This study considers replacing the externally accessible void spaces inside a disposal package containing a spent nuclear fuel assembly (SNFA) with high heat conducting metal to increase the effective thermal conductivity of the package and simplify the heat transfer mechanism inside the package by reducing it to a conduction dominant problem. The focus of the study is on preventing the gaps adjacent to the walls of the package components, produced by solidification shrinkage of poured liquid metal. We approached the problem by providing a temporary coating layer on the components to avoid direct build-up of thick metal oxides on their surface to promote metallic bonding at the interfaces under a non-inert environment. Laboratory scale experiments without SNFA were performed with Zn coated low carbon steel canisters and Zamak-3 void filler under two different filling temperature conditions – below and above the melting point of Zn (designated BMP and AMP respectively). Gap formation was successfully prevented in both cases while we confirmed an open gap in a control experiment, which used an uncoated canister. Minor growth of Al-Fe intermetallic phases was observed at the canister/filler interface of the sample produced under the BMP condition while their growth was significant and showed irregularly distributed morphology in the sample produced under the AMP condition, which has a potential to mitigate excessive residual stresses caused by shrinkage prevention. A procedure for the full-scale application was specified based on the results.

This is the final report on a project to update and improve the conceptual design of deep boreholes for high level nuclear waste disposal. The effort was concentrated on application to intact US legacy LWR fuel assemblies, but conducted in a way in which straightforward extension to other waste forms, host rock types and countries was preserved. The reference fuel design version consists of a vertical borehole drilled into granitic bedrock, with the uppermost kilometer serving as a caprock zone containing a diverse and redundant series of plugs. There follows a one to two kilometer waste canister emplacement zone having a hole diameter of approximately 40-50 cm. Individual holes are spaced 200-300 m apart to form a repository field. The choice of verticality and the use of a graphite based mud as filler between the waste canisters and the borehole wall liner was strongly influenced by the expectation that retrievability would continue to be emphasized in US and worldwide repository regulatory criteria. An advanced version was scoped out using zinc alloy cast in place to fill void space inside a disposal canister and its encapsulated fuel assembly. This excludes water and greatly improves both crush resistance and thermal conductivity. However the simpler option of using a sand fill was found adequate and is recommended for near-term use. Thermal-hydraulic modeling of the low permeability and porosity host rock and its small (≤ 1%) saline water content showed that vertical convection induced by the waste’s decay heat should not transport nuclides from the emplacement zone up to the biosphere atop the caprock. First order economic analysis indicated that borehole repositories should be cost-competitive with shallower mined repositories. It is concluded that proceeding with plans to drill a demonstration borehole to confirm expectations, and to carry out priority experiments, such as retention and replenishment of in-hole water is in order.

This paper focuses on the improvement of the longevity and robustness of materials for sealing and plugging the upper portion of a deep borehole used for permanent isolation of high-level nuclear waste. Analytical models of porous and laminar flows show that even when materials have low intrinsic permeability, micron-sized cracks and gaps between the plug and rock (formed via chemical reaction, shrinkage, osmotic consolidation, etc.) significantly diminish the plug's sealing properties. On this basis, materials such as asphalt, traditional cements, and pure bentonite which crack or shrink under certain conditions are unfavorable. An ongoing test program has formulated expanding cement mixtures containing MgO to prevent such bypass flow. Furthermore, these findings support using stable, malleable, and low-permeability plug material (k <= 10(-16) m(2)) such as a crushed rock (70%) and bentonite (30%) mixture. Alternative clays such as sepiolite could be blended with the bentonite to further reduce the potential negative effects of salinity on bentonite permeability. A bounding and analytical model of a scenario where radionuclide escape is determined by advection through the plug (and assuming a large and constant driving pressure) shows that a plug permeability of 10(-16) m(2) is sufficiently low to prevent advective transport of radionuclides from a depth of 2 to 3 km to the suiface within the timescale of interest (similar to 1 million yr). Purely diffusive transport over the same distance, whether through the plug or host rock, is conservatively estimated to be significant only for a time >850 000 yr.

