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Summary Response of the Nuclear Control Institute to Comments on
"The Sea Transport of Vitrified High-Level Wastes:
Unresolved Safety Issues"

Edwin S. Lyman, PhD
Nuclear Control Institute
January 1998


In December 1996, the Nuclear Control Institute (NCI) released a report titled "The Sea Transport of Vitrified High-Level Radioactive Wastes (VHLW): Unresolved Safety Issues" [NCI (1996)]. The report analyzed the potential consequences of a severe accident in coastal waters that resulted in damage to the transport ship and the VHLW shipping cask, causing the cargo to sink to a depth of 200 meters (typical of the maximum depth of the continental shelf).

The report's chief finding was that such an event could result in significant health consequences to consumers of marine products from the region. This contradicts the claims of the companies in charge of shipping the waste, as well as the results of a highly publicized study by the Central Research Institute of the Electric Power Industry of Japan (CRIEPI), which found that radiation doses to humans would be extremely small.1 NCI (1996) showed that the CRIEPI results were not consistent with other studies, such as one by the Nuclear Energy Agency (NEA) of the OECD, that analyzed similar accidents and found that doses could be hundreds of thousands of times larger.2

NCI (1996) was submitted to the Marine Environment Protection Committee (MEPC) of the International Maritime Organization (IMO) by Greenpeace International, and to the U.S. Government by several members of Congress. The International Atomic Energy Agency (IAEA) submitted a highly critical commentary [IAEA (1997)] on the report to the Maritime Safety Committee (MSC) of the IMO.3 Subsequently, Sandia National Laboratories (SNL), at the request of the U.S. Department of Energy (DOE), issued a similarly critical report [SNL (1997)].4

The IAEA conceded that NCI (1996) "...in general does a mathematically correct analysis of the conditions it analyzes," but it called the report "fundamentally flawed and ... easily misleading" because (a) it considers a highly improbable scenario but does not attempt to calculate the probability of the scenario; and (b) it does not "establish the sequencing of events necessary to lead to the conditions that are analysed."

The SNL report contains similar arguments: "the scenario analyzed is so improbable that it is of very little or no concern," and that "even if, against all odds, the scenario endpoint were somehow to be reached," the consequences would be extremely minor.

A detailed review of IAEA (1997), prepared by NCI, finds that the IAEA report provides little specific information to support its conclusions, and contains numerous misleading statements as well as several outright errors. Space limitations restrict what can be included here; the interested reader is referred to the longer paper for more details.5 While IAEA (1997) and SNL (1997) overlap significantly, the latter document contains additional points which merit discussion; these will be addressed separately below.

Overall, the IAEA's critique of NCI (1996) can be characterized as "muddying the waters": instead of trying to increase understanding and reduce uncertainties, IAEA (1997) overcomplicates the issues at hand without resolving them.

However, the IAEA does not dispute the fundamental conclusions of NCI (1996); namely, that the consequences of the specified accident could be severe. In fact, an IAEA Chairman's Report, which became public after NCI (1996) was released, acknowledged that "if a large irradiated fuel package were to be lost on the continental shelf, some large exposures could result"6 (a comment which applies equally well to a VHLW package). Yet the IAEA remains conspicuously silent about the patently false claims circulated by British Nuclear Fuels Limited (BNFL) that even if VHLW became "directly exposed to the sea ... the effect of such a scenario would be negligible."7

Since the IAEA admits that severe accidents can result in large exposures, it is incumbent on it to demonstrate that the probabilities of such accidents are acceptably small. However, the IAEA has not done this. Little quantitative information exists on the frequency of sea accidents that could create environments more severe than those simulated by Type B tests, or on the resistance of Type B packages to such accidents. The IAEA invokes the principle of "graceful failure" --- that Type B packages are so resilient that they can tolerate much more severe accidents than those they are designed to survive --- but this is largely accepted on faith, and has not been verified for the package used to ship VHLW by sea.

In an attempt to resolve these issues, the IAEA is now undertaking a Coordinated Research Program (CRP) on the issue of sea transport. According to the aforementioned Chairman's Report, "the question of whether [a] sea-mode dependent package requirement might be needed ... appears to depend upon the final results of the CRP." The present lack of information undermines the IAEA's assurance that such accidents are unworthy of regulatory consideration. For emergency planning purposes, it is prudent (and consistent with IAEA's own policy) to assume that severe accidents are possible and evaluate their consequences.

