<\/p>\n
<\/p>\n
<\/p>\n
<\/p>\n
<\/p>\n
\n\nSUMMARY<\/b><\/h4>\n
<\/p>\n
High neutron economy, on-power refuelling, and a simple bundle design provide a high degree of flexibility that enables\u00a0CANDU<\/small>\u00a0(CAN<\/span>ada\u00a0D<\/span>euterium\u00a0U<\/span>ranium; registered trademark) reactors to be fuelled with a wide variety of fuel types.<\/span><\/b><\/p>\n
Near-term applications include the use of slightly enriched uranium (SEU<\/small>), and recovered uranium (RU<\/small>) from reprocessed spent Light Water Reactor (LWR<\/small>) fuel.<\/span><\/b><\/p>\n
Plutonium and other actinides arising from various sources, including spent\u00a0LWR<\/small>\u00a0fuel, can be accommodated, and weapons-origin plutonium could be destroyed by burning in\u00a0CANDU<\/small>.<\/span><\/b><\/p>\n
In the\u00a0DUPIC<\/small>\u00a0fuel cycle, a dry processing method would convert spent Pressurized Water Reactor (PWR<\/small>) fuel to\u00a0CANDU<\/small>\u00a0fuel.<\/span><\/b><\/p>\n
The thorium cycle remains of strategic interest in\u00a0CANDU<\/small> to ensure long-term resource availability, and would be of specific interest to those countries possessing large thorium reserves, but limited uranium resources.<\/span><\/b><\/p><\/blockquote>\n
\n1. INTRODUCTION<\/b><\/h6>\n
Fuel cycles have always been of key strategic importance<\/span><\/b>\u00a0to the nuclear industry (Green and Boczar (1)).\u00a0Keen interest in fuel cycles<\/span><\/b>\u00a0that improve uranium utilization was originally driven by a belief that uranium resources would not support the requirements of a growing nuclear system.\u00a0Reprocessing technology was developed to provide plutonium<\/span><\/b>\u00a0for fast breeder reactors to extend fuel resources. Similarly, early work in the thorium fuel cycle was motivated by uranium resource considerations. Interest in effective uranium utilization is now motivated by other considerations, such as environmental concerns far the front- and back- end of the fuel cycle, and national policies to secure the maximum benefit from nuclear energy resources, or to increase energy self-reliance.\u00a0Reprocessing, and recycling the recovered uranium and plutonium back into thermal reactors<\/span><\/b>, is a means of increasing the energy derived from the original mined uranium.<\/p>\n
The fuel cycle is being increasingly viewed in the context of the overall waste management strategy. Hence, there is currently interest in actinide burning and transmutation as waste management options, even though disposal concepts, such as geological disposal, have been shown to effectively eliminate the radiological risk from long-lived actinides. A related interest in fuel cycles stems from the end of the cold war, and\u00a0the enticing possibility of burning weapons-origin plutonium<\/span><\/b>\u00a0or high enriched uranium in nuclear power stations to generate electricity while enhancing world security.<\/p>\n
Of course, economics remains an important consideration in assessing fuel cycle options. The fuel cycle can impact both the fuelling and capital costs of a nuclear plant.<\/p>\n
In Canada, the desire to burn natural uranium fuel (so as not to depend on enrichment facilities) necessitated a reactor design (the\u00a0CANDU<\/small>) having high neutron economy, achieved through the use of heavy-water moderation, low parasitic absorption in the core (in the reactivity control devices and in structural materials), and on-power refuelling. While the capability of using natural uranium is still an important attribute in some markets, these features also provide an unsurpassed degree of flexibility that enables\u00a0CANDU<\/small>\u00a0to utilize\u00a0a wide variety of fuel types to meet local market requirements<\/span><\/b>.<\/p>\n
This paper examines some of the\u00a0CANDU<\/small>\u00a0fuel cycle options that are currently of interest to\u00a0AECL\u00a0<\/small>and its customers<\/span><\/b>.<\/p>\n
2. SEU \u00a0 [\u00a0S<\/span>lightly\u00a0E<\/span>nriched\u00a0U<\/span>ranium ]<\/b><\/h6>\n
The use of\u00a0SEU<\/small>\u00a0in\u00a0CANDU<\/small>\u00a0has many attractions (Boczar et al (2)). Enrichment would reduce the quantity of spent fuel produced in\u00a0CANDU<\/small>, which may be perceived by the public as addressing the first of the environmental three-R’s: reduce, recycle, reuse. A\u00a0U-235<\/small>\u00a0content of\u00a01.2<\/small>\u00a0percent would increase the burnup in\u00a0CANDU<\/small>\u00a0by a factor of three, and hence result in a three- fold reduction in the quantity of spent fuel produced. Enrichment would alleviate pressure on interim storage requirements at the reactor as well.<\/p>\n