The United States is in need of a new and more adaptive long-term strategy for spent nuclear fuel. In this communication, we outline the fundamental reasons why deep borehole disposal should receive more detailed investigation, alongside traditional shallow mined repositories. This potential solution is supported by advancing drilling technologies and an improving understanding of extremely long fluid residence times in deep bedrock. Radionuclide isolation is supported by verifiable and stable geologic barriers such as long transport distances to aquifers, low permeability, and reducing chemical conditions. The modular nature of implementing deep borehole disposal could offer unique programmatic and economic advantages. Experience with a pilot borehole program will be required to confirm the feasibility of drilling and emplacement operations, and key chemical and hydraulic conditions.

The purpose of this study was to demonstrate the neutronic feasibility and competitive fuel cycle economics of sodium fast reactors operating with uranium metal (UZr) fuel on a once-through fuel cycle. Uranium startup fast reactors (USFRs) decouple their deployment from that of expensive reprocessing and recycle facilities. This could facilitate and speed up the deployment of conventional fast reactors, which, in their traditional designs, heavily depend on the availability of reprocessing facilities for transuranic fuel production. The uranium requirement and fuel cycle cost of studied USFR core designs are calculated to be comparable to those of typical light water reactors. The main design constraint is the fast neutron fluence imposed on the cladding material, which is required to be below 5.0 X 10(23) n/cm(2) even for advanced oxide dispersion strengthened steels. Therefore, moderators need to be inserted in the fuel assemblies to lower the fast neutron flux so that the fuel residence time limited by neutron fluence can be extended to match the reactivity limited fuel residence time. In this study, magnesium oxide is used for reflectors as well as for the moderator.

While extensive stress field data are available from crystalline rock boreholes drilled in France (Soultz), Germany (KTB), and the USA (Cajon Pass, Monticello Reservoir), Canada and Sweden, the data and methods used to analyze them have yet to be applied to very deep geologic disposal facilities. Typically, to alleviate the stress fields that are intensified in a borehole, muds (mixtures of clay and water) are a critical component and are widely used in the drilling industry. In the first portion of this paper, we review the available mechanical data and analysis methods. Based on the most applicable measurements (in Canada and Germany), we propose values of stress fields and rock properties to be used for generic assessment of deep borehole disposal. The minimum horizontal stress can be approximated as Sh=23.2D, the maximum horizontal stress as SH=43.1D, and the vertical stress as SV=27.5D, where the stresses are in MPa and D is depth in km. This analysis also incorporates the effect of thermal stresses (relieved) by the cool drilling mud. Using an average uniaxial compressive strength results (C=212 MPa) and conservatively neglecting the increase in strength of crystalline rock under polyaxial conditions, a stable borehole can be drilled to 4.55 km with mud density of 1020 kg/m3. This is based on a stability limit such that at the bottom of the hole, a significant portion of the wall (180°/360°) reaches a critical state of stress (i.e., experiences spalling). Using relations developed for shallow mines (which may be overly conservative) the spalled zone is estimated to have a radius that is approximately twice that of the borehole. To reach 5 km, the mud density should be raised to 1420 kg/m3, or be actively cooled (90°C) below the ambient temperature of the rock (∼135°C) at that depth.

Reprocessing of Light Water Reactor (LWR) spent fuel to recover plutonium or transuranics for use in Sodium cooled Fast Reactors (SFRs) is a distant prospect in the U.S.A. This has motivated our evaluation of potentially cost-effective operation of uranium startup fast reactors (USFRs) in a once-through mode. This review goes beyond findings reported earlier based on a UC fueled MgO reflected SFR to describe a broader parametric study of options. Cores were evaluated for a variety of fuel/coolant/reflector combinations: UC/UZr/UO{sub 2}/UN;Na/Pb; MgO/SS/Zr. The challenge is achieving high burnup while minimizing enrichment and respecting both cladding fluence/dpa and reactivity lifetime limits. These parametric studies show that while UC fuel is still the leading contender, UO{sub 2} fuel and ZrH 1.7 moderated metallic fuel are also attractive if UC proves to be otherwise inadequate. Overall, these findings support the conclusion that a competitive fuel cycle cost and uranium utilization compared to LWRs is possible for SFRs operated on a once-through uranium fueled fuel cycle. In addition, eventual transition to TRU recycle mode is studied, as is a small test reactor to demonstrate key features. (authors)

