https://hdl.handle.net/1721.1/67472 2026-04-09T02:27:22Z https://hdl.handle.net/1721.1/77615 Feasibility of Breeding in Hard Spectrum Boiling Water Reactors with Oxide and Nitride Fuels Feng, Bo; Kazimi, Mujid S.; Forget, Benoit This study assesses the neutronic, thermal-hydraulic, and fuel performance aspects of using nitride fuel in place of oxides in Pu-based high conversion light water reactor designs. Using the higher density nitride fuel hardens the neutron energy spectrum and results in higher breeding ratios. The state-of-the-art high conversion light water reactor, the Resource-renewable Boiling Water Reactor (RBWR), served as the template core upon which comparative studies between nitride and oxide fuels were performed. A 1/3 core reactor physics model was developed for the RBWR using the stochastic transport code MCNP. The code was coupled with a lumped channel thermal-hydraulics 5-channel model for steady-state analyses. The depletion code MCODE, which links MCNP with ORIGEN, was used for all burnup calculations. Select physics parameters were calculated and with the exception of the void coefficients, agreed with reported data. The void coefficients of the coupled core were calculated to be slightly positive using two different methods (10% power increase and 5% flow reduction). The standard RBWR assembly designs, which use tight lattice hexagonal fuel rod arrays, with oxide fuel were then replaced with various nitride fuel assembly designs to determine the potential increase in breeding ratio, the potential to breed with pressurized water, and the potential to improve the critical power ratio with a wider pin pitch. Without changing the assembly geometry or discharge burnup, using nitride fuel resulted in a breeding ratio of 1.14. Using single-phase liquid water, the nitride fuel RBWR assembly resulted in a conversion ratio of 1.00. Another nitride fuel assembly design with boiling water maintained a 1.04 breeding ratio while increasing the pitch-to-diameter ratio from 1.13 to 1.20. This modification increased the hot assembly critical power ratio from 1.22 to 1.36, as calculated using the Liu- 2007 correlation. A high-porosity nitride fuel is recommended for high burnup conditions, to accommodate the nitride fuel’s higher swelling and less favorable mechanical properties compared to the oxide fuel. The high porosity allows additional volume for pressure-induced densification, alleviating swelling and subsequent cladding strain. To predict the performance of high-porosity nitride fuel, fission gas and fuel behavior mechanistic models were developed for high burnup and low-temperature conditions. These models were validated with reported irradiation data and implemented, along with fuel material properties, into the steady-state fuel behavior code FRAPCON-EP. Under simulated RBWR conditions, a fuel density no more than 85% of theoretical density is recommended to maintain satisfactory fuel performance. 2011-06-01T00:00:00Z https://hdl.handle.net/1721.1/77614 PROLIFERATION RESISTANT, LOW COST, THORIA-URANIA FUEL FOR LIGHT WATER REACTORS Kazimi, Mujid S.; Driscoll, Michael J.; Ballinger, Ronald G.; Clarno, K. T.; Czerwinski, Kenneth R.; Hejzlar, Pavel; LaFond, P. J.; Long, Y.; Meyer, J. E.; Reynard, M. P.; Schultz, S. P.; Zhao, X. 1. Summary Project Objectives: Our objective is to develop a fuel consisting of mixed thorium dioxide and uranium dioxide (ThO[subscript 2]-UO[subscript 2]) for existing light water reactors (LWRs) that (a) is less expensive overall than the current uranium-dioxide (UO[subscript 2]) fuel, (b) is very resistant to nuclear weapons-material proliferation, (c) results in a more stable and insoluble waste form, and, (d) generates less spent fuel per unit energy production. This project is being conducted in collaboration with INEEL. This annual report presents the MIT progress in the investigations from October 1998 up to June 1999. 1999-06-01T00:00:00Z https://hdl.handle.net/1721.1/75738 High Performance Fuel Design for Next Generation PWRs: 11th Quarterly Report Kazimi, Mujid S.; Hejzlar, Pavel; Feng, Dandong; Kohse, Gordon E.; Morra, Paolo; Ostrovsky, Yakov; Saha, Pradip; Xu, Zhiwen; Yuan, Yi; Carpenter, David M.; Feinroth, Herbert; Lahoda, Edward J.; Sundaram, Ramu K.; Hamilton, Holly I. Technical Narrative: The overall objective of this NERI project is to examine the potential for a high performance advanced fuel for Pressurized Water Reactors (PWRs), which would accommodate a substantial increase of core power density while simultaneously providing larger thermal margins than current PWRs. This advanced fuel will have an annular geometry that allows internal and external coolant flow and heat removal. The project is led by the Massachusetts Institute of Technology (MIT), with collaboration of four industrial partners – Gamma Engineering Corporation, Westinghouse Electric Corporation, Framatome ANP (formerly Duke Engineering & Services), and Atomic Energy of Canada Limited. Quarterly Report for Project DE-FG03-01SF22329 April 2004 – June 2004 2004-07-01T00:00:00Z https://hdl.handle.net/1721.1/75737 Flexible Conversion Ratio Fast Reactor Systems Evaluation Final Report Todreas, Neil E.; Hejzlar, Pavel; Fong, Chris J.; Nikiforova, Anna; Petroski, Robert; Shwageraus, Eugene; Whitman, Joshua Executive Summary: The goal of this project is to develop the conceptual designs of fast flexible conversion ratio reactors using lead and liquid salt coolants and to compare the results with a gascooled fast reactor developed in an MIT NERI project and a sodium-cooled reactor under development at ANL. To maintain the scope of the study manageable within the 2-year time frame and funding constraints, core designs that fit in the same reactor plant were executed for two limiting conversion ratios: (1) near zero, to transmute legacy waste and (2) near unity, to operate in a sustainable closed cycle. To reap the benefits of economy of scale, a large power rating of 2400MWt was set as the target thermal power for both reactor designs. In addition, the achievement of inherent reactor shutdown in unprotected accidents (without scram) was set as a desirable goal. Project DE-FC07-06ID14733 2008-06-01T00:00:00Z https://hdl.handle.net/1721.1/75736 High Performance Fuel Design for Next Generation PWRs: Final Report Kazimi, Mujid S.; Hejzlar, Pavel; Carpenter, David M.; Feng, Dandong; Kohse, Gordon E.; Lee, Won Jae; Morra, Paolo; No, Hee Cheon; Ostrovsky, Yakov; Otsuka, Yasuyuki; Saha, Pradip; Shwageraus, Eugene; Xu, Zhiwen; Yuan, Yi; Zhang, Jiyun; Feinroth, Herbert; Hao, Bernard; Lahoda, Edward J.; Mazzoccoli, Jason P.; Sundaram, Ramu K.; Hamilton, Holly This summary provides an overview of the results of the U.S. DOE funded NERI (Nuclear Research Energy Initiative) program on development of the internally and externally cooled annular fuel for high power density PWRs. This new fuel was proposed by MIT to allow a substantial increase in power density (on the order of 30% or higher) while maintaining or improving safety margins. A comprehensive study was performed by a team consisting of MIT (lead organization), Westinghouse Electric Corporation, Gamma Engineering Corporation, Framatome ANP (formerly Duke Engineering) and Atomic Energy of Canada Limited. The study involved the evaluation of the new fuel in terms of thermal hydraulic, neutronics, fuel performance including first scoping irradiation tests at the MIT reactor, fuel manufacturing and economics. Project DE-FG03-01SF22329 2006-01-01T00:00:00Z https://hdl.handle.net/1721.1/75735 High Performance Fuel Design for Next Generation PWRs Appendices B-I to FY-02 Annual Report Kazimi, Mujid S.; Hejzlar, Pavel; Ballinger, Ronald G.; Carpenter, David M.; Feng, Dandong; Kohse, Gordon E.; Lee, Won Jae; No, Hee Cheon; Ostrovsky, Yakov; Otsuka, Yasuyuki; Stahl, Peter; Xu, Zhiwen; Yuan, Yi; Zhang, Jiyun; Feinroth, Herbert; Hao, Bernard; Lahoda, Edward J.; Mazzoccoli, Jason P.; Sundaram, Ramu K.; Hamilton, Holly B.1.1 VIPRE modeling of PWR core with annular fuel: Optimization studies in the first year used an isolated channel and models for MDNBR analyses. These analyses provided sufficient knowledge of potential thermal hydraulic performance of annular fuels to select the 13x13 array as the most promising configuration. To obtain more realistic and accurate MDNBR, a whole core model is necessary. In particular, the major concern is correct representation of channel flow rate. The earlier models used the core-average mass flux, which does not account for flow rate reduction in the hot channels due to increased pressure drop in this channel as a result of higher subcooled, or possibly, saturated boiling. Therefore, it is expected that the MDNBR obtained from the full core VIPRE-01 model will be smaller than the values obtained from the isolated channel model. Progress Report for Work August 2001 through July 2002 2003-08-01T00:00:00Z https://hdl.handle.net/1721.1/75734 High Performance Fuel Design for Next Generation PWRs 2nd Annual Report Kazimi, Mujid S.; Hejzlar, Pavel; Ballinger, Ronald G.; Carpenter, David M.; Feng, Dandong; Kohse, Gordon E.; Lee, Won Jae; No, Hee Cheon; Ostrovsky, Yakov; Otsuka, Yasuyuki; Stahl, Peter; Xu, Zhiwen; Yuan, Yi; Zhang, Jiyun; Feinroth, Herbert; Hao, Bernard; Lahoda, Edward J.; Mazzoccoli, Jason P.; Sundaram, Ramu K.; Hamilton, Holly The overall objective of this NERI project is to examine the potential for a high performance advanced fuel design for Pressurized Water Reactors (PWRs), which would accommodate a substantial increase of core power density while simultaneously providing larger thermal margins than current PWRs. This advanced fuel employs an annular geometry that allows internal and external coolant flow and heat removal. The project is led by the Massachusetts Institute of Technology (MIT), with the collaboration of four industrial partners – Gamma Engineering Corporation, Westinghouse Electric Corporation, Framatome ANP DE & S (formerly Duke Engineering & Services), and Atomic Energy of Canada Limited. The project is organized into five tasks: 1. Task 1 Assess the thermal hydraulic performance of the internally and externally cooled annular fuel to identify the configuration with the highest potential for power density increase while maintaining ample thermal margins, as well as key aspects of mechanical design to ensure that new fuel will not perform outside established hydraulic and mechanical constraints, 2. Task 2 Determine the neutronic performance of the new fuel, and the design that will minimize fuel cycle cost and assures that reactor physics safety parameters are as good or better than those of current PWRs, 3. Task 3 Explore various methods of manufacturing of this advanced fuel, including new innovative fabrication processes to produce annular fuel elements with the required product characteristics, 4. Task 4 Evaluate fuel cycle cost and capital cost implications of high power density to determine the economic viability of the high-performance fuel, and 5. Task 5 Analyze fuel performance of the new UO2 annular fuel obtained by various production technologies including irradiation testing in the MIT reactor. Progress Report for Work August 2002 through July 2003 2003-08-01T00:00:00Z https://hdl.handle.net/1721.1/75291 Vented Inverted Fuel Assembly Design for an SFR Vitillo, Francesco; Todreas, Neil E.; Driscoll, Michael J. The goal of this work is to investigate the feasibility of a vented inverted fuel assembly for a sodium-cooled fast reactor. The inverted geometry has been previously investigated for application in Gas-cooled Fast Reactors since it improves thermal-hydraulic and neutronic performance of those reactors. Venting is a concept studied during the past and its major past application in sodiumcooled fast reactors was in the Dounreay Fast Reactor in the United Kingdom. In this work the inverted assembly approach was adopted because it allows high fuel volume fraction, reduction of the coolant void reactivity, less neutron leakage, the reduction of the enrichment and lower pressure drop for the same channel length because grids nor wire wraps are no longer necessary. However all results of this work apply also to venting of conventional fuel pins. Performance criteria for vented fuel assemblies in term of materials, thermalhydraulics and venting systems have been investigated in order to set design goals. In particular, for the materials, a limit for maximum cladding surface temperature, cladding and other core internal structure fluence and maximum fuel temperature in the hot channel has been identified. For the thermal-hydraulic analysis, the goals are increasing fuel volume fraction, keeping the fuel and the cladding surface temperature as low as possible compared with those of a similar power rating core and minimizing core pressure drops. Regarding the venting system the design goals are retaining as much 137Cs in an upper plenum and keeping the overall assembly height within the values of current technology for a reactor of similar size. Therefore the height of an upper plenum (which must contain sodium bond volume expelled due to fuel thermal expansion, sodium bond volume due to its thermal expansion and the cesium volume of a single assembly if the cesium is completely released into the plenum) has been determined. Investigation of physical and chemical behavior of volatile fission products in sodium is presented, in order to determine the maximum activity inventory which would eventually be released into the primary sodium. Assumptions for the simplified approach adopted are discussed. Results of this analysis show that the most troublesome radionuclides in terms of propensity to escape from the venting system (due to their half-life being longer than a threshold time chosen based on physical behavior of escaping fission products: bubbling out for gases and pure diffusion for other volatile elements) are noble gases (85Kr and 133Xe), cesium (134Cs and 137Cs) and tritium (3H). For the thermal-hydraulic analysis a comparison between a pin-type fuel assembly and three inverted fuel assemblies with different parameters has been made, in order to demonstrate benefits of such a concept and to determine the best configuration. In particular attention is on core pressure drop, fuel and cladding temperature given the mass flow rate and assembly power. The results show that the best configuration has the same core pressure drop and hence pumping power and the same total active fuel length of a similar performance pin-type core. A final vented inverted fuel assembly design is proposed, which meets all the design goals. Such a configuration lets volatile radionuclides with short enough half-lives completely decay before release or be released in a negligible quantity after an infinite time of diffusion in sodium. Longer lived fission products will be released into the coolant, while fission gases will be vented first into the sodium and eventually to the cover gas after bubbling up through the sodium itself. Methods for purifying cover gas and coolant from vented radionuclides are proposed as well as storage systems for radioactive materials from the purification process. Results show that charcoal is the best absorber for noble gases whereas cold traps can be usefully used to remove cesium and tritium from primary sodium. Noble gases are produced in a (conservatively estimated) quantity of 38 m3/year (at STP) at core end-of-life and can be stored in adsorbent packed cylinders. Materials in cold traps are chemically treated to obtain liquid waste. Hence they can be converted into a solid and then stored in Pyrex glass. Finally a review of materials with regard to increasing the coolant core outlet temperature is given: in particular HT9 cladding and various ex-core structural materials. It has been shown that, with regard to cladding material limits, venting can provide at least a 20°C increase in the core outlet temperature since venting decreases mechanical stress on the cladding due to fission gas pressure. Also, based on current designs and experience high-chromium steels are very promising candidates for ex-core structural material (e.g piping), together with ODS (if their chemical compatibility with liquid sodium and weldability are verified): the latter can operate at about 600°C still keeping a margin of 100°C from the upper temperature limit. Based on the present analysis is that the ex-core structural material limit is a more limiting factor than the cladding material limit with regard to increasing the coolant core outlet temperature. In conclusion it has been demonstrated that the vented inverted fuel assembly configuration is an interesting and valuable concept to take into account for future investigation in order to improve the performance of sodium-cooled fast reactors. 2011-06-01T00:00:00Z https://hdl.handle.net/1721.1/75290 Effects of Surface Parameters on Boiling Heat Transfer Phenomena Truong, Bao Hoai; Hu, Lin-wen; Buongiorno, Jacopo; McKrell, Thomas J. Nanofluids, engineered colloidal dispersions of nanoparticles in fluid, have been shown to enhance pool and flow boiling CHF. The CHF enhancement was due to nanoparticle deposited on the heater surface, which was verified in pool boiling. However, no such work has been done for flow boiling. Using a cylindrical tube pre-coated with Alumina nanoparticles coated via boiling induced deposition, CHF of water was found to enhance up to 40% compared to that of the bare tube. This confirms that nanoparticles on the surface is responsible for CHF enhancement for flow boiling. However, existing theories failed to predict the CHF enhancement and the exact surface parameters attributed to the enhancement cannot be determined. Surface modifications to enhance critical heat flux (CHF) and Leidenfrost point (LFP) have been shown successful in previous studies. However, the enhancement mechanisms are not well understood, partly due to many surface parameters being altered at the same time, as in the case for nanofluids. Therefore, the remaining objective of this work is to evaluate separate surface effect on different boiling heat transfer phenomena. In the second part of this study, surface roughness, wettability and nanoporosity were altered one by one and respective effect on quenching LFP with water droplet was determined. Increase in surface roughness and wettability enhanced LFP; however, nanoporosity was most effective in raising LFP, almost up to 100ºC. The combination of the micro posts and nanoporous coating layer proved optimal. The nanoporous layer destabilizes the vapor film via heterogeneous bubble nucleation, and the micro posts provides intermittent liquid-surface contacts; both mechanisms increase LFP. In the last part, separate effect of nanoporosity and surface roughness on pool boiling CHF of a well-wetting fluid, FC-72, was investigated. Nanoporosity or surface roughness alone had no effect on pool boiling CHF of FC-72. Data obtained in the literature mostly for microporous coatings showed CHF enhancement for well wetting fluids, and existing CHF models are unable to predict the enhancement. 2011-06-01T00:00:00Z https://hdl.handle.net/1721.1/75289 General Analysis of Breed-and-Burn Reactors and Limited-Separations Fuel Cycles Petroski, Robert C.; Forget, Benoit A new theoretical framework is introduced, the “neutron excess” concept, which is useful for analyzing breed-and-burn (B&B) reactors and their fuel cycles. Based on this concept, a set of methods has been developed which allows a broad comparison of B&B reactors using different fuels, structural materials, and coolants. This new approach allows important reactor and fuel-cycle parameters to be approximated quickly, without the need for a full core design, including minimum burnup/irradiation damage and reactor fleet doubling time. Two general configurations of B&B reactors are considered: a “minimum-burnup” version in which fuel elements can be shuffled in three dimensions, and a “linear-assembly” version composed of conventional linear assemblies that are shuffled radially. Based on studies of different core compositions, the best options for minimizing fuel burnup and material DPA are metal fuel (with a strong dependence on alloy content), the type of steel that allows the lowest structure volume fraction, and helium coolant. If sufficient fuel performance margin exists, sodium coolant can be substituted in place of helium to achieve higher power densities at a modest burnup and DPA penalty. For a minimum-burnup B&B reactor, reasonably achievable minimum DPA values are on the order of 250-350 DPA in steel, while axial peaking in a linear-assembly B&B reactor raises minimum DPA to over 450 DPA. By recycling used B&B fuel in a limited-separations (without full actinide separations) fuel cycle, there is potential for sodium-cooled B&B reactors to achieve fleet doubling times of less than one decade, although this result is highly sensitive to the reactor core composition employed as well as thermal hydraulic performance. 2011-02-01T00:00:00Z https://hdl.handle.net/1721.1/75288 Design of a Functionally Graded Composite for Service in High Temperature Lead and Lead-Bismuth Cooled Nuclear Reactors Short, Michael P.; Ballinger, Ronald G. A material that resists lead-bismuth eutectic (LBE) attack and retains its strength at 700°C would be an enabling technology for LBE-cooled reactors. No single alloy currently exists that can economically meet the required performance criteria of high strength and corrosion resistance. A Functionally Graded Composite (FGC) was created with layers engineered to perform these functions. F91 was chosen as the structural layer of the composite for its strength and radiation resistance. Fe-12Cr- 2Si, an alloy developed from previous work in the Fe-Cr-Si system, was chosen as the corrosion-resistant cladding layer because of its chemical similarity to F91 and its superior corrosion resistance in both oxidizing and reducing environments. Fe-12Cr-2Si experienced minimal corrosion due to its self-passivation in oxidizing and reducing environments. Extrapolated corrosion rates are below one micron per year at 700°C. Corrosion of F91 was faster, but predictable and manageable. Diffusion studies showed that 17 microns of the cladding layer will be diffusionally diluted during the three year life of fuel cladding. 33 microns must be accounted for during the sixty year life of coolant piping. 5 cm coolant piping and 6.35 mm fuel cladding were produced on a commercial scale by weld-overlaying Fe-12Cr-2Si onto F91 billets and co-extruding them, followed by pilgering. An ASME certified weld was performed followed by the prescribed quench-and-tempering heat treatment for F91. A minimal heat affected zone was observed, demonstrating field weldability. Finally, corrosion tests were performed on the fabricated FGC at 700°C after completely breaching the cladding in a small area to induce galvanic corrosion at the interface. None was observed. This FGC has significant impacts on LBE reactor design. The increases in outlet temperature and coolant velocity allow a large increase in power density, leading to either a smaller core for the same power rating or more power output for the same size core. This FGC represents an enabling technology for LBE cooled fast reactors. 2010-10-01T00:00:00Z https://hdl.handle.net/1721.1/75281 Conceptual Design of an Annular-Fueled Superheat Boiling Water Reactor Ko, Yu-Chih; Kazimi, Mujid S. The conceptual design of an annular-fueled superheat boiling water reactor (ASBWR) is outlined. The proposed design, ASBWR, combines the boiler and superheater regions into one fuel assembly. This ensures good neutron moderation throughout the reactor core. A single fuel design is used in the core. Each annular fuel element, or fuel tube, is cooled externally by boiling water and internally by steam. Fuel pellets are made of low enrichment UO2, somewhat higher than the traditional BWR fuel enrichment. T91 and Inconel 718 are selected as candidates for the cladding material in view of their excellent physical properties and corrosion resistance. The fuel-cladding gap is filled with pressurized helium gas, like the existing lighter water reactor fuels. The ASBWR fuel assembly contains sixty annular fuel elements and one square water rod (occupying a space of four fuel elements) in an 8 by 8 square array. Annular separators and steam dryers are utilized and located above the core in the reactor vessel. Reactor internal pumps are used to adjust the core flow rate. Cruciform control rods are used to control the reactivity of the core, but more of them may be needed than a traditional BWR in view of the harder spectrum. The major design constraints have been identified and evaluated in this work. The ASBWR is found promising to achieve a power density of 50 kW/L and meet all the main safety requirements. This includes a limit on the minimum critical heat flux ratio, maximum fuel and cladding operating temperatures, and appropriate stability margin against density wave oscillations. At the expected superheated steam of 520 °C, the plant efficiency is above 40%, which is substantially greater than the efficiency of 33 to 35% that today’s generation of LWRs can achieve. In addition to generating electricity, the ASBWR may also be useful for liquid fuel production or other applications that require high temperature steam. The uncertainties about this design include the performance of cladding materials under irradiation, the attainment of desirable heat transfer ratio between the external and􀀃internal coolant channels throughout the fuel cycle, and the response to the traditional transients prescribed as design basis events. 2010-10-01T00:00:00Z https://hdl.handle.net/1721.1/75280 Use of Response Surface for Evaluation of Functional Failure of Passive Safety System Fei, Tingzhou; Golay, Michael W. Passive safety systems are more vulnerable to their environment and initial condition due to the typical low driving forces, e.g., that of the natural circulation. We investigate the merits of different methods for analysis of the probability of passive safety system “functional failure”. Variation of the coolant flow condition due to complex thermal hydraulic phenomena may cause a passive safety system to be unable to perform its function. In this report the RELAP5 code is used with normally distributed input parameters to estimate the functional failure probability of the passive system. Response surfaces are generated from RELAP5 results using different sampling techniques. Comparison between response surface and RELAP5 shows that the standard deviations are different. We identify sufficient levels of simulation effort required for accurate estimates. 2010-03-01T00:00:00Z https://hdl.handle.net/1721.1/75279 Thermal-Hydraulic Analysis of Innovative Fuel Configurations for the Sodium Fast Reactor Memmott, Matthew J.; Hejzlar, Pavel; Buongiorno, Jacopo The sodium fast reactor (SFR) is currently being reconsidered as an instrument for actinide management throughout the world, thanks in part to international programs such as the Generation-IV and especially the Global Nuclear Energy Partnership (GNEP). The success of these programs, in particular the GNEP, is dependent upon the ability of the SFR to manage actinide inventory while remaining economically competitive. In order to achieve these goals, the fuel must be able to operate reliably at high power densities. However, the power density of the fuel is limited by fuel-clad chemical interaction (FCCI) for metallic fuel, cladding thermal and irradiation strain, the fuel melting point, sodium boiling, and to a lesser extent the sodium pressure drop in the fuel channels. Therefore, innovative fuel configurations that reduce clad stresses, sodium pressure drops, and fuel/clad temperatures could be applied to the SFR core to directly improve the performance and economics. Two particular designs of interest that could potentially improve the performance of the SFR core are the internally and externally cooled annular fuel and the bottle-shaped fuel. In order to evaluate the thermal-hydraulic performance of these fuels, the capabilities of the RELAP5-3D code have been expanded to perform subchannel analysis in sodiumcooled fuel assemblies with non-conventional geometries. This expansion was enabled by the use of control variables in the code. When compared to the SUPERENERGY II code, the prediction of core outlet temperature agreed within 2%. In addition, the RELAP5-3D subchannel model was applied to the ORNL 19-pin test, and it was found that the code could predict the measured outlet temperature distribution with a maximum error of ~8%. As an application of this subchannel model, duct ribs were explored as a means of reducing core outlet temperature peaking within the fuel assemblies. The performance of the annular and bottle-shaped fuel was also investigated using this subchannel model. The annular fuel configurations are best suited for low conversion ratio cores. The magnitude of the power uprate enabled by metal annular fuel in the CR = 0.25 cores is 20%, and is limited by the FCCI constraint during a hypothetical flow blockage of the inner-annular channel due to the small diameters of the inner-annular flow channel (3.6 mm). On the other hand, a complete blockage of the hottest inner-annular flow channel in the oxide fuel case results in sodium boiling, which renders the annular oxide fuel concept unacceptable for use in a SFR. The bottle-shaped fuel configurations are best suited for high conversion ratio cores. In the CR = 0.71 cores, the bottle-shaped fuel configuration reduces the overall core pressure drop in the fuel channels by up to 36.3%. The corresponding increase in core height with bottle-shaped fuel is between 15.6% and 18.3%. A full-plant RELAP5-3D model was created to evaluate the transient performance of the base and innovative fuel configurations during station blackout and UTOP transients. The transient analysis confirmed the good thermal-hydraulic performance of the annular and bottle-shaped fuel designs with respect to their respective solid fuel pin cases. 