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NBS

PUBLICATIONS

NBSIR 81-2409

Bibliography of Literature on

Underground Corrosion of Metals

and Alloys Considered for Use in

the Construction of Containers

for Nuclear Waste

U.S. DEPARTMENT OF COMMERCE

National Bureau of Standards

Center for Materials Science

Materials Chemistry Division

Washington, DC 20234

October 1 981

Issued January 1982

Prepared for mI.S. Department of Energy

i/ashington, DC 20585

31-2409

1981

c. 2
!l
NBSIR 81-2409

BIBLIOGRAPHY OF LITERATURE ON

UNDERGROUND CORROSION OF METALS

AND ALLOYS CONSIDERED FOR USE IN

THE CONSTRUCTION OF CONTAINERS

FOR NUCLEAR WASTE

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FEB 3 1982

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B. T. Sanderson and J. Kruger

U.S. DEPARTMENT OF COMMERCE

National Bureau of Standards

Center for Materials Science

Materials Chemistry Division

Washington, DC 20234

October 1 981

Issued January 1982

DISTRIBUTION OF THIS DOCUMENT IS UNLIMITED

Prepared for U.S. Department of Energy

Washington, DC 20585

U.S. DEPARTMENT OF COMMERCE, Malcolm Baldrige, Secretary

NATIONAL BUREAU OF STANDARDS, Ernest Ambler, Director

Bibliography of Literature on Underground Corrosion

of Metals and Alloys Considered for Use

in the Construction of Containers for Nuclear Waste

by

B. T. Sanderson and J. Kruger

Chemical Stability and Corrosion Division

Center for Materials Science

National Measurement Laboratory

National Bureau of Standards

Washington, D. C. 20234

Introduction

The citations and annotations from the literature presented in this

bibliography consist of articles, papers, talks, books, reports, proceedings of conferences and patents that deal with the corrosion of metallic

container materials that are proposed for use for the containment of

nuclear waste in underground repositories. Both author and subject

indices are provided. The author index is listed by the first author.

The subject index uses key words found in the individual bibliography

items. This bibliography contains five kinds of metallic corrosion data

that is aimed at serving the needs of those who are concerned with the

nuclear waste container design:

1 . General underground corrosion data of metals that may be

applicable to the use of metallic nuclear waste containers . This data

describes the performance of metals and alloys that may be used or

considered for use in underground environments.

2. Data on corrosion considerations in package design . These

data consist of patents and designs of actual corrosion resistant metallic containers for use in underground environments and include provisions

to resist corrosion.
3.

Data on metallic corrosion in geothermal brines . Since some

of the proposed underground environments for buried nuclear waste include

salt deposits and deep ocean sediments, a cursory sampling of data

pertaining to the corrosion performance of metals and alloys in geothermal brines is included. An effort was made during compilation of this

information to concentrate on those metals that might have possible

application to the use of metallic nuclear waste containers.

4. Data on the internal corrosion of nuclear waste containers to

be used in underground environments . While the primary concern of this

bibliography is the compliation of data regarding the interaction of

external surfaces of nuclear waste containers with underground environments, the importance of the internal corrosion resistance of the container to its nuclear waste contents is quite obvious. Therefore, some

but not a complete set of data concerned with internal corrosion of

containers for underground use have been included. Also some of the

individual references cited relate to both the anticipated internal

corrosion problems as well as the external corrosion problems.,

5. Data on external corrosion of nuclear waste containers for

underground use . These data concern the reactions between the outer

surfaces of a radioactive nuclear waste container and its underground

environment such as salt, seabed sediments, rock or soils.

Abstracts are included when available and some lengthy abstracts

have been shortened or condensed. When no abstract was abailable or an

article was quite brief, key words have been included where possible to

better describe the subject material. This compilation was performed

utilizing the following computer data bases: National Technical Information Service data base. Engineering Index data base. Department of
3

Energy data base, and Chemical Abstracts data base. In addition to the

computer research, a hard copy search of references was accomplished at

the library of the National Bureau of Standards. Another invaluable

source of information has been the personal contacts with the individuals

at organizations such as the Battel le Pacific Northwest Laboratory,

Sandia National Laboratory, Oak Ridge National Laboratory, and the U. S.

Bureau of Mines and proved quite helpful. Finally, the 70 odd years of

work by NBS in the field of underground metallic corrosion has been

utilized extensively.

We gratefully acknowledge the support of the preparation of the

Bibliography by the Department of Energy National Waste Terminal Storage

Program.
AUTHOR INDEX

Abrego, L. , Braithwaite, J. W. , Corrosion of Titanium in

Saline Nuclear Waste Isolation Environments, 1980 Annual Mtg.,

American Nuclear Society, Las Vegas, NV, 9-12 June 1980,

p. 34.

The corrosion of the Ti alloys Ti-50A and TiCode-12 is being

studied to determine their useful lifetime as canister materials

for the isolation of nuclear waste. The long-range goal of

this program is to qualify a material to survive the first 300

to 600 years of emplacement in a chloride-containing environment.

Types of corrosion that could lead to canister failure include

uniform attack, stress corrosion cracking, and pitting/crevice

corrosion. It was shown in a prior study that the uniform

corrosion rates of the above two Ti alloys are extremely small

in 250 °C brines (# <0.01 mm/yr) and that pitting attack did

not occur. The subject of the present investigation is the

crevice corrosion and stress corrosion cracking susceptibility

of Ti-50A and TiCode-12 in both nonirradiated and gammairradiated saturated brine and sea water environments.

Ahlstrom, P. E., Ceramic and Pure-Metal Canisters in Buffer

Material for High-Level Radioactive Waste, Nucl . Chem. Waste

Manage. , 1(1), 77-88 (1980).

Within the Swedish KBS project, a great effort has been devoted

to the development of canisters for the encapsulation of highlevel vitrified waste or spent unreprocessed fuel. The canisters

will, upon final disposal in deep geological formations,

provide an extra engineered barrier against the dispersal of

radioactive nuclides in the groundwater. Comprehensive studies

have been made on two types of pure metal canisters; i.e., PbTi for the high-level vitrified waste and pure Cu for spent

fuel. The final disposal is made in hard crystalline rock,

where the waste canisters are embedded in a buffer material--a

clay with good long-term stability. For the Pb-Ti canister, a mixture of quartz sand and bentonite is proposed whereas for

the Cu canister highly compressed bentonite is preferred.

Careful evaluations predicted a lifetime of at least thousands

of years for the Pb-Ti canisters and probably tens of thousands

of years in the actual environment. For the Cu canister, it

is realistic to expect a lifetime of hundreds of thousands of

years. Comprehensive studies have also been made of a ceramic

canister of alumina for spent fuel. Through hot isostatic

pressing, it is possible to make a completely tight and jointfree alumina container. Alumina has a very high chemical and

mechanical resistance over very long periods of time. Preliminary studies have also been made of a glass ceramic material

of the beta-spodumene type.

Akol'zin, P. A., Tskhvirashvi 1 i , D. G., Vardigoreli, 0, Sh.,

Corrosion of Copper and Its Alloys in Geothermal Waters, Prot.

Metals. , 9(1 ), 56-58 (1973).

Tests were conducted on specimens from copper and brass tubes,

used in water-to-water heat exchangers. The rate of corrosion,

initially very considerable, decreased with tim$ and then

became stable. No change in the rate of corrosion was observed
5

in half-a-year due to the formation of protective films on tube surfaces. Pure copper samples showed corrosion rates

three to ten times greater than copper alloys. To maintain

adequate corrosion performance, a water flow rate of about

three m/sec is recommended for brass tubing.

[4] Allard, B. , Kipatsi, H. , Rydberg, J., Radiolysis of Buffer

Materials, Kaernbraenslesaekerhet , Rept. No. KBS-TR-56, p. 22

(1977).

The effect of gamma irradiation on the properties of the

possible buffer material bentonite (bentoni te:quartz, 10:90)

has been discussed, mainly in qualitative terms. Changes in

ion exchange properties and lattice structure of the clay

mixture due to irradiation in an aqueous solution have been

studied. The production of gas (hydrogen) due to radiolysis

has been determined as a function of the radiation dose. The

resistance to corrosion of titanium in a radiation field has

been demonstrated.

[5] Anderson, K. , Neretnieks, I., Diffusion of Nuclides of Low

Solubility From a Copper Capsule After Penetration Through a Hole in the Capsule, Kaernbraenslesaekerhet, Rept. No. KBS-TR80, p. 11, 1978.

Diffusion of nuclides of low solubility from a copper capsule

after its penetration has been computed. It is assumed that

by some mechanism the copper has corroded, leaving a hole

equal to the inner diameter of the capsule. The uranium oxide

is slowly dissolved, and the nuclides diffuse out through the

hole. The diffusivity of the uranium ions is taken to be

10"^°m2/S in the copper corrosion products, as well as in the

porous mass in the capsule UO 2 , zircaloy corrosion products,

lead corrosion products). The dissolution of the uranium

oxide matrix is governed by the solubility of UO 2 and the

distance it has to diffuse. The solubility is assumed to be

very high--1070g/m3—which is the value obtained under oxidizing conditions and in a water with very high carbonate

content (550 nig/1). Initially, the transport rate of uranium

is two g/year. It decreases rapidly as the uranium oxide is

dissolved and as the diffusion distance increases. It will

take 2.8 million years to dissolve and transport away a mass

of uranium oxide in the two meters nearest the assumed hole in

the capsule. If this hole has only ten percent of the inner

area of the copper capsule, the time increases to 7.1 million

years. When 0.2 m of the uranium oxide nearest the hole have

been dissolved, the dissolution rate is smaller than the

transport from the repository. The rate of diffusion from the

capsule may then become the rate determining step.

[6] Angerman, C. L. , Rankin, W. N, , Durability of Containers for

Storing Solidified Radioactive Wastes, Corrosion/77 NACE Mtg.,

San Francisco, CA, 14 March 1977.

Most concepts for the disposal of highly radioactive waste

involve converting the waste to a solid form, like concrete or
6

glass, and storing this solid form in metal containers. Two

major factors in the final selection of materials for these

containers are the compatibility between waste form and con- tainer material and the durability of the material at temperatures and stresses expected during service and possible accidents. Currently, AISI 1020 carbon steel appears to be a better material than other alloys, such as Cor-Ten A, Type 304L

stainless steel, or Inconel 600, considered. This choice is

based on the results of 10,000 hours of heating tests that

showed container compatibility with both concrete and glass

waste forms. The selection is also based on: (1) analyses of

the strengths and (2) oxidation resistances of the alloys

under the conditions expected during 100 year storage in air

and in various impact and thermal accidents. The thinner wall

thickness required for satisfactory performance of the stronger,

more-oxidation-resistant alloys is offset by their higher cost

per pound.

[7] Anticorrosion Methods Materials, Article, Buried Nuclear Waste

must be Enclosed in Au to Resist Corrosion, Anticorr. Methods

Matls. , ^(6), 12-13 (1977).

Keywords: Radioactive waste; waste disposal; gold plating;

Corrosion Prevention.

[8] Asphahani, A., Corrosion Resistance of High Performance Alloys,

Matls. Perform. , ]_9(12), 33-43 (1980).

A review with 88 references on highly alloyed stainless steels,

Ni alloys, and Co alloys for the processing industries and

pollution control equipment. The resistance to pitting/crevice

corrosion is examined, and a ranking of alloys is established.

Stress corrosion cracking is examined in the presence of NaCl

and/or H 2 S.

[9] Baboian, R. , Haynes, 6. S. , Corrosion Resistance of Copper

Clad Stainless Steel Cable Shielding Materials, NACE, Inti.

Corr. Forum, Chicago, IL, Preprint paper 89, 4-8 March 1974.

Copper clad stainless steel (Cu/430 SS/Cu) and solid copper

behave similarly in all soils. Results with panels and shieldings in cables were in good agreement. After extended exposure,

copper and copper alloy perforates (such as in acid soils),

whereas the stainless steel acts as a corrosion barrier in the

Cu/430 SS/Cu shielding, thus preventing corrosion of the inner

copper layer and maintaining shielding integrity. Data from

potentiodynamic polarization curves in soil water extracts are

useful in determining the mechanisms of corrosion in various

soil environments.

[10] Banning, L. H. , Oden, L. L. , Corrosion Resistance of Metals in

Hot Brines: A Literature Review, US Bur. Mines, Info. Circ.

No. 8601, (1973).
7

With the ever-increasing demand for energy in all forms,

electical energy generated from geothermal resources could

become an important energy source in the next few decades.

Geothermal fluids, which contain appreciable quantities of

dissolved salts and gases, are generally corrosive to materials of construction. Since data on corrosion of materials

in geothermal brines is limited, the literature was surveyed

through February 1972 to obtain a compilation of information

on corrosion resistance of aluminum and its alloys, copper or copper-base alloys, carbon steel and cast iron, stainless

steels, chromium-iron alloys, nickel- and cobalt-base alloys,

and titanium in heated brines, including sea water.

[11] Bednar, L. , Young, J. W. , New Corrosion Resistant Material for

URD Transformer Cases, Matls. Prot. Perform., 1_2(3), 21-25

(1973).

A composite material, fabricated from a thin layer of stainless

steel, metal lurgical ly bonded on each side to a heavy layer of

carbon steel, offers high resistance to pitting corrosions and

is less expensive than Type 409 stainless. The cathodic

corrosion mechanism is evaluated and illustrated sample of

corrosive attack of various metals compared with the composite.

These materials are used in underground environments.

