The United States is at an important crossroads in its management of high-level radioactive waste (HLW). It is presently in the process of changing over from a strategy that was developed during World War II, which involves storing HLW in a relatively mobile liquid form, to a strategy of the 1990s, which involves removing and isolating potentially harmful radionuclides and immobilizing them into solid and inert forms, more specifically, borosilicate glass. The resulting waste glass products will then be permanently disposed of by deep burial, within stable geologic formations, where they become one element of a multibarrier waste-isolation system. This barrier system is designed to retain radionuclides so they can be permanently isolated from the public and from the accessible environment. Important contributing factors to the success of this strategy include the excellent stability and technical performance of waste glass forms and the ability of the glass, as well as waste package materials, to retain radionuclides even when exposed to potential leachants within a repository environment.

The generation of volatile organic compounds (VOCs), hydrogen, and carbon oxides due to alpha particle irradiation of polyethylene, polyvinylchloride, hypalon, and neoprene, is being investigated. A wide diversity of VOCs was found including alkenes, alkanes, alcohols, ketones, benzene derivatives, and nitro compounds. Their yields however, were quite low. The relative amounts of these compounds depended on the material, atmosphere present, and the absorbed dose. This investigation will help evaluate the effect of ionizing radiation on the long-term performance assessment and regulatory compliance issues related to the Waste Isolation Pilot Plant (WIPP).

We conducted in situ experiments with simulated, i.e., nonradioactive, contact-handled trans-uranic (CH TRU) waste drums at the Waste Isolation Pilot Plant (WIPP) facility for a period of about four years. We performed these tests in two rooms in rock salt, at WIPP, with drums surrounded by crushed salt or 70 wt. % salt/30 wt. % bentonite clay backfills, or partially submerged in a NaCl brine pool. Air and brine temperatures were maintained at ~40°C. These full-scale (210-L drum) experiments provided in situ data on: backfill material moisture-sorption and physical properties in the presence of brine; waste container corrosion adequacy; and, migration of chemical tracers (nonradioactive actinide and fission product simulants) in the near-field vicinity, all as a function of time. Individual drums, backfill, and brine samples were removed periodically for laboratory evaluations. Waste container testing in the presence of brine and brine-moistened backfill materials served as a severe overtest of long-term conditions that could be anticipated in an actual salt waste repository. We also obtained relevant operational-test emplacement and retrieval experience. All test results are intended to support both the acceptance of actual TRU wastes at the WIPP and performance assessment data needs. We provide an overview and technical data summary focusing on the WIPP CH TRU environmental overtests involving ] 74 waste drums in the presence of backfill materials and the brine pool, with posttest laboratory materials analyses of backfill sorbed-moisture content, CH TRU drum corrosion, tracer migration, and associated test observations.

Multi-year, simulated remote-handled transuranic waste (RH TRU, nonradioactive) experiments are being conducted underground in the Waste Isolation Pilot Plant (WIPP) facility. These experiments involve the near-reference (thermal and geometrical) testing of eight full-size RH TRU test containers emplaced into horizontal, unlined rock salt boreholes. Half of the test emplacements are partially filled with bentonite/silica-sand backfill material. All test containers were electrically heated at about 115 W/each for three years, then raised to about 300 W/each for the remaining time. Each test borehole was instrumented with a selection of remote-reading thermocouples, pressure gages, borehole vertical-closure gages, and vertical and horizontal borehole-diameter closure gages. Each test emplacements was also periodically opened for visual inspections of brine intrusions and any interactions with waste package materials, materials sampling, manual closure measurements, and observations of borehole changes. Effects of heat on borehole closure rates and near-field materials (metals, backfill, rock salt, and intruding brine) interactions were closely monitored as a function of time. This paper summarizes results for the first five years of in situ test operation with supporting instrumentation and laboratory data and interpretations. Some details of RH TRU waste package materials, designs, and assorted underground test observations are also discussed. Based on the results, the tested RH TRU waste packages, materials, and emplacement geometry in unlined salt boreholes appear to be quite adequate for initial WIPP repository-phase operations.

In 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.

Several 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.

A 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.

A mixture of bentonite clay and quartz sand is being considered for use as a waste-package backfill, the material placed between a radioactive-waste canister and the repository host rock. Compacts of bentonite/quartz with weight-percent ratios of 100/0, 70/30, 50/50 and 30/70 were made at room temperature under a pressure of 100 MPa (15 ksi). Upon initial heating, the thermal conductivity of the 70/30 compact rose from 1.20 W/m·K at 298 K to 1.32 W/m·K at 373 K. After further heating to 473 K, it fell to 1.10 W/m·K, reflecting the loss of interlamellar water from the bentonite. The conductivity of the now-dehydrated compact was reproducible through several heating and cooling cycles, varying from 1.15 W/m·K at 573 K to 1.03 W/m·K at 298 K. The other mixtures were qualitatively similar to the 70/30: the 100/0, 50/50 and 30/70 dehydrated compacts displayed conductivities of 0.59, 1.06 and 0.83 W/m.K,respectively, at 298 K. Measured densities ranged from 1.98 to 2.12 g/cc. Combined geometric-mean and Maxwell models for thermal conduction in composite systems predict the measured results reasonably well. An analysis of the impact of backfills on high-level waste (HLW) package design indicates that no significant thermal penalty is imposed.

The first phase of a complex high-level waste (HLW) package interactions test in a salt environment has been completed. The test system consisted of PNL 76–68 HLW glass (loaded with inactive fission products and 238U) surrounded by a stainless steel waste canister, a TiCode- 12 overpack, a bentonite/sand backfill, excess brine leachant, and a bedded rock salt container, all heldwithin a 19-liter autoclave. All components were physically compromised in order to force wasteform-barrier-salt interactions to occur during this 95-day, 250°C overtest. Analyses of leachant, wasteform, and all barrier surfaces were performed. Test data included the synergistic effects between barriers and confirmed previous analyses of simpler systems. The glass wasteform exhibited some surface alteration but was not dissolved to any significant degree. The TiCode-12 overpack showed minimal uniform corrosion and no localized attack. Observed mineralogical alteration of the backfill was minimal.