THE IMPORTATION AND STORAGE OF HIGH-LEVEL RADIOACTIVE WASTES AT ROKKASHO-MURA: SAFETY CONCERNS

April 16, 1996

Introduction

On 26 April 1995, the ship Pacific Pintail arrived in Mutsu-Ogawara port. The Pintail was carrying a shipment of 28 canisters of vitrified (glassified) high-level radioactive wastes (VHLW) which had originated in La Hague, France, and was on its way to a storage facility in Rokkasho-mura. This shipment of extremely hazardous material, which was the first of dozens being planned to return VHLW generated during the overseas reprocessing of Japanese- origin spent fuel, ignited considerable controversy at many points along its sea route.

Governor Kimura of Aomori Prefecture, invoking a resolution which granted him the authority to refuse the entry into port of any vessel should he feel that the assurance of safety and the provision of information are inadequate, initially refused to permit the Pintail to dock. The standoff was resolved a day later, after he had extracted a promise from the Science and Technology Agency (STA) that it would not authorize a final disposal site for HLW in Aomori without the governor's consent.

Although this agreement may have resolved for now the question of whether Aomori will become an official final disposal site for VHLW without local approval, it does not address the issue of whether the interim storage facility will become an unofficial one, should STA be unable to site a geologic repository elsewhere. Given the delays which have plagued every other aspect of the Japanese nuclear program, it is probable that the schedule for development of a final repository site will undergo slippage as well. However, there is no indication that the authorities have studied the safety of VHLW storage at Rokkasho for an "interim" period which could well be far longer than the 30-50 year design lifetime of the storage facility.

The experience of spent fuel management in the United States should serve as a warning to communities in Japan, such as the citizens of Aomori Prefecture, who accept "promises" by authorities concerning the disposal of nuclear wastes. In the U.S., spent nuclear fuel generated by electric utilities is currently being stored at reactor sites. The U.S. Department of Energy (DOE) had formally agreed to take possession of all commercial spent fuel by early 1998. However, because of delays in the geologic repository program, as well the inability to find a location which would accept an interim storage facility for spent fuel, DOE has acknowledged that it cannot honor its promise. Thus spent fuel in the U.S. may remain at reactor sites for the indefinite future.

This paper discusses some of the safety concerns pertaining to the importation and storage at Rokkasho of VHLW conditioned in Europe. Most of these concerns originate from the fact that Japan has little oversight of the foreign VHLW production process and the quality control methods being employed. Thus it is not able to closely monitor the operation to ensure that the VHLW produced abroad meets Japanese domestic safety criteria. The series of inspection tests carried out by Japan Nuclear Fuels Limited (JNFL) on the returned VHLW canisters are not sufficiently rigorous to alleviate these concerns.

These issues, some of which were raised in reports by this author over a year ago,1 have still not been satisfactorily resolved. On March 22, 1996, the Expert Advisory Committee to the Aomori Government released a safety report which directly addresses some of these issues. However, this report, which attempts to justify the safety of current practices rather than critically evaluating them, raises more questions than it answers. In fact, the inadequacy of the official response to these outstanding questions shows clearly that the two criteria of safety assurance and openness of information have still not been met for the return to Japan of VHLW produced overseas, and provides a sound basis for the Aomori governor to refuse any further importation of VHLW into the prefecture until they have been answered.

What are Vitrified High-Level Wastes (VHLW)?

High-level wastes (HLW) are the highly radioactive liquid residues which are generated when spent nuclear fuel is reprocessed to extract the plutonium it contains. Japan has been shipping spent fuel to reprocessing plants in France and the U.K. for many years. Under the terms of the contracts, the HLW must be returned to Japan for disposal.

The current industrial practice is to condition these wastes for storage, transport and disposal by converting them into solid glass blocks through a process called vitrification. In this process, HLW is blended with glass-forming materials and melted at a high temperature (about 1150C). The melt is then poured into thin-walled (5 millimeter-thick) stainless steel canisters and slowly cooled until it hardens. This product is referred to as vitrified high-level waste (VHLW).

Because HLW generates considerable heat from radioactive decay, the VHLW canisters will remain quite hot for decades after they are produced. During normal conditions of storage at La Hague, the centerline glass temperature can be as high as 510C. During transport, the temperatures at the center and surface of a VHLW canister will typically be as high as 400C and 325C, respectively, under normal conditions.2

It is important to note that the VHLW glass contained within the canister is not a single, monolithic block. During the cooling process, the glass cracks extensively, and very small particles (known as "fines") are produced where the glass comes in contact with the canister. In addition, at typical transport and storage temperatures, radioactive gases are released at a low but significant rate from the glass. Therefore, the stainless steel canister plays an essential role in containing radioactivity during normal conditions. In the event of an accident, such as a canister drop or a fire, the integrity of the canister is crucial, since the glass itself does not have a high resistance to mechanical impact or heat.