2009-08-01T00:00:00Z https://hdl.handle.net/1721.1/75278 Critical Heat Flux Enhancement via Surface Modification Using Colloidal Dispersions of Nanoparticles (Nanofluids) Truong, Bao H.; Hu, Lin-Wen; Buongiorno, Jacopo Nanofluids are engineered colloidal dispersions of nanoparticles (1-100nm) in common fluids (water, refrigerants, or ethanol…). Materials used for nanoparticles include chemically stable metals (e.g., gold, silver, copper), metal oxides (e.g., alumina, zirconia, silica, titania) and carbon in various forms (e.g., diamond, graphite, carbon nanotubes). The attractive properties of nanofluids include higher thermal conductivity, heat transfer coefficients (HTC) and boiling critical heat flux (CHF) than that of the respective base fluid. Nanofluids have been found to exhibit a very significant enhancement up to 200% of the boiling CHF at low nanoparticle concentrations. In this study, nanofluids were investigated as an agent to modify a heater surface to enhance Critical Heat Flux (CHF). First, the CHF of diamond, Zinc Oxide and Alumina water-based nanofluids at low volume concentration (<1 vol%) were measured to determine if nanofluid enhances CHF as seen in literature. Subsequently, the heaters are coated with nanoparticles via nucleate boiling of nanofluids. The CHF of water was measured using these nanoparticle precoated heaters to determine the magnitude of the CHF enhancement. Characterization of the heaters after CHF experiments using SEM, confocal, and contact angle were conducted to explain possible mechanisms for the observed enhancement. The coating thickness of the nanoparticle deposition on a wire heater as a function of boiling time was also investigated. Finally, theoretical analyses of the maximum CHF and HTC enhancement in term of wettability were performed and compared with the experimental data. The CHF of nanofluids was as much as 85% higher than that of water, while the nanoparticle pre-coated surfaces yielded up to 35% CHF enhancement compared to bare heaters. Surface characterization of the heaters after CHF experiments showed a change in morphology due to the nanoparticles deposition. The coating thickness of nanoparticle was found to deposit rather quickly on the wire surface. Within five minutes of boiling, the coating thickness of more than 1 μm was achieved. Existing CHF correlations overestimated the experimental data. 2008-06-01T00:00:00Z https://hdl.handle.net/1721.1/75277 Methods for Comparative Assessment of Active and Passive Safety Systems Oh, Jiyong; Golay, Michael W. Passive cooling systems sometimes use natural circulation, and they are not dependent on offsite or emergency AC power, which can simplify designs through the reduction of emergency power supplying infrastructure. The passive system approach can lead to substantial simplification of the system as well as overall economic benefits, and passive systems are believed to be less vulnerable to accidents by component failures and human errors compared to active systems. The viewpoint that passive system design is more reliable and more economical than active system design has become generally accepted. However, passive systems have characteristics of a high level of uncertainty and low driving force for purposes of heat removal phenomena; these characteristics can result in increasing system unreliability and may raise potential remedial costs during a system’s lifetime. This study presents a comprehensive comparison of reliability and cost taking into account uncertainties and introduces the concept of flexibility using the example of active and passive residual heat removal systems in a PWR. The results show that the active system can have, for this particular application, greater reliability than the passive system. Because the passive system is economically optimized, its heat removal capacity is much smaller than that of the active system. Thus, functional failure probability of the passive system has a greater impact on overall system reliability than the active system. Moreover, considering the implications of flexibility upon remedial costs, the active system may be more economical than the passive system because the active system has flexible design features for purposes of increasing heat removal capacity. 2008-02-01T00:00:00Z https://hdl.handle.net/1721.1/75276 Stability Analysis of Natural Circulation in BWRs at High Pressure Conditions Hu, Rui; Kazimi, Mujid S. At rated conditions, a natural circulation boiling water reactor (NCBWR) depends completely on buoyancy to remove heat from the reactor core. This raises the issue of potential unstable flow oscillations. The objective of this work is to assess the characteristics of stability in a NCBWR at rated conditions, and the sensitivity to design and operating conditions in comparison to previous BWRs. Two kinds of instabilities, namely Ledinegg flow excursion and Density Wave Oscillations (DWO), have been studied. The DWO analyses were conducted for three oscillation modes: Single Channel thermal-hydraulic stability, coupled neutronics regionwide out-of-phase stability and core-wide in-phase stability. Using frequency domain methods, the three types of DWO stability characteristics of the NCBWR and their sensitivity to the operating parameters and design features have been determined. The characteristic equations are constructed from linearized equations, which are derived for small deviations around steady operating conditions. The Economic Simplified Boiling Water Reactor (ESBWR) is used in our analysis as a reference NCBWR design. It is found that the ESBWR can be stable with a large margin around the operating conditions by proper choice of the core inlet orifice scheme, and for appropriate power to flow ratios. In single channel stability analysis, neutronic feedback is neglected. Design features of the ESBWR, including shorter fuel bundle and use of part-length rods in the assemblies, tend to improve the thermal-hydraulic stability performance. However, the thermalhydraulic stability margin is still lower than that of a typical BWR at rated conditions. In neutronic-coupled out-of-phase as well as in-phase stability analysis, the perturbation decay ratios for ESBWR at our assumed conditions are higher than that of a typical BWR (Peach Bottom 2) at rated conditions, due to its lower thermal-hydraulic stability margin and higher neutronic feedback. Nevertheless, the stability criteria are satisfied. To evaluate the NCBWR stability performance, comparison with BWR/Peach Bottom 2 at both the rated condition and maximum natural circulation condition has been conducted. Sensitivity studies are performed on the effects of design features and operating parameters, including chimney length, inlet orifice coefficient, power, flow rate, and axial power distribution, reactivity coefficients, fuel pellet-clad gap conductance. It can be concluded that the NCBWR and BWR stabilities are similarly sensitive to operating parameters. 2007-10-01T00:00:00Z https://hdl.handle.net/1721.1/75275 Selection of Correlations and Look-Up Tables for Critical Heat Flux Prediction in the Generation IV "IRIS" Reactor Romano, A.; Todreas, Neil E. In order to fulfill the goals set forth by the Generation IV International Forum, the current NERI funded research has focused on the design of a Gas-cooled Fast Reactor (GFR) operating in a Breed and Burn (B&B) fuel cycle mode. B&B refers to a once-through fuel cycle where low enriched uranium (less than 5 w/o 235U in U) subcritical assemblies are loaded into the core in equilibrium, yet in-situ plutonium breeding carries the fuel through a discharge burnup on the order of 150 MWD/kgHM. The B&B fuel cycle meets the GenIV goals of sustainability, economics, and proliferation resistance by increasing fuel burnup without the need for spent fuel reprocessing, recycle, or reuse of any kind. The neutronic requirements for B&B are strict and require an ultra-hard neutron spectrum. Therefore, the GFR is ideally suited for this fuel cycle. In the present work the B&B GFR concept evolved into two practical reactor designs, both of which build on extensive previous gas-cooled reactor design experience. 2000-06-01T00:00:00Z https://hdl.handle.net/1721.1/75274 Feasibility of Very Deep Borehole Disposal of US Nuclear Defense Wastes Dozier, Frances E.; Driscoll, Michael J.; Buongiorno, Jacopo This report analyzes the feasibility of emplacing DOE-owned defense nuclear waste from weapons production into a permanent borehole repository drilled ~4 km into granite basement rock. Two canister options were analyzed throughout the report: the canister currently used by the DOE for vitrified defense waste and a reference canister with a smaller diameter. In a thermal analysis, the maximum temperatures attained by the rock surrounding the waste, waste form, canister, liner, and gaps during the post-emplacement period were calculated. From this data, simple analytic equations were formed that can be used to calculate the maximum temperature differences for both defense waste and spent fuel when one does not want to repeat the analysis. Canister corrosion and waste form dissolution analyses were performed using Pourbaix diagrams. Finally, the cost and time for drilling the borehole and emplacing the defense waste were calculated. The temperature change in the granite is 15.1°C for the reference canister and 45.7°C for the DOE Canister. The resulting maximum temperature at the bottom of the borehole is 135.1°C (reference canister) and 165.7°C (DOE canister) for the bounding defense waste. The centerline temperature for the borosilicate glass waste package is approximately 150°C for the reference canister and 207°C for the DOE canister. Because of the thermodynamic properties, overall corrosion resistance, and reasonable cost, pure copper was shown to be the best borehole outer canister material. High-chromium stainless steel could also be a good option for borehole canisters because it has been shown to be highly corrosion-resistant in environments similar to predicted borehole environments. Cesium ion was found to have the highest concentration in the borehole environment. However, the relatively low half life of the most abundant cesium isotope suggests that the cesium would decay before the canister is breached. For the reference canister, the drilling and emplacement costs are not expected to exceed $46/kg of vitrified waste and the total disposal cost was found to be $153/kg of vitrified waste. The total cost of disposal of defense waste in DOE containers is not expected to exceed $53/kg of vitrified waste. Based on these analyses, disposal of vitrified defense waste in deep boreholes is expected to be technically and economically feasible. 2011-06-01T00:00:00Z https://hdl.handle.net/1721.1/75273 Cross Section Generation Strategy for High Conversion Light Water Reactors Herman, Bryan R.; Shwageraus, Eugene; Forget, Benoit; Kazimi, Mujid S. High conversion water reactors (HCWR), such as the Resource-renewable Boiling Water Reactor (RBWR), are being designed with axial heterogeneity of alternating fissile and blanket zones to achieve a conversion ratio of greater than one and assure negative void coefficient of reactivity. This study assesses the generation of few-group macroscopic cross sections for neutron diffusion theory analyses of this type of reactor, in order to enable three-dimensional transient simulations. The goal is to minimize the number of energy groups in these simulations to reduce computational effort. A two-dimensional cross section generation methodology using the Monte Carlo code Serpent, similar to the traditional deterministic homogenization methodology, was used to analyze a single RBWR assembly. Results from two energy group and twelve energy group diffusion analyses showed an error in multiplication factor over 1000 pcm with errors in reaction rates between 10 and 60%. Therefore, the traditional approach is not sufficiently accurate. Instead, a three-dimensional homogenization methodology using Serpent was developed to account for neighboring zones in the homogenization process. A Python wrapper, SerpentXS, was developed to perform branch case calculations with Serpent to parametrize few-group parameters as a function of reactor operating conditions and to create a database for interpolation with the nodal diffusion theory code, PARCS. Diffusion analyses using this methodology also showed an error in multiplication factor over 1000 pcm. The three-dimensional homogenization capability in Serpent allowed for the introduction of axial discontinuity factors in the diffusion theory analysis, needed to preserve Monte Carlo reaction rates and global multiplication factor. A one-dimensional finite-difference multigroup diffusion theory code, developed in MATLAB, was written to investigate the use of axial discontinuity factors for a single RBWR assembly. The application of discontinuity factors on either side of each axial interface preserved multiplication factor and reaction rate estimates between transport theory and diffusion theory analyses to within statistical uncertainty. Use of this three-dimensional assembly homogenization approach in generating few-group macroscopic cross sections and axial discontinuity factors as a function of operating conditions will help further research in transient diffusion theory simulations of axially heterogeneous reactors. 2011-06-01T00:00:00Z https://hdl.handle.net/1721.1/75272 A Drop-In Concept for Deep Borehole Canister Emplacement Bates, Ethan A.; Buongiorno, Jacopo; Driscoll, Michael J. Disposal of high-level nuclear waste in deep boreholes drilled into crystalline bedrock (i.e., “granite”) is an interesting repository alternative of long standing. Work at MIT over the past two decades, and more recently in collaboration with the Sandia National Laboratory, has examined a broad spectrum of design aspects associated with this approach. For emplacement, past reports suggest using steel cables to lower each canister into the borehole. This process would require many years to complete and precise control to safely lower the canisters thousands of meters. The current study evaluated a simple, rapid, “passive” procedure for emplacement of canisters in a deep borehole: free-fall release into a water-flooded borehole. The project involves both analytic modeling and 1/5th scale experiments on a laboratory mockup. Experiments showed good agreement and validated the model. Depending on the inputs used for the mass and dimensions of the full scale canister and the viscosity of water, the model predicted terminal velocities of 2.4-2.6 m/s (4.5-5.8 mph). Further experiments showed that this could be reduced by 50% by making the surface hydraulically rough. Based on these predictions and a structural analysis, there seems to be little risk of damage when a canister reaches the bottom of the borehole or impacts the stack of previously loaded canisters. For reference, dropping the canister in air from a height of only 0.3 m (1 ft) would result in an impact velocity of 2.44 m/s. Cost estimates for the conventional drill string based method were developed, and the drop-in method was concluded to reduce emplacement costs and time by a minimum of 70%, down to $700,000 per borehole. It is concluded that a simple drop-in procedure deserves serious consideration for adoption as a standard procedure for borehole loading. 2011-06-01T00:00:00Z https://hdl.handle.net/1721.1/75271 PWR Cores with Silicon Carbide Cladding Kazimi, Mujid S.; Dobisesky, J.; Carpenter, David M.; Richards, J.; Pilat, Edward E.; Shwageraus, Evgeni The use in present generation PWRs of fuel clad with silicon carbide rather than Zircaloy has been evaluated as an aid to reaching higher discharge burnups and to operation at higher reactor power levels. A preliminary fuel design using fuel rods with the same dimensions as Westinghouse RFA fuel assemblies but with fuel pellets having 10 vol% central holes has been adopted. The central holes mitigate the higher fuel temperatures that occur due to the lower thermal conductivity of the silicon carbide, and the open gap between the fuel and cladding that persists over most of the irradiation. With this fuel design, it has been found possible to achieve 18 month cycles that meet present-day targets for peaking, boron concentration and shutdown margin while allowing average discharge burnups up to 80 MWD/KgU, as well as power uprates of 10% and possibly 20%. For non-uprated cores, the silicon carbide clad fuel has a clear economic advantage that increases with increasing discharge burnup. Even for comparable discharge burnups, there is a fuel cost savings of several million dollars per cycle as long as it does not increase the cost of fabrication by more than 50%, which seems highly unlikely. With 10-20% power uprates, the economics of the fuel cycle will improve, but the total value of such an uprate depends on the cost of needed plant modifications. Modifications to the control rod configuration or absorbing material may also be required to meet the shutdown margin criterion, particularly for the 20% uprate. Silicon carbide’s ability to sustain higher burnups and higher duty than Zircaloy also allows the design of a licensable two year cycle that has a fuel cost comparable to that of the reference 18 month Zircaloy core, and will furthermore reduce the average annual outage time. 2011-04-01T00:00:00Z https://hdl.handle.net/1721.1/75270 A Generalized Optimization Methodology for Isotope Management Massie, Mark Edward; Forget, Benoit This research focuses on developing a new approach to studying the nuclear fuel cycle: instead of employing the trial and error approach currently used in actinide management studies in which reactors are designed and then their performance is evaluated, the methodology developed here first identifies relevant fuel cycle objectives–like minimizing decay heat production in a repository, minimizing Pu-239 content in used fuel, etc.–and then uses optimization to determine the best way to reach these goals. The first half of this research was devoted to identifying optimal flux spectra for irradiating used nuclear fuel from light water reactors to meet fuel cycle objectives like those mentioned above. This was accomplished by applying the simulated annealing optimization methodology to a simple matrix exponential depletion code written in Fortran using cross sections generated from the SCALE code system. Since flux spectra cannot be shaped arbitrarily, the second half of this research applied the same methodology to material composition of fast reactor target assemblies to find optimal designs for minimizing the integrated decay heat production over various timescales. The neutronics calculations were performed using modules from SCALE and ERANOS, a French fast reactor transport code. 2010-09-01T00:00:00Z https://hdl.handle.net/1721.1/75269 Plugging of Deep Boreholes for HLW Disposal Jensen, K. G.; Driscoll, Michael J. This is a progress report covering work through July 2010 under a Sandia-MIT contract dealing with design and siting/licensing criteria for deep borehole disposal of spent nuclear fuel or its separated constituents. The principal focus is on conceptual design of a m ultilayer plug for our reference case borehole. It is similar to configurations recommended earlier by Swedish and Russian specialists. A secondary set of contributions update previous work in the areas such as multi-branch borehole configurations, use of cast iron canister inserts to resist crushing, and prospects for faster drilling. 2010-07-01T00:00:00Z https://hdl.handle.net/1721.1/75268 Deep Boreholes Attributes and Performance Requirements Driscoll, Michael J.; Jensen, K. G. This is a progress report covering work through mid-May 2010 under a Sandia-MIT contract dealing with design and siting/licensing criteria for deep borehole disposal of spent nuclear fuel or its separated constituents. It consists of additional short technical notes which scope out the performance-related requirements of a deep borehole repository. The most important changes since our last report are reversion to a single-branch vertical borehole (as recommended in the March 15 Workshop), and the consequential recommended adoption of a cast iron canister to alleviate the resulting bottom canister crushing threat. The case is also made that post closure nuclear criticality is not a credible scenario, with a very large margin of safety, even under very conservative assumptions. 2010-05-01T00:00:00Z https://hdl.handle.net/1721.1/75265 Assessment of helical-cruciform fuel rods for high power density Conboy, Thomas M.; McKrell, Thomas J.; Kazimi, Mujid S. In order to significantly increase the power density of Light Water Reactors (LWRs), the helical-cruciform (HC) fuel rod assembly has been proposed as an alternative to traditional fuel geometry. The HC assembly is a self-supporting nuclear fuel configuration consisting of 4-finned, axially-twisted fuel rods closely packed against one another in a square array. Within the LWR core, HC fuel would in theory possess several inherent advantages over traditional fuel, potentially allowing for operation at a higher power density. Chief among these advantages are a larger surface-to-volume ratio, a shorter radial heat conduction path, and improved mixing characteristics. In previous work, computational models of the HC fuel assembly have been of limited accuracy due to the absence suitable correlations. To address needs within these subchannel analysis models, experimental measurements of rod bundle coolant mixing have been conducted with 4x4 arrays of HC test rods. The tests used the technique of a hot water tracer injection (at 95°C) into a bulk flow of cold water (at 25°C). Downstream temperature measurements were used to judge the rate of lateral cross-flow within the HC rod bundle. These tests were conducted at atmospheric pressure, and encompassed a range of mass fluxes from 1000 kg/m2s to 3500 kg/m2s, HC rod twist pitches of 200cm, 100cm, and 50cm, and different hot water injection velocities and mixing lengths. Data from over 300 tests was analyzed, yielding a best fit correlation for use with any twist pitch, rod length, or coolant flow rate. Compared to the bare rod bundle, this correlation implies an enhancement in the intensity of turbulent interchange of 40% brought about by the HC geometry, and a 1.6% forced diversion of axial flow per subchannel, per quarterturn along the rod length. These parameters fit all data points considered within a standard deviation of 24%. Stochastic error was limited to ±16% by the use of precise temperature sensors. By applying this empirical mixing model to the subchannel representation of a BWR core featuring the HC rod design, a need to increase the flow area of the edge subchannels was demonstrated. This prompted a slight re-design of the HC fuel rod cross-section in order to make room for small spacer protrusions at the duct wall, to increase flow to peripheral subchannels. The modification was accomplished by reducing fin length, but increasing the inner diameter to maintain the reference fuel volume. The water rod region was also adjusted to maintain the reference assembly hydrogen to uranium atom ratio. With this modification, the model predicted a 24% allowable power uprate for the 200cm twist pitch HC core. Inlet and exit enthalpies were maintained from the reference cylindrical-rod core. When applied to a PWR core of HC rods, also with a fixed power to flow ratio, this empirical mixing model predicted an allowable power uprate of 47%, using traditional CHF correlations for cylindrical fuel. In subcooled conditions, CHF is known to be more sensitive to peaked areas of non-uniform heat-flux than in saturated two-phase flow conditions. Therefore power density gains will likely be dependent on the degree to which the rod twist would disrupt of nascent pockets of vapor; this effect should be further investigated experimentally. In order to further ascertain the potential gain in power density for the new design, an experiment must be carried out to obtain CHF data for the HC rod bundle. Two facilities with this aim were designed in great detail for BWR conditions: the first would operate using high pressure water at 7MPa, and the alternate would use a relatively low pressure refrigerant at equivalent conditions. The appropriate scaling laws were applied, which resulted in the choice of R134a as the simulant fluid. The R134a facility was found to be possible to construct at a greatly reduced cost. 2010-05-01T00:00:00Z https://hdl.handle.net/1721.1/75263 Development of a Bayesian Network to Monitor the Probability of Nuclear Proliferation Holcombe, Robert; Golay, Michael W. Nuclear Proliferation is a complex problem that has plagued national security strategists since the advent of the first nuclear weapons. As the cost to produce nuclear weapons has continued to decline and the availability of nuclear material has become more widespread, the threat of proliferation has increased. The spread of technology and the globalization of the information age has made the threat not only more likely, but also more difficult to detect. Proliferation experts do not agree on the universal factors which cause nations to want to proliferate or the methods to prevent countries from successfully developing nuclear weapons. Historical evidence also indicates that the current nuclear powers pursued their nuclear programs for different reasons and under different conditions. This disparity presents a problem to decision makers who are tasked with preventing further nuclear proliferation. Bayesian Inference is a tool of quantitative analysis that is rapidly gaining interest in numerous fields of scientific study that have previously been limited to purely statistical methods. The Bayesian approach removes the statistical limitations of large-n data sets and strictly numerical types of data. It allows researchers to include sparse and rich data as well as qualitative data based on the opinions of subject matter experts. Bayesian inference allows the inclusion of both the quantitative data and subjective judgments in the determination of predictions about a theory of interest. This means that contrary to classic statistical methods, we can now make accurate predictions with reduced information and apply this probabilistic method to problems in social science. The problem of nuclear proliferation is one that lends itself to a Bayesian analysis. The data set is relatively small and the data is far from consistent from country to country. There is however, a wide body of literature that seeks to explain proliferation factors and capabilities through both quantitative and qualitative means. This varied field can be brought together in a coherent method using Bayesian inference and specifically Bayesian Networks which graphically represent the various causal linkages. This work presents the development of a Bayesian Network describing the various causes, factors, and capabilities leading to proliferation. This network is constructed with conditional probabilities using theoretical insights and expert opinion. Bayesian inference using historical and real time events within the structure of the network is then used to give a decision maker an informed prediction of the proliferation danger of a specific country and inferences about which factors are causing it. 2010-04-01T00:00:00Z https://hdl.handle.net/1721.1/75259 HLW Deep Borehole Design and Assessment: Notes on Technical Performance Jensen, K. G.; Driscoll, Michael J. This is a progress report covering work through mid-April 2010 under a Sandia-MIT contract dealing with design and siting/licensing criteria for deep borehole disposal of spent nuclear fuel or its separated constituents. It consists of a collection of short technical notes which scope out the performance-related requirements of a deep borehole repository. Taken together the results highlight the need to focus on water transport as the dominant phenomenon. In this regard, I-129 is singled out as the likely limiting species because of its high, water chemistry-independent, solubility and long half life. Host rock thermal conditions are also examined, but found not likely to be a limiting constraint. They do, however, argue in favor of using a cluster of shorter multibranch boreholes rather than a much deeper single hole. 2010-04-01T00:00:00Z https://hdl.handle.net/1721.1/75256 A Framework for Performance Assessment and Licensing of Deep Borehole Repositories Jensen, K. G.; Driscoll, Michael J. This is the initial progress report under a Sandia-MIT contract dealing with development of engineering and geological siting criteria for deep borehole disposal of spent nuclear fuel or its separated constituents. Appendix C to this report reproduces the statement of work. The basic conceptual design used as the basis for assessment is presented, followed by screening for features, events and processes of special relevance, using criteria previously developed in the US for mined repository assessment: specifically those identified in the Environmental Impact Statement and Total System Performance Assessment protocols. Transport of radionuclides dissolved in water through highly impervious igneous bedrock (“granite”) is reaffirmed as the dominant mechanism of concern. The important beneficial role of deep-down water chemistry is also highlighted, in that low solubility under reducing conditions, retardation by adsorption, and inhibition of buoyancy and colloid formation by salinity, are all keys to assurance of effective sequestration. These insights are brought to bear to structure our future work scope. 2010-03-01T00:00:00Z https://hdl.handle.net/1721.1/75252 Regional Examples of Geological Settings for Nuclear Waste Disposal in Deep Boreholes Sapiie, B.; Driscoll, Michael J.; Jensen, K. G. This report develops and exercises broad-area site selection criteria for deep boreholes suitable for disposal of spent nuclear fuel and/or its separated constituents. Three candidates are examined: a regional site in the Baltic Fennoscandian shield for the fourteen nation European Repository Development Organization (ERDO) group of small European users; an area in the Arabian shield for newly announced reactor programs in several nations of the Middle East; and, following the same theme, a US site in Minnesota based on exploitation of the Canadian Shield. The criteria applied are restricted to technical, geological aspects and do not address the significant sociopolitical constraints faced by all repository programs. It is concluded that the subject sites all pass first-level technical criteria, and would thus be eligible for in-the-field follow-up, if so desired, by the cognizant organizations. 