Conclusions and Recommendations
In this study we have analyzed the possibilities for (1) using nuclear reactors to process surplus weapons plutonium (WPu) into forms resistant to diversion or theft for reuse in nuclear weapons, (2) using reactors to consume the plutonium altogether, and (3) rendering the plutonium resistant to diversion or theft by immobilizing it in waste forms similar to those contemplated for disposal of high-level wastes originating in nuclear reactors. This work has been motivated by the direct and indirect security dangers that would be posed by long continuation of the status quo in management of the increasing stocks of surplus plutonium from dismantled nuclear weapons, which consists of guarded and monitored storage of plutonium pits (the nuclear-explosive cores of the weapons) at various sites in Russia and at the Pantex site (near Amarillo, Texas) in the United States. The direct dangers of the prolonged storage of the pits are the risks of their diversion (by the possessor states) or theft (by or for other states and subnational groups) for use in nuclear weapons; the indirect dangers are the adverse effects on incentives for further arms control, and for nonproliferation, that will arise if the United States and Russia fail to put additional barriers in the way of reuse of this plutonium in their own arsenals.
In considering the possibilities for using nuclear reactors or immobilization technologies to provide such barriers, we have tried to address, on a comparative basis, technological readiness, institutional requirements, economics, and environment, health, and safety characteristics. We have given by far the greatest weight, however, to the security characteristics of the various possibilities— their capacity to reduce, in a timely way, the direct and indirect security dangers
posed by prolonged storage of the plutonium as pits, while minimizing the new security problems arising from the disposition operations themselves. This chapter summarizes our main conclusions and recommendations. (This is also done more briefly in the Executive Summary at the beginning of the report.)
DISPOSITION OPTIONS AND END-POINTS
Options entailing the incorporation of surplus plutonium into fuel that is then irradiated in nuclear reactors (for brevity, "reactor options") can be subdivided by the type of reactor to be used and by the condition of the plutonium at the conclusion of the disposition operation (end-point). Under reactor types, the categories of possible interest are:
electric-power reactors of currently operating commercial types, using fuels of current designs, or evolutionary adaptations of these reactors and fuels (for brevity, "current-reactor types");
electric-power reactors of more advanced varieties ("advanced-reactor types");
current or future naval propulsion reactors;
current or future research reactors; and
new reactors and/or fuels designed specifically for plutonium disposition, or for plutonium disposition in combination with tritium production.
With respect to the end-points of reactor options for plutonium disposition, there is a continuous spectrum of possibilities in terms of residual plutonium quantity, isotopic composition, and quantity of accompanying fission products, but it is useful to distinguish three general classes of outcomes with somewhat different purposes, as follows:
The "spent fuel" outcome is the result of a once-through fuel cycle in which a moderate fraction of the WPu is destroyed and the remainder is embedded in spent fuel that is similar—in bulk, radioactivity, and isotopic composition of the contained plutonium—to the spent fuel that already exists, in considerably larger quantities, from civilian nuclear-power generation. Net destruction of plutonium in this option, taking into account all plutonium isotopes, can range from slightly negative (in that more plutonium is produced from uranium-238 than is consumed from the initial stock of WPu) to as high as 50 percent in advanced light-water reactors and 80 percent in gas-cooled reactors. The purpose of this approach is to create substantial physical, chemical, and radiological barriers to use of the WPu in nuclear explosives by the original owner of the material or by others. This way of doing so would reduce the WPu management problem, at the end-point, to a modest
part of the reactor spent fuel management task that will exist in any case.
The "spiking" outcome results when briefer irradiation of WPu in a nuclear reactor processes the plutonium more rapidly (but with correspondingly less destruction of the plutonium and smaller changes in the isotopic composition of the remainder) into an irradiated fuel form with enough radioactivity to provide a moderate degree of protection for a few years after discharge. Net destruction of plutonium in this option is small. The purpose of this approach is to achieve a degree of physical, chemical, and radiological protection for the entire stock of surplus WPu more quickly, or with fewer reactors, than in the spent fuel option. It achieves this aim at the cost that the resulting degree of protection is not as high as in the spent fuel approach.
The "elimination" outcome results from the use of fission and transmutation through multiple recycles of plutonium-bearing fuel to convert a very high fraction of the initial surplus WPu into other elements. The purpose of this approach is to eliminate as completely as practicable the possibility of reuse of the WPu in nuclear explosives.
The options that entail immobilizing the WPu in waste forms similar to those contemplated for disposal of fission-product wastes (for brevity, "immobilization options") are somewhat less diverse than the reactor options, but they can vary in the type of waste form, in its dimensions, in the content of fission products, and in weight percent of contained plutonium. A particularly important distinction is between (1) immobilization options that add the plutonium to a waste form heavily laden with fission products, as planned in connection with the stabilization of U.S. defense high-level wastes in borosilicate glass, and (2) immobilization options in which the plutonium is incorporated into a waste form without fission products. The first class of options would serve a similar purpose to that of spent fuel reactor options, that is, to create substantial physical, chemical, and radiological barriers to further use of the WPu in nuclear explosives. (This way of doing so, unlike the reactor spent fuel approach, would not change the isotopic characteristics of the plutonium.) Vitrification options without fission products would offer quicker, easier processing at the cost of lower barriers to reuse in weapons; in that sense these options are analogous to the "spiking" variants of the reactor options.
NARROWING THE RANGE OF OPTIONS
We have concluded that an appropriate goal for WPu disposition operations to be undertaken over the next few decades is to convert the surplus WPu into forms approximately as resistant to diversion or theft for reuse in weapons as is the plutonium in spent fuel from commercial nuclear reactors. Achieving less
resistance than this would mean that the WPu remained a unique security hazard. Achieving more resistance than this would not bring much gain in security until a comparable degree of increased resistance was provided to the commercial spent fuel as well. That is so because the plutonium in commercial spent fuel is, despite some disadvantages of its isotopic composition for nuclear-weapon purposes, nonetheless usable for crude but quite powerful nuclear weapons by unsophisticated bomb-makers and for still more powerful weapons by sophisticated ones. If WPu were harder to obtain than plutonium from commercial spent fuel, then the latter—which exists in considerably larger quantities—would become the dominant security risk.
