Retrieval and Analysis of Simulated Defense HLW Package Experiments at the WIPP

1988 ◽  
Vol 127 ◽  
Author(s):  
Martin A. Molecke ◽  
N. Rob Sorensen
Keyword(s):  
Rock Salt ◽  
Waste Form ◽  
High Level Waste ◽  
General Corrosion ◽  
Uniform Corrosion ◽  
Geochemical Environment ◽  
Waste Package ◽  
Barrier Materials ◽  
Backfill Materials

ABSTRACTIn situ waste package performance experiments involving simulated (non-radioactive) defense high-level waste (DHLW) containers have been in progress since late 1984 at the Waste Isolation Pilot Plant (WIPP) facility. These experiments involve full-size, simulated DHLW containers of several metals and designs emplaced in the WIPP bedded rock salt. These test containers are surrounded by granular backfill (packing) materials, have in many cases been intentionally injected with brines, and are heavily instrumented. A majority of the test packages also contain nonradioactive DHLW borosilicate glass waste form, either within the container and/or outside of it. The primary purpose of these WIPP simulated DHLW experiments is to evaluate the in situ durability and performance of all waste package engineered barrier materials, and to perform package concept validation testing.Twelve of the test DHLW containers, emplaced in WIPP test Room B, have been in heated operation since 1985 and had a maximum surface temperature of about 190°C. These containers were recently retrieved, after about 3 years of heated exposure, for detailed posttest laboratory analyses of: general corrosion and metallurgical degradation, waste form and backfill materials alterations, and other rock salt-brine-barrier materials near-field interactions with the “repository” geochemical environment. Test canisters and overpacks made of ASTM Grade-12 titanium showed essentially no visible degradation in either the base metal or welds; cast mild steel A216/WCA over-packs have suffered some uniform corrosion. Significant degradation of the removed instruments and associated test apparatus has been found: pieces of stainless steel (both 304L and 316) apparatus have undergone extensive stress-corrosion cracking failure and non-uniform attack; Inconel 600-sheathed instruments have undergone both extensive uniform and localized (pitting) attack. Granular backfill materials have been significantly compacted by creep closure to about a density of 2 kg/m. Laboratory analyses are still in progress. Further details on these materials results plus instrumentation data and other in situ WIPP waste package test observations are discussed.

In-Situ Corrosion Studies on Cast Steel for a High-Level Waste Packaging in a Rock Salt Repository

1988 ◽  
Vol 127 ◽  
Author(s):  
W. Schwarzkopf ◽  
E. Smailos ◽  
R. Koster
Keyword(s):  
Electron Beam ◽  
Rock Salt ◽  
Corrosion Behaviour ◽  
Cast Steel ◽  
Maximum Temperature ◽  
High Level Waste ◽  
General Corrosion ◽  
Vertical Temperature Profile ◽  
Corrosion Studies

ABSTRACTPrevious corrosion studies performed on a number of materials have shown that unalloyed steels are promising materials for long-term resistant packagings to be used in disposal of heat-generating wastes in rock salt formations. This is the reason why those steels are the subject of more detailed investigations. This paper reports an in-situ experiment conducted in the Asse salt mine in which the influence of selected characteristics (welding, shape) of containers on the corrosion behaviour of cast steel was studied. The material was tested in NaCl brine which might intrude into an HLW borehole in an accident scenario. For this, an electron beam welded cast-steel tube was stored for 18 months in a 2-m deep heated borehole and the annular gap between the tube and the borehole wall was filled with saturated NaCl brine. The vertical temperature profile in the borehole was in the range from 90°C to 200°C; the maximum temperature occurred in the center of the heated zone and the minimum temperature in the upper parts of tube.Under the testing conditions cast steel was subjected to general corrosion at a maximum corrosion rate of 120 μm/a. Considering this magnitude of the corrosion rates, the resulting corrosion allowances are technically acceptable for a packaging having long service-lives. Pitting and crevice corrosion as well as stress-corrosion cracking did not occur in cast steel, and electron beam welding did not exert a noticeable influence on cast-steel corrosion. With these results available, cast steel continues to be considered as a promising HLW packaging material.