While the natural-uranium-fueIled\u00a0CANDU<\/small>\u00a0is the most neutron-efficient of all commercial reactors in operation today,\u00a0SEU<\/small>\u00a0would further improve the uranium utilization (the energy derived from the mined uranium). An improvement of about\u00a030<\/small>\u00a0percent in uranium requirements is achieved for an enrichment of\u00a01.2<\/small>\u00a0percent. Enrichment would preserve the advantage in uranium utilization that\u00a0CANDU<\/small>\u00a0enjoys over the\u00a0LWR<\/small>, as the latter goes to higher burnups, employs uranium-conserving fuel-management techniques, or employs lower tails enrichment that could eventually be made economical through advanced enrichment technology. Uranium utilization is an important consideration for some countries that have few indigenous uranium resources, and which have a strategic interest in energy self- reliance.<\/p>\n
In operating\u00a0CANDU<\/small>\u00a0stations, significant cost reductions can be achieved by using\u00a0SEU<\/small>\u00a0fuel. Fuel cycle costs are minimized at an enrichment of around\u00a01.2<\/small>\u00a0percent, and are about\u00a030<\/small>\u00a0percent lower than with natural uranium fuel. Both front- and back-end fuel cycle costs would be reduced with\u00a0SEU<\/small>.<\/p>\n
SEU<\/small>\u00a0offers greater flexibility in reactor design. In new reactors, or in existing reactors where there is sufficient heat removal capacity,\u00a0SEU<\/small>\u00a0can be used to uprate reactor power without exceeding existing limits on bundle or channel power, by flattening the channel power distribution across the reactor core (Chan and Dastur (3)). This option involves trading-off the extra burnup potential of\u00a0SEU<\/small>\u00a0for more power. In a new reactor design, the use of power flattening to obtain more power from a given-sized core has an advantage in capital costs over simply adding more channels to the reactor. One option for\u00a0AECL<\/small>‘s new\u00a0CANDU 9<\/small>\u00a0reactor uses enrichment of around\u00a00.9<\/small>\u00a0percent to flatten the channel power distribution in the core to obtain\u00a01030 MW<\/small>e from a\u00a0480<\/small>-channel, Darlington-size core, nominally rated at\u00a0935 MW<\/small>e (Hart (4)). SEU<\/small>\u00a0could be used to reduce the capital cost of new plants by increasing the pressure tube thickness to upgrade the primary heat transport system (PHTS<\/small>) conditions, thereby achieving higher thermodynamic efficiency, or by reducing the moderator inventory by decreasing the moderator and reflector volumes (Dastur and Chan (5)).<\/p>\n
The use of enrichment in\u00a0CANDU<\/small>\u00a0also offers greater flexibility in fuel bundle design. One example is the Low Void Reactivity Fuel (LVRF<\/small>) bundle, in which the use of enrichment and neutron absorber materials allows any value of void reactivity and discharge burnup to be designed (Boczar et al. (6)). This has the potential for increasing the degree of passive safety in the\u00a0CANDU<\/small>\u00a0design, as well as reducing capital costs (by allowing a simplification of the\u00a0PHTS<\/small>).<\/p>\n
CANDU<\/small>‘s on-power refuelling offers flexibility in fuel management that facilitates the use of\u00a0SEU<\/small>\u00a0and other advanced fuel cycles. This flexibility extends from the equilibrium core, where, for example, different fuel management strategies could be used to accommodate different levels of enrichment, to the transition from one type of fuel (such as natural uranium) to another (such as\u00a0SEU<\/small>). Fuel management strategies have been identified for both the equilibrium core, and for the transition from natural uranium to\u00a0SEU<\/small>\u00a0(Charles Dastur (7); Younis and Boczar (8, 9); Boczar et al. ())<\/p>\n
No reactor physics obstacles have been identified, and no significant changes are required to accommodate\u00a0SEU<\/small>\u00a0in\u00a0CANDU<\/small>. An advanced fuel bundle is being developed as the optimum carrier of enriched fuels in\u00a0CANDU<\/small>\u00a0(Lane et al. (11)). This new bundle, called\u00a0CANFLEX<\/small>, (CAN<\/span><\/b>DU\u00a0FLEX<\/span><\/b><\/small>ible Fuelling) is more subdivided than other\u00a0CANDU<\/small>\u00a0bundles, having\u00a043<\/small>\u00a0elements with two pin sizes. When operated at current bundle powers, peak linear element ratings are reduced by\u00a015-20<\/small>\u00a0percent, depending on burnup. The lower ratings, along with optimized internal fuel-element design, will facilitate the achievement of extended burnup in\u00a0CANDU<\/small>\u00a0by reducing fuel temperatures, and hence fission-gas release within the fuel element. Critical heat flux and critical channel power will be increased, due to the optimization of the number and location of flow-disturbing appendages in the bundle. This feature can also be used to increase operating margins in operating reactors.<\/p>\n