Contrary to the claims of the IAEA and SNL, the following postulated chain of events resulting in a VHLW cask being damaged and lost at sea is quite plausible. For each event, outstanding technical questions are discussed that illustrate why quantification of their probabilities is such an uncertain undertaking. It is therefore premature at the least to conclude that such shipments do not pose unacceptable risks to the environment.

1) Involvement of the VHLW transport vessel in a beyond-design-basis collision.

The Pacific Nuclear Transport Limited (PNTL) ships used to transport VHLW are designed so that if struck by a vessel traveling at 15 knots with a displacement of 23,000 tonnes, the striking vessel will not penetrate the cargo area. It is unclear how they would fare against larger, faster ships that are common on the high seas. A ship with 50,000 tonnes displacement traveling at 24 knots would have a kinetic energy more than five times greater than, and could penetrate twice as deeply as, the vessel the VHLW transport ship is designed to resist.

IAEA (1997) incorrectly implies that the safety features of PNTL ships are comparable to those of ships which the previously cited OECD report judged could reduce the probability of occurrence of a severe accident to a negligible level. The reference ship considered in detail in the OECD report, which was designed "specifically to reduce the risks from transportation accidents," is considerably more robust than the PNTL ships. The OECD reference vessel is strong enough to prevent penetration into the cargo hold by a vessel of any size with a speed less than 24 knots. In addition, the ship has a large quantity of urethane foam insulation that would provide protection from an engulfing, 982C fire for three days and would prevent the ship from sinking even if it were cut in two.

The OECD report briefly considered another type of ship which resembled the PNTL vessels (purpose- built but essentially conventional in design), but did not analyze it in detail because it was recognized that it would provide a lesser degree of safety than the ship described above. The OECD estimated that the probability that one of these ships could lose its cargo as a result of an accident was between 0.12% and 0.16% per year -- hardly a negligible value.

2) A breach of the cargo area, subjecting the VHLW casks to an accident environment more severe than the IAEA Type B test, followed by sinking of the vessel or expulsion of a cask into the sea.

IAEA (1997) and SNL (1997) assert that analyses sponsored by the U.S. Department of Energy show that the forces imparted to a cask during a ship collision would be "about the same order of magnitude as those produced by the regulatory tests." However, a previous SNL study indicates that the impact environments that a cask can encounter during a marine collision may be more severe than the Type B impact test. For example, it was found that following the initial collision, the shipping cask was subjected to a series of more than ten closely spaced impacts (as a result of collisions alternating between the two hulls), at speeds of up to 33.6 meters per second, and ultimately was crushed between the two hulls.8 In contrast, the Type B test involves only one impact test, at a speed of 13.3 meters per second.

Even if all the impacts were less severe than the Type B impact, the cumulative effect of the sequence could stress the package beyond the failure point. Thus the conclusion of the SNL study that "these impact forces are likely less than would be seen during a regulatory drop test" does not appear to be supported by their results. In more violent collisions than the one analyzed by SNL, impact energies would be even greater and regulatory conditions could be exceeded by even greater margins.

3) Breaching of the VHLW package as a result of excessive stress.

The IAEA's assumption that any Type B package is capable of withstanding accidents greatly exceeding the regulatory accident is not supported by the available evidence.

4) Rapid corrosion by seawater of the sensitized stainless steel VHLW canisters.

NCI (1996) argues that the stainless steel canisters which encase the VHLW are "sensitized" --- made more vulnerable to localized corrosion --- because of an incompatibility of the selected steel with the VHLW production process. It concluded that this condition would accelerate their decomposition in seawater and subsequent release of radionuclides, as a result of intergranular corrosion (IC) or intergranular stress-corrosion cracking (IGSCC).

IAEA (1997) and SNL (1997) challenge this argument. They claim not that sensitization does not occur (and provide no evidence to support such a conclusion), but only that NCI (1996) has not provided definitive proof. As noted below, the weight of the evidence clearly indicates that the VHLW canisters undergo significant sensitization. It should be the responsibility of the producers of VHLW to ensure that sensitization, which has implications for the safety of transport and storage, does not occur.