Acceptance of this "spent fuel standard" for disposition of surplus WPu over the next few decades rules out of consideration, for this purpose, the use of reactor options in the "spiking" mode and the use of immobilization variants that embed the plutonium into waste forms lacking fission products. Because of the low radiological barrier associated with the final plutonium form in these cases, they would not meet the spent fuel standard. We can imagine circumstances in which the "spiking" mode or a no-fission-product immobilization variant might be worth considering as a preliminary step prior to further irradiation in reactors or to later reimmobilization with fission products, in order to gain a modicum of protection quickly if carrying out the campaign to the spent fuel standard was expected to take a very long time; but neither of these approaches is acceptable as a stand-alone approach.
Even as preliminary steps, these low-radiation-barrier approaches have significant liabilities. For example, using reactors in the "spiking" mode increases MOX fuel fabrication capacity requirements—which are likely to be a limiting factor in any case—roughly in proportion to the speed-up in plutonium processing compared to the "spent fuel" mode. In the case of most reactor types, moreover, use of the spiking approach also significantly reduces reactor capacity factors (hence electrical output) because of increased downtime for refueling. (This is not the case for CANDU reactors, which can be refueled while they are operating.) In the case of immobilization without fission products as a preliminary step, it can be questioned whether the temporary barriers added in this way, compared to storage as pits, are sufficient to offset the security risks and economic costs of the extra handling and processing steps involved.
Acceptance of the "spent fuel standard" also effectively removes the "elimination" options from consideration as the primary disposition approach for the decades immediately ahead, although these options deserve continued study for their possible role in reducing the security hazards of all plutonium— military and civilian—in the longer term. The main reasons for this conclusion are (1) the elimination options are less developed technically and more demanding institutionally than many "spent fuel" options, and hence could not be initiated nearly as quickly; and (2) the elimination options, once started, would require a much longer operating time to achieve any reasonable elimination
standard for the surplus WPu, compared to the operating time required by spent fuel options to achieve the spent fuel standard for this material. Choosing an elimination option as the main approach to disposition, then, would be a prescription for great delay in both starting and completing the disposition campaign. We judge the direct and indirect security risks of such delay to be unacceptable. It makes far more sense to use one of the current-reactor/spent-fuel options or immobilization with fission products to bring the surplus WPu relatively quickly to a level of protection comparable to that of plutonium in commercial spent fuel and then consider, in the light of evolving technological capabilities and evolving conceptions of the nuclear-energy future, how the residual security risks of all of the plutonium at the spent fuel standard, military and civilian alike, might be subsequently reduced.
CURRENT-REACTOR OPTIONS FOR MEETING THE SPENT FUEL STANDARD
Commercial reactor types currently operating in the United States offer the technical possibility of transforming U.S. WPu into spent fuel within a few decades:
In mid-1994 the United States had 109 operating light-water reactors (LWRs) totaling 99,500 electrical megawatts (MWe) of generating capacity. Most if not all of these reactors would be capable, without significant modification, of operating with at least one-third mixed-oxide (MOX) fuel. (The remainder of the core would contain ordinary low-enriched uranium [LEU] fuel.) Assuming an initial weight fraction of 4-percent plutonium in heavy metal (uranium plus plutonium) in MOX fuel, just over 7 percent of the U.S. LWR capacity—for example, six 1,200-MWe reactors—would suffice to process 50 tons of WPu into spent fuel in 25 years of operation, assuming one-third MOX cores.
Three of the operating U.S. LWRs—the 1,221-MWe pressurized-water reactors (PWRs) at Palo Verde, Arizona—and one 1,240-MWe PWR that is 75-percent complete in Satsop, Washington, were designed to use 100-percent MOX cores. U.S. reactor manufacturers have indicated that a number of the other operating reactors could also use 100-percent MOX cores without major modification. Assuming favorable safety review of this capability, two 1,200-MWe-class PWRs using 100-percent MOX cores with 4-percent plutonium in heavy metal could process 50 tons of WPu into spent fuel in 25 years of operation.
If the use of 100-percent MOX cores were desired but the existing U.S. reactors suitable for this turned out to be unavailable for the WPu disposition campaign, modifications to one or more of the other operational or under-construction U.S. LWRs would make it possible, at tol-
erable cost, to use these in the 100-percent MOX mode (again assuming favorable safety review). The modifications would entail addition of more control absorbers and corresponding changes to the hardware at the top of the core, and if applied to reactors already operating would require a substantial shutdown period to complete.
If a plutonium loading of 6.8-percent plutonium in heavy metal in a 100-percent MOX core passed safety review, a single 1,250-MWe-class PWR could process 50 tons of WPu into spent fuel in 30 years.
The limiting ingredients on the timing of the current-reactor/spent-fuel approach in the United States would be providing the needed MOX fuel fabrication capacity (no such capacity is currently operational in the United States) and obtaining the necessary approvals and licenses (use of MOX fuel in U.S. power reactors is not now licensed).
The most expeditious solution to the MOX fabrication problem appears to be to bring to operability the partly completed government MOX fabrication facilities at the Fuel and Materials Examination Facility (FMEF) at Hanford, Washington. The panel estimates that the FMEF could be made operational as soon as 2001 at a level of 50 tons heavy metal per year (MTHM/yr) of MOX fuel fabrication for an investment of $180 ± $40 million (1992 dollars), including licensing and interest during construction. Given this start date, the last of the nominal 50 tons of WPu would be transformed into MOX fuel in 2025 and loaded into reactors in 2026. The sum of all incremental costs for disposition of 50 tons of WPu by the current-reactor/spent-fuel option—based on MOX fabrication at 50 MTHM/yr at FMEF and use of 100-percent MOX cores in two PWRs that needed no modifications for this role—are estimated at $450 ± $250 million; if the reactors needed modification, the incremental costs would be $1,500 ± $400 million. (These figures are the present values, as of the start of plutonium disposition operations at the reactors, of the net cost streams—i.e., total costs of electricity generation using WPu-MOX, minus the costs of generating the same electricity in the same reactors using LEU fuel-measured in 1992 dollars and based on a real cost of money of 7 percent per year.)
