4
Reactor Options

This chapter describes each of the reactor options examined by the panel. For each option, we offer:

  1. a description of the technology and its development status, including how the technology would be used to process plutonium;

  2. a description of the factors affecting timing of the option, including:

    1. technical uncertainty,

    2. reactor capacity and throughputs,

    3. fuel fabrication, and

    4. licensing and public acceptance issues;

  1. a description of issues related to safeguards and security during the processes required for the option;

  2. a description of issues related to accessibility for use in weapons of whatever plutonium would remain in the spent fuel;

  3. a discussion of economic issues;

  4. a discussion of environment, safety, and health (ES&H) issues; and

  5. a discussion of other issues that may be important to policy-makers in choosing between options.

We begin by discussing several options involving the use of current-generation reactors, then turn to more advanced reactor systems. The description of the first option—the use of light-water reactors of existing designs—will be the most detailed, both because it is an option of particular interest and because



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Management and Disposition of Excess Weapons Plutonium: Reactor-Related Options 4 Reactor Options This chapter describes each of the reactor options examined by the panel. For each option, we offer: a description of the technology and its development status, including how the technology would be used to process plutonium; a description of the factors affecting timing of the option, including: technical uncertainty, reactor capacity and throughputs, fuel fabrication, and licensing and public acceptance issues; a description of issues related to safeguards and security during the processes required for the option; a description of issues related to accessibility for use in weapons of whatever plutonium would remain in the spent fuel; a discussion of economic issues; a discussion of environment, safety, and health (ES&H) issues; and a discussion of other issues that may be important to policy-makers in choosing between options. We begin by discussing several options involving the use of current-generation reactors, then turn to more advanced reactor systems. The description of the first option—the use of light-water reactors of existing designs—will be the most detailed, both because it is an option of particular interest and because

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Management and Disposition of Excess Weapons Plutonium: Reactor-Related Options other reactor options can be considered in part by comparison to that basecase. More detailed comparative assessments of the options based on criteria related to security, cost, and ES&H can be found in Chapter 6. U.S. PLUTONIUM IN CURRENT-GENERATION U.S. LIGHT-WATER REACTORS Description of Technology and Status Light-water reactors (LWRs) are the most mature of any of the proposed burners of weapons plutonium (WPu). Over 100 LWRs are operating in the United States and about 400 worldwide. LWRs have over 4,000 reactor-years of operation. They supply almost 75 percent of the electricity consumed in France and about 22 percent of the electricity generated in the United States. U.S. LWRs use low-enriched uranium (LEU) fuel. Plutonium in the discharged fuel is not reprocessed for recycle. In Sweden and the United States, geologic repositories are being designed primarily for long-term disposal of spent fuel discharged from uranium-fueled LWRs. Other countries are considering such a direct-disposal fuel cycle as well. It is also possible to fuel LWRs with mixed-oxide (MOX) fuel, which combines plutonium dioxide and natural or depleted uranium dioxide as a PuO2-UO2 mixture. Work on such MOX fuels for LWRs has a long history. The U.S. Plutonium Utilization Program began in 1956, and was soon followed by related work in several European nations and Japan. The development effort was motivated in part by the potential for fuel-cycle economies perceived at the time and also as a means to reduce the consumption of uranium ore. It focused on the reprocessing of LWR discharge fuel to recover and recycle the plutonium and uranium. Several tests of partial core loadings of MOX fuel were conducted in U.S. LWRs during the 1960s and 1970s. Although plutonium recovered from LWR fuel was used in these tests, the results are generally applicable to MOX made from WPu.1 In 1963 Belgium used a partial loading of MOX fuel in its BR-3 pressurized-water reactor (PWR). After many years of experimentation, by 1986 Belgium irradiated a core with a 70-percent MOX loading. Belgium provides a significant fraction of the world's currently operating MOX fabrication capability and is considering an expansion of its MOX fabrication plant. Its MOX fabrication services are marketed in conjunction with France by the MELOX consortium. Two Belgian LWRs are licensed to burn MOX fuel. Germany tested and demonstrated MOX fuel in LWRs from 1968-1977 and began commercial use of MOX fuel in LWRs in 1981. Seven reactors in 1   See Chapter 2 for a discussion of the differences between WPu and reactor plutonium (RPu). Details on the U.S. Plutonium Utilization Program are found in USNRC (1976).

