ICEM11 Final Program 9.7.11pm_ICEM07 Final Program ... - Events

Abstracts Session
7) A NEW SAFETY CONCEPT FOR GEOLOGICAL DISPOSAL IN JAPAN (II) (wP-59357)
Kazumi Kitayama, Nuclear Waste Management Organization of Japan (NUMO) (Japan)
In a previous paper for ICEM2009 a new safety concept was proposed for the geological disposal program in Japan, which was
a development of the conventional multi-barrier system concept that takes greater account of the actual behaviour in a deep geological
environment. This paper describes progress of the study. In particular, it points out that geological disposal provides passive
defence-in-depth safety functions based on geochemical equilibrium, and that natural analogues are effective in understanding the
behaviour of engineered/natural barriers over hundreds of thousands of years and in further developing safety concepts. In addition,
a long-term safety concept based on the realistic evolution of a geological disposal system can be explained to both the general
public and experts in a straightforward and comprehensible manner.
Detailed discussion of specific long-term behaviour of engineered and natural barriers, including possible interaction between
barriers, will be reported in the paper.
8) THERMO-HYDRO-MECHANICAL SIMULATION OF A HEATING AND HYDRATION
EXPERIMENTAL STUDY (CHINA-MOCK-UP) IN UNSATURATED GMZ BENTONITE (wP-59212)
Liang Chen, Ju Wang, Yuemiao Liu, Beijing Research Institute Of Uranium Geology;
Federic Collin, University Of Liège; Jingli Xie, Beijing Research Institute Of Uranium Geology (China/Belgium)
The preliminary repository concept for disposal of high level radioactive waste in China is a shaft-tunnel model, located in saturated
zones in granite. The engineered barrier system (EBS) is composed of vitrified waste, canisters, buffer materials, backfill
and seals. Unsaturated compacted bentonite is foreseen by lots of countries as a backfill and sealing material. At the present stage,
the Gaomiaozi (GMZ) bentonite is considered as the candidate buffer and backfill material for the Chinese repository. In order to
study the behavior of this material in coupled THMC conditions, a mock-up facility, named China-Mock-up, has been installed in
Beijing Research Institute of Uranium Geology. The heater, which substitutes a loaded waste canister, is placed inside of the facility.
To simulate the impact of groundwater, the water inflow of the barrier from outer surface is also possible. This paper presents
a numerical study of the coupled THM behavior of China-mock-up test. In the paper, the basic physical characters of GMZ bentonite
and experimental facility of China-mock-up test are firstly presented. Based on the experimental studies, the numerical simulation
is realized by the program of LAGAMINE. In the simulation, the following physical phenomena are taken account: the
transport of liquid (advection) and heat ...
SESSION 33 — RECENT ADVANCES IN PROCESSING AND IMMBOLIZATION OF HLW,
FISSLE MATERIAL AND TRANSUANIC (TRU) - PART 1 OF 2 (2.14)
1) URANIUM METAL OXIDATION, GRINDING AND ENCAPSULATION IN BOROBOND:
TRU WASTE MANAGEMENT (w/oP-59279)
Kevin Cook, Larry Addington, Beth Utley, Boron Products (USA)
Hydrogen generation mitigation for K Basin sludge was examined by encapsulation of uranium metal in BoroBond®, pre-oxidation
of uranium metal with Fentons reagent and grinding of Densalloy SD170, an irradiated uranium metal surrogate. Encapsulation
in BoroBond® resulted in pressure increase rates at 60°C ranging from 0.116 torr/h to 0.186 torr/h compared to 0.240 torr/h
for a uranium metal in water standard. Samples cast with higher water content led to increased rates. A Fentons reagent system
consisting of a simple reagent mix of FeSO4•7H2O, H2O2 and HCl effectively oxidized ¼ cubes of uranium metal in under four
days at room temperature. Increased peroxide addition rate, increased FeSO4•7H2O concentration and low pH all increase the corrosion
rate. Densalloy SD170 with an average particle size of 581 µm with 7.63 % of particles less than 90 µm was milled so that
over 90 % of the Densalloy mass measured less than 90 µm in 6 hours of milling. Acceptable wear rates were seen on wear components
that were from standard materials (Nitronic SS and 440SS).
2) RESULTS OF INVESTIGATION OF HIGH LEVEL SOLID RADIOACTIVE WASTE
FROM PITS NO1 AND NO2 AT THE GREMIKHA SITE (w/oP-59390)
B.S. Stepennov, V.M. Afanasyev, A.Yu. Kazennov, N.V. Kartashev, A.V. Korolev, O.E. Kiknadze, D.V. Pavlov, Kurchatov
Institute; V.V. Eremenko, FGUP SevRAO; G. Fady, CEA; Lucien Pillette-Cousin, Areva TA (Russia/France)
One of the legacies of the Cold War period is the unsafe and uncontrolled storage of high level radioactive waste at the former
soviet navy bases. At the former naval maintenance base of Gremikha in the North-West of Russia, it has been necessary to conduct
characterization of high level waste in order to decide the best way to recover and condition it, with good radiation protection
and safety conditions. France, represented by CEA, associated with AREVA TA as technical support, funded the studies performed
by the Russian partners to define conditions of characterization of this high level waste that has been stored for tens of years in two
pits, filled with water and containing sludge. The Russian Kurchatov Institute conducted the clarification of the water contained in
the pit by removal of sludge using filtration techniques. Then, a 3D model of the content of the pits was realized including all
objects identified after water clarification. Specific methods and technologies were used for that because the dose rates near some
objects were as high as 400 Gy/h. Data obtained makes it possible now to choose the best option for the management of the identified
high level solid radioactive waste. The paper presents the techniques used for clarification of water contained in the pits, the
main results and the expected way forward to recover and condition this high level waste.
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