The 12th International Conference on Environmental ... - Events

Session 40-41 Abstracts
Previous theoretical studies carried out recently have illustrated the potential magnitude of the problem, with reference to the
fundamental nuclear data and typical isotopic compositions of wastes. Neutron multiplicity counting can, in principle, differentiate
between isotopes that undergo spontaneous fission, however in practice the uncertainties in waste assay are such that this is rarely
beneficial. More practical “compensation” techniques use combinations of different assay techniques (for example passive and
active neutron counting) and knowledge of the actinide ratios in the waste stream fingerprint.
In this paper we describe various waste assay applications as case studies. For each example we describe the nature of the challenge
and show how solutions have been developed for applications where the presence of curium has caused problems. We
describe the technical solutions, showing the limitations and assumptions of each. We also emphasise the role of robust Quality
Assurance procedures, to ensure that the techniques are implemented reliably and with predictable outcomes. Finally, we describe
the benefits that have been realised for the plant operations teams, with regard to improved measurement accuracy, avoidance of
false over-estimation of the Pu inventory and subsequent improvement in plant throughput.
4) THEORETICAL MODELLING OF NUCLEAR WASTE FLOWS - 16377
J.F. Adams, S.R. Biggs, M. Fairweather, D. Njobuenwu and J. Yao, University of Leeds (UK)
A large amount of nuclear waste is stored in tailings ponds as a solid-liquid slurry, and liquid flows containing suspensions of
solid particles are encountered in the treatment and disposal of this waste. In processing this waste, it is important to understand
the behaviour of particles within the flow in terms of their settling characteristics, their propensity to form solid beds, and the resuspension
characteristics of particles from a bed. A clearer understanding of such behaviour would allow the refinement of current
approaches to waste management, potentially leading to reduced uncertainties in radiological impact assessments, smaller
waste volumes and lower costs, accelerated clean-up, reduced worker doses, enhanced public confidence and diminished grounds
for objection to waste disposal.
Mathematical models are of significant value in nuclear waste processing since the extent of characterisation of wastes is in
general low. Additionally, waste processing involves a diverse range of flows, within vessels, ponds and pipes. To investigate experimentally
all waste form characteristics and potential flows of interest would be prohibitively expensive, whereas the use of mathematical
models can help to focus experimental studies through the more efficient use of existing data, the identification of data
requirements, and a reduction in the need for process optimisation in full-scale experimental trials. Validated models can also be
used to predict waste transport behaviour to enable cost effective process design and continued operation, to provide input to
process selection, and to allow the prediction of operational boundaries that account for the different types and compositions of particulate
wastes.
SESSION 41 - TRANSPORTATION AND STORAGE OF HLW, FISSILE, TRU, AND SNF
1) EXPERIENCE WITH DRY CASK STORAGE TECHNOLOGY IN GERMANY - 16416
Heinz Geiser, Jens Schroeder, Dietrich Hoffmann, GNS mbH (Germany)
In Germany spent fuel will exclusively be disposed of in deep geological formations. Until the availability of a final repository,
spent fuel assemblies (SFA) are mainly stored in casks of the CASTOR® type which constitute dual-purpose casks for transport
as well as for storage. For the storage of these casks, interim storage facilities (ISF) have been constructed on the sites of each
nuclear power plant (NPP) in Germany. In the meantime GNS has loaded more than 1000 casks at 25 different sites in Germany.
This paper will discuss the experience gained during the first decade of using dry cask storage technology.
2) CONTINGENCY OPTIONS FOR THE DRY STORAGE OF MAGNOX SPENT FUEL IN THE UK - 16330
Jenny E. Morris, Galson Sciences Limited (UK); Stephen Wickham, Phil Richardson, Galson Sciences Ltd. (UK);
Colin Rhodes, NDA (UK); Mike Newland, UKAEA (UK)
<strong>The</strong> UK Nuclear Decommissioning Authority (NDA) is responsible for safe and secure management of spent nuclear fuel.
Magnox fuel is held at some Magnox reactor sites and at Sellafield where it is reprocessed using a number of facilities. It is intended
that all Magnox fuel will be reprocessed as described in the published Magnox Operating Programme (MOP). In the event, however,
that a failure occurs within the reprocessing plant, the NDA has initiated a programme of activities to explore alternative contingency
options for the management of wetted Magnox spent fuel.
Magnox fuel comprises metallic uranium bar clad in a magnesium alloy, both of which corrode if exposed to oxygen or water.
Consequently, contingency options are required to consider how best to manage the issues associated with the reactivity of the metals.
Questions such as whether Magnox spent fuel needs to be dried, how it might be conditioned, how it might be packaged and
held in temporary storage until a disposal facility becomes available, all require attention.
During storage in the presence of water, the corrosion of Magnox fuel produces hydrogen (H2) gas, which requires careful
management. When uranium reacts with hydrogen in a reducing environment, the formation of uranium hydride (UH3) may occur,
which under some circumstances can be pyrophoric, and might create hazards which may affect subsequent retrieval and/or repackaging
(e.g. for disposal). Other factors that may affect the choice of a viable contingency option include criticality safety, environmental
impacts, security and Safeguards and economic considerations.
3) WIPP: A PERSPECTIVE FROM TEN YEARS OF OPERATING SUCCESS - 16189
Phillip C. Gregory, Washington TRU Solutions (USA)
<strong>The</strong> Waste Isolation Pilot Plant (WIPP) located 35 miles east of Carlsbad, New Mexico, USA is the first and, to the author’s
knowledge, only facility if the world for the permanent disposal of defense related transuranic (TRU) waste. Soon after plutonium
was first synthesized in 1940 by a team of scientists at the University of California’s Berkley Laboratory the need to find a permanent
repository for plutonium contaminated waste was recognized due to the 24,100 year half-life of Plutonium-239. In 1957 the
National Academy of Sciences published a report recommending deep geological burial in bedded salt as a possible solution. However,
more than 50 years passed before the solution was realized when in 1999 WIPP received the first shipment of TRU waste from
the Los Alamos National Laboratory. Ten years later, more than 6,000 shipments of TRU waste have been disposed of in rooms
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