If the MOX fuel for the same reactors, under the same assumptions about fuel loading and burnup, were fabricated at a newly constructed MOX plant with capacity sufficient for processing 50 tons of WPu in 25 years, transformation of the WPu into MOX could begin as soon as 2003, the last WPu would then be transformed into fuel in 2027, and the last batch of WPu-MOX would be loaded into reactors in 2028. If the reactors used did not require modification for this role the incremental costs (net present value at start of reactor operation, 1992 dollars) would be $900 ± $300 million if the new fuel fabrication plant did
not pay property taxes and insurance and $1,100 ± $300 million if it did; if the reactors required significant modification, these costs would become $1,900 ± $400 million and $2,100 ± $400 million, respectively.
The problem of approvals and licenses for the use of MOX reactor fuel might become easier if the U.S. government chose to purchase and convert to federal facilities, for the purpose of WPu disposition, commercial reactor installations that were either never completed or whose continued operation is becoming economically less attractive for their utility operators. One such option would be the purchase of the mothballed, 75-percent complete Washington Public Power Supply System WNP-3 reactor, which is of the type and size capable of processing all 50 tons of surplus U.S. WPu in 30 years, and/or the purchase of the 63-percent complete WNP-1 reactor, located at the Hanford site, which is of the same size and could be modified during the completion work, if necessary, to permit its operation with 100-percent MOX. Purchase and use of one or both of these reactors in conjunction with MOX fuel fabrication at the Hanford FMEF would have the advantage of confining the handling of unirradiated MOX fuel to a single federal site (if WNP-1 were purchased) or to two federal sites in the same state (if WNP-3 or both reactors were purchased). WPPSS has decided to cease maintaining these reactors and to sell the parts, so timely government action would be required to keep this option open. Given a timely decision to proceed with this option, it should be possible for the WNP reactor or reactors to begin loading MOX fuel in 2002, consistent with the schedule assumed for fuel fabrication at the FMEF. We have estimated the incremental costs of this option (net present value at start of reactor operation, 1992 dollars) at $1,300 ± $1,600 million if one of the reactors is used at 6.8 percent plutonium in heavy metal to load the 50 tons of WPu in 30 years (last WPu-MOX loaded in 2031) and $2,200 ± $3,000 million if both reactors are used at 4.0 percent plutonium in heavy metal to load the 50 tons of WPu in 25 years (last WPu-MOX loaded in 2026).1
If, for some reason, no combination of currently operating and partly completed U.S. LWRs was deemed attractive for the WPu disposition mission, it
If, in the two-reactor case, the reactors are operated for another 5 years on LEU after completing 25 years of WPu-MOX operation, the maximum profit (in our 70-percent confidence range) increases to $0.8 billion and the maximum loss shrinks to $5 billion. As discussed in detail in Chapter 6, the cost uncertainties in the case of reactors that would not otherwise operate are very large, because they must take into account sales of electricity that would not otherwise be produced, and future electricity prices are uncertain enough that at the high end, such reactors might show a net profit, while at the low end, the net cost could run to billions of dollars.
would be possible at the cost of a few years' delay to construct a new dual-purpose (plutonium disposition/electricity generation) reactor or reactors on a government site. The logical choice of reactor type for this function, given adoption of the spent fuel standard and given the desirability of minimizing the delay, would be an evolutionary LWR. Given a timely decision to proceed, fuel loading in such a reactor or reactors could begin as soon as 2005. As examples of this approach, we evaluated cases involving two reactor types: use of a 1,256-MWe ABB-Combustion Engineering System-80+ PWR using 100-percent MOX with 6.8 percent plutonium in heavy metal at an average burnup of 42.2 megawatt-days per kilogram of heavy metal (MWd/kgHM) (loading the 50 tons of WPu in 30 years) and use of two 1,300-MWe General Electric (GE) advanced boiling-water reactors (ABWRs) using 100-percent MOX with 3.0 percent plutonium in heavy metal at an average burnup of 37.1 MWd/kgHM (loading the 50 tons of WPu in 29 years). We estimated costs of the ABB-CE PWR option to range from $1,600 ± $1,800 million if the fuel is fabricated at FMEF and the facilities do not pay property taxes and insurance to $3,200 ± $1,900 million if the fuel is fabricated in a new plant and the reactor and reprocessing plant do pay property taxes and insurance (present values in 1992 dollars as of start of reactor operation, net of electricity sales); the corresponding cost estimates for the GE ABWR option are $2,600 ± $3,600 million and $5,500 ± $3,800 million for the cases with and without property taxes and insurance.2
Heavy-water-moderated reactors in commercial operation in Canada (known as CANDU reactors, where CANDU stands for Canadian deuterium-uranium) appear to be compatible, without physical modification, with the use of 100-percent MOX fuel. (Favorable regulatory review of the safety of such operation would of course be required.) Two typical currently operating CANDUs of 769 MWe each could transform 50 tons of WPu into spent fuel somewhat less radioactive than that from U.S. LWRs in about 24 years of operation. Canada has 20 CANDU reactors totaling about 14 GWe (46 gigawatt-thermal; GWt). As with U.S. LWRs, the pacing elements of a plutonium disposition scheme based on existing CANDU reactors would be provision of the needed MOX fuel fabrication capacity and obtaining the needed permissions and licenses for burning such fuel. Canada has no MOX fuel fabrication capacity; fabricating MOX fuel for CANDUs at the FMEF is technically feasible and would be the most expeditious approach. The fabrication capacity needed for a 24-year campaign in two CANDUs using fuel with 1.2 percent plutonium in heavy metal is 170 MTHM/yr, which is within the capability envisioned for FMEF for this fuel type. Given a timely decision to proceed, the FMEF would
Note that possibilities for negative net costs (i.e., profits) occur only in the cases where no property taxes or insurance are paid and then, as examination of Table 6-16 reveals, only in cases where the value of electricity at the busbar is above 5.5 cents per kilowatt-hour (1992 dollars). The profits made in these cases would be less than those that would be earned by operating the same reactors on LEU (again, see Table 6-16).
be able to begin operation producing CANDU fuel as soon as 2001, and in the two-reactor scenario just described the last of the 50 tons of U.S. WPu would be loaded into the reactors in 2025. The panel estimates the incremental costs of this CANDU disposition option at about $1 billion. Because of their continuous refueling capabilities, CANDUs also offer the possibility of very rapidly "spiking" the entire WPu inventory and then reirradiating the fuel on a longer time scale to bring it to typical spent fuel burnup levels, all without adverse impact on electrical output. This approach would require MOX fuel fabrication capacity substantially larger than could be readily provided at the FMEF facility, however.