A series of papers issued by the Central Research Institute of the Electric Power Industry of Japan (CRIEPI) has analyzed the sensitization and corrosion behavior of a variety of stainless steels, including Type 309S, which COGEMA and BNFL now say that they use. The CRIEPI reports show that Type 309S will sensitize under typical VHLW cooling conditions.13 Moreover, cooling curves specific to the production of COGEMA glass in publicly available literature show that the canister temperature does remain in the sensitization range for several hours (Figure 1).14 SNL (1997) concedes that under these conditions, "sensitization of the VHLW canisters is likely."

To conclusively resolve this issue, COGEMA and BNFL should perform chemical analysis on samples of stainless steel from actual VHLW canisters in storage and make the results and methodology available for public review. This course of action was recommended by NCI in 1994, and, to our knowledge, never carried out.

IAEA (1997) describes several conditions that facilitate the occurrence of IGSCC, such as stress, sensitization and elevated temperature. Although IAEA (1997) claims otherwise, these conditions would all be present in the event that a VHLW canister were exposed to sea water following the sinking of a ship.

IAEA (1997) says that IGSCC is unlikely because "industry sources have stated that very little operational stress is present in the canisters as presented for transport." However, it is not "operational stress" that is of concern, but "residual stress." VHLW canisters are frozen into a state of high tensile stress because of the differential thermal contraction of the stainless steel and glass during cooling. In addition, the thermal stresses induced by the welding of the canister lid can also be significant.15

For sensitized stainless steels, IGSCC has been observed for temperatures as low as 20C,16 which could be experienced by the contents of a severely damaged VHLW cask lost in coastal waters. A more serious condition would apply if the shipping cask were to remain largely intact (for example, if damage were limited to seal failure), since the temperature of the water infiltrating into the insulated cask cavity would be quickly heated to 100C or greater (as is assumed in the CRIEPI study), at which rapid IGSCC can take place, leading to canister penetration within a few months. The intense gamma radiation emitted by the canisters may also promote IGSCC because of radiolysis effects. Thus, the environmental conditions encountered in coastal waters will be severe enough to promote IGSCC of the sensitized VHLW canisters in a sunken cask.

It should be noted that the VHLW canisters can fail quickly without the occurrence of rapid corrosion. The CRIEPI study assumed that the canisters would buckle at one end as a result of the water pressure at 200 meters depth. Subsequent corrosion would enhance radionuclide releases by increasing the surface area of the VHLW directly exposed to seawater.

5) Leaching of radionuclides from the VHLW into the marine environment.

Contrary to the contentions of IAEA (1997) and SNL (1997), NCI (1996) does not assume that the shipping cask "disappears," or that it undergoes such a massive failure that it spills its entire contents onto the ocean floor. Instead, NCI (1996) conservatively assumes that the cask is damaged to the extent that there is a moderate flow of water through the cask cavity, resulting in an appreciable rate of glass corrosion.

Whether this condition will be attained merely by failure of the elastomer seals depends on the width of the gap between the lid and the body of the cask. The heat generated by the VHLW canisters will drive convection currents that will cause seawater to flow through the cask at a rate which depends strongly on the gap width (a cubic dependence). Deformation of the bolts attaching the lid to the body, which can result from a severe impact, can increase the gap and lead to a greater rate of release. A typical gap width of 0.1 millimeters will correspond to a flow rate of about 0.2 milliliters per second through the cask, corresponding to a glass dissolution rate at 90C of about 3x10-5 grams per square centimeter per day [g/(cm2-d)], which is only about one-tenth of the maximum observed rate.17

By using conservative assumptions with regard to the glass dissolution behavior, NCI estimates (see below) that a glass dissolution rate of 310-5 g/(cm2-d) could result in dose rates over 50 milliSievert (mSv), or 5 rem, per failed cask per year to consumers of marine products from the affected region, from the release of two actinides (americium-241 and curium-244) alone. Retention of actinides at the glass surface may lower this estimate by a factor of 100, but there is strong evidence that retention may not occur.

The CRIEPI study assumes that the VHLW cask is undamaged except for failure of the O-ring seal. The low dose rates it obtains (610-4 mSv per year, which is 100,000 times smaller than the above estimate) can be largely attributed to the assumption of an unrealistically small value for the gap width (0.01 millimeters), resulting in a flow rate one thousand times smaller than the one given above. Small variations in gap width can lead to large variations in the consequences. However, the CRIEPI report does not discuss the sensitivity of their results to variations in the input parameters.