[12] Bird, J. M. , Ringwood, A. E., Container for Radioactive Nuclear

Waste Materials, US 4,192,765 (Cl. 252-301. IW; G21F1/08)

11 March 1980, Appl . 878,113, 15 February 1978. Chemical

Abstracts, Vol. 93, 1980.

Long-term storage of nuclear reactor wastes in underground

rock formations requires container alloys thermodynamically

stable against degradation in geochemical environments.

Container alloys are proposed on the basis of naturally oc- curring awaruite and josephinite as stable mineral alloys,

which contain Ni 60-90, Fe 10-40, Cu 0-5, and Co 0-5 percent.

Their stoichiometric phase is Ni^Fe.

[13] Boase, D. G. , Vandergraaf, T. T. , The Canadian Spent Fuel

Storage Canister: Some Materials Aspects, Nucl . Tech., 32(1 )

,

60-71 (1977).

Concrete canisters for interim dry storage of spent, irradiated

Canadian D-U (CANDU) fuel are being developed by Atomic Energy

of Canada, Ltd. The canisters are designed to contain fuel

safety for periods of 50 to 100 years in C-steel baskets

sealed inside a steel- and Pb-lined concrete shield. A demonstration

program at the Whiteshell Nuclear Research Establishment is

utilizing four instrumented canisters to establish the canister

structural integrity when exposed to the thermal stresses

generated by the decay heat of the stored fuel. A review of

other potential materials problems identified three areas of

concern: corrosion of the fuel basket and canister lining,

fuel sheath oxidation, and UO 2 oxidation. Preliminary analysis

suggests that the first of these will be minimized by the

migration of moisture to the outside of the canister under the

influence of the temperature gradient, and the second is
8

predicted to be insignificant for periods < 100 years. The

third area was less well understood, and a'detailed experimental

study was, therefore, undertaken. Initial canister designs

conceived the use of airfilled fuel baskets, with UO 2 fuel

temperatures initially in the 200-300 ° range. Oxidation of

the UO2 in defected fuel could cause contamination of the

basket and complicate subsequent fuel retrieval. The rates

and mechanisms of UO 2 oxidation were studied using powders,

sintered pellets, and intentionally defected fuel elements.

In the present demonstration canisters, the possibility of

oxidation of the fuel has been eliminated by storing it in Hefilled baskets.

[14] Braithwaite, J. W. , Molecke, M. A., High-Level Waste Canister

Corrosion Studies Pertinent to Geologic Isolation, Conf. on High-Level Radioactive Solid Waste, Denver, CO, 19 December 1978,

p. 35.

The compatibility of candidate high-level waste (HLW) canister

materials with deep geologic isolation environments is addressed.

Results are presented which are applicable to the following

repositories or test facilities: bedded and domed salt, subseabed sediment, and various types of hardrock. Such studies are an essential portion of the technological basis for terminal

waste management. These studies identify HLW canister or overpack materials satisfying appropriate requirements for

barrier lifetime. Mechanical properties, as well as constraints

on cost and consumption of critically limited materials, are also selection criteria. Lifetime objectives range from a minimum of several years for retrievabi 1 i ty constraints up to

several hundred years for retardation of near-field interactions (e.g., waste form leaching with potential radionuclide

release to the geosphere) during the period of greatest HLW

thermal output. A review of present and prior applicable

corrosion results is presented. However, emphasis is on the

results obtained from current laboratory and in-situ HLW

cani ster/corrosion programs at Sandia Laboratories. The

effects of multiple variables on corrosion susceptibility and

rates are briefly discussed, and some applicable data are given. It is possible to provide a canister/overpack barrier

which can survive geologic isolation environments for periods

of several hundred years.

[15] Braithwaite, J. W. , Magnani , N. J., Munford, J. W., Titanium

Alloy Corrosion in Nuclear Waste Environments, Corrosion/80

NACE Mtg., Chicago, IL, 6 March 1980.

A corrosion study has been conducted on the titanium alloys

Ti-50A, TiCode-12, and Ti-Pd to evaluate their suitability as

a long lifetime (300 to 600 year) canister material for the

isolation of nuclear wastes. TiCode-12, selected as the

primary candidate material, is shown to be very resistant to

environmental attack. Results of electrochemical, general

corrosion, and stress corrosion cracking experiments are presented. These data, including those from severe overtests,

have shown that TiCode-12 is a viable candidate material for
9

long-term waste isolation.

[16] Braithwaite, J. W. , Magnani , N. J., Corrosion Considerations

for Nuclear Waste Isolation Canisters, Annual Mtg. Matls.

Rsch. Soc., Boston, MA, 28 November 1978.

The environments in bedded salt, sub-seabed sediments, and

hardrock are described. A corrosion study is being conducted

on various candidate overpack alloys in deoxygenated sea water

and oxygenated sea water and brine; some rate data are given

for 14 to 50 days. Results show that deoxygenated saturated

brine is more corrosive than deoxygenated sea water plus

sediments, that oxygen increases the corrosion rate, especially

in sea water, and that localized attack in deaerated media is

minimal, even for 304L stainless steel.

[17] Braithwaite, J. W. , Corrosion Behavior of Metals in Terminal

Nuclear Waste Storage Environments, NACE Mtg. on Corrosion,

Houston, TX, 6 March 1978.

A laboratory and field experimentation program has been undertaken to determine the magnitude of and the remedy for corrosion

problems which might be encountered during terminal nuclear

waste storage. The investigation has centered upon canister

and instrumentation material selection for the following three

geologic environments being considered at Sandia Laboratories

for terminal storage: (1) bedded salt, (2) deep sub-sea

sediments, and (3) hardrock. A wide range of metal alloys are being considered as candidate materials because of the fluidity

of the design requirements. Ambient and 90 °C corrosion data

have been collected in sea water and saturated brine. However,

the depth of burial, along with the residual heat production

of each loaded canister, produces a high- temperature, highpressure, corrosive environment, which can only be simulated

or produced in the field (in-situ) or in laboratory autoclaves.

Possible corrosion mechanism and rate determinations are being

made using an autoclave equipped with internal electrochemical

instrumentation capabilities. Preliminary investigations are considering the effects of the following variables on corrosion

rate: (1) temperature, (2) pressure, (3) dissolved oxygen

concentration, and (4) brine chemistry. Field experiments,

using electrically heated canisters, are to begin shortly in

two shale formations.

[18] Branch, H. C. , Corrosion Resistant Materials for URD Equipment,

Matls. Prot. Perform., 12(3), 9-13 (1973).

More economical solutions to underground residential equipment

corrosion problems are the use of highly corrosion-resistant

alloys to replace cathodic protection and organic coating

techniques. Materials evaluated include silicon bronze,

copper-nickel. Types 304 and 316 stainless and highly alloyed

metals, such as alloys 20 and 825.

[19] Carter, J. P. , McCawley, F. X., Cramer, S. D. , Needham, P. B.,

Jr., Corrosion Studies in Brines of the Salton Sea Geothermal
10

Field, Report RI8350, Bureau of Mines, 1979.

The Bureau of Mines, US Department of the Interior, has conducted

in-si tu corrosion studies at the Sal ton Sea Known Geothermal

Resources Area (KGRA) in the Imperial Valley, CA, to evaluate

and characterize materials of construction for geothermal resources recovery plants. General-, pitting-, and crevicecorrosion characteristics of 13 commercially available alloys

were investigated for periods of 15 and 30 days in seven process environments expected to be found in typical geothermal resouces plants. Stainless steel alloy 29-4, Inconel 625, and

the Hastelloys G, S, and C-276 were the most resistant to

general corrosion, did not pit, and exhibited little susceptibility

to crevice corrosion. Stainless steel alloys 430, E-Brite 26-

1, and 6X had low general corrosion rates, but pitted and were susceptible to crevice corrosion. Carbon and 4130 steels had

high corrosion rates, pitted, and had high susceptibilities to

crevice corrosion. The major scale-forming mineral on the

corrosion samples in most of the process environments studied

was galena mixed with lesser amounts of other minerals.

[20] Carter, J. P., Cramer, S. D. , Field Stress Corrosion Tests in

Brine Environments of the Sal ton Sea Known Geothermal Resource

Area, Matls. Perform., 19(9), 13-16 (1980).

Stress corrosion tests (30-day) were conducted at the Sal ton

Sea Known Geothermal Resource Area on seven Fe-base and Nibase alloys in four brine and steam process streams by using

wellhead brine from geothermal well Magmammax 1. Transgranular

cracking of AISI 316L stainless steel and intergranular and

transgranular cracking of AISI 430 stainless steel was observed

in all four process streams. E-Brite 26-1 exhibited intergranular

and transgranular cracking in three of the four process streams.

AISI 1020 carbon steel. Inconel 625, Hastelloy G, and Hastelloy C276 showed no evidence of stress corrosion cracking.

[21] Carter, J. P. , Cramer, S. D. , Conrad, R. K. , Corrosion of

Stainless Steels in the Geothermal Environments of the Salton

Sea Known Geothermal Resource Area, Natl. Assn. Corr. Eng.

Inter. Forun Toronto, Canada, Apr. 1981.

The U. S. Department of the Interior, Bureau of Mines, in

support of its geothermal mineral and metal recovery program,

has conducted a series of field corrosion and electrochemical

tests in the geothermal environments of the Imperial Valley of

California. This paper presents the results of a 45day field

corrosion test conducted in 1978 on six stainless steel alloys

(>10 Cr) exposed to brine and steam environments of well

Magnamax #1 on the Salton Sea field and electrochemical data

on fourteen stainless steel alloys (>10 Cr) exposed to wellhead

brine. Stainless steels have excellent to good resistance to

general corrosion, but are highly susceptible to pitting and

crevice corrosion. Type 316 L stainless steel was found to

undergo intergranular corrosion and transgranular stress

corrosion cracking. Increasing the molybdenum content reduced

general corrosion, while higher contents of nickel and molybdenum
n

reduced pitting.

[22] Carter, J. P. , Cramer, S. D. , Conrad, R. K., An Overview of

Materials and Corrosion in Geothermal Systems, NACE Energy

Sources, Tech. Conf., New Orleans, LA, 1980. Bureau of

Mines has been involved in the development of the geothermal

mineral resources of the Imperial Valley, CA. The brines

produced from geothermal reservoirs in this area are extremely

corrosive. Furthermore, the brines readily form complex

scales during processing that complicate the corrosion process

and interfere with process operations. The Bureau has supported

a research program with the objective of identifying optimum

materials of construction for geothermal resource recovery

plants. This program consists of both laboratory and field

studies which include evaluating the corrosion resistance of a wide variety of engineering alloys in typical geothermal

process environments, applying electrochemical corrosion

measurement techniques in scale-forming geothermal brines,

using isokinetic sampling in studying two-phase geothermal

fluids, characterizing the kinetics and mineralogy of scale

deposition from hypersaline brines, and determining the solubility

of oxygen, methane, and carbon dioxide in brines.

[23] Casteels, F. , Tas, H. , Naets, J., Brabers, M., Corrosion of

Materials in a Clay Environment, Sci. Basis Nucl. Waste Manage.,

2, 38593 (1980).

Corrosion of canister materials for radioactive waste storage

was determined in a clay deposit which may be used for waste

disposal. Materials tested included: A1 and its alloys, C

steel, Cr steel, austenitic and ferritic stainless steels, Ni

alloys, Ni superalloys, Ti, and Ti alloys. At high temperatures,

S and S compounds are the main corrosive agents in the clay.

SO 2 is formed mainly due to the reaction of pyrite in the clay

with humidity. Excellent corrosion resistance was found for

Hastelloy B.

[24] Christenssen, H. , Consideration of Radiolysis of Ground Water,

Kaernbraenslesaekerhet, p. 26 (1978).

Computer calculations of the radiolysis of ground water have

been carried out with regard to the final disposal of highly

active waste. Radiolysis of water outside the container of

waste encapsulated in glass results in an equilibrium situation

after less than 50 h. The equilibrium concentrations of

oxygen, hydrogen, and hydrogen peroxide are low on the order

of 20, 10, and 20 ppB, respectively. Radical concentrations

are much lower, for example, C(0H) - 2 x lO'^^m. For fuel

elements directly deposited in a copper container, the radiolysis

of vyeter inside the container has been studied by calculations.

The purpose has been to determine whether hydrogen could be

formed in concentrations exceeding the solubility limit and

also to calculate the amount of copper corrosion. Assuming
12

that oxygen reacts diffusion controlled with the copper surface,

it was found that hydrogen was formed in concentrations far

below the solubility limit, even after several million years.

Zirconium may possibly react with water below 100 °C, resulting

in the formation of hydrogen at a rate of 0.0001 mole/year.

The corrosion of copper is less than 50 mg per liter of water

present.

[25] Claiborne, H. C. , Rickertsen, L. D., Graham, R. F., Expected

Environment in High-Level Nuclear Waste and Spent Fuel Repositories

in Salt, ORNL/TM-7201 , p. 206, Oak Ridge National Laboratories,

TN (1980).

The expected environments associated with high-level waste (HLW)

and spent fuel (SF) repositories in salt formations are described.

These environments include the thermal, fluid, pressure, brine

chemistry, and radiation field predicted for the repository

conceptual designs. Canistered waste packages containing HLW

in a solid matrix or SF elements are emplaced in vertical

holes in the floor of the rooms. The emplacement holes are backfilled with crushed salt or other material and sealed at some later time. Sensitivity studies are presented to show

the effect of changing the areal heat load, the canister heat

load, the barrier material and thickness, ventilation of the

storage room, and adding a second row to the emplacement

configuration. The calculated thermal environment is used as input for brine migration calculations. The vapor and gas

pressure will gradually attain the lithostatic pressure in a sealed repository. A computer code (REPRESS) was developed

assuming that these changes occur slowly (equil. conditions).