Currently operating commercial reactors in Russia and Ukraine also have the technical capacity to implement the "spent fuel" option in a timely way for a nominal 50 tons of surplus WPu from the stockpile of the former Soviet Union. For safety reasons, if this option is selected the only currently operating Soviet-designed reactors that should be used are the 3,130-MWt VVER-1000 LWRs, which are similar to Western PWR designs. Russia has six such reactors in operation (of which, according to one official of the Russian Ministry of Atomic Energy, only four are suitable for MOX fuel) and Ukraine has nine. Depending on the results of a safety analysis of high plutonium loadings in these reactors, and on the acceptability of high plutonium content in the spent fuel, it might be necessary to bring into operation some of the additional VVER-1000 reactors currently standing unfinished in Russia if the goal is to process 50 tons of WPu in 30 years of reactor operations within that country alone.3 The potentially limiting factors governing the timing of plutonium disposition in VVER-1000 reactors thus include this possible need to complete additional plants, as well as the successful upgrading of the safety features of these reactors and the provision of MOX fuel fabrication capacity. No such capacity is currently operating in the former Soviet Union, although a facility with an intended capacity of about 100 MTHM/yr-enough to feed 12 VVER-1000s using one-third MOX cores-stands unfinished at the Chelyabinsk-65 site. The timing of disposition of 50 tons of Russian plutonium in VVER-1000s could, in principle, be similar to that envisioned above for the disposition of 50 tons of U.S. plutonium in LWRs of currently operating commercial types.
The "spent fuel" option could also be implemented for U.S. and former Soviet Union WPu in currently operating Japanese and European LWRs that already use or are licensed to use one-third MOX fuel. (As of 1993, eight LWRs in France, seven in Germany, and two in Switzerland are using MOX fuel, and more are licensed to do so; Belgium and Japan plan to begin loading MOX fuel
in commercial reactors later in the decade.) This approach would be of (at best) limited value if the input of WPu to these reactors merely displaced the use of separated civilian plutonium, which is usable for bomb-making and probably not as well guarded as WPu. If WPu is to be fabricated into MOX in the facilities producing MOX for these reactors, its use should be phased in with, rather than replacing, the consumption of already separated civilian plutonium; either WPu should be fabricated using MOX capacity that would otherwise be idle (such as the Hanau plant in Germany, which might be opened specifically for the purpose of plutonium disposition), or existing reprocessing contracts should be renegotiated to delay separation of additional plutonium until existing stocks of civilian plutonium and WPu are consumed.4 This approach would also require international agreements and safeguards for the considerable international shipment of WPu that would be entailed. The timing of plutonium disposition in this mode could be similar to that envisioned above for the use of U.S. LWRs of currently operating commercial types.
Experimental and prototype liquid-metal fast reactors exist in a few countries and offer some near-term capacity for plutonium disposition with the “spent fuel" option—without reprocessing—although most are not in operation at the present time. The largest such capacity is the French Superphenix reactor (3,000 MWt), which in principle could process 50 tons of WPu into spent fuel in about 20 years. MOX fuel fabrication capability sufficient to support this operation exists in France and Belgium. In practice, however, operating Superphenix in this plutonium-burning mode would require modifications that could take many years to design and implement; and controversy can be expected about whether it should be restarted at all. Reaching agreement on the desirability and terms of the associated international transfer of the plutonium could also prove difficult. Russia and Kazakhstan have two liquid-metal fast reactors totaling about two-thirds the capacity of Superphenix, but they are too old to complete the plutonium disposition mission in their expected lifetimes, and questions have been raised about their safety.
Use of reactors on U.S. naval vessels for disposition of WPu is not practical on the time scales of interest. Such reactors now use high-enriched uranium fuel that is replaced only at very long intervals, if at all. To switch naval reactors to plutonium would first require a long (probably 10-20 year) program to develop and certify plutonium fuels for such use. Because of the extremely high reliability requirements for naval reactors and the uncertainties introduced by such a change, the navy would oppose it. The current rate of loading new fuel into naval reactors is essentially zero, moreover, meaning that the capacity for WPu
disposition in this way is negligible at present; and there are no prospects for a net increase in the number of nuclear-powered naval vessels.
Existing research reactors, similarly, do not offer an attractive option for the disposition of WPu. These reactors are generally small in capacity and in duty factor, they refuel only rarely or not at all, and they are highly dispersed geographically and often located in institutional settings that would be difficult to safeguard. Given the availability of many more attractive possibilities, then, research reactors do not deserve serious consideration for the disposition mission.
ADVANCED REACTORS AND SPECIALTY FUELS
If the "spent fuel standard" is adopted, there is no need to develop and deploy an advanced-reactor type or nonfertile fuel type to achieve that aim. As indicated above, the numbers and characteristics of existing reactor types, using ordinary MOX fuels, are more than adequate to carry out the spent fuel option, and the limitation on the reactor-spiking option is fuel fabrication capacity, not inadequacies in the numbers or characteristics of reactors or the characteristics of fuels. Advanced reactors and nonfertile fuels do not offer sufficient advantages over existing reactor and fuel types, for achieving the spent fuel end-point, to offset the liabilities of longer lag times and additional costs before loading of WPu into reactors could begin. Advanced-reactor types and nonfertile fuels thus are of interest for the WPu mission only if the "elimination" end-point is to be sought in the longer term. (As noted above, such an approach would bring large security gains only if it were applied to reactor plutonium as well as to WPu.)
To illustrate more specifically the timing and cost penalties associated with the use of advanced reactors for the spent fuel mission, we constructed scenarios and cost estimates for using, in a once-through mode, the three advanced-reactor types that would be the least difficult to bring to the point of operation if a decision were made to do so:
As an example of an advanced light-water reactor (ALWR), we took the Westinghouse PDR-600, of which two 610-MWe units using 100-percent MOX cores with 5.5 percent plutonium in heavy metal and average burnup of 40 MWd/MTHM could load the nominal 50 tons of WPu in 34 years. 5 The panel estimates that loading of WPu-MOX fuel
in these reactors could begin, given a timely decision, as early as 2008. We estimated costs of this option to range from $3,100 ± $2,100 million if the fuel is fabricated at FMEF and the facilities do not pay property taxes and insurance to $5,100 ± $1,900 million if the fuel is fabricated in a new plant and the reactor and reprocessing plant do pay property taxes and insurance (present values in 1992 dollars as of start of reactor operation, net of electricity sales).