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Management and Disposition of Excess Weapons Plutonium: Reactor-Related Options Germany are using MOX, five others have a license to do so, and six more have submitted license applications (Thomas 1994, Wilcox 1994). A modest-scale MOX fuel fabrication plant at Hanau operated for several years before losing its license, and a larger plant was nearly completed before encountering licensing difficulties that have so far (late 1994) prevented it from operating. France began fueling PWRs with MOX fuel on a commercial basis in 1985, and is building a substantial MOX fabrication facility. Sixteen French reactors are licensed to use MOX, and seven of them were doing so as of late 1994 (Nigon and Golinelli 1994). Japan is considering use of MOX fuel in roughly 10 LWRs by approximately the year 2000 (Yamano 1994) and is constructing a substantial MOX fabrication facility and a large commercial reprocessing plant for recycle of plutonium as MOX fuel. Britain, while not having a domestic MOX use program, is building a large MOX plant for foreign customers to complement the reprocessing services it already offers.2 Government-funded work on commercial fuel reprocessing and plutonium recycle in the United States was terminated by presidential directive in the mid-1970s, reflecting concerns that worldwide commercial plutonium recycle might stimulate the proliferation of nuclear weapons, and was also uneconomical (see Carter 1977). While the Reagan and Bush administrations did not take a similarly negative view, U.S. industry has concluded that the high cost of a commercially owned U.S. reprocessing plant makes near-term deployment of reprocessing and plutonium recycle uneconomical in the United States. For that reason, there are now no commercial facilities in the U.S. for reprocessing or for MOX fuel fabrication, and there are no commercial reactors in the United States licensed to use MOX fuel. From the considerable world experience there is a mature technology adequate to implement the use of WPu as MOX fuel in LWRs. The simplest concept is once-through irradiation of MOX fuel to burnups similar to those used with uranium fuel in LWRs. No reprocessing would be required, and the rate-limiting step would probably be MOX fuel fabrication. Even a small subset of U.S. commercial LWRs would suffice to absorb the nominal 50 tons of excess WPu in this way at the WPu-MOX fuel fabrication rates likely to be attainable. The following sections present specifics on the technical issues involved in using MOX fuel in LWRs and the rates at which various LWR configurations could process the WPu. Technical Issues in LWR MOX Fueling Early plans for MOX fueling in LWRs, in the mid-1960s to mid-1970s, were to reprocess all fuel discharged from a given reactor and to recycle all of 2   As of late 1994, the target opening date for this plant was 1997 (Wilcox 1994). For more information on civilian plutonium fuel programs, see NAS (1994, Appendix B) and Albright et al. 1993.

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Management and Disposition of Excess Weapons Plutonium: Reactor-Related Options the recovered plutonium as MOX fuel for subsequent reloads for that reactor. This was known as self-generated recycle. The neutronic characteristics of LWRs are such that the plutonium produced would be sufficient to fuel about a third of the reactor core when steady-state recycle was reached. The remainder of the core would continue to be fueled with LEU. The one-third core loading was also thought at that time to be a practical limit for most of the LWRs that had been designed on the basis of uranium fueling. There are several reasons why the fraction of a reactor core that can utilize plutonium fuel without compromising safety margins may be limited, unless the reactor has been designed or modified specifically for plutonium fuel use: the lower delayed-neutron fraction in plutonium fuel as compared with uranium fuel, the higher average neutron energy in plutonium fuel, and the 0.3-electron volt neutron-absorption resonance of plutonium-239 combine to put extra demands on a reactor's reactivity-control systems; the more energetic neutron spectrum and higher gamma-ray flux in plutonium fuels increase radiation damage and thermal stresses in reactor internals; and the higher radioactive-decay energy of plutonium fuels puts greater demands on post-shutdown (including emergency) core-cooling capabilities (see Chapter 2, “Some Differences Between Plutonium- and Uranium-Based Fuels"). For these reasons, and because self-generated recycle could only provide enough plutonium for one-third of the cores of the reactors involved, the designs of most present LWR cores were not intended to offer the capability to burn plutonium fuel in more than one-third of the core. For disposition of excess separated plutonium, however, using MOX in all of the reactor core rather than only one-third would be highly desirable, as it would reduce the number of reactors or the time required by a factor of three. This could substantially decrease the amount of transportation and the number of sites required to process the excess plutonium, reducing risks of theft during transport and reducing the political and licensing liabilities of involving more of the reactor industry in the plutonium-processing operation. There are also some technical incentives to move in the direction of full-MOX cores. Because the MOX fuel rods have a higher fission cross-section than do the all-uranium rods, the two different kinds of fuel rods must be carefully distributed within the reactor core to avoid local overheating. Also, the reactivity of MOX fuel tends to change more rapidly during an irradiation cycle than that of uranium fuel. There is greater spatial self-shielding of neutron flux in fuel assemblies composed of MOX fuel rods. Design problems are aggravated by the presence of both uranium and MOX fuel rods in the same core. Thus, a reactor core fueled entirely with MOX fuel would be easier to design and to program for fuel reloads. During the 1970s the U.S. nuclear industry did envision an alternative to self-generated recycle of plutonium with one-third MOX cores. Some utilities planned to dedicate some of their LWRs to operate as "plutonium burners," to be fueled entirely with MOX fuel. An all-MOX plutonium burner would receive