The use of enrichment is the logical first step from natural uranium fuel in\u00a0CANDU<\/small>.<\/p>\n
3. CANDU \/ PWR \u00a0 SYNERGISM<\/b><\/h6>\n
The basis for the synergism between\u00a0CANDU<\/small>\u00a0and\u00a0PWR<\/small>\u00a0arises from the fundamental characteristics of the two reactor types:\u00a0PWR<\/small>\u00a0fissile requirements are higher than for\u00a0CANDU<\/small>, because of the good neutron economy of the latter (Boczar and Dastur (l2)). The higher parasitic loads in the\u00a0PWR<\/small>\u00a0lattice need to be compensated for by extra fissile material, both in the fresh and spent fuel. As a result,\u00a0PWR<\/small>\u00a0spent fuel has a high fissile content — about\u00a00.9<\/small>\u00a0percent by weight of\u00a0U-235<\/small>, and about\u00a00.6<\/small>\u00a0percent fissile plutonium, depending on the initial enrichment and exit burnup. In\u00a0CANDU<\/small>, the fissile content in the fresh fuel is low because of good neutron economy. Moreover, the initial\u00a0U-235<\/small>\u00a0and the self-generated plutonium are burned to low levels in\u00a0CANDU<\/small>.\u00a0CANDU<\/small>\u00a0fresh natural uranium contains\u00a00.7<\/small>\u00a0percent\u00a0U-235<\/small>, while the spent fuel contains\u00a00.2<\/small>\u00a0percent\u00a0U-235<\/small>\u00a0and\u00a00.2<\/small>–0.3<\/small>\u00a0percent fissile plutonium. Even with\u00a01.2<\/small>\u00a0percent\u00a0SEU<\/small>, there is only about\u00a00.4<\/small>\u00a0percent fissile material in the spent\u00a0CANDU<\/small>\u00a0fuel. Hence, spent\u00a0PWR<\/small>\u00a0fuel has about\u00a01.5<\/small>\u00a0percent fissile material, compared to about\u00a00.4<\/small>\u00a0percent fissile material in spent\u00a0CANDU<\/small>\u00a0fuel. Spent\u00a0PWR<\/small>\u00a0fuel, therefore, can be viewed as a source of fissile material for\u00a0CANDU<\/small>.<\/p>\n
CANDU<\/small>‘s excellent neutron economy means that about twice as much energy can be extracted from the fissile material in spent\u00a0PWR<\/small>\u00a0fuel by recycling it in\u00a0CANDU<\/small>\u00a0rather than in a\u00a0PWR<\/small>\u00a0(Boczar et al. (13); Hastings et al. (14)). In conventional reprocessing, fission products are removed, and the uranium and plutonium are separated. The plutonium can be mixed with uranium (either natural, depleted, or the recovered uranium from the reprocessing plant) to form\u00a0MOX<\/small>\u00a0fuel, which can be effectively utilized in\u00a0CANDU<\/small>.\u00a0AECL\u00a0<\/small>has performed extensive studies on the use of\u00a0MOX<\/small>\u00a0fuel in\u00a0CANDU<\/small>\u00a0in collaboration with an overseas client. No technical obstacles have been identified, and in fact there is considerable potential for optimizing the plant design to reduce capital costs through the use of\u00a0MOX<\/small>, as with\u00a0SEU<\/small>.<\/p>\n
The uranium from reprocessing is referred to as “recovered uranium” (RU<\/small>). It has a\u00a0U-235<\/small>\u00a0content of around\u00a00.9<\/small>\u00a0percent, and its use in\u00a0CANDU<\/small>\u00a0without re-enrichment is a very attractive fuel cycle option; it is discussed in greater detail in the next section.<\/p>\n
A chemical decontamination process could be used to separate fission products and unwanted actinides from the unseparated uranium\/plutonium mixture, which would then be co-converted into\u00a0MOX<\/small>\u00a0fuel, and used either as-is in\u00a0CANDU<\/small>, or diluted with natural or depleted uranium (depending on the desired burnup). This is the conventional\u00a0TANDEM<\/small>\u00a0fuel cycle. The advantage of chemical decontamination over conventional reprocessing lies in the potential of a cheaper, simpler process that is more proliferation-resistant and easier to safeguard, since plutonium is not separated from uranium.\u00a0AECL\u00a0<\/small>and the Korean Atomic Energy Research Institute (KAERI<\/small>) investigated the\u00a0TANDEM<\/small>\u00a0cycle in the early 1980’s.<\/p>\n
