Treatment of Irradiated Graphite from French Bugey Reactor - 13424
Creators
- 1. Studsvik, Inc., 5605 Glenridge Drive NE, Suite 705, Atlanta, GA (United States)
- 2. electricite de France, 154 Avenue Thiers, CS 60018, 69458 Lyon Cedex 06 (France)
Description
Beginning in 2009, in order to determine an alternative to direct disposal for decommissioned irradiated graphite from EDF's Bugey NPP, Studsvik and EDF began a test program to determine if graphite decontamination and destruction were practicable using Studsvik's thermal organic reduction (THOR) technology. The testing program focused primarily on the release of C-14, H-3, and Cl-36 and also monitored graphite mass loss. For said testing, a bench-scale steam reformer (BSSR) was constructed with the capability of flowing various compositions of gases at temperatures up to 1300 deg. C over uniformly sized particles of graphite for fixed amounts of time. The BSSR was followed by a condenser, thermal oxidizer, and NaOH bubbler system designed to capture H-3 and C-14. Also, in a separate series of testing, high concentration acid and peroxide solutions were used to soak the graphite and leach out and measure Cl-36. A series of gasification tests were performed to scope gas compositions and temperatures for graphite gasification using steam and oxygen. Results suggested higher temperature steam (1100 deg. C vs. 900 deg. C) yielded a practicable gasification rate but that lower temperature (900 deg. C) gasification was also a practicable treatment alternative if oxygen is fed into the process. A series of decontamination tests were performed to determine the release behavior of and extent to which C-14 and H-3 were released from graphite in a high temperature (900-1300 deg. C), low flow roasting gas environment. In general, testing determined that higher temperatures and longer roasting times were efficacious for releasing H-3 completely and the majority (80%) of C-14. Manipulating oxidizing and reducing gas environments was also found to limit graphite mass loss. A series of soaking tests was performed to measure the amount of Cl-36 in the samples of graphite before and after roasting in the BSSR. Similar to C-14 release, these soaking tests revealed that 70-80% Cl-36 is released during roasting tests. (authors)
Availability note (English)
Available from: WM Symposia, 1628 E. Southern Avenue, Suite 9-332, Tempe, AZ 85282 (US)Additional details
Publishing Information
- Imprint Pagination
- 6 p.
- Report number
- INIS-US--13-WM-13424
Conference
- Title
- International collaboration and continuous improvement
- Acronym
- Waste Management 2013 - WM2013 Conference
- Dates
- 24-28 Feb 2013
- Place
- Phoenix, AZ (United States)
INIS
- Country of Publication
- United States
- Country of Input or Organization
- France
- INIS RN
- 45042351
- Subject category
- S38: RADIATION CHEMISTRY, RADIOCHEMISTRY AND NUCLEAR CHEMISTRY; S12: MANAGEMENT OF RADIOACTIVE WASTES, AND NON-RADIOACTIVE WASTES FROM NUCLEAR FACILITIES;
- Resource subtype / Literary indicator
- Conference, Non-conventional Literature
- Descriptors DEI
- CARBON 14; CHLORINE 36; CONCENTRATION RATIO; DECONTAMINATION; GASIFICATION; GRAPHITE; HEAT EXCHANGERS; IRRADIATION; MASS TRANSFER; RADIOACTIVE WASTE DISPOSAL; ROASTING; STEAM; TRITIUM; VAPOR CONDENSERS
- Descriptors DEC
- BETA DECAY RADIOISOTOPES; BETA-MINUS DECAY RADIOISOTOPES; BETA-PLUS DECAY RADIOISOTOPES; CARBON; CARBON ISOTOPES; CHEMICAL REACTIONS; CHLORINE ISOTOPES; CLEANING; DIMENSIONLESS NUMBERS; ELECTRON CAPTURE RADIOISOTOPES; ELEMENTS; EVEN-EVEN NUCLEI; HYDROGEN ISOTOPES; ISOTOPES; LIGHT NUCLEI; MANAGEMENT; MINERALS; NONMETALS; NUCLEI; ODD-EVEN NUCLEI; ODD-ODD NUCLEI; OXIDATION; RADIOACTIVE WASTE MANAGEMENT; RADIOISOTOPES; THERMOCHEMICAL PROCESSES; WASTE DISPOSAL; WASTE MANAGEMENT; YEARS LIVING RADIOISOTOPES
Optional Information
- Notes
- 5 refs.