The brine chemical environment is outlined in terms of brine

chemistry, corrosion, and compns. The nuclear radiation

environment is the stored energy that can be released as a result of radiation damage or crystal dislocations within

crystal lattices.

[26] Colombo, P. , Neilson, R. M. , Jr., Properties of Radioactive

Wastes and Waste Containers. Progress Report No. 6, JulySeptember 1977, p. 47 (1978). Brookhaven National Laboratory,

Upton, NY.

A survey was made of the available literature on the corrosion

of steels in soil environments. The effect of the physical

and chemical properties of the soil on the rate and type of

corrosion of steels has been evaluated. The corrosion rates

of ferrous alloys are influenced primarily by the soil environment;

alloy compositional and microstructural considerations are secondary. However, as the steel is made more resistant to

uniform corrosion, it becomes more susceptible to pitting

corrosion. The various tests and criteria proposed to measure

the corrosiveness of soil are reviewed.

[27] Colombo, P. , Neilson, R. M. , Jr., Kendig, M. W., Analysis and

Evaluation of a Radioactive Waste Package Retrieved from the

Atlantic 2800 Meter Disposal Site, Report 1979, Brookhaven

National Laboratories , NY.
13

During the 1976 Atlantic 2800 m radioactive waste disposal

site survey, the first recovery of a radioactive waste package

from a radioactive waste disposal site was performed by the

EPA Office of Radiation Programs. The retrieved 80 gal.

radioactive waste package was transported to Brookhaven National

Lab., where container corrosion and matrix leach rate and

degradation studies were conducted.

[28] Cramer, S. D. , Carter, J. P. , Laboratory Corrosion Studies in

Low- and High-Salinity Geobrines of the Imperial Valley, CA.,

BUMINES-RI-8415, Bureau of Mines, Avondale, MD, 1980.

Corrosion research is being conducted by the Federal Bureau of

Mines to determine suitable construction materials for geothermal resource recovery plants. As part of this research, the

corrosion resistance of 31 iron-, nickel-, aluminum-, copper-,

titanium-, and molybdenum-base alloys was characterized and

evaluated in laboratory corrosion studies in low- and highsalinity geobrines representative of those found in the Imperial

Valley, CA. General, crevice, pitting, weld, and stress

corrosion were measured at 105 and 232 C in deaerated brines

and brines containing dissolved 02, C02, and CH4.

[29] Cramer, S. D. , Needham, P. B., Jr., Linear Polarization Measurements at High Temperatures in Hypersaline Geothermal Brines,

Rept. #RI-8308, Bureau of Mines, Avondale, MD.

The Bureau of Mines conducted a series of in situ linear

polarization measurements in high-temperature, high-pressure

hypersaline geothermal brines at the Bureau of Mines Geothermal

Test Facility in the Imperial Valley of California. The

measurements represented an evaluation of the linear polarization technique for obtaining instantaneous corrosion rates of

materials of construction in flowing hypersaline hydrothermal

fluids that rapidly form scales on exposed surfaces. A

special method was devised for use with the linear polarization technique that resulted in obtaining corrosion rates for

1020 carbon steel, 316 and 430 stainless steel, titanium,

various nickelbased alloys, and aluminum 6061 under strong

scaleforming conditions. The measurements also provided

information on scale-deposition rates in various geothermal

environments. Exploratory in situ potentiostatic polarization

measurements were made in the flowing brines to qualitatively

determine scaledeposi tion effects on the electrochemical

measurements

.

[30] Cramer, S. D. , Carter, J. P., Corrosion in Geothermal Brines

of the Sal ton Sea Known Geothermal Resource Area, ASTM Special

Tech. Publ . 717, Philadelphia, PA, 1981.

Corrosion research is being conducted by the Bureau of Mines,

U. S. Department of the Interior, to determine suitable con- struction materials for geothermal resource recovery plants.

High chromium-molybdenum ironbase alloys, nickelbase and

titaniumbase alloys, and a titaniumzirconiummolybdenum alloy

(TZM) exhibited good resistance to general, crevice, pitting.
14

and weld corrosion and stress corrosion cracking in laboratory

tests in deaerated brines of the Sal ton Sea known geothermal resource area (KGRA) type at 232 °C and in brine containing

dissolved carbon dioxide and methane. Only titanium-base

alloys were resistant to corrosion in oxygenerated Sal ton Sea

KGRA- type brine. Copper adversely affected the resistance to

general corrosion of lowalloy steels in deaerated brine,

whereas chromium, nickel, silicon, and titanium improved it.

Carbon steel. Type 4130 steel, and Types 410 and 430 stainless

steels exhibited poor corrosion resistance in field tests in

five brine and steam process streams produced from geothermal

well Magmamax No. 1.

These alloys were highly susceptible to pitting and crevice

corrosion. General corrosion rates were high for the carbon

and Type 4130 steels.

[31] Cramer, S. D. , The Solubility of Oxygen in Brines from 0 to

300 °C, I&EC Process Design and Development, April 1980 by the

American Chem. Soc. solubility of oxygen in water, in

sodium chloride brines, and in two geothermal brines typical

of the Imperial Valley, CA., was determined for temperatures

from 0 to 300 °C and for brine concentrations up to 5.69 m in

dissolved salts. Measurements were made in a high-pressure,

stirred autoclave by the technique of gas extraction. The

solubility, expressed in terms of the Henry's Law constant, k, was described with a standard deviation of six percent or less

by the empirical equation.

In k = ao + ai/T + a 2 /T 2 + as/T^ + a 4 /T^;

where T is the absolute temperature in K, and the coefficients,

ag through a 4 , depend on the concentration of dissolved salts

in the brine. A minimum in the solubility occurred in the

temperature range 60 to 100 °C. The salting-out coefficient

for sodium chloride brines varied by a factor of two over the

temperature range 0 to 300 °C with a minimum at 155 °C. The

effect of temperature on solubility in the two geothermal

brines was different from that in the sodium chloride brines

in the mid- and high-temperature ranges.

[32] Davis, R. B., Munir, Z. A., Corrosion Susceptibilities of

Various Metals and Alloys in Synthetic Geothermal Brines, J.

Matls. Sci., 11(9), 109013 (1977).

The corrosion susceptibilities of various pure metals and

alloys were investigated in synthetic geothermal fluids.

Rates of corrosion of AISI 1010 steel, types 304 and 316

stainless steels, Monel 400, and nickel were determined at

three temperatures (296, 333, and 368 K) ; and those of the

molybdenum, niobium, and titanium were determined at 368 K only. Type 304 stainless steel appears to undergo an activepassive transition at a temperature range between 333 and 368

K. In the passive state, type 304 steel has essentially the same corrosion rate as type 316. At 368 K, the corrosion rate

of pure nickel was approximately 2.5 times that of Monel,
15

which in turn was twice that of type 316 stainless steel. The

corrosion rates of Mo, Nb, and Ti were less than one mdd at

the highest experimental temperature.

[33] deSoultrait, B. , Comparative Study of the Resistance of A1 and

Cu Alloys to Corrosion for Underground and Open-Air-Mains,

Corrosion, Traitements, Protection, Finition, ^(7), 46773

(1972).

Keywords: Aluminum base alloys, corrosion, copper base alloys,

corrosion resistance.

[34] Du Pont de Nemours, Waste Management Quarterly Report, JulySeptember 1977, Savannah River Lab., Department of Energy, p. Ill, 1977.

Status of the following studies is reported: new reference

process for defense waste solidification, sludge washing

studies, spray calciner feasibility test, viscosity of glass

melts, electrical resistivity of glass melts; tank farm model

for Savannah River Plant, pilot-scale offsite tritiated wastes;

corrosion of carbon steel by concentrated synthetic nuclear

wastes, chemical form of plutonium in groundwater from the

burial ground, and groundwater transport of ext 129 I from

seepage basins.

[35] Du Pont de Nemours, Waste Management Quarterly Report, AprilJune 1978, Savannah River Lab., Department of Energy, p. 167,

1978.

The following were studied: environmental impact of longterm

management of highlevel radioactive waste at SRP (draft programmatic statement). Defense Waste Solidification Facility, cost

analysis of defense waste processing facilities, improved

reference process for defense waste supernate clarification,

removal of exp 90 Sr from waste supernate, removal of A1 from

sludges by caustic washing, ion chromatographic determination

of nitrite, demonstration of largescale calciner, radiolytic

gas production from TRU waste in concrete, incorporation of

TRU incinerator ash into glass, improved glass frit, test of

smallscale jouleheated melter, evaluation of candidate alloys

for incan melting canisters, residual solids from Tank 16

annulus, waste monitors, chemical form of Pu in burial ground

soil, alpha waste trench coring, corrosion of buried drums,

radionuclide release and migration from buried wastes, groundwater flow pattern from burial ground wells, and radionuclide

uptake by pine trees from buried irradiated scrap metal.

[35a] Corrosion Studies on Copper and titanium-lead cannisters-(or

Nuclear Waste Disposal. Ekbom , 1 . ^. , Flannerz , K, . Henri kson ,

J<. S^. , (Natl. Def. Res. I,st., Stockhom, Swed.). Underground

Disposal Radioact. Wastes, Proc. Symp. 1979. (Pub. 1980). 1,

503-16 (Eng). IAEA: Vienna, Austria. The Nuclear Fuel

Safety Project (KBS) has proposed that spent non-processed

nuclear fuel shall be disposed of by enclosing it in Cu canisters

or alternatively that reprocessed and vitrified waste shall be

enclosed in a Ti canister with a Bp lining. The canisters are
16

to be placed in vertical drill holes in rock, 500 m below

ground and embedded in a buffer of sand and bentonite. The

purpose of this arrangement is to raise several obstacles

against fission products reaching the biosphere. The thickwalled canister is one of these obstacles, which is proposed

to be a barrier for a considerable period of time. Corrosion

is the limiting factor of the canister durability. The rate

of corrosion is dependent on the amt. and transport of corrosion

reactants to the surface of the canister. The thermodn.

possibilities for various corrosion reactions on Cu and Pb

under the prevailing conditions were studied, also with regard

to bacterial influence. Entrapped atm. 0 and S^” by diffusion

was calcd., and hence the greatest possible corrosion. The

corrosion attack may start as pitting but will penetrate into

the thick metal wall at a decreasing rate. An expert group

arrived at the conclusion that under given conditions the

canisters will last for a very long time (hundreds or thousands

of years for Cu canister). To verify the expected high corrosion

resistance of Ti, lab. tests were carried out in environments,

which must be considered to provide accelerated rather than

simulated tests. No localized corrosion or H pick-up was detected, and the oxidn. rate was extremely low, corresponding

to a wal 1 -thinning of Ti of only 0.01 ym/yr. The results of

potential meausrements during y-i^adiation showed that the

passivity of Ti is not affected by the radiation.

[36] Escalante, E. , Gerhold, W. F. , Galvanic Coupling of Some

Stainless Steels to Copper Underground, Inti. Corr. Forum on Prot. and Perform, of Matls: Corrosion/75, Toronto, Ont.

,

April 14-18, 1975, p. 10, NACE, Houston, TX, 1975.

This paper is primarily concerned with the effects on the

underground corrosion of three types of stainless steels when

galvanically coupled to copper. It has been found that the

effect on corrosion of galvanically coupling copper to 26Cr6.5Ni alloy, type 304, and type 409 stainless steels is small.

There was no observable increase in the corrosion of the

coupled stainless steel compared to the same material uncoupled

at any of the six sites after three and four year exposure.

In fact, the evidence is that in some instances, the copper

protected the stainless steel.

[37] French Patent, 2209983, Confinement Chamber for Radioactive

Products or Waste, P29, January 18, 1977.

Keywords: Reinforced concrete and stainless steel, design,

tanks, underground disposal, chromium alloys, iron alloys,

corrosionresistant alloys, building materials.

[38] French Patent, 2446529, Container for Strongly Radioactive

Materials, January 12, 1979.

The title container is made of thick sheet steel cylinders,

bolted at the ends and soldered along the sides to make a water-tight unit of several concentric rings with inside

diameter % 1.20 m and length 5 m. The innermost cylinder was
17

made rust-proof with a 4 mm thick stainless steel coating.

Steel cylinders (5) are fitted one within the other, soldered

together along the sides, and bolted at the ends so that none

of the fasteners are directly in line with the others.

[39] Gerhold, W. F. , Escalante, E., Sanderson, B. T., The Corrosion

Behavior of Selected Stainless Steels in Soil Environments,

NBS-IR 812243, 1981.

In order to obtain more definitive information regarding the

corrosion and stress corrosion of stainless steels in soil

environments, NBS in cooperation with the Committee of Stainless

Steel Producers, AISI, initiated in 1970 a soil burial program

in representati ve soil environments. Test materials included

coated and uncoated sheet specimens in the annealed and sensitized

condition, uncoated welded tubing specimens and galvanically

coupled and uncoupled stressed and unstressed specimens. To

date approximately 10,000 specimens have been buried at six

soil test sites. This report contains the results obtained

for specimens buried for up to approximately eight years.

[40] Gerhold, W. F. , Fink, J. L., Corrosion Evaluation of Underground

Telephone Cable Shielding Materials, NBS-IR 812243, 1981.

Corrosion data is given on the performance of base and plastic

coated metals intended for use as cable shields for buried

telephone cable. The materials investigated on specially

prepared specimens were buried for periods up to six years in

six different soil environments. Metals tested included

homogeneous plasticbonded and metal lurgicallybonded laminates.

Some specimens were exposed bare (uncoated), while others had

plastic coatings or other types of coatings on either one or both sides. Metals studied included aluminum, copper, low

carbon steel and stainless steel alloys.