As an example of a modular high-temperature gas-cooled reactor (MHTGR), we took a General Atomics design employing conventional steam turbines, in which eight 169-MWe units using a nonfertile, particle fuel containing 100-percent plutonium and burning it to an average of 580 MWd/kgHM could load the nominal 50 tons of WPu in 28 years. The net plutonium destruction in this mode would be about 65 percent, about two times higher than the corresponding figure for typical LWR options; at the higher burnups that may be attainable in this fuel type, once-through net plutonium destruction fractions could be as high as 80 percent. The isotopic composition of this plutonium would be less attractive to bomb-makers than that of typical LWR plutonium, but it would still be usable in bombs. In short, the quantities and isotopic quality of the plutonium remaining in spent fuel from the MHTGR would still represent a significant security risk and would require safeguards comparable to those required for spent fuel from LWRs. Even given an early decision to use this option, we do not think the start of plutonium fuel loading into MHTGRs could occur before 2013 (Table 6-2). We estimate the cost of this option at $3,900 ± $2,700 million if the reactor and its fuel fabrication plant do not pay property taxes and insurance and $5,800 ± $3,200 million if they do (present value of net cost stream, after subtracting electricity revenues, in 1992 dollars as of start of reactor operation).
As an example of an advanced liquid-metal reactor (ALMR), we considered a GE design of which four 303-MWe units using fuel containing an average of 10.5-percent plutonium in heavy metal and burning it to an average irradiation of 69.1 MWd/kgHM could load the nominal 50 tons of WPu in 36 years. In this case there would be a net production of plutonium amounting to an increase of a few percent of the amount of plutonium loaded, and the isotopic quality of the discharged plutonium for weapon purposes would be higher than that of typical LWR plutonium. Our estimates of the earliest possible start of fuel loading into the reactors, and of the costs of this option, are essentially identical to our corresponding estimates for the MHTGR.
These findings are consistent with the view, expressed above, that use of the advanced-reactor options in a once-through mode to achieve the spent fuel
end-point would be significantly slower and costlier than accomplishing much the same thing with reactors of currently operating commercial types. We find, further, that other advanced-reactor options—such as the molten-salt reactor, particle-bed reactor, and accelerator-based convertor—would entail even longer development times and greater development investments, with no apparent prospect of gains in performance in the spent fuel mode that could offset the liabilities in timing and cost. Concerning the potential merits of all these advanced reactor types as contributors to national and world electricity supply in the future, we have made no investigation and offer no judgments; that is a question that deserves study in the context of the wide range of considerations appropriate to energy planning. Our conclusion is simply that we have not identified a need to develop and deploy these reactors for the purpose of bringing surplus WPu to the spent fuel standard, and indeed that there would be significant security costs (in the form of delay), as well as monetary costs, in choosing the advanced reactors over currently operating types for this mission.
Concerning the possible use of advanced reactors and/or advanced fuels to pursue the elimination of the WPu, which as noted above might eventually be deemed attractive as a step to follow the transformation to the spent fuel standard and to be applied to plutonium of civilian as well as military origin, we note that:
Plutonium elimination fractions significantly above 80 percent appear attainable only with the help of fuel reprocessing and plutonium recycle.
With suitable reprocessing and recycle capability, both thermal and fast reactors can be used for the "elimination" option. While some actinide isotopes cannot be fissioned in a thermal spectrum, thermal reactors can transmute most of them into isotopes that can. The repeated reprocessing and recycle that would be necessary for the elimination option could raise significant safeguards and security concerns.
ALMRs are superior to current LMR types for the reprocessing-based plutonium elimination mission mainly in that some ALMRs employ integral reprocessing, which does not significantly change the net elimination rate but does offer safeguards advantages over separate reprocessing.
Accelerator-based conversion (ABC) systems have been under study as a means of fissioning actinides and transmuting fission products in order to reduce the longevity of radioactive wastes. Development of this option is only at the early paper-study stage, and both the proposed subcritical-reactor technology and its fuel-cycle technology are extremely challenging and unproven. If the estimated performance could be attained, however, such systems-subcritical reactors with the needed additional neutrons provided by an accelerator-driven spallation
system—could eliminate plutonium at rates (per thermal gigawatt of capacity) comparable to those estimated for the best of the other elimination-oriented options. This approach's continuous online reprocessing would offer some advantages in waste reduction and in safeguards against plutonium theft (but not against diversion by the system's operators)—shared in varying degrees by other advanced systems that use such reprocessing. The availability of this option to receive plutonium is decades away.
Particle-bed and molten-salt reactor concepts have also been proposed as eliminators of plutonium. These concepts are both in the preliminary stages of development, and their performance as plutonium eliminators does not appear better than that of other options that could be brought into operation substantially sooner. Like the ABC concept, the molten-salt reactor does offer some safeguards advantages against plutonium theft by virtue of its online reprocessing, and one version of ABC indeed uses a molten-salt subcritical assembly.
We believe that WPu immobilization by vitrification in borosilicate glass represents a feasible technology that could meet the spent fuel standard, could be available in the relatively near future (within about a decade hence), and could potentially immobilize all of the nominal 50 tons of U.S. excess WPu in glass in a relatively short time once the vitrification campaign had begun (i.e., in a few years, very likely less than 10).
The vitrification option on which we base this view would entail mixing the WPu with high-level radioactive wastes in the course of vitrifying those wastes preparatory to long-term storage or geologic disposal. The glass logs produced by the vitrification scheme would be resistant to theft by virtue of their large size and mass (3 meters long and nearly 2,200 kg for the large-log variant, 0.5 m long and 250 kg for the small-log variant), and their high radioactivity levels; additional barriers to theft eventually would be provided by isolation in a waste repository and, perhaps, intermixing with outwardly similar waste logs that do not contain plutonium; and, even with a plutonium-bearing log or logs in one's possession, substantial chemical processing in a shielded facility would be required in order to extract the plutonium from it.