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Management and Disposition of Excess Weapons Plutonium: Reactor-Related Options make-up plutonium produced in other uranium-fueled LWRs. The all-MOX plutonium burner would be designed with the extra control absorbers needed for reactivity control, and other modifications would be needed to compensate for the increased fast-neutron and gamma fluxes, the increased decay heat, and the smaller delayed-neutron fraction. Such a plutonium-burner PWR was designed by Combustion Engineering (Shapiro et al. 1977, Anderson and Klinetob 1981). Called the System-80, this reactor was designed for flexibility to use large plutonium loadings, up to and including a full loading of MOX fuel. As compared to a PWR designed only for uranium fueling, the System-80 plutonium burner has additional control rods and drives and increased boron concentration in the reactor coolant. The cooling systems for the reactor and spent fuel storage are sized for higher long-term decay heat. Core internals are designed to improve cooling and reduce thermal stress that would otherwise result from increased gamma-ray heating. The thickness of the core support barrel is increased to mitigate the increased fast-neutron flux. Three System-80 reactors, each generating 1,300 megawatt electric (MWe) (3,750 megawatts-thermal; MWt), are in operation in the United States at the Palo Verde Nuclear Generating Station in Arizona. Four 1,000-MWe units are being constructed in Korea, including a forerunner of the evolutionary System-80+ design. One of the two uncompleted reactors of the Washington Public Power Supply System (WPPSS), designated WNP-3 (Washington Nuclear Project), is a System-80 reactor. This reactor is 75 percent complete and has been kept in mothball status. In addition, recent studies by the major U.S. reactor vendors funded by the U.S. Department of Energy (DOE) have concluded that contrary to past expectation, many (though not all) existing LWRs (both PWRs and boiling-water reactors) could use WPu MOX in 100 percent of their reactor cores with little if any modification.3 Existing margin in the control capabilities of these reactors would, it is reported, allow them to operate within existing safety envelopes with 100-percent MOX cores. In a number of the existing-reactor cases, however, the maximum safe enrichment of plutonium in the fuel would be lower than it would be in new reactors designed specifically for plutonium use, potentially increasing total fuel fabrication costs and the time required to carry out the plutonium disposition mission. It appears likely that existing reactors could handle higher plutonium loadings if modifications were made conceptually similar to those previously believed necessary to achieve full-core operation at all. This issue—including 3   ABB-CE (1994), GE (1994), and Westinghouse (1994). Indeed, it appears that full-core MOX loading is more easily achievable with WPu than it would be with RPu. The higher concentrations of higher isotopes of plutonium in RPu cause additional neutron absorption, further reducing control-rod worth compared to the WPu case (see ABB-CE 1994, Section 5).

