The Effect of Bromide on Oxygen Yields in Homogeneous α-radiolysis

MRS Advances ◽  
2017 ◽  
Vol 2 (13) ◽  
pp. 711-716 ◽  
Author(s):  
Lovisa Bauhn ◽  
Christian Ekberg ◽  
Patrik Fors ◽  
Kastriot Spahiu

ABSTRACTIn a scenario where ground water enters a canister for spent nuclear fuel in a deep geological repository, the presence of dissolved ions in the water could possibly influence the fuel dissolution due to effects on radiolysis yields. One species of particular interest in this context is bromide, which has a proven ability to scavenge hydroxyl radicals much faster than molecular hydrogen does. As a result, bromide could inhibit the beneficial effect of dissolved hydrogen, which has been shown in γ-radiolysis experiments. However, already a few hundred years after repository closure, α-decay starts to dominate in the radiation field from the spent fuel. Hence, the effects of α-radiolysis are expected to govern the fuel dissolution over the geological timeframes of the repository. In the present work, α-radiolysis experiments have been performed to determine the effect of bromide ions on the yield of hydrogen peroxide by mass spectrometric measurement of its decomposition product oxygen. The use of high activity 238Pu solutions has made it possible to study this effect during pure α-radiolysis from a homogeneously distributed radiation field. To simulate deep bedrock repository conditions, and to minimize the influence of in-leaking O2 from air, the studies were performed using graphite sealed stainless steel autoclaves with an initial atmosphere of 10 bar H2. The results show that addition of 1 mM Br- to the solution gives no significant effect on the O2 yield for radiation doses up to 2 MGy. This lack of effect is most likely explained by the limited radical escape yields from radiation tracks in pure α-radiolysis.

1999 ◽  
Vol 556 ◽  
Author(s):  
V. V. Rondinella ◽  
Hj. matzke ◽  
J. Cobos ◽  
T. Wiss

Abstractα-decay will constitute almost entirely the radiation field in and around spent nuclear fuel after a few hundred years in a geological repository. Pellets of UO 2 containing ˜0.1 and ˜10 wt. % 238Pu were fabricated using a sol-gel method and characterized, comparing their properties to those of undoped UO2. The α-radiation fields of different types of commercial LWR spent fuel are of the same order of magnitude as the fuel with the lower Pu-concentration used in this work. The results of static batch leaching tests at room temperature in demineralized water under anoxic atmosphere showed that the amounts of U released during leaching were higher in the case of UO2 containing 238pu than for undoped UO2. Relatively large amounts of Pu were released after the longest leaching times. Lattice parameter measurements using XRD and hardness measurements by Vickers indentation showed a relatively rapid build-up of α-decay damage in the material stored at ambient temperature with the higher concentration of dopant, while for the material with ˜0.1 wt. % Pu no clear variations were detected during the same time intervals.


Author(s):  
Lara Duro ◽  
Abel Tamayo ◽  
Jordi Bruno ◽  
Aurora Marti´nez-Esparza

Source term models are widely used to assess the behaviour of spent nuclear fuel after final disposal. However, most models do not take into account some phenomena which are expected to control the transport of radionuclides through the near field. Some uncertainties arise from this fact, thus making it difficult to obtain proper simulations of radionuclide behaviour in the near field. In this work, we have used a compartmental code to build up an integrated source term model in an attempt to overcome the abovementioned drawbacks. The model developed takes into account radiolytically-mediated matrix dissolution, radioactive decay chains, diffusive transport, and retardation by sorption and secondary phase precipitation, among other processes. In addition, this model has been used to estimate radionuclide mobility from spent fuel located in a conceptual clay geological repository.


2019 ◽  
Vol 133 ◽  
pp. 02005
Author(s):  
Markéta Camfrlová

Nuclear energy accounts for a significant part of the total energy production in the Czech Republic, which is currently facing a problem dealing with the high-level radioactive waste (HLW) and the spent nuclear fuel (SNF). Deep repository is the safest option for storage of HLW. Rock environment of the area must guarantee the stability of the deep geological repository for at least 100,000 years. The aim of the research is a long-term evaluation of the climatic changes of the hypothetical area of interest, which corresponds to the candidate sites for deep geological repository in the Czech Republic. The occurrences of endogenous and exogenous phenomena, which could affect site stability, were evaluated. Concerning exogenous processes, research focuses mainly on the assessment of climatic effects. The climate scenarios for the Central Europe were examined – global climate change, glaciation, and the depth of permafrost as well as CO2 increase.