Another notable feature is that the canister material is in a state of tensile stress (stretching) as a residual effect of the cooling process. This is because once the temperature is decreased below about 450C, the steel canister tends to contract faster than the glass. This stress can be quite large; for the case of the La Hague VHLW canisters, one can show it can be as high as 200 megapascals (MPa),3 or about the same as the yield strength of the Type 309 stainless steel used for the VHLW canisters at the relevant temperature (205 MPa at 315C). (The Expert Advisory Committee report estimates that the stresses will be even larger: from 216-235 MPa). A residual tensile stress of this magnitude is very significant with respect to the occurrence of a certain corrosion phenomenon known as stress corrosion cracking, as discussed below.

Vitrification at La Hague and the Problem of "Sensitization"

Given the importance of the metal canister in ensuring the safety of transport, handling and storage of VHLW, it is obvious that the canister material should be chosen to provide a high degree of confidence in its integrity during the storage period.

However, at the R7-T7 vitrification plant at La Hague, the VHLW canister material being used does not provide such confidence. In fact, one can show that this material, known as Type SUH 309 austenitic stainless steel, undergoes a phase transformation while the VHLW canisters are being cooled. This phenomenon, known as sensitization, greatly reduces its resistance both to certain types of corrosion and to mechanical impact.

Prior to the release of the aforementioned studies by this author, the sensitization problem with the VHLW canisters produced at La Hague was apparently not known to the Japanese authorities (and probably not known to the French either). The Expert Advisory Committee report tacitly acknowledges that these VHLW canisters are likely to have been sensitized. However, the report attempts to minimize the significance of this fact by taking an inappropriately narrow view.

Sensitization of austenitic stainless steels occurs when the steel is held for a certain period of time at a temperature in the range of approximately 400- 850C. The time necessary to cause extensive sensitization depends on the composition of the steel, and in particular will decrease as the carbon content increases. For instance, one experiment found that while a 3 millimeter (mm) sample of stainless steel with a carbon content of 0.03% underwent 100% sensitization after 10 hours, one with a carbon content of 0.08% sensitized completely after only 30 minutes.4 Type SUH 309 has an even higher carbon content (0.15 weight-percent) and therefore will be completely sensitized in less time.5 Furthermore, it has been observed that stressed materials (such as the VHLW canisters) undergo sensitization more rapidly than unstressed ones.

COGEMA data shows that as the VHLW canisters are being cooled after being filled with glass, the canister temperature remains within the sensitization range for about 7 hours. Thus there is little doubt that the canisters being produced at La Hague are extensively, if not completely, sensitized.

A number of stainless steels have been developed which are resistant to sensitization. Partly for this reason, one of these steels, Type 304L, is being used as the canister material at the HLW vitrification plant now operating in the United States. Type 304L is also the steel that was chosen for use at the domestic HLW vitrification plants in Japan. However, one should note that even if Type 304L is used, sensitization may occur to a limited extent during VHLW production. (There are stainless steels which show even greater resistance to sensitization, known as the "stabilized" grades.) Why COGEMA initially chose (and continues to use) a type of steel highly susceptible to rapid sensitization, apparently without prior consultation with Japanese authorities, is by no means clear.

Safety Implications of VHLW Canister Sensitization

Now that it has been established that the stainless steel VHLW canisters produced at La Hague are sensitized, the consequences for the safe handling and storage of VHLW need to be understood. The canister sensitization issue was not considered at all in the initial certification by STA of the Rokkasho VHLW storage facility. In fact, the data contained in the reference documents attached to the facility application pertained to VHLW encased in Type 304L, not Type 309, stainless steel.

The Expert Advisory Committee report takes the view that canister sensitization is not an important consideration for the safety of VHLW storage. However, their assessment contains flaws which, rather than diminishing the significance of sensitization, illustrate why it is so important. In this section, some of the impacts of sensitization on the safety of long-term VHLW storage are discussed.

Accelerated corrosion.

Austenitic stainless steels exhibit a high degree of resistance to uniform corrosion (corrosion that takes place uniformly across a surface). However, when they are exposed to certain chemical and thermal environments, they can undergo localized corrosion processes such as intergranular corrosion (IC) and intergranular stress-corrosion cracking (IGSCC).6 Localized corrosion can typically be two or three orders of magnitude more severe than uniform corrosion, and has often led to unexpected, catastrophic failures of materials.