2010-01-01T00:00:00Z https://hdl.handle.net/1721.1/75248 Impact of Alternative Nuclear Fuel Cycle Options on Infrastructure and Fuel Requirements, Actinide and Waste Inventories, and Economics Guérin, Laurent; Kazimi, Mujid S. The nuclear fuel once-through cycle (OTC) scheme currently practiced in the U.S. leads to accumulation of uranium, transuranic (TRU) and fission product inventories in the spent nuclear fuel. Various separation and recycling options can be envisioned in order to reduce these inventories while extracting additional energy and sending the ultimate waste to a repository. Choosing one of these options has direct implications for the infrastructure requirements, natural uranium consumption, actinide inventories in the system, waste repository needs and costs. In order to account for the complexity of the nuclear enterprise, a fuel cycle simulation code has been developed using system dynamics (CAFCA). An economic module was added using spreadsheets. Four main advanced fuel cycle schemes are assessed here within the context of the US market: 1) the twice-through cycle scheme (TTC): single-pass plutonium recycling in thermal spectrum LWRs using Mixed OXide (MOX) fuel; 2) Multi-recycling of TRU in sodium-cooled fast spectrum burner cores, characterized by a fissile conversion ratio lower than 1 (FBu); 3) Multi-recycling of TRU in sodium-cooled fast breeders with a conversion ratio of 1.23 (FBr); and 4) A two-tier scenario: a TTC scheme is practiced as a transition scheme to fast reactors. The base case scenario assumes annual nuclear energy demand growth rate of 2.5% from 2020 on. The technologies for plutonium separation as well as MOX fuel fabrication are assumed to be available in 2025 while the first commercial fast reactors, as well as the possibility to recycle their spent fuel, are assumed to be available in 2040. For fast reactors, the cores are assumed to be TRU fueled, and the technology to separate the minor actinides is supposed to be available at the latest 5 years before deployment of fast reactors. Limits are applied on the building rate of reprocessing plants, which are also subject to a 80% minimum life-time loading factor requirement. It is found that, despite its higher cost, at the end of the century, the TTC scheme (single Pu-MOX recycle) does not lead to large improvements in terms of natural uranium consumption (16%), repository needs (considering both fission products and MA from reprocessing facilities, and spent MOX fuel) and TRU inventory reduction (although some shifting of TRU from storage to reactors occurs). This is especially significant because it is the only advanced fuel cycle option that can be deployed in large scale in the next few decades. However, if the primary reason for introduction of the more expensive fast reactors is resource enhancement and/or control of TRU in the nuclear waste, thermal reactor recycling allows the introduction of fast reactors to be delayed by 20-25 years. Moreover, once fast reactors are introduced, their deployment is accelerated compared to a 1-tier FR scenario. However, the two-tier scheme is the most expensive scheme as it combines the requirements of both the MOX technology and the FR technology. Sensitivity analyses were performed in order to assess the impact of secondary parameters. It is found that whatever the growth rate assumed, LWRs remain a significant part of the system at the end of the century, decades after fast breeders are introduced. The reason is the fissile materials required for fabrication of start-up cores considerably affect the rate at which fast reactors can be deployed. As a result, the choice of the core design (compact core vs. large core) may be as significant as the choice of the conversion ratio. For example, the breeder scenario (CR=1.23) may lead to the same cumulative natural uranium consumption reduction (by 2100) as the self-sustaining reactors (CR=1.0) while leading to larger TRU inventory in the system and requiring greater fast reactor fuel reprocessing capacity. Allowing fast reactors to start with uranium only cores was not considered, as it will likely limit resource enhancement benefits of fast reactors. Still, in general, the higher the conversion ratio, the greater the fast reactor installed capacity, hence the greater the savings in natural uranium. Conversely, the best reduction in TRU from the OTC amount is obtained by the lower conversion ratio (45% for a pure burner with conversion ratio 0.0 by 2100). Doubling the minimum cooling time before reprocessing for all fuel types from 5 years to 10 years slows down the deployment of the fast reactors and therefore reduces their share in the total installed capacity. This is almost equivalent to replacing breeders with fast reactors with a conversion ratio of 0.75. Finally, the results show that starting the separation of the TRU 10 years prior to introduction of the fast reactors instead of 5 years provides a mid-term advantage (faster initial deployment) that vanishes within 25 years. In the long term, the fast reactor penetration results are insensitive to the assumed industrial capacity to build reprocessing facilities for the base case or at lower nuclear energy growth rates. However, the assumed industrial capacity can be a real constraint if the nuclear energy growth rates are 4% or higher. 2009-09-01T00:00:00Z https://hdl.handle.net/1721.1/75247 UPDATE ON THE COST OF NUCLEAR POWER Du, Yangbo; Parsons, John E. We update the cost of nuclear power as calculated in the MIT (2003) Future of Nuclear Power study. Our main focus is on the changing cost of construction of new plants. The MIT (2003) study provided useful data on the cost of then recent builds in Japan and the Republic of Korea. We provide similar data on later builds in Japan and the Republic of Korea as well as a careful analysis of the forecasted costs on some recently proposed plants in the US. Using the updated cost of construction, we calculate a levelized cost of electricity from nuclear power. We also update the cost of electricity from coal- and gas-fired power plants and compare the levelized costs of nuclear, coal and gas. The results show that the cost of constructing a nuclear plant has approximately doubled. The cost of constructing coal-fired plants has also increased, although perhaps just as importantly, the cost of the coal itself has spiked dramatically, too. Capital costs are a much smaller fraction of the cost of electricity from gas, so it is the recent spike in the price of natural gas that has contributed to the increased cost of electricity. These results document changing prices leading up to the current economic and financial crisis, and do not incorporate how this crisis may be currently affecting prices. 2009-05-01T00:00:00Z https://hdl.handle.net/1721.1/75245 A Benchmark Study of Computer Codes for System Analysis of the Nuclear Fuel Cycle Guérin, Laurent; Feng, Bo; Hejzlar, Pavel; Forget, Benoit; Kazimi, Mujid S.; Van Den Durpel, Luc; Yacout, Abdellatif; Taiwo, Temi; Dixon, Brent W.; Matthern, Grechen; Boucher, Lionel; Delpech, Marc; Girieud, Richard; Meyer, Maryan As use of nuclear energy is expected to expand in different parts of the world, several codes that describe the nuclear fuel cycle system are currently under development, featuring a range of capabilities and different levels of flexibility and automation. Such codes model the addition or retirement of reactors, the demand for fresh fuel, and the need for spent fuel storage and recycling facilities as the production of nuclear energy varies with time. The codes enable analysis of various scenarios for the evolution of the nuclear energy system, and the timing of deployment of new facilities. Outputs may also include fuel material mass flows, economic analysis and metrics related to spent fuel or waste assessment. The study reported here is the first attempt for benchmarking the MIT code CAFCA against three independently developed fuel cycle simulation codes. It is also among the first publicly available benchmark exercises. Some reviews of the existing codes were previously reported, but focused mostly on their theoretical capabilities. Benchmarking studies, generally involving two or three codes, have been done over the last few years, but most remain unpublished. The codes included in this study are: CAFCA (developed at MIT), COSI (developed at CEA, France), DANESS (developed at ANL) and VISION (developed by DOE laboratories for the AFCI program). The purpose of this benchmark study is to evaluate the degree of convergence of the current versions of these codes and to compare their basic methodologies. This benchmark is not a comprehensive analysis of all the codes’ capabilities but constitutes a first step towards a more complete benchmark study. In order to compare all 4 codes, only the common capabilities were considered and assessed. Those capabilities are essentially those of CAFCA, as it is the simplest code. Consequently, some of the advanced capabilities of the other codes were disabled, and their complete features were not reflected in this benchmark study. In addition, economic evaluation of the fuel cycles was not considered, even though it is a capability common to the four codes. Furthermore, the initial runs showed that a degree of freedom should be removed to ease the comparison. For that reason reprocessing capacity profiles were provided by CAFCA and used as input by the other codes. Following a description of the codes, the report presents the four scenarios selected as the benchmark cases, including initial conditions. The time period of the simulation covers the 21st century. Those scenarios differ from each other in either the nuclear energy supply growth rate (0%, 1.5% or 3%) or the type of advanced technology introduced in the midterm. The options for advanced reactors were: the “self-sustaining” fast reactor (with a fissile conversion ratio of one), fast burners of transuranics (TRU), or a combination of plutonium recycling (as mixed oxide) in the thermal light water reactors (LWRs) and fast burners. The scenarios, and hence the results, are for benchmarking purposes only and should not be considered realistic for policy studies or forecasts about the future of nuclear power. The set of constraints specified is minimal and only intended to provide a common framework for the simulations. The results are presented and commented on for each case. The first case, characterized by a very constraining zero energy growth rate, shows an excellent agreement among the codes, with identical ratios of fast reactors/thermal reactors over time. This excellent agreement was the iv result of the particular efforts made in order to get very close results (several iterations were performed to allow for adjustments). This case eventually shows how the models can produce very close results if sufficiently tuned to adhere to the same basic assumptions. This case also allowed us to identify a number of minor apparent discrepancies and explain them. In particular, it made obvious differences in results between COSI, which was tuned to track fuel batches (“discrete-flow code”) and CAFCA, DANESS and VISION, which deal with annual mass flows (”continuous-flow codes). The treatment of discrete batches of fuel by COSI, instead of timeaveraged quantities in the other codes, result in somewhat oscillatory flows and inventories of materials. Another factor leading to discrepancies among the codes is the time assumed to exist between the separation of transuranics and the manufacturing of fast reactor fuels. CAFCA speeds up the fuel manufacturing, to avoid the presence of separated transuranics in large quantities, while the other codes do not (as Pu-containing fuel has a very limited shelf-life at the initial fissile content due to Pu 241 decay with a half life of about 14.4 years). Unlike the first cases, there was in the three other cases no attempt, beyond the common set of assumptions, to iterate to get the results of the other cases to converge. Therefore, these three cases are more a reflection of how the codes actually operate and show the level of variation in results that should be considered normal. These three cases were of great interest for comparing the different strategies for fast reactor deployment and their dependence for fuel on available TRU from the operation of light water reactors. Overall, the benchmark shows that, although the codes exhibit reassuring consistency, both internally and among themselves, differences still exist. The fact that reprocessing capacity profiles were externally provided may have disturbed some of the codes designed to internally calculate this variable. Although limitations inherent in the codes exist, the differences in the results generally do not reveal major flaws, but rather reflect differing assumptions and constraints embedded in the methods and approximations of the calculations. This benchmark reveals (or reminds us) that there is no single profile for a fuel cycle scenario but several profiles that depend on industrial practices with regards to manufacturing, storage and reprocessing of nuclear fuels. These practices may aim at differing priorities of reducing stocks of stored spent fuel, avoiding the presence of separated TRU to reduce proliferation risks, ensuring sufficient supply of fresh fuel for advanced reactors and spent fuel for reprocessing plants, and minimizing some costs. Such choices can either be intrinsic to the code (through built-in assumptions) or through user choices (for example, the level of conservatism in the algorithm ensuring fuel supply for fast reactors is a user input). Moreover, the parameters left to the user’s discretion are generally not the same from one code to another, or are expressed in different terms. Finally, complete consistency between the codes is difficult to obtain. Two major conclusions can be drawn from this benchmark. First, the overall results show good consistency and similar trends. Hence, utilization of various codes is likely to lead to similar conclusions. Second, one must not expect the various fuel cycle system simulation codes to provide identical outputs. Therefore, users must keep in mind that, although the results are internally consistent and meet each code’s set of requirements, they do not project unique scenarios for meeting such requirements. 2009-04-01T00:00:00Z https://hdl.handle.net/1721.1/75242 MCODE, Version 2.2: An MCNP-ORIGEN DEpletion Program Xu, Z.; Hejzlar, Pavel MCODE Version 2.2 is a linkage program, which combines the continuous-energy Monte Carlo code, MCNP-4C, and the one-group depletion code, ORIGEN2, to perform burnup calculations for nuclear fission reactor systems. MCNP is used as the advanced physics modeling tool providing the neutron flux solution and detailed reaction rates in the pre-defined spatial burnup zones. ORIGEN, in turn, carries out multi-nuclide depletion calculations in each region and updates the corresponding material composition in the MCNP model. The MCNP/ORIGEN coupling follows the predictor-corrector approach. During a burnup timestep, end-of-timestep material compositions are first predicted based on the flux solution at the beginning-of-timestep. Using the predicted end-of-timestep material compositions, an MCNP run is performed to compute the neutron flux and detailed reaction rates, which are then used in a corrector burnup step. The final end-oftimestep material compositions are obtained as the average value of the results from the predictor and corrector steps. As a stand-alone code written in ANSI C, MCODE-2.2 is portable between Windows personal computers (PC’s) and UNIX/Linux machines. There are three utility programs in MCODE-2.2: (1) preproc to pre-process MCNP/ORIGEN libraries; (2) mcode as the console to run steady-state burnup/decay calculations; and (3) mcodeout to collect results from scattered data files under temporary directory and produce a detailed output. Further, there is an auxiliary program called mcnpxs, which is for the purpose of preparing a nuclide summary table of continuous energy MCNP cross section libraries. The routine usage of MCODE-2.2 only requires a tandem running of the three utility codes. The auxiliary code, mcnpxs, is intended to help users during the code installation/setup. Compared to other similar linkage codes, MCODE-2.2 emphasizes functionality, versatility and usability. Several features of the code follow: (1) The execution of MCNP and ORIGEN is in an automatic fashion. (2) All standard nuclear reaction types in ORIGEN2 are considered: capture, fission, (n,2n), (n,3n), (n,p), and (n,α). Therefore, both the nuclear fuel depletion and material irradiation/activation (e.g., boron-10 irradiation) can be handled. (3) A power history can be specified, i.e., power level at each timestep. The default depletion option is constant power depletion. Meanwhile, an iterative robust flux depletion scheme is available. In addition, decay calculations are also possible. (4) With appropriate ORIGEN one-group cross section libraries, users can rely on MCODE- 2.2 to automatically select important nuclides based on absorption ranking from ORIGEN isotope reservoir for MCNP calculations. (5) The enhanced predictor-corrector approach (consistent with CASMO-4) increases the accuracy with negligible computational cost increase. From the user’s point of view, MCODE-2.2 is an extension of normal MCNP criticality (kcode) calculations. The MCNP input inherits the MCODE-2.2 input in the form of a fourth paragraph (added at the end of the MCNP input deck) containing the burnup-related data and MCNP/ORIGEN calculation controls. A user-supplied equilibrium MCNP source file can also be provided, which might save CPU time by reducing the number of initial MCNP inactive cycles. iii The Monte Carlo burnup code has some unique characteristics, one of which is that all results are in nature stochastic. The statistical uncertainty passing through burnup calculations is one concern, which is believed by some people as the weakness or even indication of the Monte Carlo limitations to perform burnup calculations. Using a multiregion Gd-poisoned BWR 8×8 assembly depletion problem, it is shown that the random statistical uncertainties are benign and cancel each other with the burnup. In addition, a single PWR unit cell benchmark problem is documented. Comparison of results against CASMO-4 yields satisfactory agreement. 2008-12-01T00:00:00Z https://hdl.handle.net/1721.1/75238 A System Dynamics Study of the Nuclear Fuel Cycle with Recycling: Options and Outcomes for the US and Brazil Busquim e Silva, R.; Kazimi, Mujid S.; Hejzlar, Pavel A system dynamics simulation technique is applied to generate a new version of the CAFCA code to study mass flows in the nuclear fuel cycle, and the impact of different options for advanced reactors and fuel recycling facilities on the inventory and distribution of transuranics (TRU). Several nuclear fuel cycle options are studied for U.S. and Brazil markets, and special consideration is given to potential collaboration between the two countries. This includes the impact of advanced nuclear technologies, under a prescribed growth in demand for nuclear electricity, on demand for uranium resources, uranium enrichment, and fuel reprocessing facilities, and on total cost of nuclear electricity over the next few decades. Introduction of fuel recycling reduces the growing demand for uranium, and the long-term need for storage of radioactive spent fuel. However, the timing of introduction of recycling is important for proper technology development, and the rate of deployment is restrained by the industrial capacity as well as the desire for high utilization factor of the deployed facilities over their life time, and that is reflected in the assessments. The nuclear fuel cycle is modeled as a high level structure diagram, which provides an overview of the interconnections among its blocks without showing all details, and as a structure-policy diagram which details the decision rules applied to the structure. The high level structure diagram represents the nuclear fuel cycle; the fleet of thermal and fast reactors; the separation and reprocessing plants; the waste repository; the spent fuel storage; and the paths for the fuel and waste mass transfer. In addition, an economic model is added to study different cases under the same assumptions. The economic model is based on the forecasted need for advanced reactors and recycling facilities, assuming that all costs are recovered within the nuclear energy system. Different recycling technology options are included in the code: (1) Thermal recycling in LWRs using Combined Non-Fertile and UO2 Fuel (CONFU), (2) Recycling of TRU in fertile-free fast cores of Actinide Burner Reactors (ABR); and (3) Fast recycling of TRU with UO[subscript 2] in self-sustaining Gas-cooled Fast Reactors (GFR). Case studies for different advanced technology introduction dates and for distinct TRU depletion rates are examined. In particular, the code is 3 equipped to simulate the introduction of two recycling technology options with a prescribed allocation of the TRU supply between them. The simulation results show that early introduction of the GFR recycling scheme leads to the most significant reduction in uranium consumption and enrichment requirements, thus delaying the eventual depletion date of uranium ore. The GFR requires less uranium resources due to the use of TRU as recycled fuel and near unity fissile conversion ratio. However, in a nonbreeding reactor system, the consumption of U continues to grow, and the TRU needed to start fast reactors will be growing at a constrained rate. On the other hand, the CONFU recycling scheme keeps the TRU inventory in the entire system well below other schemes, and guarantees equilibrium between the generation and consumption of transuranics without investments in fast reactors. CONFU incinerates more TRU than the GFR and ABR schemes during the simulation period. Also, it reduces the TRU sent to the repository for disposal by orders of magnitude. The ABR scheme does the same but requires the introduction of fast reactors. Nevertheless, the CONFU and ABR schemes have no significant impact on the amount of uranium resources consumption or enrichment requirements. Economic analysis indicates that the CONFU technology is more attractive at current uranium prices, and that fast recycling becomes as attractive as thermal recycling at higher uranium prices. The results also show that if a nuclear fuel cycle state/reactor state collaboration with Brazil is started, there will be a significant impact on the U.S. cumulative TRU inventory at interim storage, enrichment requirements, uranium consumption, and number of advanced fuel facilities. The results show that a nuclear partnership without the introduction of advanced nuclear technologies would not have advantages for the U.S. Furthermore, a nuclear collaboration allows a higher ratio of fast reactors to total installed nuclear electric capacity in the U.S. 2008-11-01T00:00:00Z https://hdl.handle.net/1721.1/75223 Thermal Hydraulic Analysis of a Low Enrichment Uranium Core for the MIT Research Reactor Ko, Yu-Chih; Hu, Lin-Wen; Kazimi, Mujid S. The MIT research reactor (MITR) is converting from the existing high enrichment uranium (HEU) core to a low enrichment uranium (LEU) core using a high-density monolithic UMo fuel. The design of an optimum LEU core for the MIT reactor is evolving. The objectives of this study are to benchmark the in-house computer code for the MITR, and to perform thermal hydraulic analyses in support of the LEU design studies. The in-house multi-channel thermal-hydraulics code, MULCH-II, was developed specifically for the MITR. This code was validated against PLTEMP for steady-state analysis, and against RELAP5 and temperature measurements for the loss of primary flow transient. The benchmark analysis results showed that the MULCH-II code is in good agreement with other computer codes and experimental data, and hence it is used as the main tool for this study. Various fuel configurations are evaluated as part of the LEU core design optimization study. The criteria adopted for the LEU thermal hydraulics analysis in this study of the limiting safety system settings (LSSS), are to prevent onset of nucleate boiling during steady-state operation, and to avoid a clad temperature excursion during the loss of flow transient. In ranking the LEU core design options, the primary parameter is a low power peaking factor in order to increase the LSSS power and to decrease the maximum clad temperature during the transient. The LEU fuel designs with 15 to 18 plates per element, fuel thickness of 20 mils, and a hot channel factor less than 1.76 are shown to comply with the thermal-hydraulic criteria. The steady-state power can potentially be higher than 6 MW, which was requested in the power upgrade submission to the Nuclear Regulatory Commission. 2008-01-01T00:00:00Z https://hdl.handle.net/1721.1/75222 PWR Transition to a Higher Power Core Using Annular Fuel Beccherle, J.; Hejzlar, Pavel; Kazimi, Mujid S. The internally and externally cooled annular fuel is a new type of fuel for PWRs that enables an increase in core power density by 50% within the same or better safety margins as traditional solid fuel. Each annular fuel assembly of the same side dimensions as the 17x17 solid fuel assembly has 160 annular fuel rods arranged in a 13x13 array. Even at the much higher power density, the fuel exhibits substantially lower temperatures and a Minimum Departure From Nucleate Boiling (MDNBR) margin comparable to that of traditional solid fuel at nominal (100%) power. The major motivation for such an uprate is reduction of electricity generation cost. Indeed, the capital cost per kWh(e) of a new reactor would be smaller than the standard construction of a new reactor with solid fuel. Elaborating on previous work, we study the economic payoff of an uprate of existing PWRs given the expected cost of equipment and also cost of money using different assumptions. The fate of the already bought solid fuel is investigated. It is demonstrated that the highest return on investment is obtained by gradually loading annular fuel in the reactor core such that immediately before shutting the reactor down for the uprate construction, two batches in the core are of the annular fuel. This option implies running a core with a mixture of both annular and solid fuel assemblies. In order to prove the technical feasibility of such an option, the thermalhydraulics of this mixed core is investigated and the MDNBR is found to be either unaffected or improved. Consequently, a neutronic model is developed to verify and validate the neutronic feasibility of the transition from solid to annular fuel. This involvements assessment of the peaking factors and capability to provide control poisons within allowable concentrations The overall conclusion of this work is that annular fuel is a very promising option for existing reactors to increase their power by 50%, as it enables a significant uprate with an attractive return on investment. We show that, by a smart management of the transition, an internal return on investment of about 22–27% can be achieved. 2007-09-01T00:00:00Z https://hdl.handle.net/1721.1/75219 Sensitivity of Economic Performance of the Nuclear Fuel Cycle to Simulation Modeling Assumptions Bonnet, Nicephore; Kazimi, Mujid S. Comparing different nuclear fuel cycles and assessing their implications requires a fuel cycle simulation model as complete and realistic as possible. In this report, methodological implications of modeling choices are discussed in connection with development of the MIT fuel cycle simulation code CAFCA. The CAFCA code is meant to find the recycling facilities deployment rate that minimizes the time by which spent fuel in storage today is used up in order to lead to a nuclear fuel cycle with minimum inventory of transuranic elements. The deployment is constrained by the construction capacity of the recycling plants and by the economic requirement that their loading factor never drops below a certain level. First, through a simplified fuel cycle model, it is analytically proven that an optimum solution is to build recycling plants at full speed up to a certain point in time b, then to suspend construction until interim storage is completely depleted. The shape of the optimum solution is injected into an algorithm based on a complete model of the fuel cycle. An iterative process yields the value of b assuring depletion and satisfactory loading factors. Besides providing rigorous optimization, the analytical solution underpinning the CAFCA algorithm is expected to reduce considerably the vulnerability of the results to numerical discontinuities. Degradation of fuel quality with time in interim storage occurs due to the decay of Pu241 into Am241. While an obvious approach to track such effects is to couple the fuel cycle code with a neutronics/decay code (ORIGEN for example), it is more efficient to derive explicit equations from a simplified irradiation and decay model, allowing for analytical tracking of the fuel composition. This approach was implemented in CAFCA. All fuel cycle simulation refinements do not present the same level of importance. One should focus on the dominant parameters, i.e., those contributing most to results sensitivity. The important parameters are determined through a sensitivity study using a novel U.S. thermal recycling scenario called CONFU as a reference case. The CONFU technology is assumed to be commercially introduced 15 years from now, with an industrial capacity allowing the construction of one 1000 MT/year spent fuel separation plant every two years. Additionally, it is assumed that discharged CONFU batches remain in cooling storage for 6 years, reactors have a 60-year lifetime and economic recovery period of 20 years, and are half financed by equity with a rate of return of 15%. It is found that the cost of electricity is most sensitive to the reactors lifetime, since taking it back to a nominal value of 40 years would result in a 44% increase in the cost of electricity. Next in importance is the financing structure of the fleet. The addition of three points to the rate of return on equity would increase the cost of electricity by 14%. While scale effects are locally very beneficial in that they substantially reduce recycling plants operation costs, they prove to be of limited interest from an overall fuel cycle point of view. Using the scale effect model in CAFCA-II, doubling the separation plants capacity yields a 3% reduction of the cost of electricity. The fuel cycle presents good robustness with respect to fuel decay time degradation. Increasing CONFU batches cooling time to 18 years causes a 2% increase in the cost of electricity. 