Technical basis-Spent Nuclear Fuel project radiation and contamination trending program

2003 ◽  
Author(s):  
R E ELDER
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Technical Basis

Recent Development of Code Case on Use of Ductile Cast Iron for Transport and Storage Cask for Spent Nuclear Fuel

2008 ◽  
Author(s):  
Taku Arai ◽  
Toshiari Saegusa ◽  
Roland Hueggenberg
Keyword(s):  
Fracture Toughness ◽  
Cast Iron ◽  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Fatigue Curve ◽  
External Pressure ◽  
Ductile Cast Iron ◽  
Fracture Toughness Test ◽  
Technical Basis ◽  
And Storage

Code Case N-670 “Use of Ductile Cast Iron Conforming to ASTM A874/A 874M-98 or JIS G5504-1992 for Transport Containments, Section III, Division 3” which permits use of ductile cast iron for transport containments of spent nuclear fuel was revised to the Code Case N-670-1, “Use of Ductile Cast Iron Conforming to ASTM A874/A 874M-98 or JIS G5504-2005 for Transport and Storage Containments, Section III, Division 3”. Items revised were as follows: (a) Scope was expanded to use for transport and storage, and changed to conform year edition of JIS G5504, (b) The elongation requirement was deleted form the code case to reflect the change of year edition of JIS G5504, (c) Temperature condition of −40 °C was clearly provided for fracture toughness test, (d) Design fatigue curve was re-established, (e) External pressure chart was re-established. Technical basis of the revised code case are described in this paper.

Technical Basis Summary for Code Case N-860: Inspection Requirements and Evaluation Standards for Spent Nuclear Fuel Storage and Transportation Containment Systems

2021 ◽  
Author(s):  
John E. Broussard
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Degradation Mechanism ◽  
Evaluation Criteria ◽  
Sample Population ◽  
Spent Fuel ◽  
Evaluation Standard ◽  
Fuel Storage ◽  
Size Evaluation ◽  
Technical Basis

Abstract ASME Boiler and Pressure Vessel Code Section XI Code Case N-860 provides inspection requirements and evaluation standard for welded stainless steel canisters used for spent nuclear fuel storage. The Code Case defines an initial inspection interval and populations then defines examination requirements that are based on the primary degradation mechanism, chloride induced stress corrosion cracking (CISCC). Additional examination requirements are based on the results of the initial screening examination. This paper summarizes the technical basis for the examinations and the evaluation criteria defined in Code Case N-860. Technical basis information for topics related to the inservice inspection and the flaw evaluation of canisters are described. The topics related to inservice inspection include: 1) the reasons for and the basis of requirements for the site susceptibility of the canister installations, 2) the inspection intervals required by the Code Case, 3) the inspection sample population required by the Code Case, 4) the methods and acceptance criteria for visual examinations required by the Code Case, and 5) the size and location of the required inspection region for supplemental examinations. The topics related to the flaw evaluation include: 1) a summary of the crack growth rate technical basis, and 2) background related to flaw size evaluation for spent fuel canisters.

Convective processes in spent nuclear fuel canisters

2002 ◽  
Author(s):  
Glenn E. McCreery ◽  
Keith G. Condie ◽  
Randy C. Clarksean ◽  
Donald M. McEligot
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Convective Processes ◽  
Nuclear Fuel Canisters

A new approach to spent nuclear fuel recycling for light water reactors in the frame of remix concept

2020 ◽  
Vol 2020 (1) ◽  
pp. 67-77
Author(s):  
Nikita Vladimirivich Kovalyov ◽  
Boris Yakovlevich Zilberman ◽  
Nikolay Dmitrievich Goletskiy ◽  
Andrey Borisovich Sinyukhin
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Light Water Reactors ◽  
New Approach ◽  
Light Water ◽  
Water Reactors

Development of Criticality Accident Alarm System to Ensure Safe Retrieval of Spent Nuclear Fuel from Lepse Floating Maintenance Base

ANRI ◽  
2020 ◽  
pp. 45-53
Author(s):  
A. Lachugin ◽  
M. Kocherygin ◽  
A. Gayazov ◽  
Yury Martynyuk ◽  
A. Vasil'ev
Keyword(s):  
System Design ◽  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Alarm System ◽  
Criticality Accident ◽  
Hazardous Area ◽  
Key Features

The paper presents basic results of development of a criticality accident alarm system to ensure safe retrieval of the spent nuclear fuel from the Lepse Floating Maintenance Base. The key features and engineering aspects of the system design are described. Locations of criticality detector units and selected alarm level settings are justified, hazardous area boundaries were identified, and parameters to identify inadequately protected zones were calculated. The SRKS-01D criticality accident alarm system by SPC “Doza” was selected as base equipment. The system was commissioned in 2019 and has been successfully operated for more than 6 months.

Criticality Consequence Analysis Roadmap for Spent Nuclear Fuel Canister in a Repository

2018 ◽  
Author(s):  
Kaushik Banerjee ◽  
Thomas M. Evans ◽  
Gregory G. Davidson ◽  
Steven P. Hamilton
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Consequence Analysis
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