[41] Goldberg, A., Comments on the Use of 316 L Stainless Steel

Cladding at the Geothermal Niland Test Facility, UCID17113

University of California, Livermore, CA, 30 April 1976.

[42] Goldberg, A., Geothermal Materials Studies. Metallurgy Division

Quarterly Report, April-June 1976. Lawrence Livermore Lab.,

Livermore, CA.

Progress in evaluating materials for Total Flow Process geothermal

energy conversion systems, particularly materials for a 100 KW,

brine-tolerant turbine, is reported. In the Total Flow Process,

the total flow of the wellhead is expanded through convergingdiverging nozzles to obtain lowtemperature, high-velocity

streams of two-phase fluids for driving impulse turbines. The

turbine materials must be corrosion resistant to the high

salinity acid brines and exhibit resistance to erosion caused

by the impingement of high velocity brine droplets and entrained

materials. Programs for studying erosion in nozzle and turbine

materials, stress corrosion cracking in turbine materials;

e.g., Ti-, Fe-, Ni-, and Co-base alloys, and corrosion and

erosion failures in wellhead casing and surface pipes are
18

summarized.

[43] Gray, L. W. , Donnan, M. Y., Okamoto, B. Y., Chemical Characterization of SRP Waste Tank Sludges and Supernates, Savannah

River Lab. , p. 141 (1979)

.

Most high-level liquid wastes at the Savannah River Plant (SRP) are by-products from plutonium and enriched uranium recovery

processes. The high-level liquid wastes generated by these

separations processes are stored in large, underground, carbonsteel tanks. The liquid wastes consist of: supernate (an

aqueous solution containing sodium nitrate, nitrite, hydroxyl,

and aluminate ions), sludge (a gelatinous material containing

insoluble components of the waste, such as ferric and aluminum

hydroxides, and mercuric and manganese oxides), and salt

cake (crystals, such as sodium nitrate, formed by evaporation

of water from supernate). Analyses of SRP wastes by laserRaman spectrometry , atomic absorption spectrometry , spark- source mass spectrometry, neutron activation analysis, colorimetry,

ion chromatography, and various other wet-chemical and radiochemical

methods are discussed. These analyses are useful in studies

of waste tank corrosion and of forms for long-term waste

storage.

[44] Haijtink, B., Corrosion of Potential Materials for Containers

and Repository Structures for the Disposal of Highly Radioactive

Waste in Geological Formations, Workshop on Radioactive Waste

Management, Brussels, Belgium, June 12, 1979, ^(2), p. 597-

617, 1980.

Under the R & D programme on "Radioactive Waste Management and

Storage" (indirect action 1975-1979) of the Commission of the

European Communities, the Commission participates in various

studies concerning the corrosion aspects of potential materials

to be used in the conditioning, storage, and disposal of

radioactive waste. The materials to be used for the containers,

borehole casings, and repository linings must be to a high

degree resistant to different corrosive media under a wide

range of conditions concerning temperature, pressure, humidity,

etc. and for long periods (500-1000 years). Special criteria

are being established for the selection of materials and

programmes to test corrosion behavior in laboratories , as well

as in situ , are planned or already started.

[45] Haynes, G. S. , Baboian, R. , Comparative Study of the Corrosion

Resistance of Cable Shielding Materials, Matls. Perform.

,

18(2), 45-56 (1979).

Data are given on the three year performance of A1 , solid Cu,

and copper clad stainless steel (Cu/430 stainless steel/Cu) at

four sites with varying resistivities. Data are given also on six year results from a Rural Electrification Administration--

National Bureau of Standards test program involving cable

shielding materials with and without various dielectric coatings.
19

Effects of resistivity, pH, aeration, and dissolved salts are discussed. Data are presented on weight loss, perforations,

and galvanic effects. In general, localized corrosion of

aluminum occurred due to differential aeration, even when

plastic coatings were applied. Copper corroded at predicted

rates, but the copper clad stainless steel provided additional

protection due to the stainless steel barrier mechanism.

[46] Hehemann, R. F. , Troiano, A. R., AbuKhater, B., Ferrigno, S.,

Hydrogen Sulfide Stress Corrosion Cracking in Materials for

Geothermal Power, Electrochem. Soc. Mtg., Las Vegas, NV,

October 17, 1976, p. 34.

Studies to evaluate the performance of alloys used in geothermal

power systems are reported. Alloys which are commercially

available and those which have modified metallurgical structures

and/or composition modifications were tested to determine the

corrosive effects of the H 2 S and thermal environments in

geothermal fluids. Hydrogen embrittlement and sulfide stress

corrosion cracking were tested. Test results showing the

effects of alloy composition, tempering temperatures, fluid

temperature, and salt content, and aging on sulfide stress

cracking are tabulated.

[47] Henrikson, S. , de Pourbaix, M. , Corrosion Testing of Unalloyed

Titanium in Simulated Disposal Environments for Reprocessed

Nuclear Waste, Kaernbraenslesaekerhet, Rept. # STUDSVIK/El79/83, p. 34 (1979). Stockholm, Sweden.

On the commission of the Nuclear Safety Project (KBS), cor- rosion tests have been carried out on unalloyed titanium,

which is planned to be used as an outer corrosion resistant

canister for reprocessed nuclear waste. Tests were conducted

for 300 days in a corrosive medium of modified Baltic water at

100 and 130 °C with high (eight ppm) or low (< 10 ppb) oxygen

content. The tests at low oxygen content and 100 °C were continued for 600 days. Very low oxidation rates (0.01-0.1 mu m/year) were obtained, corresponding to a lifetime of ten to

hundred thousands of years for a six mm thick titanium canister.

In spite of the considerably more severe corrosive conditions (higher

temperature, higher chloride, and fluoride contents, as well

as lower pH) compared with those estimated for the final

disposal, no signs of localized corrosion were found, nor could any hydrogen pick-up be detected.

[48] Henriksson, S. , Pettersson, K. , Suitability of Titanium as a Corrosion Resistant Canister for Nuclear Waste, Kaernbraenslesaekerhet,

p. 55 (1977). Stockholm, Sweden.

A literature study and inventory of experience has been carried

out, aimed at assessing the possibilities of unalloyed and Pdalloyed titanium withstanding corrosion for 1000-10000 years

in contact with Baltic Sea water at 100 °C and pH 4-10. The

following assessment can be made: (1) Pitting, crevice corrosion,

stress corrosion cracking, and corrosion fatigue constitute no problem if the canister is made of unalloyed titanium corresponding
20

to ASTM Grade 1; (2) linear extrapolation of reported corrosion

rates for oxidation and general corrosion gives a life of

between 1000 and 10000 years for a five mm thick canister; and

(3) Hydrogen embrittlement resulting from hydrogen pick-up

from the deposition environment should not occur. Delayed

failure caused by a redistribution of the hydrogen initially

present in the titanium can be avoided if its concentration is

maximized to 20 ppm.

[49] Hoffman, T. L. , Corrosion Monitoring of Storage Bins for

Radioactive Calcines, Matls. Perform., j_5(l) (1976).

Corrosion coupons exposed in aluminum and zirconium calcined

radioactive waste from nuclear reactors using highly enriched

/SUP 235/U have been withdrawn and average losses calculated.

Calculated corrosion losses for 500 years are 4, 5, 20, and

35 mils for AISI types 304L, 304, and 405 and 1025 carbon

steel, respectively, in aluminum calcinate and 5, 7, 40, and

50 mils, respectively, for the same materials in Zr calcinate.

Revision of coupon withdrawal invervals to 10, 100, 250, and

450th year is contemplated. Types 304LC and 405 bins are underground in concrete vaults with provision for monitoring

leakage from or into the bins and vaults.

[50] Hoffman, T. L. , Corrosion Evaluation of Stainless Steels

Exposed in ICPP High-Level Radioactive Waste Tanks, Corrosion/75,

Toronto, Ont., April 14-18, 1975, p. 15.

Corrosion studies were conducted on welded stainless steel

types 304L, 316, 316ELC, and 348 and unwelded type 304L in

five different waste solutions from the reprocessing of nuclear

fuels at the Idaho Chemical Processing Plant, ICPP. Corrosion

rates were determined by evaluation of test specimens of the

various stainless steels immersed in the actual waste tanks

for periods of up to 19 years. As part of the over-all survei lance program on the ICPP Waste Tank Farm, the corrosion of some of these tanks has been monitored at five different

times

.

[51] Holzworth, M. L., Girdler, R. M. , Costas, L. P., Rion, W. C.,

How to Prevent Stress-Corrosion Cracking of Radioactive Waste

Storage Tanks, Matls. Protect., _7(1 ), 36-38, (1968).

Keywords: Carbon steel, corrosion, storage tanks, corrosion,

stress corrosion cracking, stress relieving, radioactive

waste, corrosion prevention.

[52] Johnson, A. B., Jr., Spent Fuel Storage Experience, Nucl

Tech. , ^(2), 165-73 (1979).

Irradiated nuclear fuel has been stored in water pools at

essentially all nuclear reactors, beginning with the earliest

plants in 1943. Fuel from water-cooled power reactors is clad

either with Zircaloy or with stainless steel. Zircaloy-clad

fuel has been stored in the US pools since 1959. Some experimental stainless-steel-clad fuel was stored for 12 years in
21

the US before reprocessing. Canadian Zircaloy-clad fuel has

been stored since 1962. There has been no evidence that the

fuel has degraded during pool storage, based principally in

visual observations and radiation monitoring of pool air and

water. However, several fuel rods have been subjected to

metal lographic examination after pool exposures up to 11 years, also with no evidence that the fuel cladding has degraded in the pool. The favorable storage experience, demonstrated technology, successful handling of fuel with reactorinduced defects, benign storage environments, and corrosionresistant materials offer sufficient bases to proceed with

expanded storage capacities and extended fuel storage until

questions regarding fuel reprocessing and final storage of

nuclear wastes have been resolved. Some surveillance is

justified to detect degradation if it becomes significant.

Surveillance programs are already under way in several countries .

[53] Jorda, R. M. , Ellis, R. C., State of the Art in Well Completion Technology as Applied to Geothermal Development. Preliminary Report, Including: Part I. Corrosion and Metal

Problems. Part II. Scale Deposition and Control. Part III.

General Production Interval Completion Techniques. Part IV.

All-Liquid Heat Recovery System Model, Completion Technology

Co., Houston, TX, p. 160 (1977).

The preliminary phase of a state of the art report concerning

well completions for geopressured and geothermal energy source

and injection wells is presented. The report covers corrosion

and metal problems, scale deposition and control, general

production interval completion techniques, and a model of an

all -liquid heat recovery system. Both hot water geopressured

source wells and water disposal (injection) wells are discussed

in some detail. Areas for further development and study with

regards to United States Gulf Coast geopressured well completions are identified.

[54] Kaernbraenslesaekerhet, Corrosion Resistance of Copper Canisters for Final Disposal of Spent Nuclear Fuel, p. 165 (1978.

Stockholm, Sweden.

The Nuclear Fuel Safety Project (KBS) has proposed that spent

non-processed nuclear fuel can be disposed of by enclosing in

copper canisters with 200 mm thick walls. The canisters are

to be placed in vertical drill-holes in rock 500 m below

ground, and embedded in a buffer of compacted bentonite. The

thermodynamic possibilities for various corrosion reactions on copper under prevailing conditions were studied, also with

regard to bacterial influence. Oxygen entrapped in the buffer

material at the closing of the storage was found to be the

oxidant of major importance for the corrosion. Sulphide in

the ground water was found to be another reactant of importance.

The supply of oxygen and sulphide mainly by diffusion was
22

calculated, and hence the greatest possible corrosion of

copper.

[55] Kaernbraenslesaekerhet, Estimation of the Corrosion Resistance

of Materials Intended for Enclosure of Nuclear Fuel Waste,

P. 154 (1977). Stockholm, Sweden.

Within the KBS project, the Swedish Corrosion Institute has

got the task to evaluate the corrosion resistance of different

materials proposed to be used in canisters for nuclear waste.

For this purpose, the Institute appointed a reference group of

specialists, mainly within the fields of corrosion and material.

KBS has proposed three different alternatives of canisters:

(1) titanium lined with lead as a canister material for reprocessed

and vitrified waste, (2) copper as a canister material for

direct disposal of spent fuel, and (3) Alumina as a canister

material for direct disposal of spent fuel. For further

evaluations, the Institute point out the need of complementary

investigations; e.g., on variations in ground water composition

at a depth of 500 m, especially the content of oxygen, chloride,

nitrite, sulfate, and organic matter.

[56] Kaernbraenslesaekerhet, Corrosion Resistance of Titanium

Canisters Lined With Lead for Final Disposal of Reprocessed

and Vitrified Waste From Nuclear Reactors, p. 80 (1978)

Stockholm, Sweden.

The nuclear Fuel Safety Project (KBS) has proposed that re- processed and vitrified waste from nuclear reactors would be

disposed of by enclosure in titanium canisters with six mm

thick walls and a 100 mm thick lead lining. The canisters are

to be placed in vertical drill-holes in rock, 500 m below

ground, and embedded in a buffer of 80-90 percent sand and 10-

20 percent bentonite. On estimation of the life of the titanium

sheath, a general corrosion rate of 0.25 mu/year has been

taken as a conservative value, which would lead to a life of

at least ten thousand years. Pitting and crevice corrosion

have been considered very unlikely at the foreseen temperatures and salt contents. Further, the risk of delayed fracture, due to hydrogen up-take, is considered as small but

cannot be completely excluded at the present state of knowledge. For this reason, the titanium sheath cannot absolutely

be guaranteed an appreciable lifetime. If the titanium sheath were penetrated due to mechanical damage or localized corrosion,

the exposed lead could suffer localized attack. The corrosion

rate would then be determined by the supply of oxygen from the

surrounding buffer to the canister surface. Conservative

calculations have shown that perforation of the 100 mm thick

lead lining would take about 4500 years. In any case, the

life of the lead lining was estimated to at least a thousand years.