The earliest possibility for implementing this vitrification option in the United States would seem to be integrating the vitrification of WPu with the currently planned campaign at the Department of Energy's Savannah River plant to vitrify that plant's defense high-level waste (HLW) onsite at its Defense Waste Processing Facility (DWPF), now expected to begin operation in 1996. Before WPu could be added to this scheme, it would be necessary to resolve
questions about the risk of criticality in the melter (a function of plutonium concentration and melter design) and to make such other modifications in the facility as would be required to handle large quantities of WPu there. These tasks probably could not be completed in much less than a decade from now. The Savannah River Site staff has proposed a schedule in which 50 tons of WPu would be vitrified with HLW during the last eight years of the currently planned 20-year DWPF campaign, approximately from 2005 to 2013. Under this schedule, all 50 tons could be emplaced, at about 1.3-percent plutonium by weight, in the 2,200 large logs already scheduled to be produced in the eight-year period. Since this estimate was made, there have been additional delays in the startup of the DWPF, and it remains possible that currently unforeseen major problems in the HLW vitrification campaign there could substantially stretch out this estimated schedule. We estimate the cost of adding the WPu to the DWPF operation at $1,000 ± $500 million (present value in 1992 dollars at the start of plutonium vitrification operations). These costs are comparable to those of the less expensive among the current-reactor/spent-fuel options.
A smaller waste-vitrification facility is under construction in West Valley, New York—the West Valley Demonstration Plant (WVDP)—with startup expected in 1995 or 1996. With a melter about half the size of the DWPF melter, but similar in technology, the WVDP is to vitrify HLW remaining at that site from the previous commercial fuel reprocessing activities there into about 300 logs similar in size and waste content to the 6,100 logs to be produced at Savannah River. While it would be possible in principle to vitrify some of the surplus WPu in this West Valley melter, it clearly could not do the whole job, and there seems little point in paying for the plutonium storage and handling facilities that would be needed at this site to do only a small part of the job.
Groundbreaking for a vitrification facility similar to Savannah River's DWPF is expected sometime this year at the Department of Energy's Hanford site. It is to be used to vitrify the military HLW now stored at that location, which is roughly comparable in quantity to that at Savannah River. By virtue of its later time schedule, the Hanford facility might be more readily and economically modifiable than the DWPF to accommodate WPu in the vitrification process, all the more so if criticality considerations prove to require extensive changes to the current DWPF design.
A waste-vitrification facility with a nominal output of 1 ton of glass per day has been in operation at the Chelyabinsk-65 site in Russia since 1987. The phosphate glass composition employed at this facility appears to be both less durable and less resistant to recriticality if plutonium is embedded in it than is the borosilicate glass planned for U.S. vitrification facilities. To our knowledge, the costs of modifying the Russian facility to make borosilicate instead of phosphate glass, and to integrate WPu with its process stream, have not been estimated, but these requirements seem unlikely to differ greatly from those we have estimated for the modifications that would be needed to U.S. vitrification
facilities. Russian authorities, however, have so far displayed strong resistance to WPu disposition schemes that "throw away" the plutonium without generating any electricity from it, irrespective of arguments that electricity generation with WPu is costlier than with LEU.
Waste-vitrification plants are operating or soon to operate in France, Great Britain, and Japan, but their use for disposition of U.S. or former Soviet Union WPu would be impeded by the logistical and political problems of international transport of this material, as well as by the cost and difficulty of add-on modifications to integrate WPu with their glass-production process streams.
Certification of plutonium- and radioactive-waste-bearing borosilicate glass as suitable for ultimate disposition in a geologic waste repository will depend on resolution of issues involving the properties of the glass itself (physical integrity, leachability) and the potential recriticality of the plutonium at distant future times if other glass constituents such as boron were to leach out preferentially. A substantial development effort, requiring a decade or more, might be necessary to decide these questions fully, although important insight into them probably can be developed in a shorter time. Depending on the results of such studies, it is possible that plutonium loadings might be constrained to percentages low enough to substantially affect the timing and cost of a vitrification campaign, or that alternative glass compositions might have to be developed.
A number of waste forms other than borosilicate glass have been proposed by various groups for consideration as alternatives for the immobilization of surplus WPu in ways that would, it is argued, meet the spent fuel standard. Besides the phosphate glass mentioned above, these candidates include synthetic rock ("synroc"), cements, and pyroprocessed metals. We do not favor further consideration of phosphate glass, synroc, or cements for the WPu disposition mission in the United States, unless unforeseen developments require it, because we believe this country has chosen borosilicate glass for the waste-immobilization function for sound (and extensively documented) reasons, the suitability of this material for containing significant quantities of plutonium in addition to HLW has already been studied rather extensively with favorable results, and a high cost in delay—hence in security as well—would be paid for reopening the question.
The pyroprocessing approach, which would use technology developed as part of the U.S. Integral Fast Reactor program, would require substantial additional engineering development and construction of major new facilities (including what would amount to a sizable LWR fuel reprocessing plant to provide feed material), and it would produce a waste form that has not been characterized at all for long-term disposition and would probably be unsuitable for emplacement in Yucca Mountain. All this strikes our panel as a prescription for long delays and big investments in pursuit of highly uncertain prospects for solving a problem for which satisfactory approaches—the current-reactor/spent-fuel and borosilicate-glass/vitrification options—are much closer to hand.
Yet another approach that has been proposed would combine the surplus WPu with beryllium so as to generate substantial neutron radiation from alpha-n reactions followed by neutron multiplication in the plutonium. This approach would yield a material with much higher plutonium concentration and lower radiation barrier than either standard spent fuel or borosilicate glass logs containing HLW, hence a much more attractive target for diversion or theft. It would also pose much greater criticality problems than the other final plutonium forms considered here, and would be unlikely to be certifiable for long-term storage in a geologic repository. Although its production might be less costly than that of WPu-bearing spent fuel or borosilicate glass logs, it does not come close to meeting the spent fuel standard and does not, in our view, deserve further consideration.