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Management and Disposition of Excess Weapons Plutonium: Reactor-Related Options the trade-offs between the costs and time required to modify reactors and the extra costs and longer times associated with using fuel of only modest enrichment—has not yet been studied in detail but should be. The importance of these recent studies is considerable, as they suggest that the job of plutonium disposition in existing U.S. LWRs—one of the main options recommended in this report and the report of the parent committee—would be significantly easier than was believed at the time the parent committee's 1994 report was written. It should be noted, however, that these analyses became available late in the panel's deliberations, and the panel has not been able to review them in detail. The panel urges that they be reviewed by the U.S. Nuclear Regulatory Commission (NRC) to ensure that the conclusion that existing LWRs could use full cores of WPu MOX without compromising existing safety margins is correct. Whatever the conclusion as to whether significant modifications would be necessary, it appears clear that existing U.S. LWRs could be adapted for full-MOX cores. Because of the considerable worldwide experience with MOX fueling in LWRs, the panel judges the technical uncertainty of the LWR MOX option—using either one-third or full-MOX cores—to be low (by comparison to other reactor options for use of plutonium fuels). Reactor Throughput, Once-Through Fuel Cycle The rate at which plutonium could be processed in a once-through MOX fuel cycle is given by the product of thermal power, capacity factor, fraction of the core that is MOX fuel, and plutonium weight fraction in fresh MOX fuel, divided by the average fuel exposure at discharge. Consider, for example, the reactor characteristics once contemplated for self-generated plutonium recycle (Shapiro et al. 1977, Hebel et al. 1978): assuming a 1,000-MWe PWR at 34.2-percent thermal efficiency, a capacity factor of 0.70, a one-third core of MOX fuel, a fuel exposure of 30,400 megawatt-days per metric ton of heavy metal (MWd/MTHM), and an initial concentration of 2.5 weight percent plutonium (Anderson and Klinetob 1981), the yearly amount of plutonium supplied as make-up fuel would be 205 kilograms (kg). Assuming a once-through fuel cycle, 50 tons of WPu could be processed by 8.1 such reactors given a nominal operating lifetime of 30 years. As noted, using MOX in 100 percent of the reactor core, rather than only one-third, would reduce by threefold the number of reactor-years required to irradiate a given initial quantity of excess WPu. Full-MOX reactors comparable to the System-80 (that is, with a capacity somewhat larger than the 1,000 MWe assumed above), operated at 75-percent capacity factor with a 100-percent MOX core, with an initial plutonium content of 2.5 percent by weight and average burnup of 31,000 MWd/MTHM, would process 828 kg of WPu per reac-

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Management and Disposition of Excess Weapons Plutonium: Reactor-Related Options tor-year, thus processing 50 tons in 60 reactor-years (for example, two reactors operating for 30 years) (Hebel et al. 1978). The initial concentration of plutonium in the MOX reloads can be increased beyond 2.5 weight percent, particularly by adding more burble poisons. ABB-Combustion Engineering (ABB-CE), for example, has suggested that in their System-80+ system, a loading of 6.8 weight percent WPu, compensated by the addition of erbia to the fuel (which also has the advantage of smoothing reactivity over the fuel's life in the reactor), would be attainable. In the existing System-80 reactors, however, which have somewhat fewer control rods, ABB-CE estimates a maximum loading of 4.5 percent (ABB-CE 1994). Estimates of the average enrichment of WPu that could be used safely in full-MOX cores in existing LWRs other than the System-80s vary. Westinghouse concludes that many existing PWRs could use full-MOX cores with an average enrichment of 4.5 percent, comparable to the enrichment possible in the System-80s. ABB-CE and General Electric conclude that their non-System-80 reactors could handle full-MOX cores with enrichments in the range of 3 percent (ABB-CE 1994, GE 1994, Westinghouse 1994). While no substantial reactor modifications are estimated to be required, the various vendors do suggest measures such as including burble absorbers in the fuel, dissolving increased quantities of boron in the cooling water, or changing the material used in the reactor control rods. What do these figures mean for how fast the plutonium disposition mission could be accomplished? A 1,200-MWe PWR using a full-MOX core with a plutonium loading of 4.0 percent, and a burnup of 42,000 MWd/MTHM would use just under a ton of WPu each year, requiring roughly 50 reactor-years for disposition of the nominal 50 tons of excess material. If a System-80 reactor with modifications could operate with an initial plutonium content of 6.8 percent, and the same average burnup, this reactor would process about 1,700 kg of WPu per reactor-year. so that one such reactor operating for 30 years would suffice to process 50 tons (see Table 6-1). As can be seen, the use of high plutonium loadings significantly increases the rate at which the WPu can be processed in a given reactor, even when the burnup is also increased. As described in detail in Chapter 6, moreover, use of such high enrichments reduces the net cost of the operation by reducing the number of kilograms of MOX fuel that need to be produced and increasing the energy value of each kilogram of that fuel. The use of such high plutonium loadings, however, does create some technical issues, similar in some respects to those involved in moving from one-third to full-MOX cores. In general, burble absorbers must be added to hold down the reactivity. These absorbers (erbium oxide, in the case of the ABB-CE proposal) also help preserve negative temperature coefficients of reactivity for the moderator and fuel. Because of the resonance in the neutron absorption cross-section of erbium-167 at energies just above the thermal resonance of plutonium-239