2007-07-01T00:00:00Z https://hdl.handle.net/1721.1/75218 AN EVOLUTIONARY FUEL ASSEMBLY DESIGN FOR HIGH POWER DENSITY BWRs Karahan, A.; Buongiorno, Jacopo; Kazimi, Mujid S. An evolutionary BWR fuel assembly design was studied as a means to increase the power density of current and future BWR cores. The new assembly concept is based on replacing four traditional assemblies and large water gap regions with a single large assembly. The traditional BWR cylindrical UO[subscript 2]-fuelled Zr-clad fuel pin design is retained, but the pins are arranged on a 22×22 square lattice. There are 384 fuel pins with 9.6 mm diameter within a large assembly. Twenty-five water rods with 27 mm diameter maintain the moderating power and accommodate as many finger-type control rods. The total number and positions of the control rod drive mechanisms are not changed, so existing BWRs can be retrofitted with the new fuel assembly. The technical characteristics of the large fuel assembly were evaluated through a systematic comparison with a traditional 9×9 fuel assembly. The pressure, inlet subcooling and average exit quality of the new core were kept equal to the reference values. Thus the power uprate is accommodated by an increase of the core mass flow rate. The findings are as follows: - VIPRE subchannel analysis suggests that, due to its higher fuel to coolant heat transfer area and coolant flow area, the large assembly can operate at a power density 20% higher than the traditional assembly while maintaining the same margin to dryout. - CASMO 2D neutronic analysis indicates that the large assembly can sustain an 18- month irradiation cycle (at uprated power) with 3-batch refueling, <5wt% enrichment with <60 MWD/kg average discharge burnup. Also, the void and fuel temperature reactivity coefficients are both negative and close to those of the traditional BWR core. - The susceptibility of the large assembly core to thermalhydraulic/neutronic oscillations of the density-wave type was explored with an in-house code. It was found that, while well within regulatory limits, the flow oscillation decay ratio of the large assembly core is higher than that of the traditional assembly core. The higher core wide decay ratio of the large assembly core is due to its somewhat higher (more negative) void reactivity coefficient. - The pressure drop in the uprated core is 17 % higher than in the reference core, and the flow is 20% higher; therefore, larger pumps will be needed. - FRAPCON analysis suggests that the thermo-mechanical performance (e.g., fuel temperature, fission gas release, hoop stress and strain, clad oxidation) of the fuel pins in the large assembly is similar to that of the reference assembly fuel pins. - A conceptual mechanical design of the large fuel assembly and its supporting structure was developed. It was found that the water rods and lower tie plate can be used as the main structural element of the assembly, with horizontal support being provided by the top fuel guide plate and core plate assembly, and vertical support being provided by the fuel support duct, which also supports the finger-type control rods. 2007-07-01T00:00:00Z https://hdl.handle.net/1721.1/75215 A PWR Self- Contained Actinide Transmutation System Shatilla, Y.; Hejzlar, Pavel; Kazimi, Mujid S. Elements of the new Global Nuclear Energy Partnership (GNEP) initiative in the US call for the expansion of domestic use of nuclear power and the minimization of nuclear waste. To achieve both goals in the short term the transmutation of trans-uranic (TRU) elements in Combined Non- Fertile and Uranium (CONFU) PWR fuel assemblies is evaluated. These assemblies are composed of a mix of standard UO[subscript 2] fuel pins and pins made of recycled TRU in an inert matrix and are designed to fit in currently deployed PWRs. A CONFU-Self-Contained (CONFU-C) assembly is shown to achieve a net TRU destruction in a self-contained TRU multi-recycling system. The system may consist of as little as one currently operating reactor that does not depend on other reactors to supply it with its inventory of recycled TRU. This is considered a major advantage of the new design over its predecessors since it eliminates the need for designating a whole fleet of CONFU reactors to produce recycled TRU for the reactor under consideration. Degradation of fissile content of the multi-recycled TRU is compensated for by drawing from legacy TRU that already comes from standard UO2 spent fuel and the usage of fresh UO[subscript 2] fuel with different enrichments depending on fuel cooling time after discharge. A recycling strategy which uses a 4.5 year period of in core irradiation, followed by one of three cooling periods (6-, 18-, and 32-year) after discharge and reprocessing is considered. Calculations have shown the equilibrium CONFU-C assembly can have a net TRU destruction of approximately 20.6 (for the 6-yr cooling) and 2.7 (for the 18-yr cooling) kg of TRU per TWhe, as compared to 11.0 kg of TRU per TWhe for the CONFU-B with a 6-yr cooling period. This represents a net burning rate of ~13% (6-yr cooling) and 3% (18-yr cooling) of the TRU loaded per assembly compared to 8% for the CONFU-B design. However, Fuel Cycle Costs (FCC) for the equilibrium CONFU-C is shown to be 12.8 (6yr-cooling) and 14.2 (18yr cooling) mills/KWhe and that for the CONFU-B to be 12 mills/KWhe. Due to the relatively long cooling period of the third option (32 yr cooling), a CONFU-C assembly could not be designed to achieve net TRU destruction in a self-contained manner. 2006-09-01T00:00:00Z https://hdl.handle.net/1721.1/75214 Fuel Cycle Options for Optimized Recycling of Nuclear Fuel Aquien, A.; Kazimi, Mujid S.; Hejzlar, Pavel The reduction of transuranic inventories of spent nuclear fuel depends upon the deployment of advanced fuels that can be loaded with recycled transuranics (TRU), and the availability of facilities to separate and reprocess spent fuel. Three recycling strategies are explored in this study: (1) Recycling in thermal Light Water Reactors (LWR) using CONFU technology (COmbined Non-Fertile and UO[subscript 2] fuel), (2) recycling of TRU in fertile-free fast cores of Actinide Burner Reactors (ABR), and (3) recycling of TRU with UO[subscript 2] in self-sustaining Gas-cooled Fast Reactors (GFR). Choosing one strategy over another involves trade-offs. The CONFU, ABR, and GFR strategies differ from each other in terms of T RU loading in the reactor, net TRU incineration, capacities of recycling facilities needed, date for technology option availability, and flexibility. The CONFU and GFR are assumed to achieve zero net TRU incineration, while the ABR is a net consumer of TRU. The TRU loading is greatest in GFR and lowest in CONFU. While both CONFU and ABR require separation (of TRU from U) and reprocessing (recycling of TRUs from fertile-free fuel), the GFR is designed to, in equilibrium, recycle TRU+U after extraction of the fission products. It is assumed that thermal recycling is available in the short-term (2015), as opposed to recycling in fast reactors (2040). Finally, thermal recycling is the most flexible as either CONFU batches or regular LWR uranium batches can be loaded; the issue of running out of TRU fuel is therefore irrelevant for this option. A fuel cycle simulation tool, CAFCA II (Code for Advanced Fuel Cycles Assessment) has been developed. The CAFCA II code tracks the mass distribution of TRU in the system and the cost of all operations. The code includes a specific model for deployment of recycling plants, with certain capacities and investment requirements. These facilities may operate with a minimum target capacity factor during the lifetime of the plant. The deployment of these facilities is also constrained by a user-specified ability to add recycling capacity within a given time interval. Finally, the CAFCA II code includes a specific model for recycling prices which reflects the economies of scale that go with increases in the nominal capacity of recycling plants. Two case studies are presented. The first explores the optimal fuel cycle and recycling plant capacities as a function of the deployment of advanced fuel cycle technologies over the next hundred years, under the assumption of the US demand for nuclear energy growing at a 2.4% annual rate. Key figures for comparison of the strategies are evaluated, including reduction of TRU interim storage requirements, maximization of TRU incineration, minimization of the size of the fleets of recycling plants and fast reactors, fuel cycle cost, and capital cost requirements. We found that it is not possible to minimize the construction rate of advanced reactors and advanced spent fuel recycling facilities simultaneously with the construction rate of separation facilities for UO[subscript 2] spent fuel. The latter was found to be more constraining than the former. Further, we found that reactor technologies with zero net TRU destruction rate can achieve total depletion of TRU inventories in spent fuel interim storage at a lower fuel cycle cost and with fewer recycling facilities than reactor technologies that incinerate TRU. However, the lower fuel cycle cost is achieved at the expense of a smaller reduction of total TRU inventories. Finally, if the construction rate of advanced nuclear technologies is large enough, the later introduction date of fast recycling schemes compared to thermal recycling schemes does not prevent the reduction of TRU inventories in interim storage by 2100. Recently, the potential benefits of multi-lateral approaches to the management of nuclear fuel have been widely discussed. These include cost attractiveness following from economies of scale, proliferation resistance, and more efficient nuclear waste treatment strategies. CAFCA II has been developed to quantify these implications for the back-end of the fuel cycle. Three partnership scenarios have been examined: first, a scenario where the “Fuel leasing/fuel take-back” concept is implemented; second, a scenario with “Limited Collaboration” at the back-end fuel cycle, where spent fuel recycling and advanced fuel fabrication are externalized in countries that have these technologies; and third, a scenario of “Full Collaboration”, under which two regions fully collaborate at the fuel cycle back-end: spent fuel inventories and advanced fuel cycle facilities are co-owned and co-managed. The second case study concentrates on optimizing the choice of (1) fuel cycle option, (2) recycling plant capacities, and (3) partnership scenario by analyzing the implications of these choices for the LWR-CONFU, LWR/ABR, and LWR/GFR strategies. The nuclear fuel cycle is simulated in a two-region context from 2005 to 2100 under the assumption that one region represents the US growing at a 2.4% annual rate and the other region represents Brazil, Indonesia, and Mexico growing at a 7.4% annual rate until 2080, and 2.4% afterwards. Under this scenario, we found that a US partnership with a region representing Brazil, Mexico, and Indonesia could be advantageous to the reduction of TRU storage in both regions if the construction rate of UO[subscript 2] spent fuel separation plants would be larger than one 1,000 MT/yr plant every two years after 2050. From the point of view of the spent fuel recycling industry, use of the largest recycling plants with the lowest construction cost per unit of installed capacity becomes optimal only with multi-national approaches to the fuel cycle back-end. 2006-06-01T00:00:00Z https://hdl.handle.net/1721.1/75213 Actinide Minimization Using Pressurized Water Reactors Visosky, M.; Kazimi, Mujid S.; Hejzlar, Pavel Transuranic actinides dominate the long-term radiotoxicity in spent LWR fuel. In an open fuel cycle, they impose a long-term burden on geologic repositories. Transmuting these materials in reactor systems is one way to ease the long-term burden on the repository. Examining the maximum possible burning of trans-uranic elements in Combined Non-Fertile and UO[subscript 2] (CONFU) PWR assemblies is evaluated. These assemblies are composed of a mix of standard UO[subscript 2] fuel pins and pins made of recycled trans-uranics (TRU) in an inert matrix, and are designed to fit in current or future PWRs. Applying appropriate limits on the neutronic and thermal safety parameters, a CONFU-Burndown (CONFU-B) assembly design is shown to attain net TRU destruction in each fuel batch through at least 9 recycles. This represents a time span of nearly 100 years of in-core residence and out-of-core storage time. In this way, when the TRU is multirecycled, only fission products and separation/reprocessing losses are sent to the repository, and the initial inventory of TRU is reduced over time. Thus, LWRs are able to eventually operate in a fuel cycle system with an inventory of transuranic actinides much lower than that accumulated to date. Three recycling strategies are considered, all using a 4.5-year in core irradiation, followed by cooling and reprocessing. The three strategies involve a short-term cooling (6-year) after discharge, a longer-term cooling (16.5-year) after discharge, or a strategy called Remix. The Remix strategy involves partitioning the Pu/Np after 6-year cooling for immediate recycle, and partitioning the Am/Cm for an additional 10.5-year cooling before remixing it into the next CONFU-B batch. At equilibrium, the CONFU-B can burn approximately 1.5 kg to 10.0 kg of TRU per TWhe depending on the recycle strategy used. This represents a net burning rate of 2- 8% of the TRU loaded per assembly, in addition to burning an amount equivalent to the TRU produced in the UO[subscript 2] pins. However, the highly heterogeneous nature of these assemblies can result in fairly high intraassembly pin power peaking. By design, an IMF pin in the assembly carries the highest power to maximize the TRU destruction. For the initial TRU loading, the highest power peaking in an IMF pin is 1.183. This is compensated by having cooler pins in the immediate vicinity. Even so, the pin peaking distribution in the assembly can result in reduced thermal margins. The assembly mentioned above has an MDNBR of 1.43, instead of 1.62 for the all-UO[subscript 2] assembly, based on a core-wide radial peak-to-average assembly power peaking of 1.50. Use of neutron poisons and tailored enrichment schemes reduces the neutronic reactivity of fresh assemblies, while improving MDNBR to 1.51. In addition, RELAP was used to evaluate the fuel behavior under large break LOCA conditions. CONFU-B performance under these conditions was comparable to the standard all-UO2 assembly. Several options for spent fuel recycling in LWRs are compared economically, and all are found to be more costly than making fresh UO2 fuel from mined ore. However, the CONFU-B strategy is less costly on a mills/kWhe basis than other thermal recycling strategies that recycle the full TRU vector. Given OECD estimates for the unit costs of each fuel type, and assuming 10% carrying charge factor, this cost is 10.0 mills/kWhe for the CONFU-B recycle, compared to 22.2 mills/kWhe for MOX-UE and 5.4 mills/kWhe for all UO[subscript 2]. Note that these FCCs assume the 2 disposal fee collected during power generation of a previous cycle can be invested while the fuel is cooling and provide a credit to the cycle that uses the fuel after reprocessing. The fuel handling challenges of multirecycling TRU in CONFU-B assemblies are compared to other multi-recycling strategies. If we assume that the spent fuel from the seventh recycle in each strategy is no longer recyclable and must be sent to the repository in its entirety, the CONFU-B strategy still places much less total burden on the repository than the once-through cycle, and even less burden than the current MOX cycle. Finally, a methodology for calculating the time integrated proliferation risk of a fuel cycle is introduced. An innovation of this methodology is the discounting of future risks to calculate an overall present value risk of a given cycle. Under this methodology, the CONFU-B presents lower risks than other multi-recycling strategies in the first 100 years. For a 10% rate of discount of risk, the CONFU-B risks are comparable to the once-through cycle. The longer term risk favors recycling due to the limited accumulation of repository risk. 2006-06-01T00:00:00Z https://hdl.handle.net/1721.1/75212 Experimental Determination of Thermal Conductivity of a Lead- Bismuth, Eutectic-Filled Annulus Carpenter, David M.; Kohse, Gordon E. In order to obtain an accurate prediction of the thermal behavior of an annular fuel assembly (see MIT-NFC-PR-048 for a description of the rods), the thermal conduction of the region from the outside of the fuel capsule to the reactor coolant (within the test assembly) must be known. The effective thermal conductivity of this composite structure is dependent on the interaction of the parts via various physical phenomena, and therefore is difficult to infer accurately from the conductivity of the constituent materials. A mock-up of the annular fuel rod containment thimble was created to allow the conductivity of the annular lead bismuth eutectic-filled gap to be measured. An electric rod heater was used to provide temperatures similar to the in-core environment, and conductivity was determined based on thermocouple temperature readings at various points across the gap. A second series of experiments substituted a steel tube for the aluminum thimble, and used a bucket of stationary water as coolant. The purpose of these changes was to increase the temperature of the eutectic and achieve a larger melted fraction, while at the same time creating a large enough temperature drop across the gap to allow reliable measurements. A third series of experiments refined the setup and were able to produce more precise measurements of the thermal conductivity. The measured conductivities were between 4 and 8 W/m-K, much lower than the reported conductivity of the lead bismuth at about 10 W/m-K. The difference must be attributed to thermal resistances at the eutectic-aluminum and eutectic-steel interfaces. This, and the inherent difficulty of measuring the interface temperature due to the finite width of the thermocouples and the existence of sharp thermal gradients makes it difficult to further reduce the uncertainty in the measured conductivity. 2005-06-01T00:00:00Z https://hdl.handle.net/1721.1/75211 Feasibility of Partial ZrO[subscript 2] Coatings on Outer Surface of Annular UO[subscript 2] Pellets to Control Gap Conductance Feinroth, H.; Yuan, Y. The viability of depositing a thin porous coating of zirconia on the outer surface of an annular UO[subscript 2] pellet is investigated experimentally. Such a layer has been proposed to buffer the contact between the fuel and cladding, thus maintaining an appropriate balance of heat transfer from the pellet to the outer and inner cladding. MIT and Gamma Engineering commissioned laboratory studies of the feasibility of depositing a controlled thickness of porous zirconia on an oxide surface. Experiments were conducted at the Thermal Spray Laboratory at SUNY-Stonybrook to produce a thin layer of Yttria Stablized Zirconia (YSZ) on alumina wafers. The experiments concluded that it is possible to use plasma spray guns to produce 50% porous layers of 15–30 micrometer thickness. Measurements were conducted at the Vitreous State Laboratory of the Catholic University of America to determine the thermal conductance of aluminazircaloy and alumina-zircaloy-YSZ sandwiches as a function of applied pressure. A relation is developed to predict the conductance at such surfaces. 2005-06-01T00:00:00Z https://hdl.handle.net/1721.1/75210 Wire Wrapped Hexagonal Pin Arrays for Hydride Fueled PWRs Diller, Peter; Todreas, Neil E. This work contributes to the Hydride Fuels Project, a collaborative effort between UC Berkeley and MIT aimed at investigating the potential benefits of hydride fuel use in light water reactors (LWRs). Core design is accomplished for both hydride and oxide-fueled cores over a range of geometries via steadystate and transient thermal hydraulic analyses, which yield the maximum power, and fuel performance and neutronics studies, which provide the achievable discharge burnup. The final optimization integrates the outputs from these separate studies into an economics model to identify geometries offering the lowest cost of electricity, and provide a fair basis for comparing the performance of hydride and oxide fuels. This work focuses on the steady-state and transient thermal hydraulic as well as economic analyses for PWR cores utilizing wire wraps in a hexagonal array with UZrH[subscript 1.6] and UO[subscript 2]. It was previously verified that square and hexagonal arrays with matching rod diameters and H/HM ratio have the same thermal hydraulic performance. In this work, this equivalence is extended to hexagonal wire wrap arrays, and verified by comparing the thermal hydraulic performance of a single hexagonal wire wrap core with its equivalent square array core with grid spacers. A separate neutronics equivalence is developed, based on the assumption that arrays with matching rod diameters and H/HM ratios will have identical neutronic performance. Steady-state design limits were separated into hard limits, which must be satisfied, or soft limits, which serve to keep the design reasonable. Design limits were placed on the pressure drop, critical heat flux (CHF), vibrations, and fuel and cladding temperature. Vibrations limits on the wire wrap assemblies were imposed for flow induced vibrations (FIV) and thermal hydraulic vibrations (THV). An analysis of the fretting wear of wire wraps indicated that wire wraps outperformed the analogous fretting wear analysis for grid spacers. A CHF study found wire wraps to outperform grid spacers. LOCA and overpower transient analyses were performed for wire wraps. The overpower transient was analyzed over a range of geometries, and found to be more limiting than the steady-state analysis. The LOCA was analyzed for various powers at the reference geometry and another geometry of interest. Through all of these analyses, it was determined that the thermal hydraulic performance of UZrH1.6 and UO2 are very similar. The optimal wire wrap designs were found to have significantly higher maximum powers than the reference core, allowing for uprates up to ~54%. This is due to improved vibrations, pressure drop, and CHF. The steady-state and transient analyses were combined with fuel performance and neutronic studies into an economics model that determines the optimal geometries for incorporation into existing PWR’s. The model also provides a basis for comparing the performance of UZrH[subscript 1.6] to UO[subscript 2] for a range of core geometries. Results presented herein show cost savings for oxide fuel with wire wraps over grid spacers of at least 0.8 mils/kWe-hr, or 4%, due to power increases predicted by the thermal hydraulic analyses. Wire wrap UZrH[subscript 1.6] has a COE savings over UO[subscript 2] of 0.7 mils/kWe-hr, or 4%. Due to the large power uprates possible, cost savings of up to 10.9 mils/kWe-hr, or 40%, can be achieved, with a UZrH[subscript 1.6] wire wrap uprate instead of building a new core. 2006-01-01T00:00:00Z https://hdl.handle.net/1721.1/75209 Thermal Hydraulic and Economic Analysis of Grid-Supported Hydride and Oxide Fueled PWRs Shuffler, C.; Trant, J.; Todreas, Neil E.; Romano, A. This report advances the Hydride Fuels Project, a collaborative effort between UC Berkeley and MIT aimed at investigating the potential benefits of hydride fuel use in light water reactors (LWRs). This effort involves implementing an appropriate methodology for design and optimization of hydride and oxide fueled cores. Core design is accomplished for a range of geometries via steady-state and transient thermal hydraulic analyses, which yield the maximum power, and fuel performance and neutronics studies, which provide the achievable discharge burnup. The final optimization integrates the outputs from these separate studies into an economics model to identify geometries offering the lowest cost of electricity, and provide a fair basis for comparing the performance of hydride and oxide fuels. This report builds on the considerable work which has already been accomplished on the project. More specifically, it focuses on the steady-state and transient thermal hydraulic and economic analyses for pressurized water reactor (PWR) cores utilizing UZrH[subscript 1.6] and UO[subscript 2]. A previous MIT study established the steady-state thermal hydraulic design methodology for determining maximum power from square array PWR core designs. In lieu of a detailed vibrations analysis, the steady-state thermal hydraulic analysis imposed a single design limit on the axial flow velocity. The wide range of core geometries considered and the large power increases reported by the study makes it prudent to refine this single limit approach. This work accomplishes this by developing and incorporating additional design limits into the thermal hydraulic analysis to prevent excessive rod vibration and wear. The vibrations and wear mechanisms considered are: vortex-induced vibration, fluid-elastic instability, turbulence-induced vibration, fretting wear, and sliding wear. Further, the transients investigated are an overpower transient, a large break loss of coolant accident (LBLOCA), and a complete loss of flow accident. In parallel with this work, students at UC Berkeley and MIT have undertaken the neutronics and fuel performance studies. With these results, and the output from the steady-state thermal hydraulic analysis with vibrations and wear imposed design limits, as well as transient thermal hydraulic analysis, an economics model is employed to determine the optimal geometries for incorporation into existing PWRs. The model also provides a basis for comparing the performance of UZrH[subscript 1.6] to UO[subscript 2] for a range of core geometries. Though this analysis focuses only on these fuels, the methodology can easily be extended to additional hydride and oxide fuel types, and will be in the future. Results presented herein do not show significant cost savings for UZrH[subscript 1.6], primarily because the power and energy generation per core loading for both fuels with square arrays supported by grid spacers are similar. Furthermore, the most economic geometries typically do not occur where power increases are reported by the thermal hydraulics. However, preliminary analysis with the lower pressure drop characteristics of wire wraps compared to grids suggest that hexagonal array cores with wire wraps will allow tight ( P over D ≺ 1.25) packing which yield significantly better power performance. This should allow hydride fuel to outperform oxide fuel since this tight core region is not accessible to oxide cores, because of neutronic constraints. 