In total, a titanium canister with a lead lining was estimated

to have a life of at least a thousand years and probably tens

of thousands of years. The expert group was unanimous in its

judgement, with the exception of Prof. Goesta Wranglen, who
23

has delivered a statement of his own.

[57] Kaernbraenslesaekerhet, Handling of Final Storage of Unreprocessed Spent Nuclear Fuel. 2. Technical, p. 325 (1978).

Stockholm, Sweden.

In this report, the various facilities incorporated in the

proposed handling chain for spent fuel from the power stations

to the final repository are described. Thus, the geological

conditions which are essential for a final repository are discussed, as well as the buffer and canister materials and

how they contribute towards a long-term isolation of the spent

fuel. Furthermore, one chapter deals with leaching of the

deposited fuel in the event that the canister is penetrated,

as well as the transport mechanisms which determine the migration of the radioactive substances through the buffer material.

The dispersal processes in the geosphere and the biosphere are also described together with the transfer mechanisms to the

ecological systems, as well as radiation doses. Finally, a summary is given of the safety analysis of the proposed method

for the handling and final storage of the spent fuel.

[58] Kamrowski, J. , Jakobs, J., Corrosion of Zn Coatings in Underground Conditions in Cu Mines, Ochr. Przed Koroz, ^(12), 310-

12 (1977).

Keywords: Galvanized steels, corrosion, chlorides, environment

underground corrosion, underground mining, inhibitors.

[59] Khanlarova, A. G. , Aliev, Ch. A., The Corrosion Resistance of

Aluminum Alloys in Stratal Water, Korroz. Zashch., 8 (1979).

Corrosion tests of 500 h duration of stressed (0.9 YS0.2) and

unstressed welded specimens of various A1 alloys were made in

a stratal water of pH 5. 4-6. 3 contg. C- 0.096, S024- 0.0011,

HCO-3 0.0018, Ca2+ 0.007, Mg2+ 0.007, and Na+1 K+ 0.085 g.

equiv. /lOO g, with and without H2S 500 g/1. The alloys

included 1911 Zn 4.5, Mg 1.4, Mn 0.3, Zr 0.2. 1915 .3. 5:2.0

Zn-Mg, D16T .Cu 3. 8-4. 9, Mn 0.3-0. 9, Mg 1.2-1. 8, and AMg3,

AMg5, and AMg6 .contg. Mg3, 5, and 6 percent resp... The AlMg alloys, stressed or unstressed, were the most corrosion

resistant to H2Scontg. stratal water. The AlZnMg alloys were

subject to significant corrosion and their use in these environments is not recommended. The use of D16T alloy in oil industry

equipment should be limited because of a tendency to conceal

subsurface corrosion.

[60] King, F. D. , Baker, W. H., Interim Storage of Spent Fuel

Assemblies, Inti. Symp. on Mgmt. of Waste From the LWR Fuel

Cycle, Denver, CO, July 11, 1976.

Spent fuel discharged from light water reactors (LWR) is

cooled at the reactor sites for at least five months to allow

shortlived radioactive isotopes to decay. Recently, spent

fuel has been considered as a possible waste form suitable for
24

interim storage or even ultimate disposal. Several alternatives have been demonstrated or proposed for retrievable

storage of spent fuel for periods of up to 100 years. These

include storage in watercooled basins, aircooled vaults,

concrete surface silos, geologic formations, or nearsurface

heat sinks. Watercooled sotrage of spent fuel in nearsurface

cells of heavily reinforced concrete lined with stainless

steel has been proven by about 30 years of operating experience

at reactor sites and fuel reprocessing plants. Nearsurface

storage with forceddraft air cooling of HTGR (High Temperature

Gas Reactor) fuels is being used by INEL (Idaho National

Engineering Laboratory) and is feasible for unpackaged LWR

fuel that has been out of the reactor at least three to four

years. Natural -draft cooling of spent fuel has also been

proposed, and demonstration programs are in progress for CANDU

(Canadian Deuterium Uranium) fuel. Spent fuel assemblies are sealed in thick lowcarbon steel containers and placed in large

cylindrical concrete housings (silos) located outdoors. The

cooling is completely passive, requiring little maintenance

and only minimal surveillance. Recent studies comparing the

economics of interim storage of spent fuel (throwaway fuel

cycle) with prompt chemical reprocessing conclude that disposal

as fuel decreases the Nation's natural resources significantly

and is not cost justified if both plutonium and uranium are recycled in the nuclear fuel cycle.

[61] Kobayashi , K. , Corrosion-Resistant Composite Material for

Radioactive Waste Disposal, Japan, Kokai Tokyo Koho, 79,

112,927, September 1979.

A composite material consists of an amorphous alloy and

hydraulic cement compounds and is resistant to corrosion,

leaching or cracking.

[62] Kopczynski, C. , Corrosion of Cu, A1 , and Zn in Subsoil Waters

of the "Machow" S Mine, Ochrona Przed Korozka, 1_5(10), 270-72

(1972).

Keywords: Copper base alloys, corrosion, aluminum base alloys,

zinc base alloys, tanks.

[63] Koplik, C. M. , Oston, S. G. , Pentz, D. L., Talbot, R., Information Base for Waste Repository Design. Vol . 3. Waste/Rock

Interactions, Rpt. No. TASC-TR-1210-l-Vol-3, p. 137 (1979).

Analytic Sciences Corp. Reading, MA.

This report is Volume 3 of a seven volume document on nuclear

waste repository design issues. This report describes the

important effects resulting from interaction between radioactive waste and the rock in a nuclear waste repository. The

state-of-the-art in predicting waste/rock interactions is

summarized. Where possible, independent numerical calculations

have been performed. Recommendations are made pointing out areas which require additional research.
25

[64] Krysko, W. W. , Lead as a Material for Atomic Waste Containers.

Analysis of the Corrosion Behavior Based on Archaeological

Finds, Metal 1., _M(5), 433-36 (1980). Germany

From archaeological finds, it has been observed that grain

size and shape is the factor governing the corrosion durability of lead. For long-term corrosion stability, the grain

size should not exceed 0.5 mm grain diameter. On the basis of

an investigation of 2000 year old Pb artifacts, conclusions

are drawn, and manufacturing conditions for Pb containers for

the burial of high-level atomic waste are recommended.

[65] Landry, B. , Curry, D. M. , Cox, J. E., Disposal of Thermal/

Radioactive Wastes in Permafrost Ground Areas, Paper No. 73-

WA/HT-8, ASTM Mtg., November 11-15, 1973, p. 12.

Thermal wastes, either hot fluids or radioactive materials,

present a disposal problem in permafrost areas due to the

delicate thermal balance in the environment. A numerical

technique is developed to simulate the heat transfer characteristics for two aspects of the underground disposal problem: (a) hot fluid injection down a wellbore into a porous

reservoir; and (b) storage of a radioactive heat source. Temperature fields and permafrost melting-front locations are presented graphically. Parametric studies include the effects

of variable thermal properties, depth of disposal region, etc.

[66] Leidheiser, H. , Jr., Corrosion of Copper, Tin, and Their

Alloys, John Wiley & Sons., Inc., NY, 1971, p. 411.

Keywords: copper base alloys, tin base alloys, underground

corrosion, stress corrosion cracking.

[67] Lovachev, V. A., Criteria for the Protection of Trunk Pipelines

Laid in Saline Soils, Tr. Vses. Nauchno-Issled. Inst. Stroit.

Magistral'n. Truboprovod. , 41, 40-56 (1977).

Abstracted from Russian Ref. Zh.; R.-Z. (Kor.) 810K213.

Experimental data from a study of the effectiveness of cathodic protection of underground trunk pipelines in saline soils

established that the corrosion rate of pipeline steel could

reach 0.6 mm/year and the pit depth, 2.4 mm/year. Parallel

specimens in approximately the same conditions could have

corrosion rates differing by an order from these values and

similar effects obtained for specimens at the same cathodic

protection potential. The maximum danger of corrosion was at

defects in the coatings on the lowest one-third of the pipe.

Temporary disconnection of cathodic polarization, or even a reduction in the potential value, could increase the corrosion

rate. It is concluded that the potential of -0.9 V (CU/CUSO4

electrode) guaranteed protection in soils only in certain

cases and that, for maximum reliability, a value of -1.0 V should be used.
26

[68] McCawley, F. X., Corrosion of Materials and Scaling in LowSalinity East Mesa Geothermal Brines, Rpt. No. BUMINES-RI8504, Bureau of Mines, p. 24 (1980).

The Bureau of Mines, in pursuing its goal of extending the

life span of strategic materials, conducted field corrosion

studies at the East Mesa Known Geothermal Resources Area (KGRA)

in the Imperial Valley, CA, to determine the optimum materials

of construction for use in geothermal mineral energy resource recovery plants. These studies included characterization of

geothermal environments and in-si tu corrosion testing. The

corrosion resistance of ten alloys exposed to five brine and

steam process environments was evaluated using the low-salinity,

high-temperature brine from geothermal well Mesa 6-1.

[69] McCright, R. D. , Corrosion Behavior of Materials Exposed to

Hypersaline Geothermal Brine, The Inti. Corrosion Forum Sponsored by the National Association of Corrosion Engineers,

April 1981, Toronto, Canada.

The corrosion rate and corrosion attack characteristics were determined for thirteen commercially available materials

exposed in a geothermal production well for three months. The

materials included carbon steels, CrMo alloy steels, martensitic

and ferritic stainless steels, highnickel alloys, and titanium.

The environment at the 1800 ft. (600 m) depth of exposure was

a single phase high salinity brine. The prevailing temperature

was 260 C and the prevailing pressure was 630 psi (4.0 MPa)

during the exposure period. Results indicated that the carbon

steels suffered intense generalized and localized corrosion.

Addition of Cr and Mo to steels imparted significant improvement in the corrosion performance in this aggressive environment. Of the stainless steels tested, the most resistant were

those containing a few per cent molybdenum.

[70] Mack, J. E., Lackey, W. J., Angelini, P., McCoy, H. E., Jr.,

Corrosion Testing of Candidate Canister Materials With Alternative Waste Forms, Trans. Ann. Conf . , Can. Nucl. Soc., 1, 60-

1 (1980).

The program of corrosion testing of metal alloy canisters to

determine their compatibility with immobilized radioactive

waste forms during geological storage is discussed.

[71] Magnani, N. J. , Braithwaite, J. W. , Corrosion Resistant Metallic Canisters for Nuclear Waste Isolation, Sci. Basis Nucl.

Waste Mgmt. , 2 , 377-84 (1980).

Studies were made to assess the ability of a canister alloy to

survive > 300 years in a repository environment. TiCode-12

and stainless steel 304 were tested in deoxygenated brines and

seawater at 250-300 The 304 stainless steel was susceptible

to localized corrosion and stress corrosion cracking in Clplus water-containing environments. TiCode-12 was corrosionresistant in bedded salt and sealed environments. No pitting
27

or crevice corrosion was observed at < 300 Uniform corrosion rates were 1-10 ym/yr.

[72] Mattson, E. , McCarthy, G. J., Corrosion Resistance of Canisters

for Final Disposal of Spent Nuclear Fuel. Scientific Basis for

Nuclear Waste Management, Ann. Mtg. Matls. Rsch. Soc., Boston, MA,

1978, p. 271-81.

A group of Swedish scientists have evaluated, from the corrosion

point of view, three alternative canister types for final

disposal of waste from nuclear reactors in boreholes in rock

500 m below ground. Titanium canisters with a wal 1 -thickness

of six mm and 100 mm thick lead lining have been estimated to

have a life of at least thousands of years, and probably tens

of thousands of years. Copper canisters with 200 mm thick

walls would last for hundreds of thousands of years. The

third type, *alpha*-alumina sintered under isostatic pressure,

is a very promising canister material.

[73] Mattsson, E. , Localized Corrosion, Br Corros. J., 1978, 13

(1), 512.

Keywords: Steels, corrosion, aluminum base alloys, copper,

corrosion pitting, underground corrosion, corrosion prevention.

[74] Mecham, W. J., Seefeldt, W. B. Steindler, M. J., An Analysis

of Factors Influencing the Reliability of Retrievable Storage

Canisters for Containment of Solid High-Level Radioactive

Waste, ANL, 76-82, August 1976.

[75] Merz, M. D. , Zima, G. E. , Jones, R. H., Westerman, R. E.,

Materials Characterization Center Workshop on Corrosion of

Engineered Barriers, p. 72 (1981). Battelle Pacific NW Laboratory, Richland, WA.

A workshop on corrosion test procedures for materials to be

used as barriers in nuclear waste repositories was conducted

August 19 and 20, 1980, at the Battelle Seattle Research

Center. The purpose of the meeting was to obtain guidance for

the Materials Characterization Center in preparing test procedures to be approved by the Materials Review Board. The

workshop identified test procedures that address failure modes

of uniform corrosion, pitting and crevice corrosion, stress

corrosion, and hydrogen effects that can cause delayed failures.

The principal areas that will require further consideration

beyond current engineering practices, involve the analyses of

pitting, crevice corrosion, and stress corrosion, especially

with respect to quantitative predictions of the lifetime of

barriers. Special techniques involving accelerated corrosion

testing for uniform attack will require development.