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Management and Disposition of Excess Weapons Plutonium: Reactor-Related Options (Pu-239), erbium can counteract the tendency of Pu-239 to contribute to positive reactivity coefficients. Other materials such as gadolinium, dysprosium, or boron may be equally effective in this role, depending on the specific application. Because the effectiveness of the burble absorbers decreases with increasing burnup, as the reactivity of the fuel declines, they contribute to smoothing out reactivity changes during the fuel cycle. Higher plutonium-loading MOX fuel will also have a higher fissile plutonium content when removed from the reactor than would ordinary LEU spent fuel. This fact will require careful attention to long-term criticality issues in preparing such fuel for geologic disposal. (See discussion below and in Chapter 6.) To summarize, the existing System-80 reactors were designed from the outset to handle full cores of plutonium fuel. Analyses by the vendors indicate that many other existing reactors could also use full cores of WPu fuel without substantial modifications to the reactors. Further analysis and review would be required to assess this conclusion. NRC review of any proposal to use plutonium fuels in U.S. reactors is likely to be intensive, and NRC review could lead to a requirement for reactor modifications. But even if significant modifications do turn out to be necessary, the panel believes that a variety of U.S. LWRs could be adapted to handle full-MOX cores safely, with sufficient enrichments to carry out the mission in a small number of reactors. Fuel Fabrication Providing adequate plutonium processing and MOX fuel fabrication capability would be an important pacing factor for processing excess WPu in U.S. LWRs. Plutonium pits would have to be shipped from Pantex (located near Amarillo, Texas), where no plutonium processing capability yet exists, to a site capable of disassembling the pits and converting the resulting metal to plutonium oxide. No facilities for carrying out the pit-processing operation on the required scale are currently operating, but facilities at Savannah River (South Carolina), Los Alamos (New Mexico), Hanford (Washington), and possibly elsewhere could be modified for this purpose—and new technologies for efficient pit conversion are being developed at the national laboratories. Using a planning figure of 4 kg of plutonium per pit (NAS 1994), processing 50 tons of plutonium in 30 years would require a capability to process more than 400 pits per year. Processing the plutonium more quickly would require correspondingly larger capabilities. There are no operating MOX fabrication facilities in the United States. LEU fuel fabrication facilities cannot readily be modified for this purpose because of the much higher radiotoxicity and safeguards requirements of plutonium fuel. Processing 50 tons of plutonium in 30 years, at a loading of 2.5 percent by weight plutonium in the fuel, would require a fabrication facility with a 67-