Even accounting for differences in the assumed flow rate, the CRIEPI result is still considerably smaller than the NCI estimate. Therefore, other assumptions in their model must also play a role. It is difficult to understand why the radiological consequences of such an accident in Japanese coastal waters would be so much smaller than those in the region considered in the OECD study (the East Coast of the U.S.).

IAEA (1997) expresses the concern that it is misleading to directly compare different studies because of their sensitivity to local geographic patterns. How then can it justify allowing BNFL to use the results of the CRIEPI study, specific to a sinking off the coast of Japan, to reassure en-route states all along the route? IAEA's point only underscores one that the Nuclear Control Institute and Greenpeace International have made repeatedly --- that site-specific environmental impact statements need to be carried out along the entire shipping route.

Response to Additional Points in the Sandia National Laboratories Report

Overall, the thrust of the Sandia National Laboratories report [SNL (1997)] is very similar to that of IAEA (1997). It argues that the scenario outlined in NCI (1996) is of extremely low probability, and moreover that NCI cannot prove the occurrence of the phenomena that it has identified as enhancing the risk of an accident with severe consequences. Otherwise, SNL (1997) seems to inordinately focus on a few minor technical points that have little bearing on the ultimate conclusions of the NCI (1996). SNL (1997) contains a good deal of speculation and does not provide sufficient evidence to justify its conclusion that there is virtually no risk from shipments of VHLW. Points specific to the Sandia report are briefly addressed below. In one instance, the report identified a numerical error in NCI (1996), for which the author is grateful. This error has now been corrected. However, it does not affect the overall conclusions of NCI (1996), as is explained below. This response is organized to respond point by point to SNL (1997).

2.0 and 3.0: Accident Scenario and Scenario Probability

CLAIM: Significant cask damage can only occur if the VHLW ship is adjacent to another ship or massive object when it is struck by an oncoming ship. This means that the probability of such an accident is proportional to the square of the probability that a single ship is involved in an accident.

RESPONSE: SNL's own studies indicate that a shipping cask aboard ship can experience accident conditions greater than those simulated by the IAEA Type B test if the ship is struck by another ship, such as a series of repeated collisions against the two hulls. Moreover, as discussed above, seal failure due to a fire or prolonged period of submersion may be sufficient to lead to significant release of radioactivity. Therefore, the probability of an accident causing damage to the cask sufficient to allow significant releases of radionuclides subsequent to submersion may be considerably greater than SNL (1997) claims.

CLAIM: The NCI scenario assumes that the ship is sunk over the continental shelf of the U.S., a region only 150 nautical miles wide.

RESPONSE: The path of the VHLW ship takes it over numerous coastal regions with depths of less than 200 meters, including (depending on the point of origin and the route) the English Channel, the Irish Sea, parts of the Carribean, the Gulf of Panama, and coastal Japan. The total distance it travels over such regions is considerably greater than 150 nautical miles. Moreover, NCI (1996) does not claim that a sinking in regions of 200-2000 meters depth would be benign. However, it has not seen any data or analysis for the consequences of accidents in this depth range.

The Sandia report attempts to show that the probability of an accident at sea occurring that will lead to a significant near-term release of radioactivity is so small as to be essentially inconceivable. However, they state that "although [probability] values are not available for many of the events in the scenario set forth ... an estimate can be made." In other words, most of the numbers they use are pulled out of thin air. Their result itself is so low that it strains credulity.

5.0 Cask Recovery (salvage).

CLAIM: If an accident resulting in the loss of VHLW canisters at sea does occur, the canisters almost certainly will be salvaged within a few months.

RESPONSE: Even if one accepts the unproven proposition that salvage of a severely damaged VHLW cask is an easy task, SNL (1997) is itself the best evidence that salvage will NOT take place. Salvage will only be attempted if the will exists to undertake a complex and potentially dangerous operation. However, SNL (1997) claims repeatedly that there is little risk, even in a worst-case accident, that there will be significant radionuclide releases in the near-term. This raises the concern that should such an accident occur, the shippers will use arguments very similar to those in SNL (1997) to justify not attempting recovery of the cargo. A similar situation occurred late in 1997 when the MSC Carla, carrying cesium chloride therapeutic radiation sources, sank in the mid-Atlantic. The French safety agency IPSN quickly released a calculation which purported to show that there would be only negligible harm from the sources if left in the ocean, and therefore there was no need to recover them.

