuss how the design criteria are incorporated into analytical techniques needed to evaluate the FEA structural simulation results for 10 CFR 71 conditions to give confidence the DPP-3 testing campaign will be successful.

A Fuel Cycle and Core Design Analysis Method for New Cladding Acceptance Criteria Using PHISICS, RAVEN and RELAP5-3D

2016 ◽  
Author(s):  
Andrea Alfonsi ◽  
George L. Mesina ◽  
Angelo Zoino ◽  
Cristian Rabiti
Keyword(s):  
Fuel Cycle ◽  
National Laboratory ◽  
Nuclear Regulatory Commission ◽  
Computational Time ◽  
Maximum Temperature ◽  
Reactor Accident ◽  
Reactor Physics ◽  
Pressurized Water ◽  
Safety Margins ◽  
Regulatory Commission

The Nuclear Regulatory Commission (NRC) has considered revising the 10 CFR 50.46C rule [1] for analyzing reactor accident scenarios to take the effects of burn-up rate into account. Both maximum temperature and oxidation of the cladding must be cast as functions of fuel exposure in order to find limiting conditions, making safety margins dynamic limits that evolve with the operation and reloading of the reactor. In order to perform such new analysis in a reasonable computational time with good accuracy, INL (Idaho National Laboratory) has developed new multi-physics tools by combining existing codes and adding new capabilities. The PHISICS (Parallel Highly Innovative Simulation INL Code System) toolkit [2,3] for neutronic and reactor physics is coupled with RELAP5-3D [4] (Reactor Excursion and Leak Analysis Program) for the LOCA (Loss of Coolant Accident) analysis and RAVEN [5] for the PRA (Probabilistic Risk Assessment) and margin characterization analysis. In order to perform this analysis, the sequence of RELAP5-3D input models had to get executed in a sequence of multiple input decks, each of them had to restart and slightly modify the previous model (in this case, on the neutronic side only) This new RELAP5-3D multi-deck processing capability has application to parameter studies and uncertainty quantification. The combined RAVEN/PHISICS/RELAP5-3D tool is used to analyze a typical PWR (Pressurized Water Reactor).

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