scholarly journals A Framework to Develop Flaw Acceptance Criteria for Structural Integrity Assessment of Multipurpose Canisters for Extended Storage of Used Nuclear Fuel

2014 ◽  
Author(s):  
Poh-Sang Lam ◽  
Robert L. Sindelar ◽  
Andrew J. Duncan ◽  
Thad M. Adams
Keyword(s):  
Stress Corrosion ◽  
Nuclear Fuel ◽  
Stress Corrosion Cracking ◽  
Structural Integrity ◽  
Storage Period ◽  
Corrosion Cracking ◽  
External Loading ◽  
Acceptance Criteria ◽  
Integrity Assessment ◽  
Used Nuclear Fuel

A multipurpose canister (MPC) made of austenitic stainless steel is loaded with used nuclear fuel assemblies and is part of the transfer cask system to move the fuel from the spent fuel pool to prepare for storage, and is part of the storage cask system for on-site dry storage. This weld-sealed canister is also expected to be part of the transportation package following storage. The canister may be subject to service-induced degradation especially if exposed to aggressive environments during possible very long-term storage period if the permanent repository is yet to be identified and readied. Stress corrosion cracking may be initiated on the canister surface in the welds or in the heat affected zone because the construction of MPC does not require heat treatment for stress relief. An acceptance criteria methodology is being developed for flaw disposition should the crack-like defects be detected by periodic Inservice Inspection. The external loading cases include thermal accident scenarios and cask drop conditions with the contribution from the welding residual stresses. The determination of acceptable flaw size is based on the procedure to evaluate flaw stability provided by American Petroleum Institute (API) 579 Fitness-for-Service (Second Edition). The material mechanical and fracture properties for base and weld metals and the stress analysis results are obtained from the open literature such as NUREG-1864. Subcritical crack growth from stress corrosion cracking (SCC), and its impact on inspection intervals and acceptance criteria, is not addressed.

Structural Integrity Assessment of WWER Internals on Stress Corrosion Cracking Criterion

2009 ◽  
Author(s):  
Valentina Fedorova ◽  
Boris Margolin
Keyword(s):  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Structural Integrity ◽  
Corrosion Cracking ◽  
Intergranular Stress Corrosion ◽  
Integrity Assessment ◽  
Dose Curve ◽  
Stress Damage ◽  
Intergranular Stress Corrosion Cracking

Stress-damage dose curve (SDDC) is introduced on the basis of the analysis of experimental data on susceptibility to intergranular stress corrosion cracking (IGSCC) of irradiated stainless steels (SS). Approaches to determination of the SDDC parameters are considered. Based on SDDC calculative procedure for estimation of reactor vessel internals (RVI) lifetime by criterion of initiation crack due to IGSCC is proposed.

Improvement of Probabilistic Fracture Mechanics Analysis Code PASCAL-SP With Regard to Primary Water Stress Corrosion Cracking

2017 ◽  
Author(s):  
Akihiro Mano ◽  
Yoshihito Yamaguchi ◽  
Jinya Katsuyama ◽  
Yinsheng Li
Keyword(s):  
Fracture Mechanics ◽  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Failure Probability ◽  
Structural Integrity ◽  
Corrosion Cracking ◽  
Integrity Assessment ◽  
Primary Water ◽  
Nickel Based Alloy ◽  
Failure Probabilities

Structural integrity assessments for cracked nuclear components are currently performed on the basis of deterministic fracture mechanics in accordance with the Rules on Fitness-for-Service for Nuclear Power Plant of the Japan Society of Mechanical Engineers. On the other hand, probabilistic fracture mechanics (PFM) is expected as a rational method for a structural integrity assessment because it can account for the uncertainties and scatters of various influencing factors and can evaluate quantitative values such as the failure probabilities of the components as the solutions. In the Japan Atomic Energy Agency (JAEA), a PFM analysis code PASCAL-SP was developed in order to evaluate the failure probability of nuclear pipe by taking into account aging degradation mechanisms such as inter-granular stress corrosion cracking (IGSCC) and fatigue in the boiling water reactor (BWR) environment. Recently, a number of cracks due to primary water stress corrosion cracking (PWSCC) have been detected in nickel-based alloy welds in the primary piping of pressurized water reactors (PWRs). Therefore, the structural integrity assessment for primary piping taking PWSCC into consideration has become important. This paper details the improvement of PASCAL-SP to evaluate the failure probability of primary pipe taking PWSCC into consideration. We introduce several probabilistic evaluation models such as crack initiation, crack growth and crack detection models related to PWSCC into PASCAL-SP. As numerical examples, the failure probabilities for circumferential and axial cracks in nickel-based alloy weld in pipe in the PWR primary water environment are calculated. We also evaluate the influence of non-destructive inspection on failure probabilities. On the basis of the numerical results, we conclude that the improved PASCAL-SP is useful for evaluating the failure probability of primary pipe taking PWSCC into account.