The Waste Package Materials Field Test in S.E. New Mexico Salt

1983 ◽  
Vol 26 ◽  
Author(s):  
Martin A. Molecke ◽  
Teresa M. Torres
Keyword(s):  
New Mexico ◽  
Field Test ◽  
High Density ◽  
Silica Sand ◽  
High Level Waste ◽  
Low Density ◽  
Waste Package ◽  
Backfill Materials ◽  
Package Materials

ABSTRACTA six-part, waste package materials field test was conducted in a halite horizon of a potash mine in southeastern New Mexico. The primary purposes of this test were to evaluate the thermophysical and geochemical performance of candidate HLW-package backfill materials emplaced in rock salt and the corrosion behavior of candidate waste canister or overpack alloys. This field test series also served as a precursor to forthcoming Waste Isolation Pilot Plant (WIPP) in situ waste package performance experiments on simulated defense high-level waste packages, serving to develop applicable testing, instrumentation, and sampling techniques. The backfill materials tested (individually, in one- to five-month tests) were: low-density bentonite clay; low-density bentonite (70 wt.%)-silica sand (30 wt.%) mixtures, both dry and brine-injected; high-density bentonite-sand annular compacts; trapped air; and finely-crushed WIPP salt. The in situ measured thermal conductivities (at a maximum canister-heater surface temperature of 150° or 250°C) for the backfills ranged from 0.25 W/mK for pure bentonite to about 1.25 W/mK for the high-density bentonite-sand. No significant backfill material degradation products were detected in post-test analyses. No appreciable corrosion of the titanium-, nickel-, or iron-based alloys embedded in the hot backfill was found; potentially significant pitting corrosion of 2 1/4 Cr-1 Mo steel and copper was detected.

Materials Interactions Relating to Long-Term Geologic Disposal of Nuclear Waste Glass

1986 ◽  
Vol 84 ◽  
Author(s):  
Ned E. Bibler ◽  
Carol M. Jantzen
Keyword(s):  
Nuclear Waste ◽  
Review Paper ◽  
Waste Glass ◽  
Waste Form ◽  
Glass Waste ◽  
Waste Package ◽  
Geologic Disposal ◽  
Nuclear Waste Glass ◽  
Backfill Materials

AbstractIn the geologic disposal of nuclear waste glass, the glass will eventually interact with groundwater in the repository system. Interactions can also occur between the glass and other waste package materials that are present. These include the steel canister that holds the glass, the metal overpack over the canister, backfill materials that may be used, and the repository host rock. This review paper systematizes the additional interactions that materials in the waste package will impose on the borosilicate glass waste form-groundwater interactions. The repository geologies reviewed are tuff, salt, basalt, and granite. The interactions emphasized are those appropriate to conditions expected after repository closure, e.g. oxic vs. anoxic conditions. Whenever possible, the effect of radiation from the waste form on the interactions is examined. The interactions are evaluated based on their effect on the release and speciation of various elements including radionuclides from the glass. It is noted when further tests of repository interactions are needed before long-term predictions can be made.

The Expected Environment for Waste Packages in a Salt Repository

1983 ◽  
Vol 26 ◽  
Author(s):  
L. R. Pederson ◽  
D. E. Clark ◽  
F. N. Hodges ◽  
G. L. Mcvpy ◽  
D. Rai
Keyword(s):  
Rock Salt ◽  
Near Field ◽  
Materials Testing ◽  
Site Specific ◽  
Waste Package ◽  
Barrier Materials ◽  
Waste Forms ◽  
Temperature Irradiation ◽  
Salt Brine ◽  
Package Materials

ABSTRACTThis paper discusses results of recent efforts to define the very near-field (within approximately 2m) environmental conditions to which waste packages will be exposed in a salt repository. These conditions must be considered in the experimental design for waste package materials testing, which includes corrosion of barrier materials and leaching of waste forms. Site-specific brine compositions have been determined, and “standard” brine compositions have been selected for testing purposes. Actual brine compositions will vary depending on origin, temperature, irradiation history, and contact with irradiated rock salt. Results of irradiating rock salt, synthetic brines, rock salt/brine mixtures, and reactions of irradiated rock salt with brine solutions are reported.

