Effects of temperature, temperature gradients, stress, and irradiation on migration of brine inclusions in a salt repository

Available experimental and theoretical information on brine migration in bedded salt are reviewed and analyzed. The effects of temperature, thermal gradients, stress, irradiation, and pressure in a salt repository are among the factors considered. The theoretical and experimental (with KCl) results of Anthony and Cline were used to correlate and explain the available data for rates of brine migration at temperatures up to 250/sup 0/C in naturally occurring crystals of bedded salt from Lyons and Hutchinson, Kansas. Considerations of the effects of stressing crystals of bedded salt on the migratin properties of brine inclusions within the crystals led to the conclusion that the most probable effects are a small fractional increase in the solubility of the salt within the liquid and a concomitant and equal fractional increase in the rate of the thermal gradient-induced migration of the brine. The greatest uncertainty relative to the prediction of rates of migration of brine into a waste emplacement cavity in bedded salt is associated with questions concerning the effects of the grain boundaries (within the aggregates of single crystals which comprise a bedded salt deposit) on brine migration through the deposit. The results of some of the estimates of rates and total amounts of brine inflow to HLW and SURF waste packages emplaced in bedded salt were included to illustrate the inflow volumes which might occur in a repository. The results of the brine inflow estimates for 10-year-old HLW emplaced at 150 kW/acre indicated inflow rates starting at 0.7 liter/year and totaling 12 liters at 30 years after emplacement. The results of the estimates for 10-year-old PWR SURF emplaced at 60 kW/acre indicated a constant inflow of 0.035 liter/year for the first 35 years after emplacement

In-Pile Corrosion Test Loops for Aqueous Homogeneous Reactor Solutions

An in-pile corrosion test loop is described which is used to study the effect of reactor radiation on the corrosion of materials of construction and the chemical stability of fuel solutions of interest to the Aqueous Homogeneous Reactor Program at ORNL. Aqueous solutions of uranyl sulfate are circulated in the loop by means of a 5-gpm canned-rotor pump, and the pump loop is designed for operation at temperatures to 300 ts C and pressures to 2000 psia while exposed to reactor radiation in beam-hole facilities of the LITR and ORR. Operation of the first loop in-pile was begun in October 1954, and since that time 17 other in-pile loop experiments were completed. Design criteria of the pump loop and its associated auxiliary equipment and instrumentation are described. In-pile operating procedures, safety features, and operating experience are presented. A cost summary of the design, fabrication, and installation of the loop and experimental facillties is also included. (auth

CORROSION OF PLATINUM BY UOsub2sub 2SOsub4sub 4 SOLUTIONS UNDER IRRADIATION

Results indicate that corrosion of the metal is accelerated to a small extent during irradiation. A summary of the results from in-pile loops and in- pile autoclaves is given. (L.M.T.

Effects of gaseous radioactive nuclides on the design and operation of repositories for spent LWR fuel in rock salt

Information relating to the identities and amounts of gaseous radionuclides present in spent LWR fuel and to their release from canistered spent fuel under plausible storage and disposal conditions was assembled, reviewed, and analyzed. Information was also reviewed and analyzed on several other subjects that relate to the integrity of the carbon steel canister in which the spent fuel is to be encapsulated and to the expected rates of transfer of gaseous radionuclides through crushed salt backfill within a disposal room in a reference repository in rock salt. The advantages and disadvantages were considered for several different canister-backfill materials, and recommendations were made regarding preferred materials. Other recommendations relate to encapsulation procedures and specifications and to needs for additional experimental studies. The objective of this work was to provide reference information, conclusions, and recommendations that could be used to establish design and operating conditions and procedures for a bedded salt repository for spent LWR fuel and that could also be used to help evaluate the safety of the repository. The results of this work will also generally apply to spent fuel repositories in domal salt. However, because the domal salt may have little or no brine inclusions within it, there may be little or no possibility that brine will migrate into open spaces around an emplaced canister. Addordingly, some of the concerns that result from the possible occurrence of brine migration in bedded salt may be of no importance in domal salt

IN-PILE RADIATION CORROSION EXPERIMENTS WITH ZIRCONIUM, TITANIUM, AND STEEL ALLOYS IN 0.17 m UOsub2sub 2SOsub4sub 4 SOLUTIONS AT 280 C