[76] Miller, R. L. , Results of Short-Term Corrosion Evaluation

Tests at Raft River, p. 95 (1977).

Four categories of short-term materials evaluation tests were conducted in geothermal fluid from Raft River Geothermal
28

Experiment, Well No. 1, to obtain corrosion data relevant to

the design of the Raft River Thermal Loop Facility. Test

programs are described and the testing philosophies are discussed. All materials and configurations which were tested

are identified and details of post-test visual examinations

are presented. The materials are then assigned to appropriate

performance categories on the basis of test behavior, and the

possible service limitations are appraised.

[77] Molecke, M. A., Gas Generation From Transuranic Waste Degradation, Symp. Sci. Basic Nucl . Waste Mgmt. , Boston, MA, November 26, 1979, p. 9.

Based on experimental data, the relative quantities of gas

generated by individual mechanism for existing forms of TRUcontaminated wastes under expected WIPP environmental conditions are ranked in order as: (1) BACTERIAL (greatest potential, but large variabi 1 ity) ; (2) THERMAL (at 40 °C , probably

insignificant at 25 °C);

(3) RADIOLYSIS (most predictable, potential short-term risks);

(4) CORROSION (under the dry conditions expected in WIPP); and

(5) ALPHA DECAY (insignificant). Results indicate that waste

processing to minimize such gas generation is not technically

required.

[78] Molecke, M. A. , Degradation of Transuranic-Contaminated Wastes

under Geologic Isolation Conditions, Rept. No. C0NF8006112,

Inti. Symp. on the Management of Alphacontaminated Wastes,

Vienna, Austria, June 1980.

An extensive experimental study of the degradation of existing

forms of defenserelated , transuraniccontaminated (TRU) wastes

is being conducted by and for Sandia Labs. Studies have been

performed under environmental conditions expected for deep

geologic isolation in the Waste Isolation Pilot Plant (WIPP),

a beddedsalt TRU waste repository planned for southeastern New

Mexico. A primary purpose of this program has been to support

the US Dept, of Energy in establishing waste acceptance criteria

for TRU wastes for isolation in the WIPP. TRU waste forms

investigated include cellulosics (paper, rags, wood), plastics,

rubbers, mixed organic composite, bitumen, a developmental

concreteTRU ash matrix, mild steel, and inorganic process

sludges. This review includes laboratory and field results

for waste degradation under a range of conditions, resultant

gas generation rates, gas compositions, corrosion rates, and

microbial formation and degradation of chelating agents.

[79] Molecke, M. A., Abrego, L., Status of Sandia HLW Canister/

Overpack Program Studies, SAND802191C. Presented at 1980 NWTS

Annual Infor. Mtg., Columbus, OH, December 1980.

[80] Nathan, J. B. , Jr., Evaluation of Materials for Underground

Exposure in Extreme Environments, SAMPE Tech. Conf. & Exhibit,

Palo Alto, CA, October 17-19, 1972, p. 37-48.
29

Results of an investigation are reported which was carried out

to define those materials most suitable for fabricating con- tainers protecting electronic hardware that can withstand

long-term burial in all geographic locations. The investigation was conducted in the four phases. Two-year underground

exposure testing in Minnesota and North Carolina substantiated

laboratory selections of 316 stainless steel, Sn63 or 55/45

Bi-Pb solder, flexible sealants, and a bituminous epoxy coating.

[81] Needham, P. B. , Jr., Cramer, S. D., Carter, J. P., McCawley,

F. X., Corrosion Studies in High-Temperature, Hypersaline

Geothermal Brines, Paper No. 59, Corrosion/79, NACE, Houston, TX.

Results for laboratory corrosion studies of 23 metals and

alloys in high- temperature, high-pressure hypersaline brines

are reported. These include the effects of O 2 , CH 4 , and CO 2 additions to the brines on general, crevice, pitting, and weld

corrosion, and stress-corrosion cracking. The results of insitu corrosion tests at a geothermal well are reported.

[82] Neretnieks, T. , Transport of Oxidants and Radionuclides Through

a Clay Barrier, Kaernbraenslesaekerhet , p. 31 (1978). Stockholm, Sweden.

The mass transfer rate for oxidants to, and radionuclides

from, a capsule in a repository has been computed. The capsule which is 0.75 m in diameter is surrounded by Montmorillonite clay. The hole is 1.5 m in diameter. For one capsule

about 1220 g, copper will corrode due to oxygen corrosion in

10,000 years. If the fissures in the rock nearest the hole

are filled with clay, the corrosion will decrease significantly.

This is valid for a case where the ground water is in equilibrium

with oxygen of 0.2 bar pressure (normal air pressure). Measurements

of the oxygen content in ground water at large depths show a more than 1000 times smaller values. The transport rate will

then be correspondingly smaller. Corrosion due to sulfate/sulfide

corrosion may reach some 590 g in the same time if there is

te!^ mg/1 of the least abundant component. The radionuclides

Sr^°, Cs^^^, Am^^^, and will decay totally in the clay

barriers. Pu^^° will be seriously hindered. The total dissolution of the uranium oxide in a capsule takes at least 1.8

million years. Nuclides with high solubilities decrease in

about 2000 years to half their original concentration. The

sodium in the Montmoril Ionite clay in the fissures is exchanged

for calcium in about 20,000 years. The exchange of the sodium

in the clay in the hole takes millions of years.

Neretnieks, I., Note on the Consequence of Hydrogen Production

in the Repository. Some notes in connection with the KBS

Studies of Final Disposal of Spent Fuel, Sept. 1978, p. 1619.

KBS Rept. # TR120.

[ 83 ]
30

Hydrogen may be produced in the copper capsule by zircalloycorrosion and by radiolysis. Hydrogen diffuses slowly in

compacted bentonite. If the production of hydrogen is fast it

will not have time to escape by diffusion at the same rate at

which it is produced. A gas bubble may form. In the present

work, it is concluded that no grave consequences are expected

if escaping hydrogen opens a channel in the clay and in the

rock

.

[84] Nomi , M. , Transportation and Disposal of Radioactive Wastes,

FAPIG, 80, 31-37 (1976). Japan.

The fuel for CAHDU reactors in Canada is natural uranium. It

is burnt up to about 7500 MWD/T in heavy water reactors, then

it is technologically stored for about 100 years. After that,

it is disposed or reprocessed. Zircaloy clads will endure

corrosion in water or in air for about 100 years, and effective

disposal means will be found within 100 years. The fuel for

light water reactors is enriched uranium, and reprocessing

must be carried out. When mixed FP waste liquid can be directly

disposed into underground structure, it is solidified and

stored intermediately, then it is disposed. The separation of

transuranic elements and extinction treatment cannot be forecase

at present, but they are the promising final disposal for

future. For solidifying high-level wastes, calcination and

vitrification are conceivable. France adopts vitrification

and storage, while in the USA, high-level wastes are calcined

and contained in canisters. Technological storage must be

adopted for the time being until the method of final disposal

and the mode of solidification will be established. The

planning and investigation of retrievable surface storage

facility are in progress, and the facility is classified into

water basin concept, air-cooled concrete vault concept, and

sealed storage cask concept. National Academy of Sciences,

USA, has made recommendations on the storing method.

[85] Paige, B. E., Si edenstrang , F. A., Niccum, M. R., Evaluation

of Hazards and Corrosion of Buried Waste Lines in NRTS (Nuclear

Reactor Testing Station) Soils, Report 1972, ICP-1013, p. 247.

Allied Chem. Corp., Idaho Falls, Idaho.

The corrosion of carbon steel, cast iron, and stainless steel

is compared for different types of NRTS soil. Detection of

leaks at the ground surface and infiltration of liquid from

corrosion-type leaks are described. Encasements, which protect

highly radioactive waste lines from external corrosion and

provide monitoring capabilities, are discussed. Cathodic

protection and existing systems and means of upgrading them

are described.

[86] Parsons, Brinckerhoff , Quade, and Douglas, Inc., Savannah

River Plant Bedrock Waste Storage Project--Status Report, Deep

Shaft Studies, p. 71 (1972).
31

A study is being made to determine problems and evaluate

solutions for the construction of deep shafts passing through

a major public water supply aquifer and designed to support

the 1000 year positive containment of radioactive wastes 2000

ft. underground. It is shown that safety and ventilation

requirements dictate two shafts connected at storage level,

the first or pilot shaft to have an I.D. of 12 ft., the second

(for tunnel excavation) to have an I.D. of 16 ft. The longevity

requirement and site conditions dictate that only natural

materials (wood and bitumen) and basic manufactured materials

(stainless steel and concrete) be used on the project.

Seismic effects are shown to be negligible. The requirement

for 1000 years of waste isolation with possibly unattended

conditions dictates a watertight shaft sealed internally from

the storage tunnel complex and constructed in such a manner

that no potential external flow paths are created. Protection

of the public water supply aquifer dictates external sealing

rings, one immediately below the equifer, another a short

distance above the tunnels leading to the storage zones, to

prevent any upward fluid migration. Provision must be made

for any long-term differential movement between soil and

shaft. A presently preferred solution is outlined.

[87] Pitman, S. G. , Griggs, B., Elmore, R. P., Evaluation of Metallic

Materials for Use in Engineering Barrier Systems, Ann. Mtg.

Matls. Rsch. Soc., Boston, MA, November 17, 1980, p. 8.

Conclusions of this work are as follows: Inconel, Incoloy,

Hastelloy C-276, and titanium alloys all had excellent cor- rosion resistance in all postulated repository environments

tested. Further work will be required to evaluate the pertinent enviro-mechanical properties of these materials; the

mechanical properties of Grade 2 titanium are better than

those of Grade 12 titanium, except the tensile and yield

strengths. These properties include fatigue-crack-growth

rate, environmental fatigue-crack-growth rate, fracture toughness, impact toughness, and dynamic fracture toughness; there

is no evidence in the current data to indicate that the simulated repository environment is aggressive to Grade 2 or Grade 12 titanium. This includes data from corrosion-fatigue,

crevice corrosion, wedge-loaded cracked specimens, and residual-stress specimens.

[88] Posey, F. A., Palko, A. A., Bacarella, A. L., Corrosivity of

Geothermal Brines. Progress Report for Period Ending September

1977. Final Report, p. 53 (1978). Oak Ridge National Lab.,

TN.

Results of studies carried out principally during FY-76 and

FY-77 on the corrosion of ferrous materials in synthetic

geothermal brines are summarized. A survey of prior work on electrochemical aspects of the corrosion of iron and carbon

steel in chloride solutions is presented, and some of the
32

results of these investigations are summarized. The principal

results of the present studies are then recapitulated. Included

are measurements of the corrosion potential, corrosion rate,

and polarization behavior of iron and carbon steel in deaerated

four M NaCl over the pH range from one to 11 at temperatures

up to 100 °C in a conventional Pyrex electrochemical cell.

The effect of pH on hydrolysis, precipitation, and electrochemical reactivity of ferrous and ferric ions in four M NaCl

at 25 °C is presented, and implications for plant operation

are discussed. Details of a refreshed, stirred titanium

autoclave system are described; the system permits electrochemical measurements to be made up to at least 200 exp 0 in

corrosive aqueous saline media. The effect of pH (from pH

equals seven to pH equals two) and temperature from 25 °C to

200 °C on the corrosion rate of type A212B carbon steel in

deaerated four M NaCl is described. A relatively simple

numerical correlation describes the data over the entire

temperature and pH range. The spontaneous corrosion potentials

and pitting potentials of types 304 and 316 stainless steel were measured in deaerated four M NaCl at pH equals five from

25 °C to 200 °C, and the data demonstrate the borderline

stability of austenitic stainless steel for brine service.

Finally, conclusions and recommendations for further studies

are presented.

[89] Rankin, W. N. , Canister Compatibility with Carlsbad Salt, Sci

.

Basis Nucl . Waste Mgmt., 2, 395-402 (1980).

Corrosion by solid NaCl from the Carlsbad (New Mexico) deposit

was studied for the following canister steel: (a) Cor-Ten A

oxidation-resistant grade, (b) low-C steel grade 70 of ASTM A516, and (c) AISI 304L stainless steel. No reaction by contact

was found at < 5000 h for sealed capsules and at < 10,000 h

for open capsules, at 80 or 225 °C. After 5000 h~, at 600 °C,

the low-C steel showed attack 20 mils deep. The canister

steels are evaluated for storage of solidified radioactive

waste in the salt bed.

[90] Rankin, W. N. , Compatibility Testing of Vitrified Waste Forms.

This paper describes an experimental program to evaluate

candidate metals for use in the fabrication of canisters for

long-term storage of vitrified radioactive wastes. The longterm compatibility of the candidate metal, both with the

contained vitrified radioactive waste and with the external

environments expected in possible final storage locations,

will be determined. Three canister materials were chosen for

testing, based on their oxidation resistance and cost: CorTen A, Type 304L stainless steel, and Inconel 600.

[91]

Rennerfelt, J., Composition of Ground Water in Deep Layers of

Bedrock, Kaernbraenslesaekerhet , p. 23 (1977). Stockholm,

Sweden

.
33

The effect of different types of dissolved solids in ground

water on corrosion and leaching is discussed. A suitable

composition of water for leaching tests is indicated. The

technique for sampling of water in bedrock at large depths is

discussed. Water analyses from different investigations are presented and a probable interval of water composition, as well as some maximum values, are given. Very low oxygen

concentrations, relatively high Fe exp 2+ concentrations, and

low levels of organic substance can be expected. Leaching of

bentonite can increase the concentration of organic substance.

[92] Romanoff, M. , National Bureau of Standards Circular 579, 1957,

p. 227, NITS PB168350 ILLC.