A fuel cycle currently under extensive consideration has an even greater degree of safeguardability than the\u00a0TANDEM<\/small>\u00a0cycle. This is the\u00a0DUPIC<\/small>\u00a0cycle (D<\/span><\/b>irect\u00a0U<\/span><\/b>se of Spent\u00a0P<\/span><\/b>WR<\/small>\u00a0Fuel\u00a0I<\/span><\/b>n\u00a0C<\/span><\/b>ANDU<\/small>), discussed in greater detail in section 5.<\/p>\n
The use of\u00a0CANDU<\/small>\u00a0to maximize the energy potential of the fissile material from spent\u00a0PWR<\/small>\u00a0fuel offers several benefits, including increased overall uranium utilization, and a reduction in the total quantity of spent fuel.<\/p>\n
4. RECOVERED URANIUM (RU)<\/b><\/h6>\n
RU<\/small>\u00a0is a by-product of conventional reprocessing of\u00a0LWR<\/small>\u00a0fuel. With a nominal\u00a0U-235<\/small>\u00a0concentration of\u00a00.9<\/small>\u00a0percent,\u00a0RU<\/small>\u00a0is a subset of\u00a0SEU<\/small>\u00a0that is particularly attractive for currently operating and future\u00a0CANDU<\/small>\u00a0reactors. Its use without re-enrichment in\u00a0CANDU<\/small>\u00a0offers many of the benefits of\u00a0SEU<\/small>. Uranium utilization (the amount of energy derived from the mined uranium used in the original\u00a0PWR<\/small>\u00a0fuel) would be improved by about\u00a025<\/small>\u00a0percent. Double the energy can be extracted from the\u00a0RU<\/small>\u00a0by burning it in\u00a0CANDU<\/small>\u00a0rather than re-enriching it as fuel for a\u00a0PWR<\/small>. Fuel burnup in\u00a0CANDU<\/small>\u00a0would be about twice that of natural uranium, resulting in a two- fold reduction in the volume of spent fuel and a commensurate reduction in back-end disposal costs. By flattening the channel power distribution across the reactor core so that all channels produce nearly the same power,\u00a0RU<\/small>\u00a0offers a power uprating capability.<\/p>\n
The suitability of\u00a0RU<\/small>\u00a0as a rector fuel for\u00a0CANDU<\/small>\u00a0was recently assessed in a joint program between\u00a0AECL\u00a0<\/small>and\u00a0COGEMA<\/small>\u00a0of France (Boczar et al. (15)). Pellets were pressed from the\u00a0RU<\/small>\u00a0powder, and both powder and pellets met\u00a0CANDU<\/small>\u00a0fuel specifications. One issue that had been identified in an earlier assessment was whether trace amounts of cesium-137<\/small>\u00a0in the\u00a0RU<\/small>\u00a0powder would be released during sintering, and if so, whether this would condense in the cold part of a sintering furnace in a commercial fuel fabrication plant, leading to a build-up in fields over time. This was assessed by sintering\u00a04000 RU<\/small>\u00a0pellets in a furnace that had been designed with a cold-trap in which volatile cesium released during sintering would condense. It was concluded that volatile cesium-l37<\/small>\u00a0would not pose a radiological problem in a commercial fuel fabrication plant.<\/p>\n
Fuel management with\u00a0RU<\/small>\u00a0should be particularly simple. A simple four-bundle shift, bi-directional fuelling scheme would result in good axial power profiles, and a refuelling rate in bundles per day that is half that for natural uranium. Alternatively, a two-bundle shift fuelling scheme could be used that would result in smaller refuelling ripples, but a higher refuelling rate. Peak channel and bundle powers would be comparable, or lower, than for natural uranium fuel. Peak element ratings with\u00a0CANFLEX<\/small>\u00a0would be below\u00a045<\/small>\u00a0kW\/m in a\u00a0CANDU 6<\/small>\u00a0reactor, which would facilitate good fuel performance at extended burnup, with low fission-gas release within the fuel element. Significant power boosting during refuelling would occur only for relatively fresh fuel, which is tolerant to power boosts. The reactivity worths of control devices would be acceptable for safety and control functions.<\/p>\n
Fuel cycle economics were recently assessed for\u00a0RU<\/small>\u00a0and\u00a0SEU<\/small>\u00a0in\u00a0CANDU<\/small>, and for re-enriched\u00a0RU<\/small>\u00a0in a\u00a0PWR<\/small>\u00a0(Boczar et al. (15)). The potential savings in\u00a0CANDU<\/small>\u00a0fuel cycle costs with\u00a0RU<\/small>\u00a0are striking. Over a range of reasonable cost assumptions, front-end fuelling costs for\u00a0RU<\/small>\u00a0are reduced relative to natural uranium by between\u00a028<\/small>\u00a0percent and\u00a067<\/small>\u00a0percent, and by\u00a015<\/small>\u00a0percent to\u00a030<\/small>\u00a0percent compared to fuelling costs for\u00a01.2<\/small>\u00a0percent\u00a0SEU<\/small>.<\/p>\n
In summary, excellent neutron economy and fuel cycle flexibility creates a niche in which\u00a0CANDU<\/small>\u00a0is uniquely suited for burning\u00a0RU<\/small>\u00a0without re-enrichment.<\/p>\n