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Management and Disposition of Excess Weapons Plutonium: Reactor-Related Options MTHM/yr capacity; at a loading of 6.8 weight percent, the required capacity would be 25 MTHM/yr. There is an existing, nearly completed MOX fabrication facility at the federal government nuclear site in Hanford, Washington, known as the Fuel Materials Examination Facility (FMEF). This facility was built in the late 1970s and early 1980s to produce fuel for the liquid-metal Fast Flux Test Facility (FFTF). It was never operated. The panel has received estimates (which may be optimistic) that this facility could be modified to produce 50 tons of LWR fuel per year or more (containing roughly 3 tons of WPu, at a 6- to 7-percent enrichment) while meeting current safeguards and ES&H standards, within roughly five years of receiving a go-ahead, for a cost in the range of $75-$150 million.4 Funds would be needed to replace or upgrade existing, outdated process computers; upgrade older facility systems such as fire protection and waste handling; upgrade other facilities such as security and radiological protection to meet current federal, state, and DOE requirements; prepare safety analysis and safety and compliance documentation; modify the facility for LWR rather than liquid-metal reactor (LMR) fuel; and provide higher throughput than was originally planned. Alternatively, a new plutonium fuel fabrication facility could be built. Estimates provided to the panel (which appear to be optimistic) indicate that such facilities could be built for between $400 million and $1.2 billion, depending on their capacity. Siting, designing, building, and licensing such a facility would probably require a decade or more. These cost and schedule issues are discussed in more detail in Chapter 6. Reactor and Institutional Options Many specific variants of MOX use in LWRs in the United States can be imagined. The reactors used could be existing, partly completed, or newly built for this purpose. Fuel fabrication could rely on partly completed or modifiable facilities or new ones. The relevant facilities could be government-owned and government-operated (GOGO); government-owned but contractor-operated (GOCO); owned and operated privately, with subsidies from the government to make the system competitive with other sources of power; or some mix of these. As described in Chapter 3, in choosing among these variants, the nation should seek to minimize: security risks (which argues for minimizing the number of sites involved and the amount of transportation of plutonium in forms vulner 4   See discussion in section "Economic Comparisons" in Chapter 6. Atomic Energy of Canada, Limited, the only vendor which has so far done a detailed study of possible use of the FMEF, concluded that adapting the facility to produce MOX for CANDU (Canadian deuterium-uranium) reactors would cost $118 million (AECL 1994).

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Management and Disposition of Excess Weapons Plutonium: Reactor-Related Options able to theft, and maximizing the use of government sites, where requisite security either already exists or is simpler to provide); costs and delays, both in construction and in gaining requisite licenses and approvals (which argues for making maximum use of existing or partly completed facilities, and again for use of government sites, where the licensing and public approval processes may be somewhat simpler); and risks to environment, safety, and health (which argues for making use of the safest and best designed facilities likely to be available for the purpose, and for choosing the scale and types of nuclear materials processing so as to minimize risks and waste streams). There are no options that perfectly meet these criteria. The following are a few of the most obvious specific candidates for this role: Currently Operating, Utility-Owned Reactors: As noted earlier, three System-80 LWRs are operational at the Palo Verde site in Arizona, which could operate with full-MOX cores without modification, with license amendments.5 The same may be possible with a variety of other operational utility-owned LWRs in the United States. In one possible approach, for example, if the utility agreed to participate, the federal government would cover any additional costs in using government-furnished MOX fuel and would provide the necessary additional safeguards and security at the site, while the utility would otherwise continue to operate the reactors much as they are operated today. Additional financial incentives might be required to convince the utility to undertake the additional political and licensing burdens involved. As several sites have more than one reactor on-site, the handling of "fresh" plutonium and MOX fuel could be limited to two sites—one where the MOX fuel would be fabricated (presumably a site within the nuclear weapons complex) and the reactor site. Exploration of utility and public reactions to this concept is still in its early stages, though more than one utility has privately expressed interest to DOE. The Washington Nuclear Project Reactors: Reactor 3 of the Washington Nuclear Project (WNP-3) in Washington state is a System-80 reactor, 75-percent complete, in the western part of the state, roughly 150 miles from the Hanford nuclear weapons complex reservation. WNP-2, complete and operating, is not a System-80, but may also be capable of handling a full core of MOX fuel without major modifications. It has the advantages of being complete, licensed, and located on the federal government's Hanford site, where the FMEF fuel fabrication facility is also located. The uncompleted WNP-1, like WNP-2, is not a System-80 but is located on the Hanford site. One or two of the three WNP reac- 5   It should be noted that the Palo Verde reactors have been experiencing steam generator problems, and one of the units, Palo Verde 1, has a poor lifetime load-factor record. See Nucleonics Week (1994), and NEI (1993), both cited in Lyman (1994).