The Circular is a report on the studies of underground corrosion

conducted by NBS from 1910 to 1955. Up to 1922, the studies were confined to corrosion due to stray-current electrolysis

and its mitigation. After it became apparent that serious

corrosion occured in soils under conditions that precluded

stray-currents as an explanation, a field burial program was initiated, in order to obtain information pertaining to the

effect of soil properties on the corrosion of metals. More

than 36,500 specimens, representing 333 varieties of ferrous,

nonferrous, and protective coating materials, were exposed in

128 test locations throughout the US. During this time, the

electrical and electrochemical aspects of underground corrosion

have been continuously studied in the laboratory. Results

from both field and laboratory investigations are presented.

[93] Romben, L. , Tuutti, K. , Final Storage in Concrete of Activated

Metal Parts, Kaernbraenslesaekerhet, p. 42 (1978). Stockholm,

Sweden.

A part of the medium-level nuclear waste problem is the disposal

of metal parts which have been activated during use in nuclear

reactors. Some alloys contain Ni-59 with a half-life or 75,000 years. A proposed method for final disposal consists

in storing the waste in rock tunnels at a depth of 300-500 m

inside containers which are surrounded by a buffer consisting

of compacted quartz-bentonite mixture. This paper deals with

the possibilities to use containers made of concrete and

designed as cubic boxes enclosing the waste material cast and

incorporated into cement mortar matrix. Special regard is

given to the containment problem in relation to the mechanical

and chemical processes that are operating. A calculation of

the rate of release of Ni^^ through the walls of a container

has given a figure of about ten ^ kg Ni per year for undeteriorated container walls. In six appendices, the sizes of

different quantities describing the mechanical and chemical

processes which are operating in this containment problem are

given.

[94] Sanderson, B. T. , Romanoff, M., Performance of Commercially

Pure Titanium in Soils, Proc. 25th Conf. Ntl . Assoc. Corr.

Eng., Houston, TX, March 10-14, 1969, pp. 2-5.
34

This paper reports the corrosion behavior of specimens of

commercially pure titanium tubes with welded seams after

exposure underground for eight years. Excellent corrosion

resistance shown by specimens in different atmospheric environments, sea water and underground is attributed to the formation

and maintenance of a stable and passive oxide film over the

surface. All details of soil test corrosion test are tabulated.

[95] Sandia Ntl . Labs., State of the Art in Well Completion Technology

As Applied to Geothermal Development, Prelim. Rpt. , Part I. Corrosion and Metal Problems. Part II. Scale Deposition and

Control. Part III. General Production Interval Completion

Techniques. Part IV. All-Liquid Heat Recovery System Model,

SAND-78-7008, p. 160 (1977).

[96] Sandia Ntl. Labs., Corrosion of Some Pure Metals in Basaltic

Lava and Simulated Magmatic Gas at 1150 exp 0 C, p. 77 (1979).

Sand791981.

[97] Schulz, W. W. , Kupfer, M. J., Solidification and Storage of

Hanford's HighLevel Radioactive Liquid Wastes, Natl. Mtg. of

the Amer. Chem. Soc., 31 Mar. 1974, Los Angeles, CA.

When the present Hanford waste management program is completed

in 1980. Approximately 140 million liters of solid salt cake,

produced by evaporation of aged radioactive liquid waste will

be stored in underground mild steel tanks. Additionally,

megacuries of Sr^° and Cs^^'^ will have been removed from

Hanford waste solutions, converted to solid Sr^°p 2 and Cs^^'^Cl,

and stored underwater in doubly encapsulated metallic containers.

Several alternative modes for longterm storage/disposal of

these highlevel wastes are presently being evaluated. For some of these modes conversion of the solids to immobile

silicates of low water solubility may be desirable. Laboratoryscale studies of both low and high temperature processes for

preparation of silicate minerals and glasses are in progress;

status of the work is reviewed.

[98] Schutz, R. W. , Covington, L. C., Titanium Applications in the

Energy Industry, Pamphlet, Metallurgical Soc., AIME (1981).

Because of its freedom from corrosion in a wide range of

aggressive environments, Ti is a leading candidate for the

material of construction in many of the new processes being

developed in the energy industry. Its use has cut maintenance

and downtime costs in many plants and, in some cases, has made

new processes practical that were previously limited by availability of suitable corrosion resistant materials. A few

examples discussed are its use in heat exchangers for prototype

OTEC units, geothermal brine testing, pressure vessels and

piping for ocean mining processes, applications as condenser

tubing in sea water-cooled surface condensers, as pressure

vessels in hydrometal lurgical mining processes, and utilization in nuclear waste disposal canisters.
35

[99] Science Applications, Inc., Technical Support for GEIS: Radioactive Waste Isolation in Geologic Formations. Vol . 22. Nuclear

Considerations for Repository Design, p. 203 (1978). Oak

Ridge, TN.

This volume is one of a 23 volume series which supplements the

"Contribution to Draft Generic Environmental Impact Statement

on Commercial Waste Management: Radioactive Waste Isolation

in Geologic Formations", Y/OWI/TM-44. The series provides a more complete technical basis for the preconceptual designs, resource requirements, and environmental source terms asso- ciated with isolating commercial LWR wastes in underground

repositories in salt, granite, shale, and basalt. Wastes are considered from three fuel cycles: uranium and plutonium

recycling, no recycling of spent fuel, and uranium-only re- cycling. Included in this volume are baseline design considerations such as characteristics of canisters, drums, casks,

overpacks, and shipping containers; maximum allowable and

actual decay-heat levels; and canister radiation levels.

Other topics include safeguard and protection considerations;

occupational radiation exposure, including ALARA programs;

shielding of canisters, transporters, and forklift trucks;

monitoring considerations; mine water treatment; canister

integrity; and criticality calculations.

[100] Scott, T. A., Issues Relevant to Nuclear Waste Canister Materials, Corrosion/80, NACE Mtg., Chicago, IL, March 6, 1980,

p. 38.

This paper focuses on the canister as a component of multi barrier systems for the containment and isolation of high-level

radioactive wastes and spent fuel from nuclear reactor burning.

It provides perspective for the role of canisters in the

overall nuclear fuel cycle with special emphasis on the multibarrier concept of waste isolation. Attention is given to the

chemical composition of wastes, waste forms, interim storage

environments, and geologic isolation repository environments;

these will all have to be considered in canister materials

selection. In addition, other factors, such as mechanical

loads, radiation effects, and radioactive decay-heat generation rates, are cited. The paper culminates with a summary of

the author's opinion of the technical issues which need to be

resolved before the canisters can be considered viable components of a multi barrier system for geologic waste isolation.

[101] Shannon, D. W. , The Role of Chemical Components in Geothermal

Brines on Corrosion, Paper No. 57, Corrosion/78, NACE, Houston,

TX, March 6-10, 1978.

[102] Shannon, D. W. , Morrey, J. R. , Smith, R. P., Use of a Chemical

Equilibrium Code to Analyze Scale Formation and Corrosion in

Geothermal Brines, Proc. Inti. Symp. Oilfield and Geothermal

Chemistry, June 27-29, 1977, p. 21.
36

[103] Shannon, D. W. , Corrosion of Iron-Base Alloys Versus Alternate

Materials in Geothermal Brines, Interim Report--Period Ending

October 1977, p. 37 (1977).

A series of 30 refreshed autoclave tests and one field test were performed to define how various chemical components in

geothermal brines affect uniform corrosion of 35 materials.

The data indicate uniform corrosion rates of carbon steels

will be satisfactory for most major components of a geothermal

power plant for low salinity, neutral to alkaline pH reservoirs,

when 20 mpy (mils per year) corrosion allowances are permitted.

Corrosion rates of carbon steels probably will be excessive

under the following conditions: carbon dioxide saturated, low

pH (less than pH six) brines at ambient to 100 exp 0 C temperature, and CO2 saturated steam condensate system; applications

near 250 exp 0 C and above in salinities above five to ten percent; thin sections, such as heat exchanger tubes (may fail by

pitting); or applications where any oxygen may be present,

such as steam condenser and waste injection systems. While

some minor alloy effects were observed among the ten carbon

steels tested, the alloy composition of the carbon steel was a second order effect compared with important brine chemistry

variables, such as pH, salinity, and temperature. Acidification of East Mesa geothermal brine to pH 4.8 increased carbon

steel corrosion three to four times in agreement with lab

tests. The corrosion rates of carbon steels were found to be

largely controlled by the composition and structure of the

corrosion product film that formed on the metal. A number of

alloys were found in the screening tests that showed negligible corrosion under all conditions tested up to 250 exp 0 C

and 22 percent salinities. Alternate materials to carbon

steels include: high chromium alloys above 23 percent Cr,

including E-Brite 26-1, 446, 29 Cr-4 Mo, 29 Cr-4 Mo-2 Ni, 26

Cr-1 Mo-1 Ti, A1 6X; nickel alloys Hastelloy C276, Inconel

625, Incoloy 825; four titanium alloys; and zirconium.

[104] Slansky, C. M. , Radioactive Waste Management, Chem. Tech.,

160-64 Mar (1975).

High-level radioactive waste is produced at Idaho Chemical

Processing Plant (ICPP) during the recovery of spent highly

enriched nuclear fuels. Liquid waste is stored safety in

doubly contained tanks made of steel. The liquid waste is

calcined to a solid and stored safely in a retrievable form in

doubly contained underground bins. The calcine can be treated

further or left untreated in anticipation of ultimate storage.

Fluidized bed calcination has been applied to many kinds of

high-level waste. The environmental impact of high-level

waste management at the ICPP has been negligible and should

continue to be negligible.

[105] Slansky, C. M. , Review of Materials of Construction in Nuclear

Fuel Reprocessing and in Radioactive Waste Treatment, Paper

No. 78, Corrosion/77, NACE, Houston, TX, March 14-18, 1977,

p. 78.
37

This paper presents a brief description of the development of

fuel reprocessing and waste management with reference to the

choice of metals and alloys used in the construction of the

various plants. Recovery processes in use employ equeous

technology, i.e., the fuels are dissolved in mineral acids and

the uranium and plutonium separated from each other and from

fission products by a solvent extraction separation process

using tributyl phosphate (TBP) as a selective solvent. The

Purex process is used to recover irradiated LWR fuel and

utilizes 30 percent TBP.

[106] Slansky, C. M. , Review of Corrosion and Materials Selection in

Radioactive Waste Handling, Rev. Coat. Corr. , 2(2-3), 79-103

(1979).

Materials of construction are reviewed for the handling of

radioactive waste. Topics covered include corrosion and

materials selection in the storage of acidic and neutralized

high-level liquid waste and the storage of intermediate-level

liquid waste; in the transport of liquid waste at the spent

fuel processing plant and of all waste by air, rail, and

truck; in the treatment of waste by evaporation, digestion,

ion exchange, calcination, and pyrometal lurgy; in the storage

of solid calcined waste, transuranic waste and miscellaneous

wastes; and a commentary on materials for the storage of waste

in geologic formations (Ni-base superalloys and stainless

steels)

.

[107] Slate, S. C. , Maness, R. F., Corrosion Experience in Nuclear

Waste Processing at Battel le-Northwest, Paper No. 81, Corrosion/77, NACE, Houston, TX, March 14-18, 1977, p. 81.

Most of the canister corrosion work has been conducted in

support of the In-Can Melter (ICM) vitrification system. The

corrosion effects of the melt, furnace atmosphere, and water

basin on stainless steel canisters are insignificant or can be

controlled. Corrosion due to the melt or its vapor is not

significant for most of the examined materials. Protection

methods will be needed to protect the canister from oxidation,

such as using an inert cover gas or surface coatings. Stress

corrosion cracking can be prevented in 304L canisters by

maintaining the chloride content of the water below one ppm

chloride and the pH in the range of nine to 12 by adding

ammonia.

[108] Smyrl , W. H., Stephenson, L. L., Braithwaite, J. W. , Behavior

of Candidate Canister Materials in Deep Ocean Environments,

Paper No. 85, Corrosion/77, NACE, Houston, TX, March 14-18,

1977, p. 85.

This paper described preliminary results of tests undertaken

to evaluate the compatibility of radioactive waste storage

canister materials with the external environments of ocean water and sediments. Bottom ocean water from the North Pacific and sediments with supernatant sea water from the Atlantic
38

were used. The materials tested were base alloys of titanium,

zirconium, and nickel. All showed corrosion rates that were

very low, even at the highest test temperatures.

[109] Sokol, J. , Cooper, M. , Radioactive Waste Management--the Need

for Multiple Barriers, Proc. 15th DoE Nucl. Air Clean Conf .

,

Boston, MA, August 7-10, 1978, pp. 998-1004.

Isolation criteria for the disposal of high-level radioactive

waste are critically examined. The results indicate that the

essential period for isolation of high-level waste is about

1000 years. Multiple barriers, such as a solidified, leachresistant waste form, a corrosion-resistant outer container,

and a stable geological formation, which limit the transport

of radioactivity into the human food chain are recommended.

The multiplicity of barriers allows for the unlikely event of

failure in one or two of the barriers while still providing

adequate isolation of the waste.

[110] Syrett, B. C. , MacDonald, D. D. , Shih, H., Wing, S. S., Corrosion Chemistry of Geothermal Brines, Ntl . Sci . Found., Grant

No. AER 76-00713, pp. 16-18 (1977). Washington, D. C.

[111] Toney, S. , Cohen, M. , Cron, C. J., Metallurgical Evaluation of

Materials for Geothermal Power Plant Applications, Geotherm.

Energy, 5(9), 8-38 (1977).