6.0: Release of Radionuclides from VHLW

CLAIM: Because the activities of Am-241 and Cm-244 are smaller than the activities of Cs-134 and Sr- 90 in the VHLW shipment, the radiological hazards of Am-241 and Cm-244 are accordingly smaller.

RESPONSE: This statement demonstrates a lack of understanding of radiological dose assessment. The radiological harm resulting from release of a particular radionuclide depends on how it is transported through the marine environment and food chain, as well as the biological hazard associated with an intake of unit activity (which is much greater for the alpha-emitters Am-241 and Cm-244 than for cesium isotopes). The conversion coefficients derived from the studies analyzed in NCI (1996) indicate that in most cases, the committed effective dose to individuals per unit release of activity into coastal waters is smaller for cesium isotopes than for the transuranic isotopes. However, the relative hazards of these isotopes is a minor consideration compared to their absolute hazards, which are significant in each case.

It should also be noted that VHLW contains substantial concentrations of other hazardous radionuclides. However, NCI (1996) did not make an exhaustive analysis of all of them, but only focused on the few for which enough information was available. Including additional radionuclides in the analysis would increase dose estimates.

CLAIM: The radionuclide leach rates NCI (1996) cites are not appropriate for the actual VHLW in the shipment, since leach rates are highly composition-dependent. NCI should include data for SON 61 glass as well as R7T7 glass.

RESPONSE: The authors of the Sandia report seem confused about the glass terminology used in the NCI report. 1021C is not a "new" glass composition, but is the numerical designation of a particular VHLW canister produced at the R7 or T7 facilities at La Hague. Each canister has a slightly different radionuclide content, but this does not affect the average leach rates. The baseline glass composition is known as R7T7 glass, which is also denoted SON 68. SON 61 glass is an earlier glass composition which is no longer employed. The application for return of VHLW to Japan refers exclusively to R7T7 glass. Therefore, the properties of SON 61 glass are not relevant to the behavior of the glass being shipped to Japan by sea.

CLAIM: The release of radionuclides from a submerged VHLW cask will be two orders of magnitude smaller than the values given in NCI (1996) and any resulting radiation exposures will be much smaller than background.

RESPONSE: SNL (1997) is correct in pointing out that the physical radionuclide release rates specified in NCI (1996) are not consistent with those in the paper by Vernaz and Godon cited by NCI. This was due to a normalization error; NCI is grateful for the correction. However, a reexamination of the source term issue indicates that the values given by Vernaz and Godon may not be appropriate for the scenario under consideration. It is possible that release rates and resulting radiation doses may be even greater than those estimated in NCI (1996).

The release rates for actinides (e.g. americium and curium) from R7T7 glass given in the 1988 OECD study are considerably higher than those given by Vernaz and Godon. This is because the former release rates were derived from the assumption that all radionuclides dissolve congruently in water (meaning that all glass components are released at rates proportional to their concentrations in the glass). However, under some conditions, such as dissolution in pure water, low-solubility actinides are preferentially retained at the glass surface. This is the effect that was observed by Vernaz and Godon.

Sediments present on the sea floor can have a large impact on radionuclide releases from VHLW. Vernaz and Godon observed, for example, that the capability of VHLW to retain americium was reduced by two orders of magnitude if a mixture of granite, sand and clay was added to the system. The OECD study conservatively assumes that the presence of sediments will inhibit saturation and re-precipitation of solubility- limited radionuclides such as the actinides. Given the large range of uncertainty in this calculation, this seems like a reasonable assumption.

The most recent volume of the series Scientific Basis for Nuclear Waste Management contains two papers which provide further evidence that the release rates of actinides (e.g. Pu, Am and Cm) from HLW glass may be greater than have been previously assumed, and that congruent dissolution of actinides may provide a more conservative model.