2006-09-01T00:00:00Z https://hdl.handle.net/1721.1/75176 Innovative Fuel Designs for High Power Density Pressurized Water Reactor Feng, D.; Kazimi, Mujid S.; Hejzlar, Pavel One of the ways to lower the cost of nuclear energy is to increase the power density of the reactor core. Features of fuel design that enhance the potential for high power density are derived based on characteristics of the pressurized water reactor (PWR) and its related design limits. Those features include: large fuel surface to volume ratio, small fuel thickness, large fuel rod stiffness, low core pressure drop and an open fuel lattice design. Three types of fuel designs are evaluated from the thermal-hydraulic point of view: conventional solid cylindrical fuel rods, internally and externally cooled annular fuel rods, and spiral cross-geometry fuel rods, with the major effort allocated to analyzing the annular fuel. Limits of acceptable power density in solid cylindrical fuel rods are obtained by examining the effects of changing the core operation parameters, fuel rod diameter and rod array size. It is shown that the solid cylindrical geometry does not meet all the desired features for high power density well, and its potential for achieving high power density is limited to 20% of current PWR power density, unless the vibration problems at the coolant higher velocity are overcome. The internally and externally cooled annular fuel potential for achieving high power density is explored, using a whole core model. The best size of fuel rods that fits in the reference assembly dimension is a 13x13 array, since the hot red will have a balanced MDNBR in the inner and outer channels. With proportional increase in coolant flow rate, this annular fuel can increase PWR power density by 50% with the same DNBR margin, while reducing by 1000 ºC the peak fuel temperature. Five issues involving manufacturing tolerances, oxide growth on rod surfaces, inner and outer gap conductances asymmetry, MDNBR sensitivity to changes in core operation parameter and resistance to instabilities were also evaluated. It is found that the main uncertainty for this design is associated with the heat split between the inner and outer channels due to differences in the thermal resistances in the two fuel-clad gaps. Annular fuel is found to be resistant to flow instabilities, such as Ledinegg instability and density wave oscillation due to high system pressure and one-phase flow along most of the hot channel length. Similar power density uprate is found possible for annular fuel in a hexagonal lattice. Large break loss of coolant accident (LBLOCA) for the reference Westinghouse 4-loop PWR utilizing annular fuel at 150% power is analyzed using RELAP, under conservative conditions. The blowdown peak cladding temperature (PCT) is found to be lower because of the low operating fuel temperature, but the flow rate from the safety injection system needs to be increased by 50% to remove the 50% higher decay heat. Loss of flow analysis also showed better performance of the annular fuel because of its low stored energy. The fuel design that best meets the desired thermal and mechanical features is the spiral 3 cross-geometry rods. The dimensions of this type of fuel that can be applied in the reference core were defined. Thermal-hydraulic whole-core evaluations were conducted with cylindrical fuel rod simplification, and critical heat flux modification based on the heat flux lateral non-uniformity in the cross geometry. This geometry was found to have the potential to increase PWR power density by 50%. However, there are major uncertainties in the feasibility and costs of manufacturing this fuel. 2005-09-01T00:00:00Z https://hdl.handle.net/1721.1/75174 NEUTRONIC AND THERMAL HYDRAULIC DESIGNS OF ANNULAR FUEL FOR HIGH POWER DENSITY BWRS Morra, P.; Xu, Z.; Hejzlar, Pavel; Saha, P.; Kazimi, Mujid S. As a promising new fuel for high power density light water reactors, the feasibility of using annular fuel for BWR services is explored from both thermal hydraulic and neutronic points of view. Keeping the bundle size similar to conventional GE 8×8 solid fuel bundles, annular fuel bundles of 5×5 and 6×6 lattices, that have increased thermal output potential, are explored. The large annular fuel rods allow both external and internal cooling, which increases the fuel surface to volume ratio and significantly reduces the fuel temperature. A methodology is developed and VIPRE code calculations are performed to select the best annular fuel bundle design on the basis of its Critical Power Ratio (CPR) performance. Within the limits applied to the reference solid fuel, the CPR margin in the 5x5 and 6x6 annular fuel bundles is traded for an increase in power density. It is found that the power density increase with annular fuel in BWRs may be limited to 23%. This is smaller than possible for PWRs due to the difference in the mechanisms that control the critical thermal conditions of the two reactors. The neutronic aspects of annular fuel in BWRs, including the reactivity history, power distribution, and burnup characteristics, are investigated. Results are compared to the conventional BWR solid fuel bundle for the same power density and total energy. In general switching to annular fuel implies smaller neutronic differences than in the PWR case. The local peaking factors are found to be similar or slightly better than those of the solid fuel bundle. To maintain the same fuel cycle length, the burnup needs to be increased even for the same bundle energy output, due to reduced fuel loading. These results are based on two dimensional bundle models using the CASMO code, whose validity has been checked using MCNP models. In summary, the annular fuel could be a profitable alternative to the solid fuel due to neutronic and thermal advantages. Further study of the trends observed in this report are needed to increase their certainty. 2004-12-01T00:00:00Z https://hdl.handle.net/1721.1/75173 Alternative Fuel Cycle Strategies For Nuclear Power Generation In The 21st Century Boscher, T.; Hejzlar, Pavel; Kazimi, Mujid S.; Todreas, Neil E.; Romano, A. The deployment of fuel recycling through either CONFU (COmbined Non-Fertile and UO[subscript 2] fuel) thermal water-cooled reactors (LWRs) or fast ABR (Actinide Burner Reactor) reactors is compared to the Once-Through LWR reactor system in terms of accumulation of actinides over the next 100 years under the assumption of a growing worldwide demand for nuclear energy. It is assumed that the growth rate is about 2.1% per year up to 2053, with alternative scenarios after that date. The transuranics (TRU) stored in temporary repositories, the TRU sent to permanent repositories, the system cost and a vulnerability index toward proliferation are calculated by the CAFCA code and taken as key figures of merit. Deployment of the ABRs is assumed to occur later (2028) than the CONFU LWRs (2015), whose technology requires less extensive additional R&D. Through 2050 the CONFU strategy performs better than the ABR strategy. The CONFU LWRs in our model yield zero net TRU incineration while the ABRs have a net consumption of TRU. Compared to the Once-Through strategy, by 2050 the CONFU (respectively ABR) strategy reduces by about 35% (respectively 9%) the total inventory of TRU in the system. This reduction corresponds to the TRU production being avoided by CONFU LWRs or being incinerated in ABRs compared to the TRU produced in the traditional LWRs used in the Once-Through strategy. By 2100, the CONFU and the ABR strategies would have reduced the worldwide TRU inventory by 62% compared to the Once- Through case with the CONFU strategy incinerating more TRU than in the ABR strategy. The three strategies are also discussed with regard to uranium ore availability, repository need, and processing plants need. It is interesting to note that with either recycling strategies the total capacity for separation of spent UO2 constituents need 10 to 12 separation plants with a capacity of 2000 MTHM/year. Furthermore, only one TRU recycling plant from fertile-free fuel would be needed at a capacity of 250 MTHM/year up to 2050. The economic analysis shows that both closed fuel cycles are more expensive than the reference Once-Through scheme. The total cost of electricity production is expected to be 3 mills/kWhe, or about 10%, larger than the Once-Through cycle case, if the spent fuel separation is paid off by the electricity sales from the resulting fuel. The timing of collection of fuel cycle costs significantly affects the cost of electricity. Paying for fuel separation by the sales of the electricity producing the spent fuel to be reprocessed later has a smaller effect on the cost of electricity in the advanced fuel cycles (between 1 and 2 mills/kWhe or between 3 and 6%) compared to the cost of electricity in the Once-Through strategy. From a policy point of view, an index of vulnerability toward proliferation is defined and gives an advantage to the advanced fuel cycles. The large amount of heavy metal in the repository and the long life time of this repository penalize the Once-Through strategy. However the results are sensitive to the accessibility factor assigned to the repository which is, as all accessibility factors, a subjective value that is not precisely defined. Moreover, worldwide cooperation to implement the two advanced strategies and the challenges this implementation could face are discussed. The use of a single behaviour mode throughout the world implies an unlikely perfect cooperation between countries that do not have the same capabilities or incentives to choose among the advanced fuel cycle strategies. Revision 1 2005-06-01T00:00:00Z https://hdl.handle.net/1721.1/75171 Combining Thorium with Burnable Poison for Reactivity Control of a Very Long Cycle BWR Inoue, Y.; Pilat, Edward E.; Xu, Z.; Kazimi, Mujid S. The effect of utilizing thorium together with gadolinium, erbium, or boron burnable absorber in BWR fuel assemblies for very long cycle is investigated. Nuclear characteristics such as reactivity and power distributions are evaluated using CASMO-4. Without thorium, the results show that gadolinium enriched in Gd-157 has the lowest reactivity swing throughout the cycle. However, the local peaking factor (LPF) in the assembly at beginning-of-life (BOL) is high. The erbium case shows more reactivity swing but the LPF is lowest of all three cases. B4C case has the highest reactivity at BOL which would have to be suppressed by control rods. The most important advantage of B4C over others is the saving of uranium inventory needed to achieve the target exposure of 15 effective full power years (EFPY). Further analysis for transient conditions must be performed to ensure meeting all transient limits. Use of thorium in place of some burnable poison makes it possible to save some uranium enrichment while achieving equivalent discharge burnup to the case without thorium, but only by about 1 %. The benefit is small because almost the same amount of burnable poison is always required for suppressing excess reactivity throughout the cycle. Since Th-232 functions more like U-238 than burnable poison, this limits the allowed thorium to extend discharge burnup. Since all fuel assembly designs in this study have the same target exposure of 15EFPY, the economic performance of each design can be compared based on the amount and enrichment of both uranium and burnable absorbers for each fuel design. The B4C-Al fuel is most economical in overall cost even with large uncertainties. The overall cost of gadolinium and erbium cases are concluded to be about the same when large uncertainties are considered. 2004-06-01T00:00:00Z https://hdl.handle.net/1721.1/75166 Optimization of the LWR Nuclear Fuel Cycle for Minimum Waste Production Shwageraus, Eugene; Hejzlar, Pavel; Kazimi, Mujid S. The once through nuclear fuel cycle adopted by the majority of countries with operating commercial power reactors imposes a number of concerns. The radioactive waste created in the once through nuclear fuel cycle has to be isolated from the environment for thousands of years. In addition, plutonium and other actinides, after the decay of fission products, could become targets for weapon proliferators. Furthermore, only a small fraction of the energy potential in the fuel is being used. All these concerns can be addressed if a closed fuel cycle strategy is considered offering the possibility for partitioning and transmutation of long lived radioactive waste, enhanced proliferation resistance, and improved utilization of natural resources. It is generally believed that dedicated advanced reactor systems have to be designed in order to perform the task of nuclear waste transmutation effectively. The development and deployment of such innovative systems is technically and economically challenging. In this work, a possibility of constraining the generation of long lived radioactive waste through multi-recycling of Trans-uranic actinides (TRU) in existing Light Water Reactors (LWR has been studied. Thorium based and fertile free fuels (FFF) were analyzed as the most attractive candidates for TRU burning in LWRs. Although both fuel types can destroy TRU at comparable rates (about 1150 kg/GWe-Year in FFF and up to 900 kg/GWe-Year in Th) and achieve comparable fractional TRU burnup (close to 50a/o), the Th fuel requires significantly higher neutron moderation than practically feasible in a typical LWR lattice to achieve such performance. On the other hand, the FFF exhibits nearly optimal TRU destruction performance in a typical LWR fuel lattice geometry. Increased TRU presence in LWR core leads to neutron spectrum hardening, which results in reduced control materials reactivity worth. The magnitude of this reduction is directly related to the amount of TRU in the core. A potential for positive void reactivity feedback limits the maximum TRU loading. Th and conventional mixed oxide (MOX) fuels require higher than FFF TRU loading to sustain a standard 18 fuel cycle length due to neutron captures in Th232 and U238 respectively. Therefore, TRU containing Th and U cores have lower control materials worth and greater potential for a positive void coefficient than FFF core. However, the significantly reduced fuel Doppler coefficient of the fully FFF loaded core and the lower delayed neutron fraction lead to questions about the FFF performance in reactivity initiated accidents. The Combined Non-Fertile and UO[subscript 2] (CONFU) assembly concept is proposed for multirecycling of TRU in existing PWRs. The assembly assumes a heterogeneous structure where about 20% of the UO[subscript 2] fuel pins on the assembly periphery are replaced with FFF pins hosting TRU generated in the previous cycle. The possibility of achieving zero TRU net is demonstrated. The concept takes advantage of superior TRU destruction performance in FFF allowing minimization of TRU inventory. At the same time, the core physics is still dominated by UO[subscript 2] fuel allowing maintenance of core safety and control characteristics comparable to all-UO[subscript 2]. A comprehensive neutronic and thermal hydraulic analysis as well as numerical simulation of reactivity initiated accidents demonstrated the feasibility of TRU containing LWR core designs of various heterogeneous geometries. The power peaking and reactivity coefficients for the TRU containing heterogeneous cores are comparable to those of conventional UO[subscript 2] cores. Three to five TRU recycles are required to achieve an equilibrium fuel cycle length and TRU generation and destruction balance. A majority of TRU nuclides reach their equilibrium concentration levels in less than 20 recycles. The exceptions are Cm246, Cm248, and Cf252. Accumulation of these isotopes is highly undesirable with regards to TRU fuel fabrication and handling because they are very strong sources of spontaneous fission (SF) neutrons. Allowing longer cooling times of the spent fuel before reprocessing can drastically reduce the SF neutron radiation problem due to decay of Cm244 and Cf252 isotopes with particularly high SF source. Up to 10 TRU recycles are likely to be feasible if 20 years cooling time between recycles is adopted. Multi-recycling of TRU in the CONFU assembly reduces the relative fraction of fissile isotopes in the TRU vector from about 60% in the initial spent UO[subscript 2] to about 25% at equilibrium. As a result, the fuel cycle length is reduced by about 30%. An increase in the enrichment of UO[subscript 2] pins from 4.2 to at least 5% is required to compensate for the TRU isotopics degradation. The environmental impact of the sustainable CONFU assembly based fuel cycle is limited by the efficiency of TRU recovery in spent fuel reprocessing. TRU losses of 0.1% from the CONFU fuel reprocessing ensure the CONFU fuel cycle radiotoxicity reduction to the level of corresponding amount of original natural uranium ore within 1000 years. The cost of the sustainable CONFU based fuel cycle is about 60% higher than that of the once through UO[subscript 2] fuel cycle, whereas the difference in total cost of electricity between the two cycles is only 8%. The higher fuel cycle cost is a result of higher uranium enrichment in a CONFU assembly required to compensate for the degradation of TRU isotopics and cost of reprocessing. The major expense in the sustainable CONFU fuel cycle is associated with the reprocessing of UO[subscript 2] fuel. Although reprocessing and fabrication of FFF pins have relatively high unit costs, their contribution to the fuel cycle cost is marginal as a result of the small TRU throughput. Reductions in the unit costs of UO[subscript 2] reprocessing and FFF fabrication by a factor of two would result in comparable fuel cycle costs for the CONFU and conventional once through cycle. An increase in natural uranium prices and waste disposal fees will also make the closed fuel cycle more economically attractive. Although, the cost of the CONFU sustainable fuel cycle is comparable to that of a closed cycle using a critical fast actinide burning reactor (ABR), the main advantage of the CONFU is the possibility of fast deployment, since it does not require as extensive development and demonstration as needed for fast reactors. The cost of the CONFU fuel cycle is projected to be considerably lower than that of a cycle with an accelerator driven fast burner system. 2003-10-01T00:00:00Z https://hdl.handle.net/1721.1/75164 MODELING THE PERFORMANCE OF HIGH BURNUP THORIA AND URANIA PWR FUEL Long, Y.; Kazimi, Mujid S.; Ballinger, Ronald G.; Meyer, J. E. Fuel performance models have been developed to assess the performance of ThO[subscript 2]-UO[subscript 2] fuels that can be operated to a high burnup up to 80-100MWd/kgHM in current and future Light Water Reactors (LWRs). Among the various issues raised in high burnup fuel applications, the pellet rim effect, fission gas release (FGR), and response to reactivity initiated accidents (RIA) were of special interest in this work. These phenomena were modeled by modifying the NRC licensing codes FRAPCON-3 for normal operation and FRAP-T6 for transient conditions. These models were verified and compared to the results of previous thorium fuel studies and high burnup uranium fuel evaluations. The buildup of plutonium in the outer rim of LWR UO[subscript 2] pellets has been observed to create a region of high fuel burnup, fission gas buildup and high porosity at the fuel rim. The power distribution of the thoria and urania fuel was calculated using a neutronics code MOCUP. Due to the lower build-up of Pu-239 (less U-238 in ThO[subscript 2]-UO[subscript 2] fuel) and flatter distribution of U-233 (less resonance capture in Th-232), thoria fuel experiences a much flatter power distribution and thus has a less severe rim effect than UO[subscript 2] fuel. To model this effect properly, a new model, THUPS (Thoria-Urania Power Shape), was developed, benchmarked with MOCUP and adapted into FRAPCON-3. Additionally a porosity model for the rim region was introduced at high burnup to account for the larger fuel swelling and degradation of the thermal conductivity. The mechanisms of fission gas release in ThO[subscript 2]-UO[subscript 2] fuel have been found similar to those of UO[subscript 2] fuel. Therefore, the general formulations of the existing fission gas release models in FRAPCON-3 were retained. However, the gas diffusion coefficient in thoria was adjusted to a lower level to account for the smaller observed gas release fraction in the thoria-based fuel. To model accelerated fission gas release at high burnup properly, a new athermal fission gas release model was developed. Other modifications include the thoria fuel properties, fission gas production rate, and the corrosion model to treat advanced cladding materials. The modified version of FRAPCON-3 was calibrated using the measured fission gas release data from the Light Water Breeder Reactor (LWBR) program. Using the new model to calculate the gas release in typical PWR hot pins gives data that indicate that the ThO[subscript 2]-UO[subscript 2] fuel will have considerably lower fission gas release beyond a burnup of 50 MWd/kgHM. Investigation of the fuel response to an RIA included: (1) reviewing industry simulation tests to understand the mechanisms involved, (2) modifying FRAP-T6 code to simulate the RIA tests and investigate the key contributors to fuel failure (thermal expansion, gaseous swelling, cladding burst stress), and (3) assessing thoria and urania performance during RIA event in typical LWR situations. ThO[subscript 2]-UO[subscript 2] fuel has been found to have better performance than UO[subscript 2] fuel under RIA event conditions due to its lower thermal expansion and a flatter power distribution in the fuel pellet (less power and less fission gas in the rim region). Overall, thoria has been found to have better performance than urania in both normal and off-normal conditions. However, calculations using the modified FRAPCON-3 showed that the internal pressure and cladding corrosion at the required high burnup of 80- 100MWd/kgHM are not acceptable with the current fuel design. Therefore, advanced fuel designs (including larger gas plenum, larger fuel grains, advanced cladding materials), and carefully designed operating strategy (i.e. decreasing power history) were assessed and the results showed that the targeted high burnup can be achieved. Further investigation of burnup issues is needed, such as the distribution of hydrogen in the cladding for heterogeneous fuels, and response of high pressure fuel pins to a loss of coolant accident, in order to assure satisfactory high burnup behavior. 2002-07-01T00:00:00Z https://hdl.handle.net/1721.1/75153 High Burnup Fuels for Advanced Nuclear Reactors Oggianu, S. M.; Christensen, Holly Colleen No; Kazimi, Mujid S. The goal of this work is to select the best candidate fuel materials to deliver high burnup in advanced light water reactors. Uranium and thorium based fuels are considered. These fuel materials must be able to withstand nearly double the burnup of current LWRs in high irradiation fields. Reactor economics, safety, proliferation resistance, fuel reprocessing and spent fuel disposal are the most important factors to be addressed. High burnup will provide the opportunity for uninterrupted operation over long periods of time, reduction of spent fuel volume and improvement of proliferation resistance. Thus, effective power cycle maintenance and fuel management and reduced fuel storage needs will lead to more economic operation. Several uranium and thorium fuel forms are analyzed to predict their capability to withstand high burnups. Their fuel cycle cost is also considered. To compare the fuel options, simple indices characterizing the behavior of the materials at high burnup are defined. Indices for the thermal stress capability, stored energy and margin for melting are derived from non-dimensional analyses. To evaluate the fuel pin lifetime, a simplified fuel performance analysis code, FUELSIM (FUEL SIMulation code) was developed. The code utilizes the VENSIM simulation system, which allows for great flexibility in the change of governing relations, permits sensitivity analysis, and facilitates graphical outputs. Based on the sensitivity analysis by FUELSIM, dominant parameters are identified and a simplified expression is developed for predicting the increase in the pin internal pressure with burnup. For each material, we obtain a maximum attainable burnup at a given smear density. Cladding strain, internal pressure and fuel melting (or phase-change) temperature are the limiting factors used to obtain these burnups. From neutronic reactivity considerations, the needed [superscript 235]U enrichment can be specified. Thus, the fuel cycle cost for each material and smear density can be estimated. Metals, oxides, carbides and nitrides of uranium and thorium were examined. Although the results show that UN provides the highest potential for attaining high burnup and economic application in once-through cycles, it has limited compatibility with water. UO[subscript 2], at 90- 95% smear density, continues being the most feasible option as a nuclear material. Also, ThO[subscript 2]/UO[subscript 2] seems to offer as good or better potential performance and economics as UO[subscript 2]. However, more reliable data on the irradiation behavior of the different materials is needed before a definitive conclusion can be drawn. Also important in the evaluation of thorium/uranium cycles are attributes that were not considered here. These include the reduction in spent fuel volume, the improvement in proliferation resistance, possible power uprates and allowance for a higher peaking factor that may be possible by taking advantage of the increased margin that results from using fuels with lower stored energies. 2001-05-01T00:00:00Z https://hdl.handle.net/1721.1/75150 A Systematic Study of Moderation Effects On Neutronic Performance of UO[subscript 2] Fueled Lattices Xu, Z.; Driscoll, Michael J.; Kazimi, Mujid S. This report addresses the physics of reactor cores that can be operated for 10 to 15 years without refueling — inspired by the objective of enhanced nuclear fuel cycle performance with regard to economics and resistance to weapon proliferation. Proliferation resistance is a primary consideration in this design. The long life operation reduces the routine access to the internals of the reactor vessel, therefore reducing the possibility for clandestine production of nuclear weapons. Additionally, reduction of reactor shutdown time can result in improved safety and economics. As a first step, the most promising fuel lattice characteristics to achieve long life from a physics point of view are studied. These studies also define the design tradeoffs involved in conceptualizing such cores. Moderation effects on UO[subscript 2] fueled lattices are analyzed systematically using state-of-the-art computer codes (CASMO-4 and MOCUP). The standard 4-loop Westinghouse pressurized water reactor (PWR) is taken as our reference core and single unit cell analysis is employed. To change the moderator-to-fuel ratio, which is characterized by the hydrogen-to-heavy-metal (H/HM) atom number ratio, various methods are adapted including varying water density, fuel density, fuel rod diameter, and fuel rod pitch. Higher burnup potential as well as longer core endurance (burnup times heavy metal mass) would be desirable. For a given initial enrichment, the results show that higher reactivity-limited burnup is achievable by either a more wet lattice or much drier lattice than normal. However, epithermal lattices are distinctly inferior performers. In terms of longer endurance, current PWR lattice parameters are about the optimum. Higher burnup and endurance can be achieved with higher initial enrichment. Characteristics of the spent fuel from high burnup UO[subscript 2] fueled lattices have been examined. The variation of isotopic mix and quantity of plutonium with moderator-to-fuel ratio for UO[subscript 2] fueled lattices has been studied to clarify the impact on its proliferation resistance. And Np production as a function of H/HM has been computed as a measure of long-term radiological hazard for high level waste disposal. It is shown that Np is mildly affected by the H/HM ratio and the current PWR lattice is close to optimum configuration. However, high burnup is significantly beneficial as a way to make the plutonium isotopic mix less attractive as a weapon material. 2001-05-01T00:00:00Z https://hdl.handle.net/1721.1/75140 ON THE USE OF THORIUM IN LIGHT WATER REACTORS Kazimi, Mujid S.; Czerwinski, Kenneth R.; Driscoll, Michael J.; Hejzla, P.; Meyer, J. E. The advantages and disadvantages of the use of thorium bearing fuel in light water reactors have been examined several times from the beginning of the nuclear energy era until the late seventies. The recent motivation for re-examining the use of thorium in light water reactors' once-through fuel cycle is enhancing the cycle proliferationresistance due to reduced plutonium production. Additionally, economic benefits from reducing the initial enrichment needs of high burnup fuels may be obtained. Similarly, it may be possible to rely on the higher melting point and higher thermal conductivity of ThO[subscript 2] to enhance the safety margin of the core. Thorium dioxide is the highest stable oxide form of thorium, which may further improve the spent fuel repository performance. The information obtained in previous studies is reviewed to assess its suitability for application to the current fuel cycle conditions. It is concluded that the thorium fuel experience of the past is insufficient to make a judgement on the feasibility and performance of the thorium bearing fuels in the reactors operating under current conditions. The needs for new research and development efforts in the areas of neutronics, fuel behavior, safety and waste performance are outlined. 1999-04-01T00:00:00Z https://hdl.handle.net/1721.1/75139 FUEL PERFORMANCE ANALYSIS OF EXTENDED OPERATING CYCLES IN EXISTING LWRs Handwerk, C. S.; Meyer, J. E.; Todreas, Neil E. An integral part of a technical analysis of a core design, fuel performance is especially important for extended operating cycles since the consequences of failed fuel are greater for this operating strategy than for current practice. This stems mainly from the fact that extended cycles offer a unique benefit by running longer without interruption; poor fuel performance, i.e. failed fuel, would degrade this benefit. The issues in this research are assessed only at the steady-state level, as a foundation for the consideration of Anticipated Operational Occurrences (AOOs) and transient conditions, which are certain to present greater challenges to nuclear fuel performance due to their more severe conditions. Even at this preliminary steady state level, extended cycle operation is found to exacerbate several fuel performance issues, resulting mainly from the fact that some fuel in an extended operating cycle is operated at higher powers over part of the core life and does not have the benefit of shuffling. In order to accurately quantify the fuel performance effects of extended cycle operation, a pseudo or "envelope" pin is created, which represents the operating characteristics of the highest power fuel rod in the core at a given pin burnup interval. This envelope pin was created for both extended cycle and current practice, so that extended cycle results could be compared to both existing licensing limits and current practice. While this approach is somewhat conservative, it is the simplest way to evaluate fuel performance in an extended cycle core where the location of the limiting fuel rod changes often and operates at higher powers for prolonged periods of time. The US Nuclear Regulatory Commission's Standard Review Plan's Sections 4.2 and 4.4 are used as the basis for the criteria that should be evaluated in this report, since these are the relevant sections of the document that prescribes the licensing limits and criteria for nuclear fuel design. From this document, ten steady state fuel performance issues are identified: (1) stress and strain, (2) fatigue cycling, (3) fretting, (4) waterside corrosion, (5) axial growth and rod bowing, (6) rod internal pressure, (7) primary hydriding, (8) cladding collapse, (9) cladding overheating, and (10) fuel centerline melt. Of these ten issues, (7) and (8) were found to be not uniquely affected by extended cycle operation. While (9) and (10) are found to not be concerns for extended cycle operation, the higher powers at which extended operating cycles can operate degrade some of the margin for transient effects, which is more of a significant concern for (9). (1) and (5) are predicted to be worse for both BWRs and PWRs when compared to current practice, and (4) and (6) are projected to present greater challenges for PWRs. Additionally, (2) is the only factor that is predicted to actually be better for extended cycle operation in both the BWR and PWR while (4) was predicted to have less of an effect in BWRs, given the comparable operating powers and shorter in-core residence time for the extended cycle case. The effects of the proposed new operating strategy on (3) were uncertain. Of all ten issues, (5) seemed to be the most problematic, as no solution was readily available. Solutions to other issues included improved assembly grid design (3), water chemistry control (4), annular fuel pellets (6), and, potentially, increasing the number of fuel rods per assembly (1,4,6,10). 