COMPARISON OF THE CURRENT-REACTOR/SPENT-FUEL AND VITRIFICATION OPTIONS
It is clear that the current-reactor/spent-fuel option and the vitrification-with-wastes option are the most attractive candidates for reducing expeditiously the security risks of surplus WPu. Our comparative evaluation of these two leading options (drawing heavily on Chapter 6 section "The Comparisons in Summary") is discussed in the following paragraphs.
In terms of the crucial timing aspect of security, the current-reactor/spent-fuel options and the vitrification-with-wastes option are roughly comparable to each other (as well as superior to all other options). Under the most optimistic assumptions that are defensible, loading of WPu into current-reactor types could begin between 2002 and 2004 and be completed between 2015 and 2025; loading of WPu into waste-bearing glass logs could begin around 2005 and be completed as early as 2013. The timing uncertainties in both cases—relating more to resolution of institutional issues in the reactor case and to resolution of technical issues in the vitrification case—are bigger than the differences in the best-case point estimates we have provided; thus it would not be meaningful to say more than that the two sets of options are comparable.
With respect to those aspects of security that depend on the details of handling, processing, and transporting various plutonium forms, vitrification-with-wastes entails fewer and somewhat simpler steps than the current-reactor/spent-fuel options, and hence may be somewhat easier to safeguard.
With respect to security of the final plutonium forms, the current-reactor options obviously meet the spent fuel standard, and we judge that the vitrification-with-wastes option meets this standard also. The plutonium in the spent fuel assembly would be of lower isotopic quality for weapon purposes than the still weapons-grade plutonium in the glass log, but since nuclear weapons could be made even with the spent fuel plutonium this difference is not decisive. Under typical assumptions, the radiological barrier presented by glass logs would be
about three times smaller than that presented by a fuel assembly (but still very high), and the mass of a glass log—containing, coincidentally, about the same amount of plutonium as a fuel assembly—would be about three times greater.6 The difficulty of separating the plutonium from the accompanying materials would be comparable in the two cases.
In terms of security, then, we believe the two options are comparable in timing; the vitrification-with-wastes option has, perhaps, a modest advantage in safeguardability of the handling, processing, and transport steps; and the current-reactor/spent-fuel option has a modest advantage in the barriers associated with the final plutonium form because of the difference in plutonium isotopics. We conclude that the two approaches are comparable in security overall, that either would be adequate, and that no other option known to us is better.
With respect to economics, our estimates indicate that the most likely costs for the vitrification-with-wastes option and for the less costly among the current-reactor/spent-fuel options are comparable at $0.5-$2 billion, expressed as the present value, in 1992 dollars as of the start of plutonium operation at the reactor or melter, of the stream of incremental costs associated with plutonium disposition by these means (subtracting electricity revenues where appropriate). The range of best estimates for all of the current-reactor/spent-fuel options extends from $0.5 billion to about $5 billion. The lowest central estimate, at about $0.5 billion, is for the MOX/spent fuel option using currently operating U.S. LWRs that need no modification to use MOX safely, with the fuel fabricated at the FMEF at the Hanford site. Four of the options studied have central-estimate costs around $1 billion: use of MOX fuel from FMEF in currently operating CANDU reactors in Canada; use of MOX from FMEF in a single, currently mothballed, partly completed PWR that would be completed for this purpose; use of MOX from an entirely new fuel fabrication plant in currently operating U.S. LWRs that need no modification to use MOX safely; and vitrification with defense HLW at the Savannah River site.
Although the central estimates in all cases considered correspond to net costs, our judgmental 70-percent confidence intervals include a possibility of profits from WPu disposition for the case in which currently mothballed, partly completed PWRs are completed for the purpose of plutonium disposition and use MOX fuel from FMEF, and for cases when new, evolutionary LWRs are
built for this purpose as government facilities (paying no property taxes or insurance) and use MOX from FMEF. These profit possibilities depend not only on the costs associated with MOX use falling at the low end of our judgmental 70-percent confidence ranges, but also on the additional electricity generated by the plutonium disposition reactors being marketable at a price of 5.5 cents per kilowatt-hour or higher (1992 dollars) at the busbar. (Given such prices and the same financial assumptions, however, a higher profit would be available from a program that completed reactors or built new ones from scratch and used them to generate electricity using LEU rather than MOX fuel, without addressing the problem of plutonium disposition.)
The range of $0.5-$5 billion (1992 dollars)—covering the best estimates of net present value, at reactor or melter startup, of most of the options considered—corresponds to $10,000 to $100,000 per kilogram of WPu, or $40,000 to $600,000 for a nominal "bomb's worth" of 4-6 kg. Even the higher figure is probably less than what this weapon material once cost to produce, as well as much less than would be spent in the attempt to recover such material if it went astray and incomparably less than would be spent to try to deter or otherwise prevent its use in the form of a bomb in the hands of a potential adversary.
With respect to environment, safety, and health (ES&H), the panel believes that options for plutonium disposition should:
comply with existing regulations, of the country in which disposition takes place, governing radioactivity and radiation from civilian nuclear-energy activities;
comply with existing international agreements and standards on the disposition of radioactive materials in the environment; and
not add significantly to the ES&H burdens that would result, in the absence of programs for disposition of WPu, from appropriate management of civilian nuclear energy generation and of the environmental legacy of past nuclear weapons production.
The panel believes that both the current-reactor/spent-fuel and vitrification-with-wastes approaches, suitably designed, can meet these criteria. Most of the ES&H impacts of WPu disposition using either of these options can be expected to represent modest additions, at most, to the routine exposures to radiation and risks of accident associated with other civilian and military nuclear-energy activities underway in the United States, and that there is no apparent reason that the activities involved in WPu disposition using either of these approaches should not be able to comply with all applicable U.S. ES&H regulations and standards.
While there are differences in detail in the ES&H challenges and risks posed by the two options in some of the activity categories—for example, a somewhat more complicated set of plutonium-handling operations for the reactor options than for the vitrification option, and a greater relative increase in
plutonium content of the final waste form for the vitrification option than for the reactor options—these differences do not consistently favor one class of options or the other, and none is large enough in relative or absolute terms to justify choosing one class of options over the other. ES&H issues that will need further attention in the next phase of study of these options include: developing and testing the systems to ensure adequate safety against criticality accidents in the melter for the vitrification option; confirming the conditions under which full-MOX cores can be used without adverse impacts on safety in reactors of currently operating commercial types; and determining the conditions that will provide adequate assurance against long-term criticality in geologic repositories containing either spent fuel or glass logs from plutonium disposition operations. These issues differ considerably in the nature and complexity of the work that will be needed to settle them to the satisfaction of the technical and regulatory communities, but it is the panel's judgment that suitable approaches exist for all of them.