Some contaminants, such as chloride salts and hydroxyl (OH- ions), can initiate intergranular corrosion of stainless steels at very low concentrations, if water is also present. This is clearly a concern with regard to the integrity of VHLW canisters: for instance, when reporters were recently shown empty stainless steel canisters at the U.S. vitrification plant, they were warned not to touch them to prevent exposure of the steel to potentially corrosive salts in their sweat.7

Because the VHLW canisters being returned to Japan are produced, stored and shipped in marine environments (all facilities are located near oceans), the ambient air concentrations of chlorides from sea salt are always high and extreme care should be taken to prevent excessive salt contamination of the canisters. It is not clear that such care is taken, however.

Stainless steel that has become sensitized is much more vulnerable to localized, intergranular corrosion than the same type of steel in the unsensitized condition. Because localized corrosion is a much more unpredictable phenomenon than uniform corrosion (e.g. it is more susceptible to small changes in environmental conditions) it is less accurate to apply the results of laboratory tests to predictions of long-term performance, or to extrapolate data from a single sample to an entire lot. Thus one consequence of using sensitized stainless steels in VHLW canisters is that uncertainties in predictions of canister performance will be greatly increased. This can only reduce confidence in the results of safety analyses.

One such uncertainty is the extent to which the stainless steel canisters will be subjected to potentially corrosive environments during storage and transport of VHLW. There are two temperature regimes which are of greatest concern. The first is a high-temperature regime, occurring early in the storage period, in which the metal may be susceptible to "hot corrosion" (otherwise known as "catastrophic oxidation") from exposure to molten salts released from the VHLW, such as cesium oxide. The second is a low-temperature regime, in which the canister surfaces have cooled sufficiently to allow moisture to condense on them, which provides an electrolyte which can enable intergranular corrosion or stress-corrosion cracking to take place.

At the La Hague storage facility, where the VHLW canisters are placed immediately after they are filled, the initial temperatures may be high enough to allow hot corrosion to occur on the inside of the VHLW canister. The extent of such corrosion depends on the particular types of salts which are released from the glass and their melting points (these salts are only corrosive if they are molten). Because such salts have been observed to cause intergranular corrosion, they may attack sensitized stainless steels more aggressively.

Two salts of concern are boron oxide and cesium oxide, which are major constituents of VHLW, have relatively low melting points (approximately 450C) and have been observed to cause hot corrosion of stainless steels. Many other constituents of the chemically complex VHLW can have a similar effect at even lower temperatures, such as selenium dioxide, which is volatile, has a melting point of 330C and may have a corrosive effect on stainless steels similar to that of its chemical cousin, sulfur dioxide.

The other storage environment relevant to corrosion is the one at low temperature (below about 100C), which will occur later in the storage period. The Expert Advisory Committee report argues that in the Rokkasho-mura storage facility, surface temperatures of the VHLW canisters will remain over 100C for most of the 50-year storage period. At the bottom of the storage column, where temperatures are the lowest, the canister temperature decreases from 230C to about 140C after 20 years of storage, and then to about 100C after 50 years, so it would appear at first glance that the canisters never get cool enough for condensation to form.

There are several reasons, however, why this is not a realistic assessment.

First, these temperatures were calculated based on a thermal loading (the rate of heat generation) of 2 kilowatts (kW) per canister. However, the average thermal loading of the first batch of VHLW canisters to be placed in the facility was only about 1.6 kW per canister, indicating the VHLW canisters have been cooled for a longer time than anticipated prior to shipment. Rough calculations indicate that the canister temperatures in the facility would be approximately 40C lower at a heat loading of 1.6 kW per canister.

Therefore, the temperatures of at least some of the canisters will fall below 100C after as little as twenty years in storage, and will then be susceptible to condensation and corrosion for a minimum of ten to thirty years (and longer if the Rokkasho facility continues to operate for more than 50 years). Corrosion may cause a loss of integrity of some of the VHLW canisters, which would render ineffective one of the barriers to radioactive releases from the facility, and increase the difficulty of handling the canisters when the time comes for them to be moved to the final disposal site.

The Expert Advisory Committee report goes on to argue that even if the temperature decreases sufficiently for condensation in the storage facility to occur, the conditions for stress-corrosion cracking will not be present. They base this on an internal Central Research Institute of [the] Electric Power Industry (CRIEPI) study (not peer-reviewed) which analyzes stress-corrosion cracking in a sample of sensitized Type 309 stainless steel. They argue that according to the results of this study, the concentration of chlorides that accumulate on the VHLW canister surfaces will not exceed the threshold necessary to induce stress-corrosion cracking, for the expected conditions of temperature, humidity and stress in the storage facility.

