Simulated Environmental Tests for Selected ATF Cladding Solutions

2021 ◽  
Author(s):  
Sami Penttila ◽  
Juha-Matti Autio ◽  
Jari Lydman ◽  
Aki Toivonen ◽  
Seppo Peltonen ◽  
...  
Keyword(s):  
High Temperature ◽  
Base Material ◽  
Operating Conditions ◽  
Advanced Technology ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
High Temperature Furnace ◽  
Normal Light ◽  
C 30

Abstract Current development on advanced technology fuel (ATF) claddings is aiming at improved high temperature integrity of new cladding solutions that are based on the existing zirconium claddings. To assess their performance for commercial use, their thorough characterization is essential. The primary requirement for the cladding materials is the ability to tolerate loss of cooling for a significant period without failing. The tests in this work were performed on different types of coated Zr-alloys in a high temperature furnace in flowing steam conditions at 1100 °C/ 60 min, 1200 °C/ 30 min and 1300 °C / 5 min. In addition, exposures were performed in pressurized water reactor (PWR) water chemistry to confirm the material viability in normal light water reactor (LWR) operating conditions. After PWR and steam tests, the exposed specimens were studied using a Zeiss Crossbeam 540 field emission gunscanning electron microscope (FEG-SEM) equipped with a semi-quantitative energy dispersive X-ray spectrometer (EDS). Most of the tested specimens indicated detached coating layer. Varying amounts of cracking in the coatings were present. Some of the cracks extended into the base material. Based on this study, further development is needed.

scholarly journals SOURCE TERM ASSESSMENT FOR 100 MWE PRESSURIZED WATER REACTOR

2020 ◽  
Vol 22 (2) ◽  
pp. 61
Author(s):  
Pande Made Udiyani ◽  
Muhammad Budi Setiawan
Keyword(s):  
Power Plants ◽  
Public Perception ◽  
Source Term ◽  
Fission Products ◽  
Operating Conditions ◽  
Normal Operation ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Starting Point

One of the barriers on the implementation of nuclear energy in Indonesia is public perception towards the safety of nuclear power plants (NPPs). Therefore, it is necessary to perform a study about the radiation impact of normal and abnormal operations of an NPP. In accordance to the program of Ministry of Research and Technology period 2020-2024, concerning the plan to build a small modular reactor (SMR)-type NPP, a radiation safety study has been performed for the 100 MWe Pressurized Water Reactor (PWR-100MWe). Source term release of radioactive substances into the environment from PWR-100MWe is a starting point in the study of the radiological consequences of reactor operation. Therefore, this paper will examine the PWR-100MWe source term under normal and abnormal operating conditions, according to the design and the design basis accident (DBA). The initial trigger of the DBA is Lost of Coolant Accident (LOCA) such as Small LOCA and Large LOCA.  Due to the limitations of available SMR data, the study of PWR-100MWe source term refers to the assumption of the release fraction of fission products per subsystem in a larger 1000MWe PWR. It is expected from this assumption that pessimistic source term will be obtained. The study begins with calculation of PWR-100MWe core inventory using ORIGEN2 code based on PWR-100MWe reactor parameters. Through the mechanistic source term model and PWR-1000MWe release parameters, source terms will be obtained for normal operation and abnormal conditions i.e. DBA. Normal source term is used to calculate the consequences of normal operation, which will be used for environmental monitoring and environmental safety analysis of the site. Whereas accident source term is the basis for calculating the radiological consequences of accidents used for SAR documents and nuclear preparedness.Keywords: SMR, PWR-100MWe, normal operation, source term, accident

ASME Section XI Appendix L Flaw Tolerance Evaluation of Pressurized Water Reactor Piping Systems to Support Second License Renewal (80-Years Operation)

2018 ◽  
Author(s):  
Anees Udyawar ◽  
Charles Tomes ◽  
Alexandria Carolan ◽  
Steve Marlette ◽  
Thomas Meikle ◽  
...  
Keyword(s):  
Crack Growth ◽  
Structural Integrity ◽  
Operating Conditions ◽  
Pressurized Water Reactor ◽  
Safe Operation ◽  
Pressurized Water ◽  
Water Reactor ◽  
Flaw Tolerance ◽  
Piping Systems ◽  
Operating Plant

One of the goals of ASME Section XI is to ensure that systems and components remain in safe operation throughout the service life, which can include plant license extensions and renewals. This goal is maintained through requirements on periodic inspections and operating plant criteria as contained in Section XI IWB-2500 and IWB-3700, respectively. Operating plant fatigue concerns can be caused from operating conditions or specific transients not considered in the original design transients. ASME Section XI IWB-3740, Operating Plant Fatigue Assessments, provides guidance on analytical evaluation procedures that can be used when the calculated fatigue usage exceeds the fatigue usage limit defined in the original Construction Code. One of the options provided in Section XI Appendix L is through the use of a flaw tolerance analysis. The flaw tolerance evaluation involves postulation of a flaw and predicting its future growth, and thereby establishing the period of service for which it would remain acceptable to the structural integrity requirements of Section XI. The flaw tolerance approach has the advantage of not requiring knowledge of the cyclic service history, tracking future cycles, or installing systems to monitor transients and cycles. Furthermore, the flaw tolerance can also justify an inservice inspection period of 10 years, which would match a plant’s typical Section XI in-service inspection interval. The goal of this paper is to demonstrate a flaw tolerance evaluation based on ASME Section XI Appendix L for several auxiliary piping systems for a typical PWR (Pressurized Water Reactor) nuclear power plant. The flaw tolerance evaluation considers the applicable piping geometry, materials, loadings, crack growth mechanism, such as fatigue crack growth, and the inspection detection capabilities. The purpose of the Section XI Appendix L evaluation is to demonstrate that a reactor coolant piping system continues to maintain its structural integrity and ensures safe operation of the plant.

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