Both from the standpoint of ES&H and from the standpoint of security, analysis and comparisons are facilitated by the circumstance that both of the leading-candidate classes of options would be adding WPu to a set of nuclear activities that would be going on in any case, and both would leave the residual WPu in a waste form—spent fuel in one case and HLW-bearing borosilicate glass logs in the other—which will exist in large quantities and will need to be safely managed whether used for WPu disposition or not. We emphasize, in this connection, that a U.S. geologic repository is not likely to be ready to receive wastes of any kind before 2015. Vitrified waste logs, with or without plutonium, will need to be stored in engineered facilities until a geologic repository is ready to receive them; and plutonium-containing spent fuel from nuclear-reactor operations, whether WPu has been incorporated in some of it or not, will need to be stored at reactor sites or at other commercial spent fuel storage facilities until a repository is ready.
Two options have emerged from our investigations as the most promising ones for the disposition of surplus WPu: the use of LWRs or CANDU reactors, employing MOX fuel in a once-through mode, to embed the WPu in spent fuel that would be emplaced eventually in a geologic repository (together with the larger quantity of such spent fuel that will exist in any case from ordinary nuclear electricity generation); and vitrifying the WPu, together with defense high-level radioactive wastes, in borosilicate glass logs of the type planned for use in the immobilization of defense HLW in any case, again for eventual emplacement in a geologic repository.
These options were selected on the basis of evaluation criteria spelled out in the 1994 report of the parent committee and in Chapter 3 of this report, above
all the importance of being able to proceed expeditiously to reduce what the parent committee characterized as “the clear and present danger" posed to national and international security by the existence of large quantities of surplus WPu in the form in which it emerges from the weapon dismantling process. The superiority of these two options to the many others in the panel's purview has been established not only by our own analyses and those of the parent committee (which drew on our preliminary findings), but also by evaluations conducted independently by the U.S. Department of Energy and by a number of other groups.
Although both of these options are technically feasible and have been recommended by the panel in part because they can be deployed comparatively rapidly, some significant uncertainties accompany both of them. The areas of uncertainty in the current-reactor/spent-fuel option are primarily in licensing and public acceptance. Those in the borosilicate-glass/vitrification option relate mainly to the technical issues of plutonium-HLW mixing and criticality. Further paper studies will not significantly reduce these uncertainties, but the experience gained in the implementation process would.
Since it is crucial that at least one of these options succeed, since time is of the essence, and since the costs of pursuing both in parallel are modest in relation to the security stakes, we recommend that project-oriented activities be initiated on both options, in parallel, at once. DOE should assign sufficient resources (both funding and personnel) to manage pursuit of both options in parallel. In connection with the current-reactor/spent-fuel option, work should be started to seek out specific reactors and MOX fabrication options that would minimize multiple plutonium transportation steps so as to reduce this aspect of security risk, to identify locations that are most amenable to public acceptance, and to ascertain the willingness and conditions of the plant owners to participate. The detailed engineering studies should be completed and licensing applications submitted to the Nuclear Regulatory Commission. From that base, the sound cost estimates, schedules, and financial plans could be prepared that are essential to considering full project authorization. In connection with the borosilicate-glass/vitrification option, laboratory work and realistic testing should be started to address the technical uncertainties. Research and development plans, program schedules, and key milestones should be defined. As the results of the research and development are obtained, detailed cost estimates and ES&H analyses can be submitted to the Department of Energy and the Nuclear Defense Facilities Safety Board in pursuit of project authorization.
The pursuit of the two options in parallel, as project-oriented tasks with near-term milestones and aggressive schedules, should be aimed at bringing both processes on-line by the end of the century or as shortly thereafter as possible. It would be a mistake to spend tens of millions of dollars and additional years of paper studies to try to demonstrate, in the absence of actual work toward deployment, which of the two options should be selected over the other. It
will likely be less expensive in the long run, and clearly superior from the security and ES&H perspectives, to proceed with both now. If either option falters due to technical, licensing, or other difficulties, the pursuit of the other option can continue without the loss of time that would have been associated with choosing early and choosing wrong. Indeed, it may prove to be desirable to implement both options for different parts of the stockpile: as one example, it may turn out that reactor approaches are preferable for the relatively pure plutonium in the components of dismantled weapons, while vitrification may prove to be the better alternative for the tons of plutonium that exist in scrap, solutions, and other forms.
We recommend, therefore, that the Department of Energy's Programmatic Environmental Impact Statement process (scheduled to be completed in 1996) should be oriented toward a decision to pursue both the current-reactor/spent-fuel option and the vitrification-with-wastes option, not toward a decision to eliminate one or the other. Subsequent preparation of Environmental Impact Statements (EISs) for both options, and the participation of the public in these processes, should proceed in parallel too. The inclusion, in the EIS and public participation processes, of the results of the specific project-oriented activities mentioned above will be essential to the success of those processes. Full project authorization for one of the options would not be granted until the EIS is completed and approved.
The fundamental objective of the WPu disposition program will not be achieved unless the Russians carry out a disposition program in parallel, on a similar time scale, and adhering to disposition standards equivalent to those of the United States. The above project-oriented activities would lend themselves to forming joint projects with the Russians to assure such a parallel approach. Joint projects will also serve to develop a technical consensus on the disposition process and standards, which, as pointed out elsewhere in our report, does not exist today. The panel recommends that the United States immediately initiate joint project-oriented activities with Russia covering both the MOX and the vitrification options.
Studies of a follow-on nature should continue on the longer-range questions of whether and how the residual security risks of weapon and other plutonium should eventually be reduced beyond the spent fuel standard. It is essential, however, that such longer-range studies not be allowed to draw resources or attention from the pursuit of the two options closest to hand for moving rapidly toward achieving that standard for the weapons plutonium that poses a "clear and present danger" today.