However, the data used by the Expert Advisory Committee to make their case are of questionable relevance to the actual conditions of VHLW storage. There are three chief reasons for this:

1) It has been observed that high gamma radiation fields promote stress- corrosion cracking in sensitized stainless steels.8 This means that stress-corrosion cracking can occur for materials in gamma fields under environmental conditions that would be "safe" for the same material in the absence of the field. Since the sensitized VHLW canisters are exposed to a gamma fields on the order of 104 Gray per hour, an accurate assessment of the potential for stress- corrosion cracking must take this into account.

2) The presence of hygroscopic (moisture-absorbing) contaminants on the canister surface can change the relationship between moisture absorbed on the surface, relative humidity and temperature. For fixed temperature, this can lower the critical humidity threshold for stress-corrosion cracking to occur. There is no indication that the sample of material studied in the CRIEPI report had surface contamination comparable to that of the actual VHLW canisters. Thus the threshold humidity levels for the VHLW canisters in storage could be lower than those in the report.

3) The Expert Advisory Committee claims that the maximum salt concentration on the VHLW canisters in storage will not exceed 1 mg of chloride per square meter, even after fifty years of storage, so that stress-corrosion cracking will not be possible. This very low value is based on the fact that the external air is filtered before it comes into contact with the canisters.

However, this value is subject to question. Airborne chloride accumulation in marine environments ranges from 5-1500 mg per square meter per day. Taking a value of 50 mg per square meter per day, and assuming that the air passes through a high-efficiency particulate air (HEPA) filter which reduces the salt concentration by 99.99%, the deposition on the canisters would still be about 2 mg per square meter per year. According to the Expert Advisory Committee's own graph, this would imply that the critical salt concentration of 70 mg per square meter (corresponding to the conditions they describe as the "worst" that may be encountered in storage) would be exceeded after 35 years. This period may be even shorter if the factors described in 1) and 2) are taken into account. Mechanisms that concentrate salts at the surface may also play a role. Thus the Expert Advisory Committee does not appear to be taking a sufficiently cautious approach to this problem.

The Expert Advisory Committee also did not take into account the possibility that salt may have accumulated on the VHLW canister surfaces before being brought to Rokkasho-mura. When they are in the La Hague storage facility, the VHLW canisters are directly exposed to cooling air drawn from the outside through coarse filters. Due to the coastal location of the facility, the cooling air contains moisture and chloride salts, which are not completely removed by the filters. Although the canisters are too hot to allow condensation of water, they may be cool enough for condensation of salts to occur.

The Expert Advisory Committee also did not consider other possible sources of corrosion. For example, it is also important that the canister may come into contact with moisture on its inside surface. Although measures are taken to prevent liquid water from being sealed in the VHLW canisters, borosilicate glasses usually contain small but significant amounts of molecular water (generally about 0.01-0.1 weight-percent, or 100-1000 parts per million).9 Radiolysis of water molecules in VHLW leads to the production of hydroxyl (OH-) ions.

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End Notes

2. H. Yamakawa et al., CRIEPI, "Demonstration Test for Transporting Vitrified High-Level Radioactive Waste: Thermal Test," Presentation to PATRAM '95 Conference, Las Vegas, Nevada, December 1995. Back to document

3. R. Rulon, "Glass to Metal Seals," in Introduction to Glass Science (L. Pye et al., eds.), Plenum Press, New York, 1972, p. 661. 4. H. Solomon, "Influence of Composition on Continuous Cooling Sensitization of Type 304 Stainless Steel," Corrosion 40 (1984) 51. Back to document

5. According to the Expert Advisory Committee report, COGEMA now says that the carbon content of the Type 309 stainless steel that it is using has been lowered to 0.08%, rather than the maximum possible value of 0.15%. It is not clear whether this has always been the case or whether the alloy was changed recently in response to the work of this author. However, this point does not change any of the conclusions of this paper, because stainless steel with 0.08% carbon is also highly vulnerable to sensitization, as the above discussion makes clear.Back to document

6. Stress-corrosion cracking is a special type of corrosion that can only occur on a surface which is under tensile stress. Although there is no generally agreed-upon minimum stress threshold below which it cannot occur, it becomes more likely for stresses greater than about 70% of the yield strength of the material. As pointed out above, the stainless steel VHLW canisters experience tensile stresses in storage at or above their yield strengths. Back to document

7. M. Wald, "Factory is Set to Process Dangerous Nuclear Waste," New York Times, March 13, 1996, p, A16. Back to document

8. J. Farmer and R. McCright, "Localized Corrosion and Stress-Corrosion Cracking of Candidate Materials for High-Level Radioactive Waste Disposal Containers in the U.S.: A Literature Review," UCRL-98756, Lawrence Livermore National Laboratory, December 1989, p.6. Back to document

9. R. Bartholomew, "Water in Glass," in the Treatise on Materials Science and Technology, Vol. 22, Academic Press, New York, 1982, p.75. Back to document

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