5. DUPIC \u00a0 [\u00a0D<\/span>irect\u00a0U<\/span>se of Spent\u00a0P<\/span>WR Fuel\u00a0I<\/span>n\u00a0C<\/span>ANDU ]<\/b><\/h6>\n
The\u00a0DUPIC<\/small>\u00a0fuel cycle exploits the\u00a0CANDU<\/small>\u00a0neutron economy and fuel cycle flexibility in a manner that maximizes the safeguardability of recovered fissile material from spent\u00a0PWR<\/small>\u00a0fuel (Keil et al. (16)). The various\u00a0DUPIC<\/small>\u00a0options do not use reprocessing or wet chemical processes, only dry processes, to utilize the energy content of spent\u00a0PWR<\/small>\u00a0fuel in\u00a0CANDU<\/small>.<\/p>\n
In\u00a01992, AECL, KAERI<\/small>\u00a0and the U.S. Department of State completed Phase I of an assessment of the\u00a0DUPIC<\/small>\u00a0cycle. Five mechanical reconfiguration options were assessed, involving rearranging the spent\u00a0PWR<\/small>\u00a0elements into\u00a0CANDU<\/small>\u00a0bundles, with or without double cladding. Two conceptual\u00a0CANDU<\/small>\u00a0fuel-bundle designs were evaluated to maximize fuel utilization:\u00a061<\/small>– and\u00a048<\/small>-element bundles having either single- or double-clad element sheaths. These bundles were chosen to make use of the smaller\u00a0PWR<\/small>-size elements while maximizing the fuel content of\u00a0CANDU<\/small>\u00a0bundles.<\/p>\n
Two powder-processing concepts were also evaluated. In the\u00a0OREOX<\/small>\u00a0option (O<\/span><\/b>xidation,\u00a0RE<\/span><\/b>duction of enriched\u00a0OX<\/span><\/b>ide fuel), spent\u00a0PWR<\/small>\u00a0pellets would be subject to successive oxidation\/reduction cycles to produce a sinterable UO2<\/sub>\u00a0powder that would be pressed into pellets, sintered, loaded into\u00a0CANDU<\/small>\u00a0sheaths, and fabricated into conventional\u00a0CANDU<\/small>\u00a0bundles. The second powder-processing option was “VIPAC<\/small>” (VI<\/span><\/b>bratory comPAC<\/span><\/b>tion), in which\u00a0PWR<\/small>\u00a0pellets would be ground into small, dense granules and vibratory-packed into sheaths.<\/p>\n
All of the options were assessed against a set of selection criteria, which included retrofitability to\u00a0CANDU<\/small>\u00a0and to\u00a0PWR<\/small>, safeguardability, licensability, reactor physics, fuel performance, fuel handling, fuel fabrication, and waste management.<\/p>\n
Both the mechanical reconfiguration options and the powder-processing options were found to be feasible. For the mechanical reconfiguration options, the low ratings (and consequently lower peak center-line temperatures) resulting from greater subdivision with the\u00a048<\/small>-element or\u00a061<\/small>-element bundles compensated- for both the greater variation in fissile composition due to axial or rod-to-rod variations in fissile content, and the greater heat resistance of double-cladding. The primary appeal of the\u00a0VIPAC<\/small>\u00a0option is its inherent simplicity, since no sintering is required (part of the fuel sinters in-core), and the specifications on the granules are much less stringent than for pellets. The main disadvantage is that it results in lower fuel densities than pellet fuel, and there is much less world-wide experience with\u00a0VIPAC<\/small>\u00a0than with pellet fuel.<\/p>\n
It was concluded that\u00a0OREOX<\/small>\u00a0is the most promising option, largely because of the homogeneity of the resultant powder and pellets. One of the advantages of this process is that it removes a high fraction of gaseous and volatile fission products, thereby improving fuel burnup. The\u00a0CANDU<\/small>\u00a0burnup with the\u00a0OREOX<\/small>\u00a0option is about\u00a018 MW<\/small>d\/kg, using spent fuel from the reference Korean\u00a0PWR<\/small>\u00a0which has an average discharge burnup of\u00a035 MW<\/small>d\/kg (initial\u00a0U-235<\/small>\u00a0enrichment of\u00a03.5<\/small>\u00a0percent).<\/p>\n
The\u00a0DUPIC<\/small>\u00a0cycle is particularly attractive in Korea, which has both\u00a0CANDU<\/small>\u00a0and\u00a0PWR<\/small>\u00a0reactors. In an equilibrium system in which the spent\u00a0PWR<\/small>\u00a0fuel would provide the fuelling needs of\u00a0CANDU<\/small>, the\u00a0DUPIC<\/small>\u00a0cycle would improve uranium utilization by about\u00a025<\/small>\u00a0percent, compared to an open cycle in which\u00a0CANDU<\/small>\u00a0was fuelled with natural uranium. In this scenario, the total quantity of spent fuel produced by both\u00a0CANDU<\/small>\u00a0and\u00a0PWR<\/small>\u00a0will be reduced by a factor of three.<\/p>\n