Author(s):  
Donald Wayne Lewis

In the United States (U.S.) the nuclear waste issue has plagued the nuclear industry for decades. Originally, spent fuel was to be reprocessed but with the threat of nuclear proliferation, spent fuel reprocessing has been eliminated, at least for now. In 1983, the Nuclear Waste Policy Act of 1982 [1] was established, authorizing development of one or more spent fuel and high-level nuclear waste geological repositories and a consolidated national storage facility, called a “Monitored Retrievable Storage” facility, that could store the spent nuclear fuel until it could be placed into the geological repository. Plans were under way to build a geological repository, Yucca Mountain, but with the decision by President Obama to terminate the development of Yucca Mountain, a consolidated national storage facility that can store spent fuel for an interim period until a new repository is established has become very important. Since reactor sites have not been able to wait for the government to come up with a storage or disposal location, spent fuel remains in wet or dry storage at each nuclear plant. The purpose of this paper is to present a concept developed to address the DOE’s goals stated above. This concept was developed over the past few months by collaboration between the DOE and industry experts that have experience in designing spent nuclear fuel facilities. The paper examines the current spent fuel storage conditions at shutdown reactor sites, operating reactor sites, and the type of storage systems (transportable versus non-transportable, welded or bolted). The concept lays out the basis for a pilot storage facility to house spent fuel from shutdown reactor sites and then how the pilot facility can be enlarged to a larger full scale consolidated interim storage facility.


Author(s):  
Hsoung-Wei Chou ◽  
Szu-Ying Wu

Abstract When a canister used for final disposal of spent nuclear fuel is in the deposition hole of deep geological repository, the loading case which would most impact the structure integrity of the canister is to postulate an earthquake induced rock shear through a deposition hole. This paper evaluates the acceptable sizes of defects in the cast iron insert of canister using fracture mechanics analysis. The submodelling technique of finite element analysis was employed to calculate the fracture behavior of the canister with postulated defects subjected to shear loads due to earthquake. At first, the stresses of the global model and the uncracked submodel were compared to check the correctness of the transferred displacements from the global model to the submodel. Then, surface and internal flaws with various shapes and depths were modeled individually in the submodel. A 5 cm shear displacement was applied on buffer and then transmitted to the canister. The calculations of J-integral of each flaw indicate that the surface semi-elliptical defect is more critical and the density of buffer material significantly affects the acceptable size of postulated defects. Present results can provide acceptance criteria of crack detection for canisters during manufacture and examination processes.


2010 ◽  
Vol 240 (3) ◽  
pp. 668-671 ◽  
Author(s):  
Dušan Ćalić ◽  
Matjaž Ravnik

2006 ◽  
Vol 932 ◽  
Author(s):  
Christophe Poinssot ◽  
Cécile Ferry ◽  
Bernd Grambow ◽  
Manfred Kelm ◽  
Kastriot Spahiu ◽  
...  

ABSTRACTEuropean Commission supported a wide research project entitled “Spent Fuel Stability under repository conditions” (SFS) within the 5th FWP, the aim of which was to develop a common understanding of the radionuclides release from spent nuclear fuel in geological disposal and build a RN release model in order to assess the fuel performance. This project achieved by the end of 2004 focuses both on the Instant Release Fraction (IRF) model and the Matrix Alteration Model (MAM).A new IRF model was developed based on the anticipated performances of the various fuel microstructures (gap, rim, grains boundaries) and the potential diffusion of RN before the canister breaching. However, this model lets the choice to the end-user about the degree of conservativeness to consider.In addition, fuel alteration has been demonstrated to be linked to the production of radiolytic oxidants by water radiolysis at the fuel interface, the oxidation of the fuel interface by radiolytic oxidants and the subsequent release of uranium under the influence of aqueous ligands. A large set of experimental data was therefore acquired in order (i) to upgrade the current radiolytic kinetic scheme, (ii) to experimentally correlate the fuel alteration rate and the fuel specific alpha activity by performing experiments on alpha doped samples, (iii) to experimentally test the potential inhibitor effect of hydrogen on fuel dissolution. Based on these results, a new MAM was developed, which was also calibrated using the experiments on inactive UO2 samples. This model was finally applied to representative granitic, salt and clayey environment to predict spent fuel long-term fuel performance.


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