The use of deep boreholes for the disposal of high-level radioactive waste is reassessed, emphasizing key enabling technical features and their strong linkage to national and international fuel cycle policy. Emplacement 2 to 4 km deep in widely available granitic continental bedrock, under a 1-km caprock layer of high-integrity bedrock, is shown in this analysis to have the potential to provide sufficiently low host rock permeability to prevent radionuclide escape by transport in water - the only plausible release mechanism. The modular nature of the concept enables multiregion siting in large user countries and is especially well-suited for small-user nations. Irretrievability can be built-in to better meet safeguards objectives, and the exceptionally high assurance of confinement makes the disposal of minor actinides (and troublesome fission products) an attractive alternative to their destruction by transmutation.

The inverted geometry (fuel outside coolant tubes) has been previously investigated at MIT for application in gas-cooled fast reactors and pressurized water-cooled thermal reactors. Venting has also been studied for conventional fuel pins and was employed for those in the Dounreay Fast Reactor. In the present work the inverted fuel approach was adopted because it allows high fuel volume fraction, reduction of the coolant void reactivity, neutron leakage and enrichment, as well as lower pressure drop for the same channel length because grids and wire wraps are no longer necessary. Furthermore most results also apply to venting of conventional fuel pins. Physical and chemical behavior of volatile fission products in sodium was investigated to determine the maximum activity inventory which would eventually be released into the primary sodium. Results of this analysis show that the most troublesome radionuclides in terms of propensity to escape from the venting system are noble gases ({sup 85}Kr and {sup 133}Xe), and cesium ({sup 134}Cs and {sup 137}Cs). A final vented inverted fuel assembly design is proposed which meets all the design goals which have been set. Additionally purification systems were devised to reduce radionuclide activity of the coolant and the cover gas to tolerablemore » levels. It is concluded that vented inverted (or vented conventional pin) fuel is a feasible concept and has sufficiently promising advantages - increasing fuel volume fraction to 50% and core outlet temperature by 20 deg. C, hence incrementing plant thermal efficiency by about 1% - to warrant serious consideration for future SFR designs. (authors)« less

Breeding blankets are of interest for a sodium fast reactor (SFR) as they allow for small cores to have positive breeding gains. However, because they breed very high-quality plutonium, core designers are not currently encouraged to employ blankets. After verifying that the ERANOS code was in good agreement with BGcore, a Monte Carlo based depletion system, it was shown that an SFR blanket design could breed less attractive plutonium than light water reactor (LWR) bred plutonium for making a nuclear explosive device. Minor actinide (MA) doping and moderator addition were the two options studied. This study shows that it is possible to build an SFR with a secure blanket with MA addition; at steady state MAs from approximately 1.5 LWRs are required per SFR [both rated at 1 GW(electric)]

A risk-informed methodology is applied to the selection of an ultimate heat sink for a Passive Secondary Auxiliary Cooling System. The reliability of the chosen design during the bounding transient, a station blackout, is calculated. The methodology considers both active component failures and the potential for inadequate cooling due to adverse thermal–hydraulic conditions. A response surface is developed as a surrogate for the thermal–hydraulic code and used for uncertainty propagation. The uncertainty introduced by the use of the response surface itself is explored. Two sensitivity studies are performed. The first study measures the sensitivity of peak clad temperature to initial ambient conditions and system degradation. The second study explores the sensitivity of system reliability to code error.

