scholarly journals Oak Ridge National Laboratory Technical Input for the Nuclear Regulatory Commission Review of the 2017 Edition of ASME Section III, Division 5, ‘High Temperature Reactors’

2020 ◽  
Author(s):  
Weiju Ren ◽  
Jude Foulds ◽  
Roger Miller ◽  
Wolfgang Hoffelner
Keyword(s):  
High Temperature ◽  
National Laboratory ◽  
Nuclear Regulatory Commission ◽  
Oak Ridge ◽  
Oak Ridge National Laboratory ◽  
High Temperature Reactors ◽  
Technical Input ◽  
Regulatory Commission

scholarly journals Experimental studies in high temperature aqueous chemistry at Oak Ridge National Laboratory

1997 ◽  
Vol 69 (5) ◽  
pp. 905-914 ◽  
Author(s):  
R. E. Mesmer ◽  
D. A. Palmer ◽  
J. M. Simonson ◽  
H. F. Holmes ◽  
P. C. Ho ◽  
...  
Keyword(s):  
High Temperature ◽  
National Laboratory ◽  
Experimental Studies ◽  
Aqueous Chemistry ◽  
Oak Ridge ◽  
Oak Ridge National Laboratory

Verification of New Capabilities of Deterministic Load Module of FAVOR 09.1

2010 ◽  
Author(s):  
Shengjun Yin ◽  
Terry L. Dickson ◽  
Paul T. Williams ◽  
B. Richard Bass
Keyword(s):  
Computational Study ◽  
National Laboratory ◽  
Nuclear Regulatory Commission ◽  
Pressurized Water ◽  
Oak Ridge ◽  
Problem Class ◽  
Wide Range ◽  
Significant Generalization ◽  
Water Reactors ◽  
Regulatory Commission

This paper describes a computational study conducted by the Probabilistic Pressure Boundary Integrity Safety Assessment (PISA) program at Oak Ridge National Laboratory (ORNL) in support of the Nuclear Regulatory Commission (NRC) sponsored verification of the new capabilities of the latest version of Fracture Analysis of Vessels – Oak Ridge (FAVOR) 09.1. The v09.1 version of FAVOR represents a significant generalization over previous versions, because the problem class for FAVOR has been extended to encompass a broader range of transients and vessel geometries. FAVOR, v09.1, provides the capability to perform both deterministic and risk-informed fracture analyses of boiling water reactors (BWRs) as well as pressurized water reactors (PWRs) subjected to heat-up and cool-down transients. In this study, deterministic solutions generated with the FAVOR v09.1 code for a wide range of representative internal/external surface-breaking flaws and embedded flaws subjected to selected thermal-hydraulic transients were benchmarked with the solutions obtained from ABAQUS (version 6.9-1) for the same transients. Based on the benchmarking analyses, it is concluded that the deterministic module implemented into FAVOR, v09.1, satisfies the criteria described in the FAVOR software design documentation.

scholarly journals Deployment and Operation of the ES-3100 Type B Shipping Container

2006 ◽  
Author(s):  
Jeffrey G. Arbital ◽  
Dean R. Tousley ◽  
Dennis B. Miller
Keyword(s):  
National Laboratory ◽  
Nuclear Regulatory Commission ◽  
Department Of Energy ◽  
Radioactive Material ◽  
Oak Ridge ◽  
Shipping Container ◽  
Enriched Uranium ◽  
Transportation Research ◽  
Regulatory Commission ◽  
The U.S

The U.S. Department of Energy (DOE) National Nuclear Security Administration (NNSA) is shipping, for disposition purposes, bulk quantities of fissile materials, primarily highly enriched uranium (HEU). The U.S. Department of Transportation (DOT) specification 6M container has been the workhorse for NNSA and many other shippers of radioactive material since the 1980s. However, the 6M does not conform to the packaging requirements in the Code of Federal Regulations (10 CFR 71) and, for that reason, is being phased out for use in the DOE secure transportation system by the end of 2006. BWXT Y-12 developed and licensed the ES-3100 container to replace the DOT 6M. The ES-3100 was certified by the Nuclear Regulatory Commission (NRC) in April 2006. The process of deploying the new package began in June 2005 and is planned to be completed in July 2006. The package will be fully operational and completely replace the DOT 6M at the Y-12 National Security Complex (Y-12) by October 2006. This paper reviews the deployment process and the mock loading station that was installed at National Transportation Research Center (NTRC) of Oak Ridge National Laboratory. Specialized equipment, tools, and instrumentation that support the handling and loading operations of the ES-3100 are described in detail. Loading options for other user sites are explored in preparation for deployment of this new state-of-the-art shipping container throughout the DOE complex and the private sector.