In-pile loop experiments L-2-15 and L-4-16 were designed to test the radiation corrosion of Zircaloy-2 and other possible reactor construction materials in UO/sub 2/SO/sub 4/ solutions. The solutions employed were 0.17 m UO/ sub 2/SO/sub 4/, 0.015 m CuSO/sub 4/, and 0.03 m H/sub 2/SO/sub 4/ in H/sub 2/O for experiment L-2-15, and 0.17 m UO/sub 2/SO/sub 4/, 0.015 m CuSO/sub 4/, and 0.025 m H/sub 2/SO/sub 4/ in H/sub 2/O for experiment L-4-16. The mainstream temperature in the experiments ranged from 278 to 280 deg C. Construction material for the loops was type 347 stainless steel. Specimens of types 347 and 309SCb stainless steels titanium-55A and -110AT, platinum, Zircaloy-2, crystalbar zirconium, and a variety of other zirconium alloys were tested. The power density at core specimens ranged from 19.8 to 4.6 w/ml in L-2-15 and from 5.7 to 1.3 w/ml in L-4-16. For loop L-2-15, the total time of hightemperature operation with UO/sub 2/SO/sub 4/ was 792 hr, during in-pile exposure, and the reactor energy was 1632 Mwh; for loop L-4-16, 1032 hr and 2325 Mwh. During both experiments most of the reactor energy was accumulated at 3-Mw power level. In general, stainless steel corrosion results from these experiments were comparable to those observed in other in-pile loop experiments. Corrosion was confined primarily to the core areas and was power-density dependent. Some variations in attack, both positive and negative, with velocity of solution flow past specimens have been observed in other experiments, but there was no apparent effect of varying velocities in the range 10 to 40 fps on either the core-channel or in- line channel specimens in the present experiments. The coreannulus steel specimens in L-2-15 corroded at rates very much greater than those in the channel. This difference may have resulted, in part, from the differences in velocities, however, it may have also been a result of galvanic actton between the steel annulus specimens and adjacent platinum specimens. In previous 250 deg C experiments the occurrence of a change in the stainless steel corrosion rate was correlated with a decrease in acidity and/or increase in the nickel concentration. The results for the oxygen consumption rates on steel during radiation exposure in the present experiments varied with radiation time in a manner qualitatively similar to that observed at the lower temperature. However, the concentration of excess acid in the present experiments probably remained fairly constant throughout the radiation exposures, and correlations similar to those obtained at the lower temperature could not be established. The acid concentration in the 280 deg C experiments was greater than the concentrations prevailing when corrosion rate changes occurred in the 250 deg C experiments. The difference in acid tolerance is probably a result of the increased temperature, since a similar beneficial effect of temperature occurs out-ofpile No overall correlation has been established for the various factors found to have influenced steel corrosion in previous experiments. Results of the present experiments provide additional evidence in support of previous findings but do not further their interpretation. Zircaloy-2 corrosion results from both loops have been discussed and correlated elsewhere in terms of the 280 deg C relationship between the corrosion rate R (mils per year, mpy), power density P (w/ml), and uranium sorption factor alpha : 1/R = 2.23/P alpha + 1/40. The data from these experiments obey this relationship. (This is only a portion of the Author abstract.

Brine migration in salt and its implications in the geologic disposal of nuclear waste

This report respresents a comprehensive review and analysis of available information relating to brine migration in salt surrounding radioactive waste in a salt repository. The topics covered relate to (1) the characteristics of salt formations and waste packages pertinent to considerations of rates, amounts, and effects of brine migration, (2) experimental and theoretical information on brine migration, and (3) means of designing to minimize any adverse effects of brine migration. Flooding, brine pockets, and other topics were not considered, since these features will presumably be eliminated by appropriate site selection and repository design. 115 references

Storage and release of radiation energy in salt in radioactive waste repositories

The results showed that appreciable amounts of gamma-radiation energy can be stored under certain exposure conditions. The results also showed that thermally activated annealing takes place at elevated temperatures and that the rates of this annealing at temperatures above about 150/sup 0/C are such that negligible amounts of energy will be stored in salt in a repository where the salt is at temperatures above about 150/sup 0/C. We are not able to show that thermally activated annealing takes place in rock salt at temperatures below about 150/sup 0/C, although it may do so. There may also be some radiation-induced annealing. The results of measurements of energy stored at irradiation temperatures between 30 and 150/sup 0/C, together with results of theoretical considerations, showed that the maximum stored energy that would be formed in salt in a repository with no annealing whatsoever would be about 50 cal/g. We did not conceive a means by which the stored energy could be released abruptly, nor was any significant hazard believed to be possible from the release if it should occur abruptly by some unforeseen mechanism. In practice, salt temperatures below about 150/sup 0/C will occur only with isolated or semi-isolated canisters in a repository, or with most canisters, after very long times during which the gamma-dose rate will have dropped off markedly. Available evidence from thermal annealing observations and from calorimetric and dissolution measurements indicated that there were no significant losses of either chlorine or sodium during or after irradiation. One effect of the stored energy which must be accounted for in a safety analysis is the generation of H/sub 2/, which takes place upon aqueous dissolution of radiation-damaged salt; about 0.1 cm/sup 3/ of H/sub 2/ is generated per calorie of stored energy. However, we have not recognized any problems arising from this effect which cannot be counteracted by appropriate design and operation of the repository

Review of information on the radiation chemistry of materials around waste canisters in salt and assessment of the need for additional experimental information

The brines, vapors, and salts precipitated from the brines will be exposed to gamma rays and to elevated temperatures in the regions close to a waste package in the salt. Accordingly, they will be subject to changes in composition brought about by reactions induced by the radiations and heat. This report reviews the status of information on the radiation chemistry of brines, gases, and solids which might be present around a waste package in salt and to assess the need for additional laboratory investigations on the radiation chemistry of these materials. The basic aspects of the radiation chemistry of water and aqueous solutions, including concentrated salt solutions, were reviewed briefly and found to be substantially unchanged from those presented in Jenks's 1972 review of radiolysis and hydrolysis in salt-mine brines. Some additional information pertaining to the radiolytic yields and reactions in brine solutions has become available since the previous review, and this information will be useful in the eventual, complete elucidation of the radiation chemistry of the salt-mine brines. 53 references

EXAMINATIONS OF SPECIMENS AND SCALES TAKEN FROM THE HRT FOLLOWING RUNS 13 AND 14

Following HRT runs 13 and 14, several metallic specimens were removed from the high pressure system and transferred to the Materials Section for examination. Samples of scale accumulation in the high pressure system were also taken after these runs and transferred to the Materials Section. Examination and analyses of these several specimens are still in progress, but some of the results are available and are reported. A possible interpretation of some of these results indicates that a considerable quantity of nickel was contained in the core scale accumulation at the end of run 13, and that part of this nicke1 was dissolved in solution during run 14. The amount of nickel which may have come from this source during run 14 roughly accounts for all of the increase in nickel in solution durirg run 14. A significant amount of uranium was also found in the sca1e accunnulation in the core after run 13. (auth