scholarly journals Preventing Stress Corrosion Cracking of Spent Nuclear Fuel Dry Storage Canisters

2019 ◽  
Vol 19 ◽  
pp. 346-361 ◽  
Author(s):  
Lloyd Hackel ◽  
Jon Rankin ◽  
Matt Walter ◽  
C Brent Dane ◽  
William Neuman ◽  
...  
Keyword(s):  
Stress Corrosion ◽  
Nuclear Fuel ◽  
Stress Corrosion Cracking ◽  
Spent Nuclear Fuel ◽  
Corrosion Cracking ◽  
Dry Storage

scholarly journals Interaction of Cyclic Loading (Low-Cyclic Fatigue) with Stress Corrosion Cracking (SCC) Growth Rate

2020 ◽  
Vol 2020 ◽  
pp. 1-10
Author(s):  
Rehmat Bashir ◽  
He Xue ◽  
Rui Guo ◽  
Yueqi Bi ◽  
Muhammad Usman
Keyword(s):  
Growth Rate ◽  
Stress Corrosion ◽  
Crack Growth ◽  
Stress Corrosion Cracking ◽  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Cyclic Fatigue ◽  
Joint Effect ◽  
Corrosion Cracking

The structural integrity analysis of nuclear power plants (NPPs) is an essential procedure since the age of NPPs is increasing constantly while the number of new NPPs is still limited. Low-cyclic fatigue (LCF) and stress corrosion cracking (SSC) are the two main causes of failure in light-water reactors (LWRs). In the last few decades, many types of research studies have been conducted on these two phenomena separately, but the joint effect of these two mechanisms on the same crack has not been discussed yet though these two loads exist simultaneously in the LWRs. SCC is mainly a combination of the loading, the corrosive medium, and the susceptibility of materials while the LCF depends upon the elements such as compression, moisture, contact, and weld. As it is an attempt to combine SCC and LCF, this research focuses on the joint effect of SCC and LCF loading on crack propagation. The simulations are carried out using extended finite element method (XFEM) separately, for the SCC and LCF, on an identical crack. In the case of SCC, da/dt(mm/sec) is converted into da/dNScc (mm/cycle), and results are combined at the end. It has been observed that the separately calculated results for SCC da/dNScc and LCF da/dNm of crack growth rate are different from those of joint/overall effect,  da/dNom. By applying different SCC loads, the overall crack growth is measured as SCC load becomes the main cause of failure in LWRs in some cases particularly in the presence of residual stresses.

scholarly journals Detecting and imaging stress corrosion cracking in stainless steel, with application to inspecting storage canisters for spent nuclear fuel

2020 ◽  
Vol 109 ◽  
pp. 102180 ◽  
Author(s):  
Marcel C. Remillieux ◽  
Djamel Kaoumi ◽  
Yoshikazu Ohara ◽  
Marcie A. Stuber Geesey ◽  
Li Xi ◽  
...  
Keyword(s):  
Stainless Steel ◽  
Stress Corrosion ◽  
Nuclear Fuel ◽  
Stress Corrosion Cracking ◽  
Spent Nuclear Fuel ◽  
Corrosion Cracking

Spent Fuel Canister Probabilistic Confinement Integrity Assessment

2017 ◽  
Author(s):  
John E. Broussard ◽  
Shannon Chu ◽  
Kevin Fuhr
Keyword(s):  
Stainless Steel ◽  
Austenitic Stainless Steel ◽  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Probabilistic Model ◽  
Population Based ◽  
Corrosion Cracking ◽  
Spent Fuel ◽  
Integrity Assessment ◽  
Induced Stress

A probabilistic model was developed that considers the likelihood of through-wall penetration of chloride-induced stress corrosion cracking (CISCC) in austenitic stainless steel canisters and compares different population-based sample inspection regimes. This paper describes the inputs and methods used to simulate multiple canisters with a range of susceptibilities. This paper also summarizes results of key illustrative cases.

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