The performance of some materials used for the manufacture of

conventional steam turbine power plants can be adversely

affected by contaminants, particularly sulfide and chloride,

in geothermal steam. Equally important, the chemical composition of geothermal is known to be site dependent. For these

reasons, on-site metallurgical evaluation of materials for

geothermal power generation equipment is often desirable.

This investigation is concerned with the determination of the

chemical composition of geothermal steam. An evaluation is

made about the effect of geothermal steam exposure on the

corrosion, stress-corrosion cracking, and fatigue behavior of

selected turbine and heat exchanger materials.

[112] Tskhrirshvi 1 i , D. , On Corrosion of Metals in Geothermal Power

Plants, Geothermics, ]_(3), 113 (1972).

[113] Tylecote, R. F. , Durable Materials for Sea Water: The Archaeological Evidence, Report 314(R), Brit. Nucl. Fuel. Ltd., 1977.

[114] Wethe, P. I., Appendix to Chapter 6: Radioactive Waste and

Transport. Vedlegg til Utredningen 'Kjernekraft og Sikkerhet',

p. 171-300 (1978). Oslo, Norway.

The appendices are here edited into a homogeneous treatment of

the whole question of high-level radioactive wastes originating from nuclear fuel and its reprocessing, and from low and

medium level radioactive wastes arising from operation of
39

nuclear facilities. The Swedish AKA (1976) and KBS (1977)

reports are directly quoted at length on a number of specific

subjects. The nature and amounts of the wastes, costs, res- ponsibilities; processing; solidification and vitrification of

high-level wastes; after-heat generation and cooling requirements; intermediate storage and final disposal, both underground in geologically stable formations and marine; transmutation; encapsulation and corrosion and leaching, with asso- ciated hazard evaluation are some of the principal items

discussed. The Oklo phenomenon is discussed in connection

with radionuclide migration. The final sections are devoted

to discussion of transport, containers and regulations.

[115] Yamazaki , H., Materials for Geothermal Power Plants, Chinetsu

Gijutsu^(l), 39-47 (1977). Japan.

A study was made on corrosion and abrasion resistance of

stainless steels in acidic hot solutions for development of

geothermal power plant materials. In a corrosion test in

0.1 M H 2 SO4 containing 0.001-0.01 M H 2 S, the corrosion loss of

SUS 304 was smaller than that of SUS 309 and SUS 310 and

corrosion products were mainly FeS, FeS 2 , and FeySs. SUS 304

and SUS 316 had excellent abrasion resistance in an abrasion

test in H 2 SO4 acidic (pH 1-2) and H 2 S-saturated solution.

[116] Zastrow, 0. W. , Copper-Corrosion in Moderate and High Resistivity Soils, Matls. Perf., J^( 8 ), 31-36 (1974).

Keywords: Copper, corrosion, underground corrosion, corrosion

tests

.
40

Subject Index

a acid resistance - 98

acid soils - 9

alloys, candidate - 35

aluminum - 10, 23, 40, 45

aluminum alloys - 10, 23, 28, 29, 33, 59, 62, 68, 73, 76

aluminum alloys - 61

archeological data - 64, 113

austenite - 76, 88

b bacteria - 54, 77

basalt - 99

basalitic lava - 96

biodegradation - 77

biosphere - 57

bismuth alloy - 80

brass - 3, 14, 76

brine chemistry - 17, 25, 101, 102, 103, 110

brines - 1, 10, 14, 16, 19, 31, 42

bronze - 18, 76

c calcium carbonates - 53

calcium compounds - 76

calcium sulfates - 53

canister, contai ner material s - 1 , 2, 12, 13, 14, 15, 17, 23, 25

26, 35a, 35 , 44 , 47 , 48, 54, 55, 56, 57, 60, 70, 71, 72, 74

79, 84, 89, 90, 98, 100, 107 , 108

capsules - 5, 24 , 47 , 55 , 82, 83, 114

carbides - 87

carbonates - 5, 53, 76

carbon dioxide - 22, 28, 76, 81, 103

carbon steel - 6 , 10 , 11 , 13, 19, 20, 23, 29, 30, 34, 40, 43, 49

51, 60, 68, 69, 76, 85, 88, 103

casks - 47, 84, 99

cast-iron - 10, 76, 85, 87

cathodic corrosion - 11 cathodic protection - 67, 85

chlorides - 1, 47, 55, 58, 76, 107, 111

chromium alloys - 10, 30, 37, 49, 60, 86, 103

chromium nickel steels - 49

chromium steels - 10, 23, 49, 68, 69

clads and composites - 2, 9, 11, 38, 41, 45, 52, 61, 84

clay - 2, 4, 23, 82, 83

coatings - 38, 39, 45, 107

cobalt alloys - 8, 10, 14, 42, 46, 106

composite materials - 2, 11, 40, 61

containers, canisters - 5, 6, 12, 15, 24, 27, 35, 35a, 37, 38, 44,

55, 60, 64, 72, 80, 84, 93, 97, 99, 100, 108, 109, 114

copper - 2, 3, 5, 9, 10, 24, 36, 40, 45, 54, 55, 72, 73, 82, 83,

87, 116
41

copper alloys - 3, 10, 14, 18, 28, 33, 62, 66, 76, 87

copper nickel alloys - 18

corrosion control - 73

corrosion fatigue - 48

corrosion monitoring or measurement - 22, 29, 49

corrosion rates - 1, 3, 14, 16, 17, 19, 26, 29, 30, 32, 48, 50,

56, 67, 69, 71, 78, 103

Cor-ten - 6, 89, 90

cost -6,11, 14, 35, 90, 114

crack propagation - 46

crevice corrosion -1,8, 19, 28, 30, 48, 56, 71, 75, 81, 87

design, respository - 63, 99

drums - 99

e E-Brite stainless steel - 68, 103, 107

embrittlement - 46

environments - 12, 14, 16, 17, 25, 65

environments saline - 1 , 68

f failure -1,5, 42, 46, 75, 109

g galvanic corrosion - 36, 45

galvanic coupling - 39, 45

galvanized steels - 58

general corrosion - 15, 19, 21, 28, 30, 45, 48, 81

geological environments - 14, 17, 57, 78

geological formations - 2, 14, 17, 44, 60, 86, 87, 99, 100, 106,

109, 114

geosphere - 57

geothermal - 19, 41, 111, 112, 115

geothermal brines, fluids - 3, 10, 19, 20, 21, 22, 28, 29, 30, 31,

32, 42, 46, 53, 68, 69, 76, 81, 88, 98, 101, 102, 103, 110

geothermal wells - 46, 53, 69, 95

gold plating - 7 granite - 99

ground water - 2, 24, 34, 35, 54, 55, 82, 87, 91, 93

h Hastealloy - 14, 19, 20, 23, 68, 87, 103, 106, 107

hydrogen - 4, 5, 24, 47, 48, 56, 75, 83

hydrogen compounds - 5 hydrogen effects - 75

hydrogen embrittlement - 46, 48

hydrogen sulfide - 8, 46, 59, 76

i in-can-melting - 35, 107

Incoloy - 14, 87, 103, 107

Inconel - 6, 14, 19, 20, 46, 68, 87, 90, 103, 106, 107

inhibitors - 58

iron alloys - 26, 28, 30, 37, 42, 49, 60, 76, 86, 103
42

1 lead - 2, 5, 13, 35a, 55, 56, 64

localized corrosion - 16, 45, 47, 56, 71, 73

low carbon steel - 60, 89

m magnesium chlorides - 87

martensite - 76

methane - 22, 28, 30, 81

mild steel - 78

molybdenum - 21, 28, 30, 32

Monel - 14, 32, 46, 76

multiple barriers - 75, 100, 109

n near surface storage - 60

nickel - 21, 32, 105, 106, 108

nickel alloys - 8, 10, 14, 19, 20, 23, 28, 29, 42, 49, 68, 69, 76,

103, 105, 108

nickel superalloys - 23, 106, 107

niobium - 32

nitrate - 43

nitrite - 43, 55

0 ocean sediments - 14, 17, 27, 108, 114

organic matter - 55, 91

oxidation - 48

oxidation rates - 47

oxidation resistance - 6

oxygen - 17, 22, 24, 28, 31, 47, 54, 55, 56, 81, 82, 91, 103

p palladium alloyed titanium - 48

passive oxide film - 3, 94

patents and designs - 12, 13, 37, 38, 99

pitting corrosion - 1, 8, 11, 19, 26, 28, 30, 42, 48, 73, 81, 88, 103

pH values - 88, 103, 107

plating - 7, 9

plutonium - 34, 35

polarization measurement - 9, 29

pressure effects - 17, 25, 31, 44, 69, 81

protective films - 3, 94

pyrite - 23

r radiolysis - 4, 24, 35, 63, 77, 78, 83

repositories - 5, 25, 44, 71, 82, 83, 87, 99, 100

residual stresses ~ 87

retrievable storage - 104

rocks - 2, 12, 14, 17, 54, 56, 63, 72, 82, 93, 99

s saline - 1 saline soils - 67

salt - 87 , 88 , 89

salt deposits - 14, 17, 25, 63, 71, 89, 99 sea water - 1, 10, 14, 15, 16, 27, 47, 48, 108, 113

sensitized steels - 39
43

shale - 17, 99

silica - 76

silicon bronze - 18

soil environment - 26, 39, 67, 85, 92

soils, corrosivity - 9, 26, 45, 85, 92, 116

solder - 38, 80

Sp/Haynes alloy - 106

Sp/Illium alloy - 106

stainless steels - 6., 8, 10, n. 14, 16, 18, 19, 20, 21, 23, 29, 30,

36, 37, 38, 39, 40, 41, 45, 46, 49, 50, 52, 60, 68, 69, 71, 76,

80, 85, 86, 87, 88, 89, 90, 93, 105,, 106 107, 115

steels - 14, 26, 32, 46, 60, 67, 73, 86, 88, 104

Stellite - 46

stratal water - 59

stress corrosion -1,8, 15, 20, 21, 28, 30, 42, 46, 48, 51, 59, 66,

71 , 75, 81 , 107, 111

stressed metals - 39, 51, 59, 99

sulfate-sulfide corrosion - 53, 55, 82

sulfides - 54, 76, 111

sulfur - 23

sulfur compounds - 23

tanks - 34, 37, 43, 50, 51, 62, 97, 104

temperature effects - 1., 6, 17, 23, 25, 29, 31, 32, 44, 46, 47, 56,

63, 65, 68, 69, 78., 81 , 87,, 88,, 89,, 103 , 108

tests corros ion - 3, 6, 16, 17, 19, 20, 21, 22, 27, 28, 39, 42, 44, 47,

50, 59, 70, 75, 76,, 80 , 81,. 87,. 88,. 92, 94, 96,, 108 , 111, 112, 115, 116

Ti-Code - 1

,

15, 71

tin alloys - 66, 80

titanium - 1 , 2, 4, 10, 14, 23, 29, 32, 35a

,

47, 48,. 55, 56, 72 , 76 , 94, 98,

105, 108

titanium alloys - 14, 15, 19, 23, 28, 30, 42, 68, 87, 103, 105, 108

transition elements - 87

tungsten alloys - 87

Udimet - 107

uniform corrosion - 1, 26, 71, 75

volcanic rocks - 87

water - 5, 9

water underground - 9, 62

weld corrosion - 28, 30, 81, 94

welded aluminum alloys - 59

welded materials - 39

welded stainless steels - 39, 50

welded titanium - 94

zinc alloy - 62

zinc coatings - 57

Zircaloy - 5, 14, 52, 83, 84

zirconium - 30, 103, 108
NBS-114A IREV. 2-80

,'.S. DEPT. OF COMM.

BIBLIOGRAPHIC DATA

SHEET (See instructions)

1. PUBLICATION OR

REPORT NO.

NBSIR 81-2409

2, Performing Organ. Report No. 3. Publication Date

January 1982

4. TITLE AND SUBTITLE '

BIBLIOGRAPHY OF LITERATURE ON UNDERGROUND CORROSION OF METALS AND ALLOYS

CONSIDERED FOR USE IN THE CONSTRUCTION OF CONTAINERS FOR NUCLEAR WASTE

5. AUTHOR(S)

B. T. Sanderson and J. Kruger

6. PERFORMING ORGANIZATION (If joint or other than NBS, see in struction s)

NATIONAL BUREAU OF STANDARDS

DEPARTMENT OF COMMERCE

WASHINGTON, D.C. 20234

7. Contract/Grant No.

8 . Type of Reports. Period Covered

9.

SPONSORING ORGANIZATION NAME AND COMPLETE ADDRESS (Street. City. State, ZIP)

U.S. Department of Energy

1000 Independence Avenue, S.W.

Washington, D.C. 20585

iO. SUPPLEMENTARY NOTES

Document describes a computer program; SF-185, FIPS Software Summary, is attached.

11.

ABSTRACT (A 200-word or less factual summary of most significant information. If document includes a significant

bi bliography or literature survey, mention it here)

This bibliography consists of references pertaining to the corrosion of

metallic container materials for nuclear waste disposal. Five kinds of

corrosion data are presented - general underground corrosion data that

may apply to metallic nuclear waste containers; corrosion considerations

in package design; metallic corrosion in geothermal brines; internal corrosion

of nuclear waste containers for underground use; and external corrosion of

nuclear waste containers for underground environments. Abstracts are provided for most references, and key words are included when there is no abstract. One hundred and sixteen references are presented.

12,

KEY WORDS (S/x to twelve entries; alphabetical order; capitalize only proper names; and separate key words by semicolons)

alloys; corrosion; corrosion data; containers; geothermal brines; metals;

nuclear waste; underground

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