1998-01-01T00:00:00Z https://hdl.handle.net/1721.1/75135 An Alternative to Gasoline: Synthetic Fuels from Nuclear Hydrogen and Captured CO[subscript 2] Middleton, B. D.; Kazimi, Mujid S. The motivation for this study stems from two concerns. The first is that carbon dioxide from fossil fuel combustion is the largest single human contribution to global warming. The use of nuclear power to produce hydrogen on a global scale for any of various possible end uses would reduce the net amount of carbon dioxide emitted into the atmosphere. The second concern is in regard to U.S. dependence on foreign oil. Over 58% of petroleum used by the US in 2002 was imported and most likely a higher fraction is being imported today. With the majority of this oil originating in highly volatile Middle Eastern countries, there is a potential threat to stability in the US energy market. This study was conducted to determine the extent to which nuclear power can contribute to a transition in the transportation sector; away from an infrastructure that places the US at risk for depending largely on foreign oil and that makes it inevitable that large quantities of carbon dioxide will be emitted into the atmosphere. Several scenarios are reviewed in this study for using nuclear hydrogen in transportation, including: • Combining hydrogen with carbon dioxide captured from fossil fired plants to produce liquid fuel • Using nuclear power to aid in the recovery of oil from tar sands or shale oil Initially, a review of the literature pertaining to the potential contribution of nuclear power to hydrogen production is performed. Two approaches for producing hydrogen from water are found that have significant literature related to the subject. These cycles are High Temperature Steam Electrolysis and the Sulfur Iodine Cycle. The UT-3 cycle is also promising but does not seem to offer the same advantages with respect to energy efficiency. This work focuses on the High Temperature Steam Electrolysis option. A review of possible nuclear reactor concepts is also performed. Many advanced concepts have been proposed, a large number of which show potential in producing hydrogen. However, there are drawbacks to many of them for several reasons. The high temperatures needed eliminate some reactors while lack of operational experience eliminates others. Ultimately, the two concepts that are proposed for hydrogen production in the literature found are the High-Temperature Gas Cooled Reactor (HTGR), which uses Helium coolant, and a modified version of the Advanced Gas Reactor (AGR) using supercritical CO[subscript 2] as the coolant (S-AGR). The reactor concepts that are chosen for aiding production of oil from tar sands are the Advanced Candu Reactor (ACR-700), the Pebble Bed Modular Reactor (PBMR), and the Advanced Passive pressurized water reactor(AP600). A detailed study of how nuclear power can contribute to production of shale oil has not been performed. Therefore, the section dealing with this particular possibility is much less in depth and more speculative. However, some preliminary calculations are performed and presented in this report. Based on the reference year 2025 case, we find that the United States will need about 6.60 billion barrels of ethanol (EtOH) or 8.77 billion barrels of methanol (MeOH) in order to replace the conventional gasoline (CG) that will otherwise be used. About 39.4% of the CO[subscript 2] that is projected to be emitted from coal plants will need to be captured to produce this much EtOH and about 41.1% of the CO[subscript 2] will need to be captured to produce the needed MeOH. For production of EtOH, we estimate that there will need to be between 700 and 900 GWth of nuclear power to produce the needed hydrogen and energy to create this amount of EtOH. By the same token, it will take between 1000 and 1400 GWth of nuclear power to aid in production of the needed MeOH. In the same year – 2025 – the entire world will require 16.87 billion barrels of EtOH or 22.49 billion barrels of MeOH to replace the CG that will otherwise be used. This would require capture of 29.5% of total emitted CO[subscript 2] for production of EtOH or 28.4% for production of MeOH. This amount of hydrogen and the associated energy requirements will demand between 1800 and 2300 GWth to produce the needed EtOH or between 2550 and 3500 GWth to produce the needed MeOH. These numbers show that there is a very wide market for using nuclear power to aid in the production of alternative fuels to aid in the transition to the hydrogen economy. The large fraction of emitted CO[subscript 2] that need to be captured shows that a benefit of this process would be to significantly decrease the total greenhouse gas emissions. A total cycle analysis reveals that the total reduction in CO[subscript 2] emissions in the U.S. will be slightly more than 20% for either ethanol use or methanol use. A second benefit would be to decrease a nation’s dependence on imported petroleum. In conclusion, it is found that the concept of alternative liquid fuels produced from nuclear hydrogen and captured carbon dioxide is viable. There is abundant CO2 for use and the hydrogen can be produced with proven technology. There is also evidence that nuclear power can be utilized in the production of oil from sand and shale. Revision 2 2007-04-01T00:00:00Z https://hdl.handle.net/1721.1/75134 A System Dynamics Model of the Energy Policymaking Process Oggianu, Stella Maris; Hansen, Kent F. Electric energy is a fundamental commodity for any aspects of the modern world. However, there are many uncertainties in the sources of electricity that are going to be used in the future. Some of these uncertainties are inherent to the electricity technologies and to the costs of fuels, but the biggest uncertainties come from the impact of future regulations and policies on capital costs, and operations and maintenance costs. Although system dynamics models have been extensively used for applications to the electric power, all the existing models are based on the supply/demand dynamics, and policies are considered as externalities. On the contrary, the energy policymaking model (the EPM model) presented in this report focuses on the complementary problem. This is, the determination of how byproducts and issues related to the adequate supply of electric energy modify the opinions and perceptions of the diverse sectors of the social/political environment; the analysis of the aspects of this environment that account for the formation of energy policies, and the assessment of how these policies are determinants of the technology used to supply electricity. The technologies considered are nuclear, fossil and windmills. The architecture of the EPM model is based on the assumption that policies are formed to minimize societal concerns regarding energy availability and price, nuclear waste, nuclear proliferation, nuclear safety, fossil emissions including greenhouse effect, acid rain, and land requirements for windmills. In this way, each technology is measured by its ability to reduce these concerns. The resulting policies impact on the economics of each of these options. At the same time, economics determines the selection of the new source of electricity. One of the most important results derived from the simulations done through the EPM model is that the revival of the nuclear industry may not be enough to prevent the increase in the production of greenhouse gases. The limited capacity of the industry to build plants is an important factor to consider. Another result is that the opening of Yucca Mountain at the earliest date means the removal of an important barrier for the future growth of the industry, as the risk premium of nuclear power plants may be reduced. Also derived from the use of the EPM model is that the electricity market should not be completely deregulated due to the likely be shortage of electricity supply, and high concerns regarding electricity availability, during peak demands. 2002-08-01T00:00:00Z https://hdl.handle.net/1721.1/75128 CATILaC: Computer-Aided Technique for Identifying Latent Conditions User's Manual, Version 1.2 Marchinkowski, K.; Weil, R.; Apostolakis, George E. 1. Overview 1.1 Introduction to the CATILaC Methodology By understanding the way that a facility coordinates the work it does, failure events can be placed into a broader organizational context. Once the organizational context is understood, steps can be taken to reduce the possibility of common-cause organizational failures. When this type of analysis is done in the context of a traditional Root Cause Analysis program, substantial insight into the possible causes of operational incidents can be obtained. This software package is designed to guide the user through the process of placing failure events into their organizational context. In doing so, the causes of the events and the human and hardware failures or deficiencies that lead to them will be better understood. Better corrective actions can be developed for all levels of the organization. The methodology involves both understanding what happened during the course of the event and identifying the hardware failures that contributed to its occurrence. To do this the analyst must identify the sequence of failures that occurred and the causes for each, locate the initiating, or trigger, event, and find the latent failures that became active during the event. Once the event is understood, the human contributions to each of the hardware factors must be identified and analyzed. During the analysis, deficient tasks within work processes are identified. By doing this, the latent conditions that led to the event can be discovered. Figure 1 shows how human contributions are linked to fallible decisions/organizational factors. CATILaC is focused on hardware failures and the human contributions that cause them rather than on operator actions that contribute to the event. Operations at a nuclear plant, especially post-trigger recovery actions, do not lend themselves to this type of work process analysis. Although it can be done using this software (see discussion of how to include operator contributions in Appendix I), there are other, more complete methods available to do that type of analysis. 2000-04-01T00:00:00Z https://hdl.handle.net/1721.1/75127 Analysis and Utilization of Operating Experience for Organizational Learning Weil, R.; Apostolakis, George E. The objective of this report is to clarify the ways that organizational factors influence nuclear power plant performance in order to improve performance. Therefore, this report studied the nuclear power plant organizational environment by identifying and detailing its important work processes. These work processes are: the Work Request Work Process; the Condition Reporting Work Process; the External Operating Experience Work Process; the Design Change Process; and the Procedure Change Work Process. Using this information, a methodology of incident investigation that targets organizational deficiencies contributing to events was developed. Using this methodology to analyze recent significant incidents, a list of important organizational factors and the context within which they influence the successful completion of tasks was identified. These factors are: 1) Communication - Pervasive – Most important between different units and departments; 2) Formalization -Execution; 3) Goal Prioritization - Prioritization; 4) Problem Identification - Planning, scheduling, and return to normal line-up; 5) Roles and Responsibilities - Execution; and 6) Technical Knowledge (job specific knowledge and broad based knowledge) - Job specific knowledge – execution/ Broad based knowledge –prioritization, planning, scheduling, and other tasks. Although safety culture and organizational learning are not listed, they are important. The reason for their exclusion is that they are not single organizational factors useful when cited in incident investigations. Rather, safety culture is a term used to describe all organizational factors, including organizational structure, that impact performance. Similarly, organizational learning was excluded because it is a collection of programs, processes, individual attitudes and culture responsible for learning. Although organizational learning was not listed, it was studied resulting in the development of the Utilization of Operating Experience Work Process. The Utilization of Operating Experience Work Process consists of the following seven steps: 1) Identification; 2) Screening/Prioritization/Dissemination; 3) Investigation/Evaluation; 4) Development; ii i 5) Implementation; 6) Closeout; and 7) Verification/Validation. Since prioritization was identified as important in the above work process and the analysis of significant events, a methodology for the prioritization of work activities at nuclear power plants was developed. This methodology produces a prioritization tool that assigns a numerical performance index to each item requiring prioritization. Applying the methodology at Seabrook Station produced a tool that allowed those who prioritize external operating experience to more efficiently and accurately do so. In addition to the success of the application at Seabrook, a workshop was held at MIT with experts in prioritizing external operating experience. These experts further validated the methodology and the resulting tool. The final piece of work in this report is an analysis of the NRC's revised oversight process as it relates to safety culture. The performance-based regulatory approach is appropriate for regulating safety culture. However, the NRC should continue the analysis of operating experience to identify additional organizational factors and the context within which they influence performance. Furthermore, they should develop performance indicators and measurement instruments for each organizational factor so that plants would be better able to take responsibility to proactively manage their safety culture. 2001-06-01T00:00:00Z https://hdl.handle.net/1721.1/75126 Nuclear Energy for Variable Electricity and Liquid Fuels Production: Integrating Nuclear with Renewables, Fossil Fuels, and Biomass for a Low- Carbon World Forsberg, Charles W. The world faces two energy challenges: (1) the national security and economic challenge of dependence on foreign oil and (2) the need to reduce carbon dioxide emissions from the burning of fossil fuels to avoid climate change. Nuclear energy as a low-carbon domestic source of energy can address both challenges. However, nuclear energy in the United States is only used for base-load electricity production—about a quarter of the total energy demand. To address the two energy challenges, we have initiated a series of studies to understand long-term nuclearrenewable energy futures for a low-carbon world that can meet all energy demands. This includes liquid fossil fuel options with low greenhouse gas releases. This is a first effort to synthesize what has been learned about hybrid energy systems. The electricity challenge is to provide variable electricity production to match demand. Today this is primarily accomplished with variable-load fossil plants burning stored coal, oil, and natural gas. It is an economic option because of the low cost of storing fossil fuels and the relatively low cost of fossil power plants. The output of nuclear and renewable electricity sources do not match electricity demand. In a low-carbon world it would be required to store electricity when excess electricity is available to meet demand at times of low electricity production. If there are restrictions on carbon dioxide emissions, economics favors nuclear for most electricity production unless renewable electricity production costs are significantly lower than nuclear electricity production costs. This is because the amount of electricity that has to be stored to match electricity production with demand is much smaller in an all-nuclear system than any renewable system1. About two-thirds of all electricity demand is base-load electricity where the steady-state electricity output of a nuclear plant matches customer demand. While there are many electricity storage technologies to help match electricity production with demand over a period of a day (smart grid, pumped hydroelectric storage, batteries, etc.), only two seasonal energy storage technologies were identified2: nuclear geothermal heat storage and hydrogen. A nuclear renewables electricity system that also produces hydrogen for industrial markets may enable an economic system for variable electricity production where a larger fraction of the electricity can be produced by wind and solar energy sources. Nuclear energy can reduce greenhouse emissions from gasoline, diesel and jet fuel by replacing fossil fuels used in the production and refining processes. In the context of increasing U.S. oil production, a primary need is for heat to recover heavy oil and shale oil. U.S. shale oil resources exceed total oil produced worldwide to date and thus their use could eliminate U.S. dependence on foreign oil. The recovery and conversion of shale oil into liquid fuels using heat from nuclear reactors may have the lowest carbon dioxide releases per liter of fuel of all the fossil fuel alternatives to conventional crude oil production. Unlike almost all other industrial processes, shale oil and heavy oil production do not require steady-state heat input. That characteristic would allow nuclear plants coupled to shale oil and heavy oil production to operate at base-load with variable heat and electricity production. The variable electricity production could help match electricity production to demand and enable the larger-scale use of renewables. Heavy oil and shale oil production are the only potential industries large enough where variable heat demand is the alternative to energy storage to match electricity production with demand. Very little research has been done on these options. There is the potential for nuclear biofuels to supply a major fraction of the liquid fuels demand. This option results in no net addition of greenhouse gases to the atmosphere. Liquid fuels from biomass are limited by the availability of biomass. Synergisms between nuclear and biofuels can enable up to three times as much liquid fuel to be produced per ton of biomass. This is achieved by using nuclear to provide heat and hydrogen to operate the biorefinery and thus avoid the use of biomass as a fuel for the biorefinery. Liquid fuels can also be made from air and water with heat and electricity from nuclear power plants. This option can provide unlimited liquid fuel and places an upper cap on the cost of liquid fuels—2 to 3 times that of the cost of electricity on a unit heat basis. Key enabling technologies for a low-carbon nuclear-renewable energy system include nuclear-geothermal gigawatt-year energy storage, high-temperature electrolysis for hydrogen production, use of nuclear heat for reservoir heating of heavy oils and shale oil, conversion of lignin (the non-cellulosic component of plants) to liquid fuels, and densification of biomass for economic transport of biomass to large biorefineries. Most applications can be met with light water reactors; but some applications require the commercialization of high-temperature reactors. A nuclear renewables energy future is possible and potentially economic. Nuclear and renewable energy sources have different characteristics and in some systems are synergistic. Allnuclear or all-renewables energy futures are more expensive and difficult to achieve. Wind and solar economics are strongly dependent on location—particularly latitude because (1) it drives variable seasonal energy demands and (2) wind and solar inputs are functions of latitude. The analysis herein is for the United States but would be generally applicable for countries at similar or higher latitudes.3 Little work has been done to develop credible low-carbon energy futures for a prosperous world of 10-billion people. The uncertainties are very large. 2011-09-01T00:00:00Z https://hdl.handle.net/1721.1/75125 Conceptual Design of Nuclear-Geothermal Energy Storage Systems for Variable Electricity Production Lee, Youho; Forsberg, Charles W. Nuclear plants have high capital costs and low operating costs that favor base-load operation. This characteristic of nuclear power has been a critical constraint that limits the portion of nuclear power plants in a grid to stay below the base-load demand. A novel gigawatt-year thermal-energy storage technology is proposed to enable base load nuclear plants to produce variable electricity to meet seasonal variations in electricity demand. A large volume of underground rock is heated with hot water (or steam or carbon dioxide) from a nuclear power plant during periods of low electricity demand, and the heat is extracted during times of high demand and converted to electricity using a standard geothermal plant (Figure 1). Among various technical options, technically mature ones were selected for the reference design; a Pressurized Water Reactor (PWR) injects hot fluid into an underground reservoir through an intermediate heat exchanger and bypass flow lines on either the primary or secondary side. The reservoir size of 500 m in each dimension at 1.5 km underneath the surface is engineered to have permeability of 2 Darcy using commercial hydraulic fracture methods, and is cyclically heated up and cooled down between the temperatures of 50°C and 250°C. Peak power electricity is produced by exploiting the stored thermal energy via an Enhanced Geothermal System (EGS) that employs a binary flash cycle. Models of a nuclear-EGS system performance, taking into account heat transfer in the reservoir, thermal front velocity in the reservoir, conductive heat & water losses, geothermal power plant electricity production performance, operating conditions and system interfaces were developed and independently compared with Computational Fluid Dynamics (CFD) simulations using FLUENT 6.3 to confirm the validity of the models. The design study with the validated models reveals that the reference nuclear-EGS system based on 2.8~6.0 GW(th) of nuclear power would have a thermal storage size of 0.7~1.5 GW(th)-year, which corresponds to 0.08~0.2 GW(e)-year with electricity round trip efficiency of 0.34~0.46. Reservoir permeability and geofluid temperature are found to be the most important design parameters that affect performance of nuclear-EGS storage systems. A grid that deploys a nuclear-EGS system will have three distinct electricity sectors: nuclear base load, EGS intermediate load, and gas turbine peak power. The nuclear-EGS storage system introduces economic benefits to a grid by leveraging economic gains arising from replacing expensive intermediate and peak electricity with cheap base-load electricity. A nuclear-EGS system has a higher capital cost than natural gas turbines; consequently, it replaces intermediate-load power plants but not all the gas turbines that operate for a small number of hours per year. It was found that the deployment of a operate for a small number of hours per year. It was found that the deployment of a Nuclear-EGS could cut the electricity production cost of the New England Independent Systems Operator (NE-ISO) by as much as 14% of the storage-free cost (Fig. 2). Economic competitiveness of nuclear power plants is the most decisive factor for the deployment of the system in a grid. Because this was the first analysis of a nuclear EGS system, we used off-the-shelf technology wherever possible to reduce uncertainties and have confidence that the system will work. Significant improvements in roundtrip efficiency and economics may be possible by development of more advanced systems. For example, existing geothermal power plants are small (megawatts) versus several hundred megawatts for a nuclear EGS system. They use double flash power systems. The larger scale may enable the use of triple-flash and other more efficient power cycles. Reservoir development methods designed explicitly for nuclear EGS systems may significantly lower the costs of reservoir development. Like any other system dependent upon geology, costs and performance will depend upon the local geology. 2011-06-01T00:00:00Z https://hdl.handle.net/1721.1/75124 Nuclear Tanker Producing Liquid Fuels From Air and Water: Applicable Technology for Land-Based Future Production of Commercial Liquid Fuels Galle-Bishop, John Michael; Driscoll, Michael J.; Forsberg, Charles W. Emerging technologies in CO[subscript 2] air capture, high temperature electrolysis, microchannel catalytic conversion, and Generation IV reactor plant systems have the potential to create a shipboard liquid fuel production system that will ease the burdened cost of supplying fuel to deployed naval ships and aircraft. Based upon historical data provided by the US Navy (USN), the tanker ship must supply 6,400 BBL/Day of fuel (JP-5) to accommodate the highest anticipated demand of a carrier strike group (CSG). Previous investigation suggested implementing shipboard a liquid fuel production system using commercially mature processes such as alkaline electrolysis, pressurized water reactors (PWRs), and methanol synthesis; however, more detailed analysis shows that such an approach is not practical. Although Fischer-Tropsch (FT) synthetic fuel production technology has traditionally been designed to accommodate large economies of scale, recent advances in modular, microchannel reactor (MCR) technology have to potential to facilitate a shipboard solution. Recent advances in high temperature coelectrolysis (HTCE) and high temperature steam electrolysis (HTSE) from solid oxide electrolytic cells (SOECs) have been even more promising. In addition to dramatically reducing the required equipment footprint, HTCE/HTSE produces the desired synthesis gas (syngas) feed at 75% of the power level required by conventional alkaline electrolysis (590 MW[subscript e] vs. 789 MW[subscript e]). After performing an assessment of various CO[subscript 2] feedstock sources, atmospheric CO[subscript 2] extraction using an air capture system appears the most promising option. However, it was determined that the current air capture system design requires improvement. In order to be feasible for shipboard use, it must be able to capture CO[subscript 2] in a system only ¼ of the present size; and the current design must be modified to permit more effective operation in a humid, offshore environment. Although a PWR power plant is not the recommended option, it is feasible. Operating with a Rankine cycle, a PWR could power the recommended liquid fuel production plant with a 2,082 MW[superscript th] reactor and 33% cycle efficiency. The recommended option uses a molten salt-cooled advanced high temperature reactor (AHTR) coupled to a supercritical carbon dioxide (S-CO[subscript 2]) recompression cycle operating at 25.0 MPa and 670°C. This more advanced 1,456 MWth option has a 45% cycle efficiency, a 42% improvement over the PWR option. In terms of reactor power heat input to JP-5 combustion heat output, the AHTR is clearly superior to the PWR (31% vs. 22%). In order to be a viable concept, additional research and development is necessary to develop more compact CO[subscript 2] capture systems, resolve SOEC degradation issues, and determine a suitable material for the molten salt/S-CO[subscript 2] heat exchanger interface. 2011-06-01T00:00:00Z https://hdl.handle.net/1721.1/75123 Nuclear-Renewables Energy System for Hydrogen and Electricity Production Haratyk, Geoffrey; Forsberg, Charles W.; Driscoll, Michael J. Climate change concerns and expensive oil call for a different mix of energy technologies. Nuclear and renewables attract attention because of their ability to produce electricity while cutting carbon emissions. However their output does not match demand. This thesis introduces a nuclear-renewables energy system, that would produce electricity and hydrogen on a large scale while meeting the load demand. The system involves efficient high temperature electrolysis (HTE) for hydrogen production, with heat provided by nuclear and electricity by the grid (nuclear and/or renewables). Hydrogen production would be variable, typically at time of low demand for electricity and large power generation from renewables. Hydrogen would be stored underground on site for later shipping to industrial hydrogen users by long-distance pipeline or for peak power production in fuel cells. A hydrogen plant was designed, and the economics of the system were evaluated by simulating the introduction of the system in the Dakotas region of the United States in both a regulated and a deregulated electricity market. The analysis shows that the system is economically competitive for a high price of natural gas ($12-13 MMBtu) and a capital cost reduction (33%) of wind turbines. The hydrogen production is sufficient to supply the current demand of the Great Lakes refineries. With today's electricity prices, a competitive production cost of $1.5 /kg hydrogen is achievable. The analysis indicates large economic incentives to develop HTE systems that operate efficiently in reverse as fuel cells to displace the gas turbines that operate only a few hundred hours per year and thus have high capital cost charges. The capital cost of the HTE system has a significant impact on system economics, with large incentives to develop reversible HTE/ FC systems to reduce those costs. Such a system would expand the use of nuclear beyond electricity generation, and allows a larger penetration of renewables by providing an energy storage media and bringing flexibility to the grid operators. 