An investigation of thermal release of carbon-14 from PWR Zircaloy spent fuel cladding-Reference-Cited by-同舟云学术

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An investigation of thermal release of carbon-14 from PWR Zircaloy spent fuel cladding

Published:1993-03 Issue:1 Volume:200 Page:128-137
ISSN:0022-3115
Container-title:Journal of Nuclear Materials
language:en
Short-container-title:Journal of Nuclear Materials

Author:

Smith H.D,Baldwin D.L

Publisher

Elsevier BV

Subject

Nuclear Energy and Engineering,General Materials Science,Nuclear and High Energy Physics

Reference13 articles.

1. 10 CFR 60;US Nuclear Regulatory Commission;US Code of Federal Regulations,1982

2. Potential Corrosion and Degradation Mechanisms of Zircaloy Cladding on Spent Nuclear Fuel in a Tuff Repository;Rothman;LLNL UCID-20172,1984

3. Proc. 9th Int. Symp. on Zirconium in the Nuclear Industry;Godlewski,1991

4. Proc. Spectrum '90 Conf.;Guenther,1990

Cited by 8 articles. 订阅此论文施引文献订阅此论文施引文献，注册后可以免费订阅5篇论文的施引文献，订阅后可以查看论文全部施引文献

1. Release and speciation of carbon from Zircaloy-4 in anaerobic and highly alkaline conditions: Comparison of simple immersion and potentiostatic corrosion tests;Radiocarbon;2018-10-03

2. Carbon 14 Distribution in Irradiated BWR Fuel Cladding and Released Carbon 14 after Aqueous Immersion of 6.5 Years;Procedia Chemistry;2016

3. A Study on the Chemical Forms and Migration Behavior of Carbon-14 Leached from the Simulated Hull Waste in the Underground condition;MRS Proceedings;2002

4. Determination of 14C in activated metal waste;Analytica Chimica Acta;2001-11

5. An approach to the assessment of high-level radioactive waste containment — II: radionuclide releases from an engineered barrier system;Nuclear Engineering and Design;2000-10

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