A Japanese study of R7T7 glass confirmed that the actinides Pu and Cm were released primarily in the form of colloids (insoluble suspensions of large particles) during glass dissolution.18 When in the colloid phase, these radionuclides do not contribute to the saturation of the solution because they are not dissolved. Consequently, total concentrations of Pu and Cm in the colloid phase were 10-100 times greater than the concentrations in the solution phase. Also, the pattern of release of these radionuclides was not characteristic of solubility-limited dissolution but of "alteration-limited" dissolution, e.g. the rate of release was controlled only by the rate of reaction of the glass with water.

A paper from the U.S. found somewhat different but also significant results.19 This study found that actinides were initially retained on the glass surface, but after several years the plated-out material began to flake off in the form of large particles that formed colloids. The result was a rapid increase in the rate of release of Pu and Am from the glass at this time. According to the paper, "the spallation of alteration phases, some of which have incorporated Pu and Am, led to total release of these elements approaching that expected from congruent dissolution of the glass."

One can estimate the worst-case consequences of an immersion accident by assuming congruent dissolution of actinides. In the worst case, assuming leaching by free-flowing water at 100C, the rate of glass corrosion is 2x10-4 g/(cm2-d). This corresponds to an actinide release per failed VHLW canister, assuming congruent dissolution, of about 0.009 TBq/day each of Am-241 and Cm-244. Using the OECD "least favorable" dose conversion coefficients, this would result in an annual dose rate to the maximally exposed individual of 17 mSv per canister, or about 340 mSv for a 20-canister shipping cask. This should be compared to the ICRP limit for public exposure to artificial radiation of 1 mSv per year. Thus the worst-case conditions are indeed quite severe.

If it is assumed that actinides are fully retained at the glass surface, the Vernaz and Godon results would imply dose rates around 0.75 mSv per year, a significant fraction of the ICRP limit. According to ICRP, the 1 mSv/year limit applies to all sources of artificial radioactivity; the dose rate from any single source should be only a fraction of this limit.

If the cask only suffers damage to the seal, lower flow rates will occur, but as discussed above, dose rates can still be quite high and very harmful.

CLAIM: NCI (1996) assumes that the surface area of the VHLW canisters increases by a factor of 10 as a result of a collision.

RESPONSE: NCI (1996) does not assume that an increase in surface area of the glass occurs as a result of the collision. Instead, the factor of 10 increase is a consequence of the well-documented fracturing that takes place when VHLW is cooled during production. For example, the CRIEPI study assumes a factor of 20 increase relative to geometric surface area for its calculations. The NCI report, on the other hand, only assumed a factor of 10 because it was closer to the factor of 9 increase in surface area reported by COGEMA for the reference R7T7 glass, and the factor of 11 assumed by the OECD study. If the authors of the Sandia report had familiarized themselves with the previous studies before attempting to critique the NCI report, they would have understood the origin of this factor.

However, the Sandia report raises an important point that should be taken into consideration in future studies. The surface area of the glass block may well increase as a result of a collision. For instance, in the case of a VHLW canister which underwent an impact test at a velocity of 35.2 m/s (about 2.5 times the Type B velocity), a factor of 40 increase in effective surface area was observed.20

7.0: Individual Doses

The seven authors of the Sandia report were unable, "despite careful review of the cited OECD report," to derive the coefficients relating peak individual doses to unit radionuclide releases presented in the NCI report. This calculation is straightforward and will be provided here for their benefit.

As an example, we derive the Cs-137 coefficient. For the "best estimate" case for a damaged canister, Table 4.1 of the OECD report gives a peak individual dose of 6.510-5 Sv/yr-MTHM (all values in the OECD report are normalized to MTHM of the original spent fuel), of which 5%, or 3.25x10-6 Sv/yr-MTHM, was attributed to Cs-137 releases. Normalizing to each VHLW canister with the correspondence 1.39 MTHM/canister, one obtains 4.5x10-6 Sv/yr-canister.

Now, the source term must be computed. The total mass leach rate from VHLW used in the OECD report for the "best estimate, damaged canister" scenario is 8.510-6 g/cm2-day. The concentration of Cs-137 in the VHLW at 50 years after vitrification (the assumed time of the accident) is 0.102 Ci/g. Therefore, the Cs-137 leach rate is 8.710-7 Ci/cm2-d = 3.210-8 TBq/cm2-d. The assumed effective surface area per canister is 19.25 m2, so the Cs-137 leach rate is 2.25 TBq/yr-canister.