2011-06-01T00:00:00Z https://hdl.handle.net/1721.1/75122 Nuclear Energy for Simultaneous Low-Carbon Heavy-Oil Recovery and Gigawatt-Year Heat Storage for Peak Electricity Production Forsberg, Charles W.; Krentz-Wee, Rebecca E.; Lee, You Ho; Oloyede, Isaiah O. In a carbon-constrained world or a world of high natural gas prices, the use of fossil-fueled power plants to satisfy variable electricity demands may be limited. Nuclear power plants operating at full capacity with large-scale energy storage systems could be employed to provide variable intermediate and peak electricity production. One storage option is to use a nuclear-geothermal system for peak electricity production. At times of low electricity demand, heat from a nuclear reactor in the form of pressurized hot water is used to heat underground rock. At times of high electricity demand, the reactor produces electricity. In parallel, cold pressurized water is injected into the bottom of the manmade hot-rock heat source, hot pressurized water is recovered, and the hot pressurized water is used with a geothermal power plant to produce peak electricity. A nuclear geothermal system for peak electricity production is a new concept with many possible configurations. This paper is an initial assessment of converting heavy oil reservoirs with a history of oil production into nuclear-geothermal systems for peak electricity production. Heavy oil is recovered by steam injection into a reservoir and raising the temperature so the heavy oil can flow to production wells. Such a reservoir may be economically attractive for conversion into a nuclear-geothermal peak electricity system because (1) the reservoir has been preheated to high temperatures that would minimize long-term heat losses from a nuclear geothermal system, (2) such geologies are likely to have reasonable permeability to water flow—a requirement for a nuclear-geothermal system, (3) much of the infrastructure is in place, and (4) the local geology is well understood—including effects of adding heat to the rock. The use of a heavy oil field as a nuclear-geothermal peak power system may significantly increase the fraction of heavy oil that is recovered and enable heavy oil recovery from deeper heavy-oil reservoirs. Total recoverable heavy oil resources may be significantly increased. The nuclear-geothermal heat storage facility acts like a washing machine on the heavy oil reservoir over time with oil extracted using the hot pressurized water. The reservoir characteristics (high porosity, etc.) for heat storage would be expected to improve as more oil is removed. The assessment is that this option is potentially attractive but there are significant uncertainties. The next step must include detailed studies of specific sites to develop a realistic understanding of the option. 2010-12-01T00:00:00Z https://hdl.handle.net/1721.1/75121 Hydrogen Production for Steam Electrolysis Using a Supercritical CO[subscript 2]- Cooled Fast Reactor Memmott, M. J.; Driscoll, Michael J.; Hejzlar, Pavel; Kazimi, Mujid S. Rising natural gas prices and growing concern over CO[subscript 2] emissions have intensified interest in alternative methods for producing hydrogen. Nuclear energy can be used to produce hydrogen through thermochemical and/or electrochemical processes. This report investigates the feasibility of high temperature steam electrolysis (HTSE) coupled with an advanced gas-cooled fast reactor (GFR) utilizing supercritical carbon dioxide (S-CO[subscript 2]) as the coolant. The reasons for selecting this particular reactor include fast reactor uranium resource utilization benefits, lower reactor outlet temperatures than helium-cooled reactors which ameliorate materials problems, and reduced power conversion system costs. High temperature steam electrolysis can be performed at conditions of 850°C and atmospheric pressure. However, compression of the hydrogen for pumping through pipes is unnecessary if electrolysis takes place at around 6 MPa. The reactor coolant at 650°C is used to heat the steam up to temperatures ranging between 250°C and 350°C, and the remaining heat is provided by thermal recuperation from product hydrogen and oxygen. Several different methods for integrating the hydrogen production HTSE plant with the GFR were investigated. The two most promising methods are discussed in more detail: extracting coolant from the power conversion system (PCS) turbine exhaust to boil water, and extracting coolant directly from the reactor using separate water boiler (WB) loops. Both methods have comparable thermal to electricity efficiencies (~43%) at 650°C. This relates to an overall hydrogen production efficiency of about 47%. The approach which utilizes separate WB loops has the added advantage of being able to provide emergency cooling to the reactor, and also the benefit of not interfering with the operation of the PCS. This makes the separate WB loop integration method a more desirable scheme for hydrogen production using HTSE. The HTSE electrolysis unit adopted for the present analysis was designed by Ceramatec in coordination with INL. In this unit the steam flows into an electrolytic cell. It is separated by electron flow from a nickel-zirconium cathode to a strontium-doped lanthanum manganite anode. The optimal conditions for stack operation have been found by INL using various modeling and experimental techniques. These conditions include a 10% by volume flow of hydrogen in the feed, a stack operating temperature of 800°C, and an operating voltage of 1.2 V. The GFR integrated with the HTSE plant via separate water boiler loops was modeled in this work using the chemical engineering code ASPEN. The results of this model were benchmarked against the Idaho National Lab (INL) process, modeled using HYSIS. Both models predict a hydrogen production rate of ~10.2 kg/sec (± 0.2 kg/sec) for a 600 MWth reactor with an overall efficiency ranging between 47%-50%. The highly recuperated HTSE plant developed for the GFR can in principle be used in conjunction with a variety of other nuclear reactors, without requiring high reactor coolant outlet temperatures. 2007-02-01T00:00:00Z https://hdl.handle.net/1721.1/75119 OPTIMIZATION OF THE HYBRID SULFUR CYCLE FOR HYDROGEN GENERATION Jeong, Y. H.; Kazimi, Mujid S.; Hohnholt, K. J.; Yildiz, Bilge The hybrid sulfur cycle (modified from the Westinghouse Cycle) for decomposing water into oxygen and hydrogen is evaluated. Hydrogen is produced by electrolysis of sulfur dioxide and water mixture at low temperature, which also results in the formation of oxygen and sulfuric acid. The sulfuric acid is decomposed into steam and sulfur trioxide, which at high temperature (1100 K) is further decomposed into sulfur dioxide and oxygen. The presence of sulfur dioxide along with water in the electrolyzer reduces the required electrode potential well below that required for electrolysis of pure–water, thus reducing the total energy consumed by the electrolyzer. Further, using only sulfuric acid for the thermochemical processes minimizes the required chemical stock in the hydrogen plant well below that required for the sulfur–iodine pure thermochemical cycle (SI cycle). In this study, ways to optimize the energy efficiency of the hybrid cycle are explored by varying the electrolyzer acid concentration, decomposer acid concentration, pressure and temperature of the decomposer and internal heat recuperation, based on currently available experimental data for the electrode potential. An optimal cycle efficiency of 43.9% (LHV) appears to be achievable (5 bar, 1100 K and 60 mol–% of H[subscript 2]SO[subscript 4] at the decomposer, 70 w–% of H[subscript 2]SO[subscript 4] at the electrolyzer). However, the ideal cycle efficiency is over 70% (LHV), which leaves room to improve the achievable efficiency with further development. For a maximum temperature of 1200 K, 47% (LHV) appears to be the maximum achievable cycle efficiency (10 bar, 1200 K and 60 mol–% of H[subscript 2]SO[subscript 4] for decomposer, 70 w–% of H[subscript 2]SO[subscript 4] for electrolyzer). The ideal cycle efficiency is over 80% (LHV). Operation under elevated pressures (70 bar or higher) results in minimized equipment size and capital cost, but there is loss in thermal efficiency. However, the loss in efficiency as pressure increases is not large at high temperature (1200 K) compared to that at low temperatures (1000–1100 K). Therefore, high pressure operation would be favored only if we can achieve high temperature. The major factors that can affect the cycle efficiency are reducing the electrode over–potential and having structural materials that can accommodate operation at high temperature and high acid concentration. 2005-05-01T00:00:00Z https://hdl.handle.net/1721.1/75098 ATTRIBUTES OF A NUCLEAR-ASSISTED GAS TURBINE POWER CYCLE Jeong, Y. H.; Saha, P.; Kazimi, Mujid S. By using a combination of a nuclear reactor, which emits no carbon dioxide, and a high efficiency natural gas turbine combined cycle (NGCC), electric utilities can reduce their generation cost as well as minimize the greenhouse gas emissions. In this work, the economic competitiveness of pure NGCC, nuclear assisted NGCC and pure nuclear power plants are studied. An advanced gas cooled nuclear reactor can be added to the conventional NGCC as a heat source for the air exiting the compressor. For this study we assumed a high temperature pebble bed modular reactor (PBMR) with reactor outlet gas temperature of 900ºC. With that temperature, the thermal contribution (fossil fuel savings and CO[subscript 2] reduction) of nuclear energy in the nuclear-assisted NGCC cycle was 46.3%. For assessing the economic competitiveness of the three options, the levelized electricity generation costs were calculated. The economics depend on the cost of natural gas and the capital cost of the nuclear reactor. Obviously, the best option for low natural gas cost is the pure NGCC, whereas the pure nuclear power is the best choice for high natural gas prices. The crossing points vary depending on the level of expected carbon tax. The pure nuclear option is not affected by the level of carbon tax. The nuclear-assisted NGCC cost is in the middle. There are several synergetic effects to using nuclear and fossil powers together. First, since the generation cost of the nuclear-assisted NGCC cycle is not as sensitive to the gas price as the NGCC, the economic risk of an NGCC plant can be minimized by adopting a nuclear-assisted NGCC cycle. Second, by introducing NGCC to nuclear power, the risk from high nuclear capital cost can be mitigated. For example, 3000 $/kW[subscript e] of nuclear capital cost can be reduced to about 1500 $/kW[subscript e]. Third, in addition to minimizing the risk from gas price fluctuation and high capital cost, even though the window is very narrow, the nuclear assisted NGCC can be more advantageous over the other two options in case of high nuclear capital costs and high gas prices. Finally, green house gas emissions can be reduced significantly using nuclear assisted NGCC. 2005-02-01T00:00:00Z https://hdl.handle.net/1721.1/75090 Conceptual Design of Molten Salt Loop Experiment for MIT Research Reactor Bean, Malcolm K.; Dewitt, Gregory Lee; Cabeche, Dion T.; Gerrity, Thomas P.; Haratyk, Geoffrey; Kersting, Alyssa R.; Lee, Youho; Virgen, Matthew M.; Lenci, Giancarlo; Lin, Christie; Metzler, Florian; Ochoukov, Roman Igorevitch; Reed, Mark; Sobes, Vladimir; Sugrue, Rosemary M.; Shwageraus, Eugene; Wisniowska, Agata Elzbieta; Youchak, Paul M. Molten salt is a promising coolant candidate for Advanced High Temperature Reactor (AHTR) Gen-IV designs. The low neutron absorption, high thermal capacity, chemical inertness, and high boiling point at low pressure of molten salt coolants could potentially lead to enhanced safety and lower cost of AHTR designs as compared with conventional Light Water Reactors. Improved economics are expected to be a result of the higher possible operating temperatures (700oC), improving thermal efficiency, availability of process heat for industrial applications, and reduced containment costs. Improved safety margins arise from the use of highly robust TRISO particles fuel in either pebble or graphite block form, greater thermal inertia, low pressure and high boiling point of molten salts relative to water cooled reactor designs. Currently, one of the main challenges associated with further advancement of AHTR design is predicting reactor core materials’ interactions with molten salt coolant over long time scales in a radiation environment. In the Fall of 2010, the Nuclear Engineering Design Project Course (22.033/33) undertook the challenge to design a molten salt test loop to be installed in the MIT Research Reactor (MITR) that would recreate anticipated AHTR operating conditions and fill the knowledge gap in understanding of materials behavior in such environment. In addition to simulating neutronic, thermal and chemical conditions similar to those of AHTR, the test loop must also meet the safety and operating requirements of the MITR. During the course, a preliminary design was developed that features an annular in core molten salt flow channel to maximize the volume available for testing materials’ samples and maintaining the salt temperature at 700oC and flow velocity at 6 m/s, while avoiding boiling at the outside surface of the loop, as prescribed by MITR safety requirements. A number of additional requirements were addressed by the students including reactivity insertion, power peaking, tritium production, shielding, and others. The design tasks were subdivided into four key areas of neutronics, thermal hydraulics, chemistry and materials, and instrumentation and control. The molten salt chosen was LiF-BeF[subscript 2] (FLiBe) with lithium enriched in [superscript 7]Li isotope up to 99.995% because this salt is the leading coolant candidate for AHTR. Hastelloy-N was chosen as the material in contact with the molten salt due to its high resistance to corrosion, good material properties at high temperature and extensive use in previous experiments. The presence of corrosion products, free fluorine and production of tritium in the molten salt were found to be important phenomena challenging the loop design. Therefore, various methods for the salt chemistry control and tritium release were evaluated and resulted in a design of multi-component system for monitoring the salt conditions, maintaining redox potential and removing the impurities and tritium from the salt. Another challenge was managing the loop operation given the relatively high freezing point of the salt at about 460oC. Procedures were developed for start-up, steady-state, shutdown and transient operation of the loop. The thermal hydraulic analyses indicate that 1.8 kW of strap heating along the loop outside the core section and a 1.5 to 2 kW pump were required, depending on final design choices. In addition, preliminary cost estimates of constructing the loop experiment at MITR were performed. The main constraints on the choice of the loop’s individual components and diagnostics were: 1) the ability to function at the designed operating temperature, pressure, and flow rates; 2) the ability to function in a nuclear radiation environment; and 3) the necessity to meet MITR safety requirements. A database of vendors for the loop’s components, instrumentation, and diagnostics was compiled. To support further work on the molten salt test loop an electronic library of references was compiled as well. Finally, a number of potential accident scenarios were examined and their effects on the safety and operation of MITR were evaluated and found to represent no danger to the public or interfere with normal operations. Minor leakages of either the reactor water or the molten salt coolant inside the loop were found to be self-sealing with little to no effects on the safety and operation of MITR. A complete failure of the loop’s heating and pumping systems was found to lead to FLiBe’s cooling and freezing inside the loop, with the freezing time ranging from several minutes to ~1 hour depending on the choice of the loop thermal insulation material. 2011-08-01T00:00:00Z https://hdl.handle.net/1721.1/75089 Design and Optimization of a High Thermal Flux Research Reactor Via Kriging-Based Algorithm Kempf, Stephanie A.; Hu, Lin-Wen; Forget, Benoit In response to increasing demands for the services of research reactors, a 5 MW LEUfueled research reactor core is developed and optimized to provide high thermal flux within specified limits upon thermal hydraulic performance, cycle length, irradiation utilization, and manufacturability. A novel fuel assembly concept which makes use of integral flux traps is postulated to meet these requirements. Each assembly can be rotated into one of three different configurations to produce flux traps of different size, shape, and neutron energy spectrum within the core. A method for predicting and guiding the search for the optimum geometry was sought. Kriging has been chosen to predict the values of eigenvalue and thermal flux at untested geometric parameters. Because kriging treats all measurements as the sum of a global deterministic function and a stochastic departure from that function, predictions come with a measurement of uncertainty. As a result, the analyst can search the design space for likely improvement, or probe areas of high uncertainty for improvements that might have been missed using other methods. The technique is used in an algorithm for constrained optimization of the design, and a set of best practices for use of this are described. The optimized design produces a peak thermal flux of 1.56 x 10[superscript 14] n/cm[superscript 2]s. Safety is demonstrated by presentation of reactivity feedback coefficients and the results of loss of flow and reactivity insertion transient analysis. A single fission target can be used to produce 96 6-day Ci of [superscript 99]Mo per week. When the reactor is oriented to take advantage of high fast flux, steels can be subjected to damage rates of 5.76 dpa per year. Silicon carbide can be damaged at a rate of 2.79 dpa/y. The concept is a safe, versatile, proliferation-resistant means of supplying current and future irradiation needs. 2011-06-01T00:00:00Z https://hdl.handle.net/1721.1/75088 Estimate of Radiation Release from MIT Reactor with Low Enriched Uranium (LEU) Core During Maximum Hypothetical Accident Plumer, Kevin E.; Newton, Thomas H., Jr.; Forget, Benoit In accordance with a 1986 NRC ruling, the MIT Research Reactor (MITR) is planning on converting from the use of highly enriched uranium (HEU) to low enriched uranium (LEU) for fuel. A component of the conversion analysis includes calculating the maximum hypothetical accident (MHA) dose implications for the two types of fuel. In this work, the dose levels at the site exclusion area boundary were calculated for the MITR MHA using both the HEU and LEU models of the MITR core. The core inventories from the reactor were calculated using the ORIGEN-S pointdepletion code linked to the MITR spectrum. The MITR spectrum was used from an MCODE simulation of the equilibrium LEU and HEU versions of the core. Release fractions from the melted fuel to containment were established using melt test data from plate-type fuel as well as modified release fractions from NRC Regulatory Guides. The dose paths considered were the same paths used in the previous work, consisting of atmospheric release through potential containment leakage as well as direct and scattered gamma dose from the containment source term. The Total Effective Dose Equivalent (TEDE) values were calculated in addition to the whole body and thyroid doses. For dose comparison the LEU thermal power was 17% higher than the HEU thermal power in order to provide equivalent total flux levels to the experimental ports. The results showed that the LEU core operating at 7 MW will yield TEDE levels 22% higher than the HEU core operating at 6 MW for equivalent release fractions. The two-hour dose at the exclusion area boundary from the LEU core operating at 7 MW using the plate-type fuel release fractions was 0.440 rem at 21 m and 0.344 rem at 8 m, while the dose from the HEU core operating at 6 MW was 0.361 rem at 21 m and 0.281 rem at 8 m. These doses are within the public dose NRC regulatory limit of 0.500 rem TEDE. 2011-06-01T00:00:00Z https://hdl.handle.net/1721.1/75087 Developing Fuel Management Capabilities Based On Coupled Monte Carlo Depletion in Support of the MIT Research Reactor Conversion Romano, Paul Kollath; Newton, Thomas H., Jr.; Forget, Benoit Pursuant to a 1986 NRC ruling, the MIT Reactor (MITR) is planning on converting from the use of highly enriched uranium (HEU) to low enriched uranium (LEU) for fuel. Prior studies have shown that the MITR will be able to operate using monolithic U-Mo LEU fuel while achieving neutron fluxes close to that of an HEU core. However, to date, detailed studies on fuel management and burnup while using LEU fuel have not been performed. In this work, a code package is developed for performing detailed fuel management studies at the MITR that is easy to use and is based on state-of-the-art computational methodologies. A wrapper was written that enables fuel management operations to be modeled using MCODE, a code developed at MIT that couples MCNP to the point-depletion code ORIGEN. To explicitly model the movement of the control blades in the MITR as the core is being depleted, a criticality search algorithm was implemented to determine the critical position of the control blades at each depletion timestep. Additionally, a graphical user interface (GUI) was developed to automate the creation of model input files. The fuel management wrapper and GUI were developed in Python, with the PyQt4 extension being used for GUI-specific features. The MCODE fuel management wrapper has been shown to perform reliably based on a number of studies. An LEU equilibrium core was modeled and burned for 640 days with the fuel being moved in the same pattern every 80 days. The control blade movement and nuclide concentrations were shown to be in agreement with what one would intuitively predict. The fuel management capabilities of REBUS-PC and the MCODE fuel management wrapper were compared by modeling the same refueling scheme using an HEU core. The element power peaking factors for the two models showed remarkable agreement. Together, the fuel management wrapper and graphical user interface will help the staff at the MITR perform in-core fuel management calculations quickly and with a higher level of detail than that previously possible. 2009-06-01T00:00:00Z https://hdl.handle.net/1721.1/75086 Design of a Low Enrichment, Enhanced Fast Flux Core for the MIT Research Reactor Ellis, T.S.; Forget, Benoit; Kazimi, Mujid S.; Newton, T.; Pilat, Edward E. Worldwide, there is limited test reactor capacity to perform the required irradiation experiments on advanced fast reactor materials and fuel designs. This is particularly true in the U.S., which no longer has an operating fast reactor but depends upon two aging thermal reactors for testing the behavior of various materials in an irradiation environment. The MIT Research Reactor is planning for a new core to end the need for highly-enriched uranium and operate the reactor with uranium enrichments under 20%. This study explores the use of the central region in the newly proposed MIT reactor core to boost the production of fast neutrons, thus making the new core more beneficial for materials testing. The Fast Flux Trap introduces a region of fissile material surrounding a central irradiation facility which is cooled by liquid lead-bismuth eutectic. This arrangement maximizes the fast neutron production by avoiding neutron moderation in the center. The fissile material, arranged in a tight hexagonal pin array, can be uranium enriched in either [superscript 235]U or [superscript 233]U, to the limit allowable for non-proliferation. Insertion of the Fast Flux Trap in the proposed low enriched uranium core operated at a 10 MW power level, can provide a 252-271% higher fast neutron flux than the previously proposed designs with low enriched fuel for the MIT research reactor and a 235%-253% higher fast neutron flux than the existing highly enriched uranium MITR-II core at 5 MW. This new core fast flux capability is within a factor of 2 to 4 of the much larger national test reactors, the Advanced Test Reactor and the High Flux Isotope Reactor, and hence can allow the MIT research reactor to be more useful for fast irradiation. The work covered both steady state and transient events involving the Fast Flux Trap, using the Monte Carlo N-Particle (MCNP) transport code. It was shown that the power distribution within the Fast Flux Trap pins as well as the plates in the rest of the core will be satisfactory; in other words, no excessive power peaking will develop. The limits of the Fast Flux Trap lifetime were found to exceed the expected licensing time of the new core. Furthermore, the reactivity implications of metallic coolant leaks, water flooding of the Fast Flux Trap and various possible test materials were all found to be acceptable. The loss of flow following a pump trip event was analyzed using the RELAP5-3D code, and found not to result in excessive temperatures with regard to materials strength and corrosion resistance. While the specific design developed in this dissertation is particular to the MIT research reactor core, the Fast Flux Trap design concept can potentially be applied in other reactor cores so that other thermal spectrum research and test reactor facilities can benefit from this enhanced capability. 2009-02-01T00:00:00Z https://hdl.handle.net/1721.1/75085 Evaluation of the Thermal-Hydraulic Operating Limits of HEU-LEU Transition Cores for the MIT Research Reactor Wan, Yunzhi; Hu, Lin-Wen The MIT Research Reactor (MITR) is in the process of conducting a design study to convert from High Enrichment Uranium (HEU) fuel to Low Enrichment Uranium (LEU) fuel. The currently selected LEU fuel design contains 18 plates per element, compared to the existing HEU design of 15 plates per element. A transitional conversion strategy, which consists of replacing three HEU elements with fresh LEU fuel elements in each fuel cycle, is proposed. The objective of this thesis is to analyze the thermo-hydraulic safety margins and to determine the operating power limits of the MITR for each mixed core configuration. The analysis was performed using PLTEMP/ANL ver 3.5, a program developed for thermo-hydraulic calculations of research reactors. Two correlations were used to model the friction pressure drop and enhanced heat transfer of the finned fuel plates: the Carnavos correlation for friction factor and heat transfer, and the Wong Correlation for friction factor with a constant heat transfer enhancement factor of 1.9. With these correlations, the minimum onset of nucleate boiling (ONB) margins of the hottest fuel plates were evaluated in nine different core configurations, the HEU core, the LEU core and seven mixed cores that consist of both HEU and LEU elements. The maximum radial power peaking factors were assumed at 2.0 for HEU and 1.76 for LEU in all the analyzed core configurations. The calculated results indicate that the HEU fuel elements yielded lower ONB margins than LEU fuel elements in all mixed core configurations. In addition to full coolant channels, side channels next to the support plates that form side coolant channels were analyzed and found to be more limiting due to higher flow resistance. The maximum operating powers during the HEU to LEU transition were determined by maintaining the minimum ONB margin corresponding to the homogeneous HEU core at 6 MW. The recommended steady-state power is 5.8 MW for all transitional cores if the maximum radial peaking is adjacent to a full coolant channel and 4.9 MW if the maximum radial peaking is adjacent to a side coolant channel. 2009-05-01T00:00:00Z https://hdl.handle.net/1721.1/75084 Pressure Drop Measurements and Flow Distribution Analysis for MIT Research Reactor with Low-Enriched Fuel Yuen-Ting Wong, S.; Hu, Lin-Wen; Kazimi, Mujid S. The MIT Nuclear Research Reactor (MITR) is the only research reactor in the United States that utilizes plate-type fuel elements with longitudinal fins to augment heat transfer. Recent studies on the conversion to low-enriched uranium (LEU) fuel at the MITR, together with the supporting thermal hydraulic analyses, propose different fuel element designs for optimization of thermal hydraulic performance of the LEU core. Since proposed fuel design has a smaller coolant channel height than the existing HEU fuel, the friction pressure drop is required to be verified experimentally. The objectives of this study are to measure the friction coefficient in both laminar and turbulent flow regions, and to develop empirical correlations for the finned rectangular coolant channels for the safety analysis of the MITR. A friction pressure drop experiment is set-up at the MIT Nuclear Reactor Laboratory, where static differential pressure is measured for both flat and finned coolant channels of various channel heights. Experiment data show that the Darcy friction factors for laminar flow in finned rectangular channels are in good agreement with the existing correlation if a pseudo-smooth equivalent hydraulic diameter is considered; whereas a new friction factor correlation is proposed for the friction factors for turbulent flow. Additionally, a model is developed to calculate the primary flow distribution in the reactor core for transitional core configuration with various combinations of HEU and LEU fuel elements. 