Near Field Radionuclide Concentrations: Sorption or Solubility Constrained?

2001 ◽  
Author(s):  
Bernhard Kienzler ◽  
Peter Vejmelka ◽  
Volker Metz
Keyword(s):  
Portland Cement ◽  
Performance Assessment ◽  
Geochemical Modeling ◽  
Coal Fly Ash ◽  
Near Field ◽  
High Level Waste ◽  
Sorption Behavior ◽  
Geochemical Environment ◽  
Backfill Materials ◽  
High Level

Abstract The amount of mobile radionuclides is controlled by the geochemical isolation potential of the repository. Many investigations are available in order to determine the maximal radionuclide concentrations released from different waste forms of specific disposal strategies for disposal in rock salt formations. These investigations result in reaction (dissolution) rates, maximum concentrations, and sorption coefficients. The experimental data have to be applied to various disposal strategies. The case studies presented in this communication cover the selection, the volumes, and the composition of backfill materials used as sorbents for radionuclides. As an example, for brown coal fly ash (BFA) - Q-brine systems, sorption coefficients were measured as well as solublilities of several actinides and other long-lived radionuclides. Dissolved CO32− was buffered to negligible concentration by the presence of high amount of Mg in solution. In the sorption experiments Pu, Th, Np, and U concentrations close or below detection limit were obtained. Concentrations in the same ranges are computed by means of geochemical modeling, if precipitation of “simple” tetravalent hydroxides (An(OH)4(am) phases) is assumed. In the case of U in a Portland cement dominated geochemical environment, measured U(VI) concentration corresponds to the solubility of hexavalent solids, such as Na2U2O7. A similar behavior of U was observed in high-level waste glass experiments. Experiments investigating sorption behavior of corroded cement showed that in the case of application of a sufficient large inventory of actinides, measured concentrations were found to be independent of the inventory. In this case, measured concentrations were controlled by solid phases. If smaller actinide inventories were applied, resulting concentrations were found to be below concentrations constrained by well-known solids. Here, a more or less pronounced sorption of the radioactive elements was observed. The radionuclide concentrations determined in the BFA “sorption” experiments are found to be close to the detection limits. For this reason, it is not possible to extrapolate the radionuclide behavior to lower concentrations. We cannot distinguish, if sorption or precipitation controls measured radionuclide concentrations. However, in the presence of reducing materials such as BFA, solubilities of tetravalent actinides and of Tc(IV) represent a realistic estimation of the maximal element concentrations needed in performance assessment studies. The concentrations of these redox sensitive elements are controlled by precipitation of An(OH)4(am) phases for disposal concepts considered in German salt formations. Under this assumption, quantities such as solid-solution ratios used in (sorption) experiments do not affect the mobilization of the radionuclides. Additional conclusions can be drawn from comparison of the findings for the redox sensitive elements in the BFA / portland cement brine systems: We can assume that expected actinide and technetium concentrations in the near-field of radioactive wastes are affected by the total inventory of radionuclides in the disposal room. Sorption will be relevant, if the total dissolved radionuclide concentration remains below the maximal solubility defined by the solid radionuclide phase which is stable in the geochemical environment. In contrast to the portland cement system, the relevant radionuclide phase are most probably tetravalent hydroxides in the BFA systems. These conclusions are of high importance to performance assessment for the radioactive waste repository systems, because they restrict the applicability of sorption models in the near field of the waste.