scholarly journals EVALUASI DAN PERKEMBANGAN PEMBUATAN BAHAN BAKAR KERNEL UO2 DI PTAPB-BATAN YOGYAKARTA

2007 ◽  
Vol 10 (1) ◽  
Author(s):  
Hidayati Hidayati ◽  
Sri Rinanti Susilowati ◽  
Didiek Herhady
Keyword(s):  
High Temperature ◽  
National Laboratory ◽  
Sol Gel ◽  
Oak Ridge ◽  
Oak Ridge National Laboratory ◽  
High Temperature Reactor

EVALUASI DAN PERKEMBANGAN PEMBUATAN BAHAN BAKAR KERNEL UO2 DI PTAPBBATANYOGYAKARTA Telah dilakukan evaluasi pembuatan bahan bakar kernel UO2 sertaperkembangannya di Bidang Kimia dan Teknologi Proses Bahan (BKTPB) – PTAPB - BATAN Yogyakarta.Pembuatan kernel UO2 telah dilakukan dengan metode gelasi internal maupun eksternal. Metode gelasiinternal dilakukan dengan cara kombinasi proses KEMA-HKFA (Keuringvan Electrotecnische Materialenat Arnhem-Hkernforchungsanlage) maupun dengan proses ORNL (Oak Ridge National Laboratory),sedangkan metode gelasi eksternal dilakukan dengan proses emulsifikasi NUKEM (Nuclear Chemie undMetalurgie Gmbh). Dengan metode gelasi internal, telah dilakukan berbagai optimasi kondisi prosesnya.Hasil sementara menunjukkan bahwa proses yang paling baik adalah proses ORNL menggunakan mediagelasi TCE (tricloro etilena). Dengan metode gelasi eksternal, telah diperoleh beberapa kondisi optimum,namun masih perlu dilakukan optimasi lebih lanjut. Untuk memilih metode gelasi internal atau eksternaltergantung pada kemudahan proses, murah secara ekonomi serta yang memberikan hasil terbaik.Pemilihan metode belum bisa diputuskan karena belum semua variabel proses dioptimasi. Penelitianmengenai pelapisan kernel UO2 menggunakan silikon karbida (SiC) maupun pirokarbon (PyC) barumerupakan tahap awal, sehingga masih diperlukan optimasi berbagai variabel prosesnya. Penelitianpembuatan kernel UO2 di BKTPB – BATAN Yogyakarta direncanakan untuk pembuatan inti elemen bakarbentuk bola untuk HTR (High Temperature Reactor) dan dikembangkan sebagai bahan awal prosespembuatan pelet (proses SGMP = Sol-Gel Microsphere Pelletization) untuk PHWR (Pressurized HeavyWater Reactor).

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.

Analysis of Circumferential Welds in BWRs for Life Beyond 60

2015 ◽  
Author(s):  
Patrick Purtscher ◽  
Simon Sheng ◽  
Terry Dickson
Keyword(s):  
National Laboratory ◽  
Computer Code ◽  
Nuclear Regulatory Commission ◽  
Boiling Water Reactors ◽  
Circumferential Welds ◽  
Oak Ridge ◽  
Original Analysis ◽  
Asme Code ◽  
Water Reactors ◽  
Regulatory Commission

This paper describes the probabilistic fracture mechanics (PFM) analyzes of the conditional probability of failure (CPF) due to brittle fracture of circumferential welds (CW) from a cold overpressurize event in boiling water reactors (BWR) operated for 72 EFPY. This analysis used the Fracture Analysis for Vessels, Oak Ridge (FAVOR) computer code, developed at the Oak Ridge National Laboratory (ORNL), under United States Nuclear Regulatory Commission (NRC) funding. Two typical vessel configurations and the associated material properties for the beltline materials, CW, axial welds (AW), and plates (PL) were used. The analyses consider the potential effects of different fabrication options, shop vs field. Shop-fabrication is mainly by submerged arc weld (SAW) process, while field fabrication used the shielded metal arc weld (SMAW) process. In either case, repairs would have required the SMAW process. The calculations show that field-fabricated vessels would have a slight increase in the CPF compared to shop-fabricated vessels, but the assumed fraction of repair welds was more significant than the fabrication option. The details demonstrate the relative importance of surface-breaking flaws vs. embedded flaws for the assumed transient. The results confirm the conclusions from the original analysis from BWRVIP-05 and BWRVIP-74, the CPF for CW is orders of magnitude less than that of PL and AW regions of the vessel; therefore, the ASME Code-required volumetric examinations of the CW every 10 years as part of the in-service inspection (ISI) program does not change the overall CPF for the vessel. In all the cases analyzed, the total CPF values of the BWRs for 72 EFPY are below the goal for safe operation.

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.

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