Although a large fraction of the gamma radioactivity would be removed from the recycled fuel, fields would still be high enough to require all refabrication and handling to be done remotely in a shielded facility. While this makes the fabrication of the\u00a0CANDU<\/small>\u00a0fuel bundles more costly and difficult, it increases the diversion-resistance of the cycle. The\u00a0OREOX<\/small>\u00a0process should result in good fuel performance, since the pellet and bundle design would be close to that of the reference\u00a0CANDU<\/small>\u00a0fuel. The safeguards assessment concluded that the proliferation risks of the\u00a0DUPIC<\/small>\u00a0cycle are relatively small, and presently known safeguards systems and technologies can be modified or adapted to meet\u00a0DUPIC<\/small>\u00a0safeguarding requirements (Pillay et al. (17)).<\/p>\n
The workscope for Phase II of the\u00a0DUPIC<\/small>\u00a0program is now being defined. This is a multi-year experimental verification program, involving optimization. of the\u00a0OREOX<\/small>\u00a0process, and fabrication of\u00a0DUPIC<\/small>\u00a0elements and bundles from spent\u00a0PWR<\/small>\u00a0fuel for subsequent test irradiation in a research reactor, followed by post-irradiation examination, development of remote fabrication technologies, and development of appropriate safeguard technology.<\/p>\n
6. ACTINIDE BURNING AND PLUTONIUM DESTRUCTION<\/b><\/h6>\n
Fuel cycle options are being proposed internationally that reduce the radiotoxicity of spent fuel arising from the long-lived actinides. Radiotoxicity is a measure of the hazard of ingesting or inhaling a substance. Radiotoxicity is not a measure of long-term risk from spent fuel in a waste management system, in which natural and man-made barriers are designed to isolate the waste from the biosphere. In fact, the environmental review of the Canadian geological disposal concept shows that the actinides pose negligible risk, because of their immobility in the disposal vault. The largest contributors to long-term dose are from the long-lived fission products,\u00a0I-129, C-14<\/small>\u00a0and\u00a0Tc-99<\/small>\u00a0(Dormuth et al. (18)), but the doses associated with these species are well below regulatory limits in a properly designed disposal vault.<\/p>\n
Nonetheless, there is interest internationally in assessing the feasibility of burning the plutonium and transuranic actinides from reprocessing in-reactor, as a waste management option. Because of its high neutron economy,\u00a0CANDU<\/small>\u00a0can be effective in this role (Dastur et al. (19)). The traces of fissile material in the transuranic mix from the reprocessing of spent\u00a0LWR<\/small>\u00a0fuel provide sufficient reactivity in a\u00a0CANDU<\/small>\u00a0lattice for use as fuel. The absence of uranium in such fuel prevents the formation of plutonium and the higher actinides. Without plutonium formation, the fissile content or the mix depletes rapidly with irradiation and constant reactor power output is maintained by using the on-power refuelling feature of\u00a0CANDU<\/small>\u00a0to shift the targets into increasing flux. The high neutron flux facilitates the transmutation and annihilation of the higher actinides. About\u00a03.6 GW<\/small>(e).a of\u00a0LWR<\/small>\u00a0actinide production could be annihilated annually in a\u00a0CANDU 6<\/small>\u00a0reactor of current design. No adverse effects on reactor dynamic behavior have been identified.<\/p>\n
High operating neutron flux, high neutron economy and on-power refuelling also make\u00a0CANDU<\/small>\u00a0particularly suitable for the annihilation of weapons-grade plutonium (Pitre and Dastur (20)). The plutonium would be irradiated in an inert matrix, such as zirconia or beryllia. The fissile content is maximized by using gadolinium to suppress-excess reactivity. Calculations show that an annihilation rate of\u00a02.5<\/small>\u00a0kg\/FPD<\/small>\u00a0(Full Power Day) can be achieved in a\u00a0CANDU 6<\/small>\u00a0reactor that is rated at\u00a0680-MW<\/small>(e). Significant fractional annihilation is achieved by shifting the fuel into locations of increasing flux level as the fissile content depletes. This is facilitated by the on-power refuelling capability of\u00a0CANDU<\/small>. Due to the abundance of heavy water, which is the main component of the\u00a0CANDU<\/small>\u00a0lattice, the effect of plutonium in the absence of\u00a0U-238<\/small>\u00a0on the reactor dynamics is shown to be acceptable with current reactor control technology. Furthermore, the reactivity coefficients of the\u00a0CANDU<\/small>\u00a0lattice can be adjusted by the judicious placement of certain burnable poisons in the fuel. This technique is used to ensure that the reactor fuel temperature and void reactivity coefficients are negative.<\/p>\n