Wipp Waste Package Testing on Simulated DHLW: Emplacement

1984 ◽  
Vol 44 ◽  
Author(s):  
Martin A. Molecke
Keyword(s):  
Thermal Loading ◽  
Accelerated Aging ◽  
Waste Form ◽  
Future Application ◽  
High Level Waste ◽  
Glass Waste ◽  
Waste Package ◽  
High Level ◽  
Artificial Introduction ◽  
Materials Behavior

AbstractSeveral series of simulated (nonradioactive) defense high-level waste (DHLW) package tests have recently been emplaced in the WIPP, a research and development facility authorized to demonstrate the safe disposal of defense-related wastes. The primary purpose of these 3-to-7 year duration tests is to evaluate the in situ materials performance of waste package barriers (canisters, overpacks, backfills, and nonradioactive DHLW glass waste form) for possible future application to a licensed waste repository in salt. This paper describes all test materials, instrumentation, and emplacement and testing techniques, and discusses progress of the various tests.These tests are intended to provide information on materials behavior (i.e., corrosion, metallurgical and geochemical alterations, waste form durability, surface interactions, etc.), as well as comparison between several waste package designs, fabrications details, and actual costs.These experiments involve 18 full-size simulated DHLW packages (approximately 3.0 m x 0.6 m diameter) emplaced in vertical boreholes in the salt drift floor. Six of the test packages contain internal electrical heaters (470 W/canister), and were emplaced under approximately reference DHLW repository conditions. Twelve other simulated DHLW packages were emplaced tinder accelerated-aging or overtest conditions, including the artificial introduction of brine, and a thermal loading approximately three to four times higher than reference. Eight of these 12 test packages contain 1500 W/canister electrical heaters; the other four are filled with DHLW glass.

Evolution of Chemistry and Its Effects on the Corrosion of Engineered Barrier Materials

2006 ◽  
Vol 985 ◽  
Author(s):  
Darrell Dunn ◽  
Yi-Ming Pan ◽  
Xihua He ◽  
Lietai Yang ◽  
Roberto Pabalan
Keyword(s):  
Environmental Conditions ◽  
Environmental Influence ◽  
Near Field ◽  
Yucca Mountain ◽  
High Level Waste ◽  
Seepage Water ◽  
Waste Package ◽  
Barrier Materials ◽  
Engineered Barrier ◽  
High Level

ABSTRACTThe evolution of environmental conditions within the emplacement drifts of a potential high-level waste repository at Yucca Mountain, Nevada, may be influenced by several factors, including the temperature and relative humidity within the emplacement drifts and the composition of seepage water. The performance of the waste package and the drip shield may be affected by the evolution of the environmental conditions within the emplacement drifts. In this study, tests evaluated the evolution of environmental conditions on the waste package surfaces and in the surrounding host rock. The tests were designed to (i) simulate the conditions expected within the emplacement drifts; (ii) measure the changes in near-field chemistry; and (iii) determine environmental influence on the performance of the engineered barrier materials. Results of tests conducted in this study indicate the composition of salt deposits was consistent with the initial dilute water chemistry. Salts and possibly concentrated calcium chloride brines may be more aggressive than either neutral or alkaline brines.

Zircaloy Corrosion in a Repository Environment

1999 ◽  
Vol 556 ◽  
Author(s):  
Jerry D. Christian
Keyword(s):  
Spent Nuclear Fuel ◽  
Near Field ◽  
Yucca Mountain ◽  
Localized Corrosion ◽  
General Corrosion ◽  
Geochemical Environment ◽  
Waste Package ◽  
Saturated Water ◽  
Radionuclide Release ◽  
Corrosion Data