The OECD study shows that the individual dose rates generated by their model are linear with respect to the rate at which radionuclides are released. Therefore, it is meaningful to divide the peak individual dose rate by the assumed radionuclide release rate to obtain the peak individual dose corresponding to unit radionuclide release. For Cs-137, one obtains (4.510-6 Sv/yr-canister)/(2.25 TBq/yr-canister) = 0.002 mSv/TBq, the value which appears in Table I of the NCI report. Values for the other listed radionuclides can be similarly obtained.

As the Sandia report points out, in this OECD scenario, peak individual doses appear 10 years after the accident. This, however, does not mean that if the packages are recovered within ten years that the resulting doses will be insignificant. Moreover, the "least favorable estimate" scenario of the OECD report finds that peak doses result within 3 years, and other models find that peak doses occur less than a year after the accident.

9.0 Sensitization of Stainless Steel VHLW Canisters

CLAIM: The exact composition of the stainless steel used for the VHLW canisters is unclear and may contain stabilizing elements. Therefore, its vulnerability to sensitization cannot be surmised.

RESPONSE: The stainless steel compositions used for the R7T7 canisters has been well documented in the import licenses issued by Japanese authorities. They are identical to ASTM Type 309 and Type 309S. There are no stabilizing elements such as niobium present.

CLAIM: Because the process conditions at the DWPF HLW vitrification plant at the Savannah River Site in the U.S. do not cause significant sensitization of the stainless steel canisters, such sensitization is not likely to occur at the R7T7 plants in France.

RESPONSE: Process parameters at DWPF at SRS are not relevant to those at the R7T7 plants, because the thermal conditions under which DWPF glass is poured are considerably less severe than those experienced with production of VHLW from commercial high-level wastes. The heat content of defense HLW is considerably less than that of commercial HLW. A DWPF canister contains, at most, one-third of the heat output of an R7T7 canister immediately after filling. Moreover, DWPF canisters are both taller and wider than R7T7 canisters, and have about three times the surface area. This means that the centerline temperatures of DWPF canisters decrease much more rapidly after they are filled and the canister wall temperature can also be reduced more rapidly without risk of thermal shock. DWPF canisters are cooled to below 100C before decontamination. In contrast, R7T7-type canisters are preheated to 1150C to minimize thermal shock and remain well over 100C when placed in storage.

Moreover, even at DWPF, it is observed that canisters remain in the sensitization range for up to two hours. This is not regarded as a problem because DWPF uses the low-carbon, sensitization-resistant Type 304L stainless steel. This period of time would be sufficient to sensitize higher carbon stainless steels such as Type 309S, which is used at the R7T7 plants.

Ultimately, Sandia acknowledges that sensitization is likely if the "assertion" of the NCI report that the VHLW canisters remain within the sensitization range for seven hours can be verified. The authors of the report are invited to inspect Figure 1 of this document, which presents the cooling curve of a COGEMA-type glass block. They are also invited to request this information from COGEMA directly and to make it publicly available.

End Notes

1. D. Tsumune et al., "Study on Method of Environmental Impact Assessment During Sea Transport of Radioactive Materials," Proceedings of the 11th International PATRAM Conference, December 1995, Las Vegas, Nevada, p. 41. Back to document

2. Nuclear Energy Agency, Feasibility of Disposal of High-Level Radioactive Waste into the Seabed, Volume 2, OECD, Paris, 1988, p. 144. Back to document

3. "Comments on MSC 68/INF.2 and MEPC 39/INF.15," submitted by the International Atomic Energy Agency (IAEA) to the 68th session of the Maritime Safety Committee of the International Maritime Organization (IMO), MSC 68/15/4, 28 February 1997. Back to document

4. J. Sprung et al., "Comments on a Paper Titled 'The Sea Transport of Vitrified High-Level Wastes: Unresolved Safety Issues,'" SAND97-1130, Sandia National Laboratories, May 1997. Back to document

5. Edwin S. Lyman, PhD, "Response of the Nuclear Control Institute to the IAEA Comments on 'The Sea Transport of Vitrified High-Level Wastes: Unresolved Safety Issues,'" May 1997. Available on request from the Nuclear Control Institute, 1000 Connecticut Ave. NW, Suite 804, Washington, DC, 20036. Back to document