Another option being proposed for disposing of weapons-grade plutonium is “spiking”: burning the plutonium in the form of a mixed oxide fuel, with the result that the radiation field from the resulting fission products is high enough to discourage diversion. Again, the on-power refuelling capability represents a significant advantage for\u00a0CANDU<\/small>\u00a0in this option.<\/p>\n
7. THORIUM FUEL CYCLES IN CANDU<\/b><\/h6>\n
Thorium is an alternate fuel to uranium, but since it has no fissile isotopes, it is necessary to provide fissile material (uranium or plutonium). The\u00a0U-233<\/small>\u00a0produced by irradiation of\u00a0Th-232<\/small>\u00a0has the highest eta value (ratio of neutrons produced to neutrons absorbed) for thermal neutron fission of any of the fissile nuclides. It is thus a very good fuel in the soft\u00a0CANDU<\/small>\u00a0spectrum. Moreover, the equilibrium concentration of\u00a0U-233<\/small>\u00a0in spent thorium fuel (about\u00a01.5<\/small>\u00a0percent\u00a0U-233<\/small>) is about five times that of fissile plutonium in spent natural uranium fuel, and so it should be a cheaper source of recycle fuel than plutonium (aithough this will be offset by higher fuel fabrication costs with recycled\u00a0U-233<\/small>, compared to recycled plutonium). The flexibility in fuel management provided by on-power fuelling is another\u00a0CANDU<\/small>\u00a0advantage in burning thorium.<\/p>\n
The fissile material can be provided in several ways, and these options define the various thorium fuel cycies (Milgram (21)). In most variants of the conventional once-through thorium cycle,\u00a0T<\/small>hO2<\/sub><\/small>\u00a0and\u00a0SEU<\/small>\u00a0are burned in separate channels, and the\u00a0U-233<\/small>\u00a0that is produced from neutron capture in\u00a0Th-232<\/small>\u00a0is burned in-situ. The conventional once-through thorium cycles require high thorium burnups,\u00a040-100 MW<\/small>d\/kg Th (compared to\u00a07 MW<\/small>d\/kg U for natural uranium fuel). Re-insertion of the spent\u00a0T<\/small>hO2<\/sub><\/small>\u00a0fuel after a cooling period can further utilize the energy from the decay of\u00a0P<\/small>a-233<\/small>\u00a0to\u00a0U-233<\/small>\u00a0while in storage. A major challenge in the once-through thorium cycles is to devise appropriate fuel management strategies.<\/p>\n
Other thorium fuel cycles employ reprocessing to optimize the energy potential from\u00a0U-233<\/small>, and these are of longer-term strategic interest. These reprocessing cycles mix\u00a0T<\/small>hO2<\/sub><\/small>\u00a0with either enriched uranium, or plutonium.\u00a0U-235<\/small>\u00a0can be provided as either high enriched uranium (around\u00a092<\/small>\u00a0percent\u00a0U-235<\/small>, as a vehicle for burning weapons-material\u00a0U-235<\/small>), or as medium enriched uranium (less than\u00a020<\/small>\u00a0percent\u00a0U-235<\/small>, for non-proliferation considerations). If plutonium were used to initiate the cycle, it would be obtained from reprocessing conventional\u00a0PWR<\/small>\u00a0or\u00a0CANDU<\/small>\u00a0spent fuel, or from dismantled weapons.<\/p>\n
If further improvements are made to the\u00a0CANDU<\/small>\u00a0neutron economy (such as removal of adjuster rods, and use of enriched zirconium for structural materials), the self-sufficient thorium cycle is feasible. This requires no fissile makeup once the equilibrium concentration of\u00a0U-233<\/small>\u00a0(1.5<\/small>\u00a0percent) has been achieved.<\/p>\n
Thorium has an additional potential benefit of lower radiotoxicity of the spent fuel than for uranium.<\/p>\n
8. CONCLUSIONS<\/b><\/h6>\n
Several options are being examined for exploiting the ability of\u00a0CANDU<\/small>\u00a0reactors to burn a variety of fuels. The direction of\u00a0CANDU<\/small>\u00a0fuel cycle development will be driven largely by local considerations, such as the availability and cost of fuel resources (uranium and thorium), the presence (or lack) of a high-technology infrastructure, and the reactor mix in the particular country. The flexibility exists with\u00a0CANDU<\/small>\u00a0technology to optimize the fuel cycle to meet the needs of our customers.<\/p>\n
9. REFERENCES<\/b><\/h6>\n
AECL-XXXX is an Atomic Energy of Canada Limited published report.<\/span><\/b><\/p>\n
\n
- GREEN, R.E. and BOCZAR, P.G., “Advanced Fuel Cycles in CANDU Reactors: Reconfirming the Need”, AECL-10156 (1990).<\/li>\n