AbstractAssessments are made of the corrosion characteristics of spent nuclear fuel Zircaloy cladding in a Yucca mountain repository environment and the potential for the cladding to provide protection against radionuclide release following waste package failure. Considerations and assumptions includes a waste package life near 10,000 years and air-saturated water contacted with waste package corrosion product goethite, based on the near-field geochemical environment evaluated in the Yucca Mountain Viability Assessment [3]. Literature corrosion data (general, pitting, and localized crevice attack) are evaluated on the basis of these conditions and the expected chemical environments that can result on the surface of the fuel. General corrosion of Zircaloy is expected to be negligible and result in a lifetime of the SNF cladding of several hundred thousand years, approaching a million years. General surface pitting is not expected. Effects of crevice localized corrosion for periods beyond 10,000 years are uncertain and require modeling development and experimental characterization. Details of the evaluations that provide the basis for the conclusions are presented.

Effect of Overpack Materials on Glass Leaching in Geological Burial

1981 ◽  
Vol 11 ◽  
Author(s):  
Lars Werme ◽  
L. L. Hench ◽  
Alexander Lodding
Keyword(s):  
Ambient Temperature ◽  
Storage System ◽  
Fission Products ◽  
Waste Form ◽  
Waste Storage ◽  
Backfill Materials ◽  
Glass Leaching ◽  
Research Institute

This is one of two papers discussing the findings in an in situ-burial experiment, presently being performed in the Stripa mine in Sweden. The purpose of the experiment is to evaluate the effects of various components in the SKBF/KBS waste storage system on the leaching of the vitreous waste form. Two configurations of glass, canister, overpack and buffer/backfill materials were designed (Figs. 1 and 2). Both configurations were inserted into 56 mm diameter boreholes in the Stripa mine and maintained at 90°C. One of the configurations (Fig. 2) was also kept at ambient temperature, 8°C. In the experiments two glass types were used, ABS 39 and ABS 41 (Table 1). These glasses, developed by Dr. T. Lakatos of the Swedish Glass Research Institute, contain 9% simulated fission products by weight and are compatible with the French AVM process.

The Strategy of the Belgian Nuclear Research Centre in the Area of High-Level Waste Form Compatibility Research

2007 ◽  
Author(s):  
Karel Lemmens ◽  
Christelle Cachoir ◽  
Elie Valcke ◽  
Karine Ferrand ◽  
Marc Aertsens ◽  
...  
Keyword(s):  
Nuclear Research ◽  
Waste Form ◽  
Research Centre ◽  
High Level Waste ◽  
Spent Fuel ◽  
National Agency ◽  
Waste Forms ◽  
In Situ Experiments ◽  
High Level

The Belgian Nuclear Research Centre (SCK•CEN) has a long-standing expertise in research concerning the compatibility of waste forms with the final disposal environment. For high level waste, most attention goes to two waste forms that are relevant for Belgium, namely (1) vitrified waste from the reprocessing of spent fuel, and (2) spent fuel as such, referring to the direct disposal scenario. The expertise lies especially in the study of the chemical interactions between the waste forms and the disposal environment. This is done by laboratory experiments, supported by modeling. The experiments vary from traditional leach tests, to more specific tests for the determination of particular parameters, and highly realistic experiments. This results in a description of the phenomena that are expected upon disposal of the waste forms, and in quantitative data that allow a conservative long-term prediction of the in situ life time of the waste form. The predictions are validated by in situ experiments in the underground research laboratory HADES. The final objective of these studies, is to estimate the contribution of the waste form to the overall safety of the disposal system, as part of the Safety and Feasibility Case, planned by the national agency ONDRAF/NIRAS. The recent change of the Belgian disposal concept from an engineered barrier system based on the use of bentonite clay to a system based on a concrete buffer has caused a reorientation of the research programme. The expertise in the area of clay-waste interaction will however be maintained, to develop experimental methodologies in collaboration with other countries, and as a potential support to the decision making in those countries where a clay based near field is still the reference. The paper explains the current R&D approach, and highlights some recent experimental set-ups available at SCK•CEN for this purpose, with some illustrating results.

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