6. C. Young, Chairman's Report, Advisory Group Meeting on "Modal Issues in the Safe Transport of Radioactive Material," International Atomic Energy Agency, Vienna, 4-6 November 1996, p. 76. Back to document

7. British Nuclear Fuels Ltd. (BNFL), "Shipment of Nuclear Material Between Europe and Japan," Press Statement, 4 December 1996. Back to document

8. V. Porter and D. Ammerman, "Analysis of a Ship-to-Ship Collision," Proceedings of the PATRAM '95 Conference, Volume III, p. 1083. This study computed the average force as a function of time on a cargo of shipping casks during a collision with a 16,750-tonne ship traveling at 30 knots. It was found that the magnitude of the initial average compressive force on the cargo was 370 meganewtons (MN), corresponding to an average acceleration of 343g for a 110 tonne VHLW cask, which is 15% greater than the 300g average acceleration measured during 9-meter drop testing of the cask. Integration of the impulse vs. time curve shows that this impact would send the cargo toward the opposite hull at a speed of 33.6 meters per second. Although the hulls are not "unyielding surfaces," as are required in the regulatory tests, they have been stiffened to provide resistance to collisions. Even if half the energy were lost to hull deformation, the energy imparted to the package would still be more than twice that associated with the regulatory test. Back to document

9. D. Ammerman and J. Bobbe, "Testing of the Structural Evaluation Test Unit," Proceedings of PATRAM '95, Volume III, p. 1123. Back to document

10. D. Brownowski and P. McConnell, "Performance Characteristics of O-Ring Seals for Radioactive Material Packages When Subjected to Extreme Temperatures," Proceedings of the 11th International Conference on the Packaging and Transport of Radioactive Materials (PATRAM '95), Volume IV, p. 1791. Back to document

11. BNFL, Cogema and FEPC, "Information Paper Submitted to the Special Consultative Meeting of the IMO by BNFL, Cogema and FEPC," undated, p. 16. Back to document

12. Y. Gomi et al., "Demonstration Test for Transporting Vitrified High-Level Radioactive Waste: Immersion Test," in the Proceedings of PATRAM '95, Volume III, p. 1145. Back to document

13. M. Mayuzumi, "Effect of Relative Humidity on Stress Corrosion Cracking Susceptibility of Candidate Canister Materials," Komae Research Laboratory Rep. No. T86050, p. 20, Table 4. Back to document

14. R. Simon, Commission of the European Communities, Brussels, "The Qualification of Waste Forms and Engineered Barriers," in Radioactive Waste Management and Disposal (L. Cecille, ed.), Elsevier Applied Science, London, 1991, p. 159, Fig 1. Back to document

15. According to the Aomori Experts' Advisory Group report, this residual stress is estimated to be 22-24 kg/mm2 (216-235 MPa). This value is 79-85% of the room-temperature yield strength of Type 309S stainless steel (275 MPa), a stress level at which IGSCC can readily occur. Back to document

16. A. John Sedriks, Corrosion of Stainless Steels, second edition, John Wiley and Sons, New York, 1996, p. 287. Back to document

17. S. Gin, "Control of R7T7 Nuclear Glass Alteration Kinetics Under Saturation Conditions," Scientific Basis for Nuclear Waste Management XIX (W. Murphy and D. Knecht, eds.), Materials Research Society, Pittsburgh, 1996, p. 189. Back to document

18. Y. Inagaki et al., "Effect of Redox Conditions of Water on Pu and Cm Leaching from Waste Glass," Scientific Basis for Nuclear Waste Management XX (W. Gray and I. Triay, eds.), Materials Research Society, Pittsburgh, 1997, p. 213. Back to document

19. J. Fortner et al., "Solution-Borne Colloids from Drip Tests Using Actinide-Doped and Fully Radioactive Waste Glasses," Scientific Basis for Waste Management XX (W. Gray and I. Triay, eds.), Materials Research Society, Pittsburgh, 1997, p. 165. Back to document

20. W. Lutze, "Silicate Glasses," in Radioactive Waste Forms for the Future (W. Lutze and R. Ewing, eds.), North-Holland, Amsterdam, 1988, p. 65. Back to document

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