- BOCZAR, P.G., MCDONNELL, F.N., LANE, A.D., FRESCURA, G.M., ARCHINOFF, G.H. and WIGHT, A.L, “Slightly Enriched Uranium in CANDU: An Economic First Step Towards Advanced Fuel Cycles”, AECL-983l (1988).<\/li>\n
- CHAN, P.S.W. and DASTUR, A.R., “The Role of Enriched Fuel in CANDU Power Uprating”, Proceedings of the 8th Annual Conference of the Canadian Nuclear Society, Saint John, New Brunswick, 1987: June 14-17.<\/li>\n
- HART, R.S., “CANDU-9 — Overview”, Proceedings of the IAEA Technical Committee Meeting on Advances in Heavy Water Reactors, Toronto, Canada, 1993: June 7-10.<\/li>\n
- DASTUR, A.R. and CHAN, P.S.W., “The Role of Enriched Uanium in CANDU Power Plant Optimization”, Proceedings of the IAEA Technical Committee Meeting on Advances in Heavy Water Reactors, Toronto, Canada 1993: June 7-10.<\/li>\n
- BOCZAR, P.G., GROENEVELS, D.C., LEUNG, L.K., DASTUR, A.R., CHAN, P.S.W., BOWSLAUGH, D.R., ALLEN, .P.J., SOEDUONO, P., CHOO, L.C., KEIL, H. and SEMOHA, R, “A Low-Void Reactivity CANDU Fuel Bundle”, Proceedings of the Third International Conference on CANDU Fuel, Chalk River, Ontario, 1992: October 4-8.<\/li>\n
- CHAN, P.S.W. and DASTUR, A.R., “Checkerboard Fuelling, the Key to Advanced Fuel Cycles in Existing CANDU Reactors”, Proc. Sixth Annual Conf. Canadian Nuclear Society, Ottawa, Canada. 1985 June.<\/li>\n
- YOUNIS, M.H. and BOCZAR, P.G., “Equilibrium Fuel-Management Simulations for 1.2 percent SEU in a CANDU 6”, AECL-9986 (1989).<\/li>\n
- YOUNIS, M.H. and BOCZAR, P.G., “Axial Shuffling Fuel Management Schemes for 1.2 percent SEU in CANDU”, AECL 10055 (1989).<\/li>\n
- BOCZAR, P.G., CHENG, N.B.Y. and THOMPSON, J.W., “The Transition from Natural Uranium to 1.2 percent SEU in a CANDU with Repositioning of Reactivity Devices”, AECL 10007 (1989).<\/li>\n
- LANE, A.D., BOCZAR, P.G., TOWNES, B.M., SUK, H.C., LEE, Y.O., JEONG, C.J. and SIM, K.S., “CANFLEX: A Fuel Bundle to Facilitate the Use of Enrichment and Fuel Cycles in CANDU Reactors”, Proceedings of the IAEA Technical Committee Meeting on Advances in Heavy Water Reactors, Toronto, Canada 1993: June 7-10.<\/li>\n
- BOCZAR, P.G. and DASTUR, A.R., ‘CANDU\/PWR Synergism”, Proceedings of the IAEA Technical Committee Meeting on Advances in Heavy Water Reactors, Toronto, Canada, 1993: June 7-10.<\/li>\n
- BOCZAR P.G., HASTINGS, I.J. and CELLI, A., “Recycling in CANDU of Uranium and\/or Plutonium from Spent LWR Fuel”, AECL-10018 (1989).<\/li>\n
- HASTINGS, I.J., BOCZAR P.G., ALLAN, C.J. and GACESA, M., “Synergistic CANDU-LWR Fuel Cycles”, AECL-10390 (1991). .<\/li>\n
- BOCZAR, P.G., SULLIVAN, J.D., HAMILTON, H., TOWNES, B.M., LEE, Y.O., JEONG, C.J., SUK, H.C. and MUGNIER, C., “Recovered Uranium in CANDU: A Strategic Opportunity”, Proceedings of the International Nuclear Congress and Exhibition, Toronto, Canada 1993: October 3-6.<\/li>\n
- KEIL, H., BOCZAR, P.G. and PARK, H.S., “Options for the Direct Use of Spent PWR Fuel in CANDU (DUPIC)”, Proceedings of the Third International Conference of CANDU Fuel, Chalk River, Ontario, 1992: October 4-8.<\/li>\n
- PILLAY, K.K.S., MENLOVE, H.O. and PICARD, RR., “Safeguardability of Direct Use of Spent PWR Fuels in CANDU Reactors”, LA- 12432-MS, 1992 October 11.<\/li>\n
- DORMUTH, K.W., GOODWIN, B.W. and WIKJORD, A.G., “Long-Term Safety Assessment of the Disposal of Nuclear Fuel Waste”, Proceedings of the International Nuclear Congress and Exhibition, Toronto, Canada 1993: October 3-6.<\/li>\n
- DASTUR, A.R., GRAY, A.S., GAGNON, N., BUSS, D.B., and VERRALL, R.A., “The Role of CANDU in Reducing the Radiotoxicity of Spent Fuel”, Proceedings of the GLOBAL’93 Conference — Future Nuclear Systems: Emerging Fuel Cycles and Waste Disposal Options, ANS Topical Meeting, Seattle, Washington U.S.A., 1993: September 12-17.<\/li>\n
- PITRE. J. and DASTUR, A.R., “The Role of the CANDU Reactor in the Disposition of Plutonium”, to be presented at the ANS Physics Topical Meeting, Knoxville, Tennessee, 1994 April.<\/li>\n
- MILGRAM, M.S., “Thorium Fuel Cycles in CANDU Reactors: A Review”, AECL-8326 (1984).<\/li>\n<\/ol>\n
<\/p>\n
\n<\/p>\n
[\u00a0Message to Jean Chr\u00e9tien<\/a>\u00a0]<\/center> <\/center> [\u00a0Plutonium Sub-Directory<\/a>\u00a0]<\/center> <\/p><\/blockquote>","protected":false},"excerpt":{"rendered":" CANDU Fuel Cycle Flexibility by D.F. Torgerson, P.G. Boczar and A.R. Dastur* AECL\u00a0Research, Chalk River Laboratories, Chalk River, Ontario, Canada,\u00a0KOJ \u00a01JO ( *\u00a0AECL\u00a0 CANDU, Mississauga, Ontario ) this paper was delivered at the 9th Pacific Basin Nuclear Conference Sydney, Australia, 1994: May 1-4. SUMMARY High neutron economy, on-power refuelling, …<\/p>\n