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Management and Disposition of Excess Weapons Plutonium: Reactor-Related Options tors could be acquired, completed (in the case of WNP-1 and WNP-3), and operated by the federal government (possibly in co-operation with a private entity) for the plutonium disposition mission. (For a discussion of the costs involved, see Chapter 6.) If the MOX fabrication capability at Hanford were used, this would have the significant advantage of confining all plutonium handling to two federal sites in the same state (or even a single large site, if only the WNP-1 and WNP-2 facilities on the Hanford reservation were used). Two groups of private companies have put forward proposals for a government/private partnership to pursue this approach.6 If these options are to be preserved, action will have to be taken soon, as the WPPSS board has recently voted to cease maintaining WNP-1 and WNP-3 and to offer the components for sale. Construction work on these reactors was halted some years ago. If these reactors were to be completed for MOX use, an intensive NRC review would be expected, which might result in requirements to upgrade some reactor systems to make them more comparable to the new reactors whose designs are now being reviewed by the NRC. Such modifications, if required, would add to the cost of completion. DOE is also considering other options involving completion of partly-completed reactor facilities for the plutonium disposition mission, such as use of three Tennessee Valley Authority reactors on which construction has been 6   One team, consisting of Battelle, Science Applications International Corporation, and Newport News, calls itself the "Isaiah Project" (after the biblical prophet who admonished the world to beat swords into plowshares). In their proposal, the private consortium they would set up would acquire and complete WNP-1 and WNP-3 at its own expense (deeding ownership of the reactors to the government) and receive revenue to pay for debt service and profit. The government would pay for reactor operations, fuel fabrication, storage, and disposal, and provide a contractual guarantee of particular quantities of steam for electricity production. Advocates for this concept have emphasized the possibility that the private entity could borrow several billion dollars against the future revenues of the project, which could be provided to the government to finance other endeavors, such as assistance for plutonium disposition in Russia. This is misleading, however, as future costs assigned to the government in this concept would come to substantially more than the sums that could be borrowed. Hence, as with other approaches, the project would involve a substantial net discounted present cost to the government, not net discounted present value. Borrowing against future revenue, with the accompanying promise of large future government expenditures, would simply amount to deficit financing by other means. This point is equally applicable to other approaches involving private financing of initial capital costs in return for government promises of later subsidies. In response to the Isaiah Project proposal, WPPSS, which built these reactors, has proposed a different concept that would use WNP-1 and WNP-2, rather than WNP-1 and WNP-3. This would keep all the operations on the Hanford reservation, which would have security advantages (because of reduced plutonium transport) and political advantages (because of the more favorable climate near Hanford than in the western part of the state). WNP-2, in the initial WPPSS proposal, would operate with only a one-third MOX core, but since it is already operating, and limited quantities of fuel might be produced in existing facilities at Los Alamos or elsewhere, it could serve as a test-bed, and begin plutonium operations before a full-MOX fabrication facility became available.

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Management and Disposition of Excess Weapons Plutonium: Reactor-Related Options lead to the growth of such large amounts of Sm-149 that it may not be restartable until the Sm-149 has been removed. If the ABC were partly fueled with thorium, as envisioned in some concepts, another potential problem is created by the isotope protactinium-233, which decays to U-233; when power is reduced from equilibrium, the amount of U-233 in the system will increase, causing an increase in reactivity. The higher the neutron flux, the larger this reactivity swing would be, raising a concern for the high-flux ABC. In short, considerable work needs to be done before there can be high assurance that the risk of potentially dangerous reactivity excursions in the proposed subcritical reactors can be eliminated. Use of Plutonium. One of the important differences between the proposed ABC molten-salt system and the MSRE is the use of plutonium as the fuel. Only a small amount of plutonium was ever introduced into the MSRE, which ran primarily on uranium. A technology development program would be needed to demonstrate that the different neutronic and chemical characteristics of the system with plutonium are adequately understood. For example, fission of uranium tends to make the chemistry of the molten salts more oxidizing, while fission of plutonium tends to make it more reducing; a reducing environment could lead to precipitation of certain metals, which could have serious effects (such as blocking flow and thereby leading to local overheating). Miscellaneous Issues. Other issues identified in the MSRE include surface cracks and ductility changes in the alloy used to contain the salt, graphite permeability to the salt, and shape stability of the graphite under intense irradiation. Warping of the graphite blocks in an ABC system could create voids where fuel might collect and stagnate, causing local overheating. The production of tritium from bombardment of the lithium in the salts is a potential ES&H concern, particularly given the permeability of the containment metals to tritium at the relevant temperatures. Similarly, the reliance on highly toxic beryllium fluoride in the MSRE might not be acceptable under today's ES&H standards. The Reprocessing The reprocessing involved in a molten-salt system, particularly one designed to transmute long-lived fission products as well as actinides, would be quite different from the aqueous reprocessing used around the world today. There is no base of experience available comparable to the experience with aqueous reprocessing. As noted above, enrichment of intensely radioactive cesium would also be required. To achieve the hoped-for goal of having the reprocessing waste qualify as Class C waste (not requiring repository disposal), unprecedented separation factors would be required. (It should also be noted

