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.

Issues Regarding Use of Master Curve Data in Resetting Initial RTNDT

2006 ◽  
Author(s):  
K. K. Yoon ◽  
J. B. Hall
Keyword(s):  
National Laboratory ◽  
Safety Evaluation ◽  
Nuclear Regulatory Commission ◽  
Crack Size ◽  
Oak Ridge ◽  
The Us ◽  
Large Populations ◽  
Asme Code ◽  
Shock Experiment ◽  
Regulatory Commission

The B&W Owners Group submitted justification for resetting the initial RTNDT for the Linde 80 weld materials using ASME Code Case N-629/N-631 to the US Nuclear Regulatory Commission and received the NRC Safety Evaluation Report with some adjustments. Two major issues were encountered during the review and approval process: 1) The pressurized thermal shock experiment data from Oak Ridge National Laboratory with full-length axial (very long) cracks fall below the Code Case curve. This observation led to the question whether there are implicit crack size limitations in the code case, and 2) For the large populations of data examined, a larger portion of data falls below the Code Case N-629/N-631 curve than the ASME KIC curve, prompting a question whether the code case is functionally equivalent to the ASME KIC curve. This paper describes these major issues and how they were addressed.

Severe Transportation Accidents: Impacts of Severe Fires on Radioactive Material Transportation

2009 ◽  
Author(s):  
Christopher S. Bajwa ◽  
Earl P. Easton ◽  
Darrell S. Dunn
Keyword(s):  
Spent Nuclear Fuel ◽  
Nuclear Regulatory Commission ◽  
Radioactive Material ◽  
Fire Investigation ◽  
Radioactive Materials ◽  
The Us ◽  
Material Transportation ◽  
Safe Transport ◽  
Regulatory Commission ◽  
Loss Of Strength

In 2007, a severe transportation accident occurred in Oakland, California in what is commonly known as the “MacArthur Maze” section of Interstate 580 (I-580). The accident involved a tractor trailer carrying gasoline that impacted an overpass support column and burst into flames. The subsequent fire burned for over 2 hours and led to the collapse of the overpass due to the loss of strength in the structural steel that supported the overpass. The US Nuclear Regulatory Commission (NRC) studied this accident to examine any potential regulatory implications related to the safe transport of radioactive materials, including spent nuclear fuel. This paper will discuss the details of the NRC’s MacArthur Maze fire investigation.

Relating ASME Flaw Evaluation to ASME Task Group on Carbon Fiber Repairs

2017 ◽  
Author(s):  
Tomas Jimenez ◽  
Eric Houston ◽  
Nico Meyer
Keyword(s):  
Carbon Fiber ◽  
Nuclear Power ◽  
Nuclear Regulatory Commission ◽  
Fiber Reinforced Polymers ◽  
Structural Factors ◽  
Operating Life ◽  
The Us ◽  
Piping Systems ◽  
Wave Passage ◽  
Regulatory Commission

As most nuclear power stations in the US have surpassed their initial 40 years of operability, the industry is now challenged with maintaining safe operations and extending the operating life of structures, systems and components. The US Nuclear Regulatory Commission (NRC), Nuclear Energy Institute (NEI), and Electric Power Research Institute (EPRI) have identified safety related buried piping systems as particularly susceptible to degradation. These systems are required to maintain the structural factors of the ASME Construction Codes under pressure and piping loads, which includes seismic wave passage. This paper focuses on evaluation approaches for metallic buried piping that can be used to demonstrate that localized thinning meets the requirements of the Construction Code. The paper then addresses a non-metallic repair option using carbon fiber reinforced polymers (CFRP) as the new pressure boundary.

scholarly journals Assessment of CTF Boiling Transition and Critical Heat Flux Modeling Capabilities Using the OECD/NRC BFBT and PSBT Benchmark Databases

2013 ◽  
Vol 2013 ◽  
pp. 1-12
Author(s):  
Maria Avramova ◽  
Diana Cuervo
Keyword(s):  
Nuclear Energy ◽  
Nuclear Regulatory Commission ◽  
Nucleate Boiling ◽  
State University ◽  
Energy Agency ◽  
Fine Mesh ◽  
The Us ◽  
Penn State ◽  
Regulatory Commission ◽  
Rod Array

Over the last few years, the Pennsylvania State University (PSU) under the sponsorship of the US Nuclear Regulatory Commission (NRC) has prepared, organized, conducted, and summarized two international benchmarks based on the NUPEC data—the OECD/NRC Full-Size Fine-Mesh Bundle Test (BFBT) Benchmark and the OECD/NRC PWR Sub-Channel and Bundle Test (PSBT) Benchmark. The benchmarks’ activities have been conducted in cooperation with the Nuclear Energy Agency/Organization for Economic Co-operation and Development (NEA/OECD) and the Japan Nuclear Energy Safety (JNES) Organization. This paper presents an application of the joint Penn State University/Technical University of Madrid (UPM) version of the well-known sub-channel code COBRA-TF (Coolant Boiling in Rod Array-Two Fluid), namely, CTF, to the steady state critical power and departure from nucleate boiling (DNB) exercises of the OECD/NRC BFBT and PSBT benchmarks. The goal is two-fold: firstly, to assess these models and to examine their strengths and weaknesses; and secondly, to identify the areas for improvement.

scholarly journals An Approach for Assessing Structural Uplifting Using Blast Motions

2008 ◽  
Author(s):  
Jinsuo Nie ◽  
Jim Xu ◽  
Charles H. Hofmayer ◽  
Syed A. Ali
Keyword(s):  
Coal Mine ◽  
Nuclear Power ◽  
Large Scale ◽  
Ground Motions ◽  
National Laboratory ◽  
Field Tests ◽  
Field Measurements ◽  
Nuclear Regulatory Commission ◽  
Brookhaven National Laboratory ◽  
Ground Acceleration

When a nuclear power plant (NPP) structure is subjected to beyond-design-basis seismic motions, a localized nonlinear effect on the soil-structure system is attributed to separations between the structure and the surrounding soils such as basemat uplift. Experiments involving field tests for real seismic events are usually difficult because of the low probability for large earthquakes at any particular site. To this end, the magnitudes of blast-induced ground motions at a coal mine have been found to be predicatable and can reach very large values. An approach has been developed to investigate whether the strong ground motions recorded at this coal mine can be used to evaluate the basemat uplift effect. This approach involves the use of a scaled ground motion to establish the relationship between the basemat uplift and the peak ground acceleration (PGA). This paper summarizes the field measurements for the ground motions at a coal mine by the Japan Nuclear Safety Organization (JNES) and a method using large scale finite element analyses for basemat uplift assessment performed by Brookhaven National Laboratory for the US Nuclear Regulatory Commission.

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).

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