Four fast reactor concepts using lead (LFR), liquid salt, NaCl–KCl–MgCl2 (LSFR), sodium (SFR), and supercritical CO2 (GFR) coolants are compared. Since economy of scale and power conversion system compactness are the same by virtue of the consistent 2400MWt rating and use of the S-CO2 power conversion system, the achievable plant thermal efficiency, core power density and core specific powers become the dominant factors. The potential to achieve the highest efficiency among the four reactor concepts can be ranked from highest to lowest as follows: (1) GFR, (2) LFR and LSFR, and (3) SFR. Both the lead- and salt-cooled designs achieve about 30% higher power density than the gas-cooled reactor, but attain power density 3 times smaller than that of the sodium-cooled reactor. Fuel cycle costs are favored for the sodium reactor by virtue of its high specific power of 65kW/kgHM compared to the lead, salt and gas reactor values of 45, 35, and 21kW/kgHM, respectively. In terms of safety, all concepts can be designed to accommodate the unprotected limiting accidents through passive means in a self-controllable manner. However, it does not seem to be a preferable option for the GFR where the active or semi-passive approach will likely result in a more economic and reliable plant. Lead coolant with its superior neutronic characteristics and the smallest coolant temperature reactivity coefficient is easiest to design for self-controllability, while the LSFR requires special reactivity devices to overcome its large positive coolant temperature coefficient. The GFR required a special core design using BeO diluent and a supercritical CO2 reflector to achieve negative coolant void worth—one of the conditions necessary for inherent shutdown following large LOCA. Protected accidents need to be given special attention in the LSFR and LFR due to the small margin to freezing of their coolants, and to a lesser extent in the SFR.

Conceptual designs of lead-cooled and liquid salt-cooled fast flexible conversion ratio reactors were developed. The performance achievable by the unity conversion ratio cores of these reactors was compared to an existing supercritical carbon dioxide-cooled (S-CO2) fast reactor design and an uprated version of an existing sodium-cooled fast reactor. All concepts have cores rated at 2400MWt. The cores of the liquid-cooled reactors are placed in a large-pool-type vessel with dual-free level, which also contains four intermediate heat exchangers (IHXs) coupling a primary coolant to a compact and efficient supercritical CO2 Brayton cycle power conversion system. The S-CO2 reactor is directly coupled to the S-CO2 Brayton cycle power conversion system. Decay heat is removed passively using an enhanced reactor vessel auxiliary cooling system (RVACS) and a passive secondary auxiliary cooling system (PSACS). The selection of the water-cooled versus air-cooled heat sink for the PSACS as well as the analysis of the probability that the PSACS may fail to complete its mission was performed using risk-informed methodology. In addition to these features, all reactors were designed to be self-controllable. Further, the liquid-cooled reactors utilized common passive decay heat removal systems whereas the S-CO2 uses reliable battery powered blowers for post-LOCA decay heat removal to provide flow in well defined regimes and to accommodate inadvertent bypass flows. The multiple design limits and challenges which constrained the execution of the four fast reactor concepts are elaborated. These include principally neutronics and materials challenges. The neutronic challenges are the large positive coolant reactivity feedback, small fuel temperature coefficient, small effective delayed neutron fraction, large reactivity swing and the transition between different conversion ratio cores. The burnup, temperature and fluence constraints on fuels, cladding and vessel materials are elaborated for three categories of material – materials currently available, available on a relatively short time scale and available only with significant development effort. The selected fuels are the metallic U–TRU–Zr (10% Zr) for unity conversion ratio and TRU–Zr (75% Zr) for zero conversion ratio. The principal selected cladding and vessel materials are HT-9 and A533 or A508, respectively, for current availability, T-91 and 9Cr–1Mo steel for relatively short-term availability and oxide dispersion strengthened ferritic steel (ODS) available only with significant development.

This paper presents the neutronic design of a liquid salt cooled fast reactor with flexible conversion ratio. The main objective of the design is to accommodate interchangeably within the same reactor core alternative transuranic actinides management strategies ranging from pure burning to self-sustainable breeding. Two, the most limiting, core design options with unity and zero conversion ratios are described. Ternary, NaCl-KCl-MgCl2 salt was chosen as a coolant after a rigorous screening process, due to a combination of favourable neutronic and heat transport properties. Large positive coolant temperature reactivity coefficient was identified as the most significant design challenge. A wide range of strategies aiming at the reduction of the coolant temperature coefficient to assure self-controllability of the core in the most limiting unprotected accidents were explored. However, none of the strategies resulted in sufficient reduction of the coolant temperature coefficient without significantly compromising the core performance characteristics such as power density or cycle length. Therefore, reactivity control devices known as lithium thermal expansion modules were employed instead. This allowed achieving all the design goals for both zero and unity conversion ratio cores. The neutronic feasibility of both designs was demonstrated through calculation of reactivity control and fuel loading requirements, fluence limits, power peaking factors, and reactivity feedback coefficients.