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Management and Disposition of Excess Weapons Plutonium: Reactor-Related Options that even if this goal is met for the fuel itself, the irradiated system components and similar wastes will almost certainly require repository disposal.) The necessary processes require development and demonstration. System Optimization Considerable further study will be required to optimize the various components of the ABC concept. One important question in that process will be: Is the accelerator that gives the concept its name needed? An alternative would be operating a similar reactor system (whichever reactor system is ultimately chosen for ABC) at full criticality, as current reactors operate (and as the MSRE operated). Does subcritical operation provide a unique safety advantage sufficient to justify consuming 10-20 percent of all the power produced by the system (meaning 10-20 percent of all its potential revenues in electricity sales) to operate the accelerator? If an accelerator is needed, would a lower-power accelerator (and a reactor closer to criticality) be sufficient? In short, the ABC technology remains quite immature. Technical feasibility of a number of the fundamental concepts, and engineering feasibility of essentially every aspect of the system, remain to be demonstrated. A development program costing several billion dollars and lasting for decades would probably be needed before a commercial-scale ABC system could be built with confidence. Other Issues The ABC concept is too immature to allow detailed treatment of issues such as licensing, ES&H, and cost, but some discussion is in order. As with other fundamentally new reactor designs, gaining required licenses and political approvals for construction and operation of an ABC system is likely to be difficult. The NRC has no experience regulating systems comparable to proposed ABC molten-salt reactors. Both the new-concept reactors and the associated reprocessing may face substantial difficulties with public acceptance. Even if all long-lived wastes could be transmuted, the storage of tanks of intensely radioactive liquid wastes would be required at the ABC site for decades or centuries, a further complication for public acceptance. If, however, claims that the system can transmute all long-lived wastes are proven, prospects for acceptance could be improved. ES&H impacts of ABC are claimed to be low. ABC proponents point in particular to the hoped-for nearly complete elimination of long-lived species that must be disposed of in geologic repositories, which would reduce the potential ES&H impacts of repository disposal. Success in a wide array of as-yet-unproven technologies will be required, however, to ensure that ES&H impacts of ABC-system operations (including reactor operations and reprocessing) will

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Management and Disposition of Excess Weapons Plutonium: Reactor-Related Options be acceptably low. As noted earlier, while molten-salt systems may offer a reduced risk of large accidents, there may also be a larger number of scenarios that could lead to radioactive contamination within the reactor building, leading to greater worker hazard. The intensely radioactive molten fuel would lead to high radioactivity levels throughout the reactor primary circuit, requiring development of remote-handling equipment, and potentially posing risks of significant worker exposures. The cost of ABC systems is quite speculative, given their current state of development. If past experience is a guide, costs will be substantially higher than now envisioned. It appears unlikely that an advanced-reactor system requiring a large accelerator and advanced reprocessing will be cost-competitive with proven LWR technology as a power producer in the next several decades, while abundant low-cost uranium continues to be available. If ABC were not cost-competitive, so that a "transmutation subsidy" of, for example, a few mils per kilowatt-hour would have to be paid to make the system economically viable, the cost to transmute the world supply of spent fuel could be very large. Some Conclusions If ABC existed and were in use for transmutation of high-level wastes and actinides, then it could readily handle the WPu. The development and validation program for ABC would take longer than for most other options, however, and it is impossible as yet to have high confidence that this approach will succeed. The role of the ABC is far from assured. First, it must be proved feasible—technically, politically, economically, and institutionally. If it is feasible, there seems no reason why its rare abilities should be wasted on the relatively easy task of converting WPu into spent fuel. Accelerator-based conversion might be of use after the WPu problem has been transformed into a small part of the civilian plutonium stock. Only if that large global stockpile were to be transmuted by ABC systems would it make sense to commit the WPu to treatment by accelerator-based conversion.

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