2008-09-01T00:00:00Z https://hdl.handle.net/1721.1/75083 Nuclear Energy Options for Hydrogen and Hydrogen-Based Liquid Fuels Production Yildiz, Bilge; Kazimi, Mujid S. Nuclear energy can be used for hydrogen production through thermochemical or electrochemical processes for splitting water (and/or steam) into its elemental parts. The overall performance of alternative routes for using nuclear energy to supply the needed heat or electricity depends on the operating temperature, efficiency of the processes involved, complexity of the systems used and capital costs of the nuclear and hydrogen technologies. In this work, we assess the economics of possible technologies to produce hydrogen using nuclear energy. The purpose of this assessment is to identify the most attractive options for further research and development and eventual application to nuclear hydrogen production. Both thermochemical processes and electrolysis require high temperatures for good efficiency. Thus, hydrogen production is best accomplished using advanced reactors that are capable of reaching much higher temperatures than today's LWRs. At temperatures above 700 [degrees]C, the options range from using steam methane reforming in the short term to the much more involved chemical cycles or steam electrolysis in the long term. The helium cooled graphite moderated reactors operating at temperatures above 850 [degrees]C have often been proposed for such purposes. However, we find the high temperature steam electrolysis process coupled to a supercritical CO[subscript 2] gas turbine cycle, possibly in a direct cycle Supercritical Advanced Gas Reactor, as more promising than other technology options. At 650 to 750[degrees]C of reactor outlet/turbine inlet/process temperatures, this technology can achieve 52 to 56% overall efficiency in converting nuclear thermal energy into energy content of hydrogen, respectively. In this work, we also evaluate the technical and economical viability of liquid fuel synthesis using nuclear hydrogen. The liquid fuel can be used in the existing mature infrastructure for transportation and combustion of liquid fuels before large scale hydrogen infrastructure becomes widely established. We propose that CO2 captured from coal plant emissions and nuclear hydrogen be the feedstock to the synthesis process. The cost of this approach would be independent of the natural gas feedstock and may prove market competitive in the near future. Considering the production cost of hydrogen, the thermochemical Sulfur-Iodide cycle coupled to the helium cooled Modular High temperature Reactor (MHR) is found to be also attractive at temperatures above 850 [degrees]C, based on the plant cost and the process efficiency estimates by the designer company. In our work, the cost of production is estimated to fall between $1.13 and 2.37/kg-H2. This range reflects the uncertainties about the operating conditions and cost of the technology in the future. 2003-09-01T00:00:00Z https://hdl.handle.net/1721.1/75021 A Unified Risk-Informed Framework to Assess the Proliferation Risk and License the Proliferation Performace of Nuclear Energy d'Oro, Edoardo Cavalieri; Golay, Michael J. In order to strengthen the current non-proliferation regime it is necessary to guarantee high standards of security for the sites that use, store, produce, or reprocess special nuclear materials (SNM). The current surge of interest in nuclear energy requires resolution of concerns about the appropriateness of the current nuclear non-proliferation regulatory framework for the threats challenging nuclear energy systems (NES). This is especially true also considering that the structure of the current industry is exposed to imminent significant changes such as the introduction of small modular reactors (SMR), and the adoption of nuclear power in countries with unstable political systems. Over recent decades, countries nominally adhering to the Non-Proliferation Treaty (NPT) violated it by building concealed facilities, by manipulating the configuration of their power plants, and by diverting material from their nuclear energy research and production sites. These events show evidence of a major paradigm shift in the area of non-proliferation, which started with the rivalry between two major opponents (each being guardian of its arsenal and technologies during the cold-war), and later reconfigured itself into the confrontation between countries hosting nuclear technologies, or networks of opponents, trying to acquire materials, knowledge and skills necessary to build a nuclear weapon. To create an appropriate response to all the above issues, and thus to strengthen back the non-proliferation regime, while confronting the shifted paradigm of nuclear proliferation, new tools and methods for evaluating the proliferation risk associated with nuclear energy systems become necessary. In this thesis, I discuss some of the fundamental traits and assumptions of the framework I developed in order to assess the proliferation risks associated with NESs. Important decisions within the proliferation domain, can be evaluated by a systematic and holistic approach. The high-level objective of the framework proposed here is to create a license process for the proliferation performance of NESs, and to provide a platform to assist the evaluations of the different alternatives than can be taken in order to strengthen the current non-proliferation regime. 2011-09-01T00:00:00Z https://hdl.handle.net/1721.1/75020 Application of the Technology Neutral Framework to Sodium-­Cooled Fast Reactors Johnson, Brian C.; Apostolakis, George E. Sodium cooled fast reactors (SFRs) are considered as a novel example to exercise the Technology Neutral Framework (TNF) proposed in NUREG-1860. One reason for considering SFRs is that they have historically had a licensing problem due to postulated core disruptive accidents. Two SFR designs are considered, and both meet the goals of the TNF that LWRs typically would not. Considering these goals have been met, a method for improving economics is proposed where systems of low risk-importance are identified as candidates for removal, simplification, or removal from safety grade. Seismic risk dominates these designs and is found to be a limiting factor when applying the TNF. 2011-09-01T00:00:00Z https://hdl.handle.net/1721.1/75019 Stress Corrosion Cracking and Crack Tip Characterization of Alloy X-750 in Light Water Reactor Environments Gibbs, Jonathan Paul; Ballinger, Ronald; Ballinger, Ronald Stress corrosion cracking (SCC) susceptibility of Inconel Alloy X-750 in the HTH condition has been evaluated in high purity water at 93 and 288°C under Boiling Water Reactor Normal Water Chemistry (NWC) and Hydrogen Water Chemistry (HWC) conditions. SCC crack growth rates of approximately 1.1x10-7 mm/s (K=28 MPa√m) under NWC conditions and 1.4x10-8 mm/s (K=28 MPa√m) under HWC in high purity water at 288°C were observed. The environmental conditions were changed from NWC to HWC during constant K loading, and the crack growth rate immediately slowed down by approximately one order of magnitude. The alloy was also tested in HWC at 93°C. No SCC crack growth was observed at K= 35 MPa√m for the length of time tested at 93°C. The fracture mode transitioned from predominantly transgranular cracking under fatigue conditions to a mixture of intergranular, pseudo-intergranular, and a small amount of transgranular fracture in constant stress intensity SCC. Pseudo-intergranular cracking is when a crack propagates directly adjacent to the grain boundary carbides and not actually on the grain boundary. The SCC crack tips were characterized with scanning electron microscopy (SEM) and 3D Atom Probe Tomography (APT). The SEM analysis was focused on the fractographic analysis and crackpropagation mode. The crack was observed to propagate adjacent to grain boundary carbides (pseudo-intergranular) and along a boundary with high coherency where no carbides were present (intergranular). The small and localized areas of transgranular cracking were occasionally seen between two regions of intergranular cracking. The APT reconstructions of the crack tips and crack wall identified several key features contributing to the SCC process: 1) Preferential oxygen transport occurs in either a finger-like or crystallographic morphology extending from the crack tip region. These regions are enriched in both oxygen and oxide with the oxide being a chromium-nickel spinel. 2) The matrix ahead of each finger-like “tunnel” is enriched in oxygen and predominantly chromium oxide. This indicates that oxygen is diffusing ahead of the crack tip into the bulk material. 3) The oxygen that penetrates directly into the base material from the crack walls in an ordered manner suggests that it is controlled by crystallographic features. 4) The main SCC crack tip is full of predominantly oxide phase and, to a lesser extent, metal atoms. The very crack tip forms a spinel of chromium and nickel oxides. Iron oxide begins to contribute to the oxide spinel approximately 25-30 nm from the actual tip. 5) The γ’ precipitates that are directly adjacent to each crack tip and crack wall were deficient in aluminum content. The aluminum content in the bulk γ ’ was approximately 6.6 at % and the near-crack γ ’ aluminum content ranged from 2.5-3.5 at %. The range of affected γ ’ was approximately 100 nm wide. 2011-09-01T00:00:00Z https://hdl.handle.net/1721.1/75018 Uncertainty And Sensitivity Analysis For Long-Running Computer Codes: A Critical Review Langewisch, D. R. 2010-01-01T00:00:00Z https://hdl.handle.net/1721.1/75017 Use of Information Theory Techniques with System Dynamics Models Middleton, Bobby D. Research was done to show how the output distribution of a system with random variables changes as the input distributions and other parameters are changed. Using Information Theory techniques, these different input and output distributions were analytically studied to try to quantify the uncertainty that exists in both the input and output distributions and to determine any relationships between the input uncertainty and output uncertainty. A simplified System Dynamics model of a generic Nuclear Power Plant Construction project was then created. The model was simple enough to allow for relative ease in analyzing data, but detailed enough to allow for changes within the model that could be interpreted as policy changes within the organization. The model was then run with various distributions for different input variables. Data was collected from these simulations for analysis. Using the results of the analytic research mentioned above, this data was analyzed to determine the feasibility of using Information Theory to enhance System Dynamics as a tool to help organizations with policy-making decisions. Results show that use of Information Theory to complement System Dynamics would be helpful in such situations. 2005-09-01T00:00:00Z https://hdl.handle.net/1721.1/75016 A System Dynamics Model of the Energy Policymaking Process Oggianu, Stella Maris; Hansen, Kent F. Electric energy is a fundamental commodity for any aspects of the modern world. However, there are many uncertainties in the sources of electricity that are going to be used in the future. Some of these uncertainties are inherent to the electricity technologies and to the costs of fuels, but the biggest uncertainties come from the impact of future regulations and policies on capital costs, and operations and maintenance costs. Although system dynamics models have been extensively used for applications to the electric power, all the existing models are based on the supply/demand dynamics, and policies are considered as externalities. On the contrary, the energy policymaking model (the EPM model) presented in this report focuses on the complementary problem. This is, the determination of how byproducts and issues related to the adequate supply of electric energy modify the opinions and perceptions of the diverse sectors of the social/political environment; the analysis of the aspects of this environment that account for the formation of energy policies, and the assessment of how these policies are determinants of the technology used to supply electricity. The technologies considered are nuclear, fossil and windmills. The architecture of the EPM model is based on the assumption that policies are formed to minimize societal concerns regarding energy availability and price, nuclear waste, nuclear proliferation, nuclear safety, fossil emissions including greenhouse effect, acid rain, and land requirements for windmills. In this way, each technology is measured by its ability to reduce these concerns. The resulting policies impact on the economics of each of these options. At the same time, economics determines the selection of the new source of electricity. One of the most important results derived from the simulations done through the EPM model is that the revival of the nuclear industry may not be enough to prevent the increase in the production of greenhouse gases. The limited capacity of the industry to build plants is an important factor to consider. Another result is that the opening of Yucca Mountain at the earliest date means the removal of an important barrier for the future growth of the industry, as the risk premium of nuclear power plants may be reduced. Also derived from the use of the EPM model is that the electricity market should not be completely deregulated due to the likely be shortage of electricity supply, and high concerns regarding electricity availability, during peak demands. 2002-08-01T00:00:00Z https://hdl.handle.net/1721.1/75015 Feasibility Investigations for Risk-Based Nuclear Safety Regulation Beer, B. C.; Golay, M. W.; Apostolakis, G. E. 2001-02-01T00:00:00Z https://hdl.handle.net/1721.1/75014 CATILaC: Computer-Aided Technique for Identifying Latent Conditions User's Manual, Version 1.2 Marchinkowski, K.; Weil, R.; Apostolakis, G. A. 2000-04-01T00:00:00Z https://hdl.handle.net/1721.1/74996 Analysis and Utilization of Operating Experience for Organizational Learning Apostolakis, G. E.; Weil, Richard The objective of this report is to clarify the ways that organizational factors influence nuclear power plant performance in order to improve performance. Therefore, this report studied the nuclear power plant organizational environment by identifying and detailing its important work processes. These work processes are: the Work Request Work Process; the Condition Reporting Work Process; the External Operating Experience Work Process; the Design Change Process; and the Procedure Change Work Process. Using this information, a methodology of incident investigation that targets organizational deficiencies contributing to events was developed. Using this methodology to analyze recent significant incidents, a list of important organizational factors and the context within which they influence the successful completion of tasks was identified. These factors are: 1) Communication - Pervasive – Most important between different units and departments; 2) Formalization -Execution; 3) Goal Prioritization - Prioritization; 4) Problem Identification - Planning, scheduling, and return to normal line-up; 5) Roles and Responsibilities - Execution; and 6) Technical Knowledge (job specific knowledge and broad based knowledge) - Job specific knowledge – execution/ Broad based knowledge –prioritization, planning, scheduling, and other tasks. Although safety culture and organizational learning are not listed, they are important. The reason for their exclusion is that they are not single organizational factors useful when cited in incident investigations. Rather, safety culture is a term used to describe all organizational factors, including organizational structure, that impact performance. Similarly, organizational learning was excluded because it is a collection of programs, processes, individual attitudes and culture responsible for learning. Although organizational learning was not listed, it was studied resulting in the development of the Utilization of Operating Experience Work Process. The Utilization of Operating Experience Work Process consists of the following seven steps: 1) Identification; 2) Screening/Prioritization/Dissemination; 3) Investigation/Evaluation; 4) Development; ii i 5) Implementation; 6) Closeout; and 7) Verification/Validation. Since prioritization was identified as important in the above work process and the analysis of significant events, a methodology for the prioritization of work activities at nuclear power plants was developed. This methodology produces a prioritization tool that assigns a numerical performance index to each item requiring prioritization. Applying the methodology at Seabrook Station produced a tool that allowed those who prioritize external operating experience to more efficiently and accurately do so. In addition to the success of the application at Seabrook, a workshop was held at MIT with experts in prioritizing external operating experience. These experts further validated the methodology and the resulting tool. The final piece of work in this report is an analysis of the NRC's revised oversight process as it relates to safety culture. The performance-based regulatory approach is appropriate for regulating safety culture. However, the NRC should continue the analysis of 2001-06-01T00:00:00Z https://hdl.handle.net/1721.1/67679 Gas Heat Transfer in a Heated Vertical Channel under Deteriorated Turbulent Heat Transfer Regime Lee, Jeongik Passive cooling via natural circulation of gas after a loss of coolant (LOCA) accident is one of the major goals of the Gas-cooled Fast Reactor (GFR). Due to its high surface heat flux and low coolant velocities under natural circulation in post-LOCA scenarios, the capability of turbulent gas flow to remove heat from the GFR core can be impaired by either a buoyancy effect or an acceleration effect. These phenomena lead to a Deteriorated Turbulent Heat Transfer (DTHT) regime. To predict accurately the cladding temperature at the hot spot, reliable heat transfer correlations that account correctly for these effects are needed. This work addresses this need by experimentally obtaining heat transfer data and developing new heat transfer correlations that can be used in system analysis codes, such as RELAP5-3D, to reduce uncertainties of predictions in these DTHT regimes. An experimental facility was designed and built using similitude analysis to match key experimental loop parameters to the GFRs' Decay Heat Removal (DHR) system operating conditions to the largest extent possible. Through a thorough literature survey two nondimensional numbers namely (1) the buoyancy parameter (Bo*) and (2) the acceleration parameter (K[subscript v]) were identified as important indicators of the DTHT regime. The experimental data was collected for a range of (1) inlet Reynolds number from 1800 to 42,700, (2) inlet Bo* up to 1x10[superscript -5] (3) and inlet Kv up to 5x10[superscript -6]. The data showed significantly higher reduction of the Nusselt number (up to by 70%) than previously reported (up to 50%). Also, the threshold at which DTHT regime occurs was found to be at smaller non-dimensional numbers than previously reported. A new phenomenon "re-turbulization", where the laminarized heat transfer recovers back to turbulent flow along the channel, was observed in the experiment. A new single phase gas flow heat transfer map is proposed based on the non-dimensional heat flux and the Reynolds number in our data, and is shown to compare well with data in the literature. Three sets of new correlations were developed, which reflect both the buoyancy and acceleration effects and have better accuracy as well as ease of numerical implementation than the existing correlations. The correlations are based on the Gnielinski correlation and replace the Reynolds number subtracting constant by a functional form that accounts for the buoyancy and acceleration effects separately, or in the combined form through a newly introduced nondimensional "DTHT" number. The three correlation types have different complexity level, with the first being the most complex and the third being the most simple and easy to apply without any need for iterations. Additional runs with natural circulation showed that the friction factor in the DTHT regime could be significantly higher than predicted by conventional friction factor correlations, although more experiments will be needed to develop reliable correlations for pressure drop in these regimes. Overall, it is concluded that due to the low heat transfer coefficient and increased friction factor in the DTHT regime, the GFR DHR system should be ideally designed to operate outside the DTHT regime to (1) avoid reduction of heat transfer capability, (2) avoid increase of pressure drop, and (3) reduce uncertainties in predictions of the cladding temperature. 2007-04-01T00:00:00Z https://hdl.handle.net/1721.1/67676 Investigation of Pressure-Tube and Calandria-Tube Deformation Following a Single Channel Blockage Event in ACR-700 Gerardi, C.; Buongiorno, J. The ACR-700 is an advanced pressure-tube (PT) reactor being developed by Atomic Energy of Canada Limited (AECL). As in conventional CANDU reactors, the PTs are horizontal. Each PT is surrounded by a calandria tube (CT), and the gap is filled with carbon dioxide gas. The space between the CTs is filled with the heavy-water moderator. One postulated accident scenario for ACR-700 is a complete coolant flow blockage of a single PT. The flow is not monitored within each individual PT, thus during the early stages of this accident the reactor remains at full power and full pressure, resulting in rapid coolant boil-off and fuel overheating. Melting of the Zircaloy (Zry) components of the fuel bundle (cladding, end plates and end caps) can occur, with relocation of some molten material to the bottom of the PT. The hot spot caused by the molten Zry/PT interaction may cause PT/CT failure due to localized plastic strains. Failure of the PT/CT results in depressurization of the primary system, which triggers a reactor scram, after which the decay heat is removed via reflooding, thus PT/CT rupture effectively terminates the accident. Clearly, prediction of the time scale and conditions under which PT/CT failure occurs is of great importance for this accident. We analyzed the following key phenomena occurring after the blockage: Coolant boil-off Cladding heat-up and melting Dripping of molten Zircaloy (Zry) from the fuel pin Thermal interaction between the molten Zry and the PT Localized bulging of the PT Interaction of the bulged PT with the CT Simple one-dimensional models were adequate to describe (a), (b) and (c), while the three-dimensional nature of (d), (e) and (f) required use of more sophisticated models including a finite-element description of the thermal transients within the PT and the CT, implemented with the code COSMOSM. The main findings of the study are as follows: Most coolant boils off within 3 s of accident initiation. Depending on the magnitude of radiation heat transfer between adjacent fuel pins, the cladding of the hot fuel pin in the blocked PT reaches the melting point of Zry in 7 to 10 s after accident initiation. Inception of melting of the UO2 fuel pellets is not expected for at least another 7 s after Zry melting. Several effects could theoretically prevent molten Zry dripping from the fuel pins, including Zry/UO2 interaction and Zry oxidation. However, it was concluded that because of the very high heat-up rate typical of the flow blockage accident sequence, holdup of molten Zry would not occur. Experimental verification of this conclusion is recommended. Once the molten Zry relocates to the bottom of the PT, a hot spot is created that causes the PT to bulge out radially under the effect of the reactor pressure. The PT may come in contact with the CT, which heats up, bulges and eventually fails. The inception and speed of the PT/CT bulging and ultimately the likelihood of failure depend strongly on the postulated mass of molten Zry in contact with the PT, and on the value of the thermal resistance at the Zry/PT interface. It was found that a Zry mass £10 g will not cause PT/CT failure regardless of the contact resistance effect. On the other hand, a mass of 100 g would be sufficient to cause PT/CT failure even in the presence of a thick 0.2 mm oxide layer at the interface. The characteristic time scales for this 100-g case are as follows: PT bulging starts within 3 s of Zry/PT contact PT makes contact with the CT in another 2 s CT bulging starts in less than 1 s CT failure occurs within another 5 s. Thus, the duration of the PT/CT deformation transient is 11 s, which gives a total duration of the accident (from PT blockage to PT/CT failure) of 18 to 21 s. 2005-11-01T00:00:00Z https://hdl.handle.net/1721.1/67675 Stability Analysis of Supercritical Water Cooled Reactors Zhao, J.; Saha, P.; Kazimi, Mujid S. The Supercritical Water-Cooled Reactor (SCWR) is a concept for an advanced reactor that will operate at high pressure (25MPa) and high temperature (500ºC average core exit). The high coolant temperature as it leaves the reactor core gives the SCWR the potential for high thermal efficiency (45%). However, near the supercritical thermodynamic point, coolant density is very sensitive to temperature which raises concerns about instabilities in the supercritical water-cooled nuclear reactors. To ensure a proper design of SCWR without instability problems, the U.S. reference SCWR design was investigated. The objectives of this work are: (1) to develop a methodology for stability assessment of both thermal-hydraulic and nuclear-coupled stabilities under supercritical pressure conditions, (2) to compare the stability of the proposed SCWR to that of the BWR, and (3) to develop guidance for SCWR designers to avoid instabilities with large margins. Two kinds of instabilities, namely Ledinegg-type flow excursion and Density Wave Oscillations (DWO), have been studied. The DWO analysis was conducted for three oscillation modes: Single channel thermal-hydraulic stability, Coupled-nuclear Out-of-Phase stability and Coupled-nuclear In-Phase stability. Although the supercritical water does not experience phase change, the thermodynamic properties exhibit boiling-like drastic changes around some pseudo-saturation temperature. A three-region model consisting of a heavy fluid region, a heavy-light fluid mixture region and a light fluid region has been used to simulate the supercritical coolant flowing through the core. New non-dimensional governing parameters, namely, the Expansion Number (Nexp) and the Pseudo-Subcooling Number (Npsub) have been identified. A stability map that defines the onset of DWO instabilities has been constructed in the Nexp-Npsub plane based on a frequency domain method. It has been found that the U.S. reference SCWR will be stable at full power operating condition with large margin once the proper inlet orifices are chosen. Although the SCWR operates in the supercritical pressure region at steady state, operation at subcritical pressure will occur during a sliding pressure startup process. At subcritical pressure, the stability maps have been developed based on the traditional Subcooling Number and Phase Change Number (also called as Zuber Number). The sensitivity of stability boundaries to different two phase flow models has been studied. It has been found that the Homogenous- Nonequilibrium model (HNEM) yields more conservative results at high subcooling numbers while the Homogenous Equilibrium (HEM) model is more conservative at low subcooling numbers. Based on the stability map, a stable sliding pressure startup procedure has been suggested for the U.S. reference SCWR design. To evaluate the stability performance of the U.S. reference SCWR design, comparisons with a typical BWR (Peach Bottom 2) have been conducted. Models for BWR stability analysis (Single channel, Coupled-nuclear In-Phase and Out-of-Phase) have been constructed. It is found that, although the SCWR can be stable by proper inlet orificing, it is more sensitive to operating parameters, such as power and flow rate, than a typical BWR. To validate the models developed for both the SCWR and BWR stability analysis, the analytical results were compared with experimental data. The Peach Bottom 2 stability tests were chosen to evaluate the coupled-nuclear stability analysis model. It was found that the analytical model matched the experiment reasonably well for both the oscillation decay ratios and frequencies. Also, the analytical model predicts the same stability trends as the experiment results. Although there are plenty of tests available for model evaluations at subcritical pressure, the tests at supercritical pressure are very limited. The only test publicly found was for the single channel stability mode. It was found that the three-region model predicts reasonable results compared with the limited test data. 2005-09-01T00:00:00Z https://hdl.handle.net/1721.1/67674 Using Risk-Based Regulations for Licensing Nuclear Power Plants: Case Study of the Gas-cooled Fast Reactor Jourdan, G.; Golay, M. W. The strategy adopted for national energy supply is one of the most important policy choice for the US. Although it has been dismissed in the past decades, nuclear power today has key assets when facing concerns on energy dependence and global warming. However, reactor licensing regulations need to be changed to get all the advantages of the most promising technologies. After reviewing the well-known drawbacks of the current regulatory system, the ongoing reforms from the Nuclear Regulatory Commission (NRC) are presented. We argue that full benefice of modern risk analysis methods could not be obtained unless adopting a more ambitious and risk-based regulatory framework. A risk-based licensing framework is then presented, based on previous research from MIT. Probabilistic Risk Assessment (PRA) analyses are used to drive the design toward more safety, and serve as a vehicle for a constructive discussion between designers and the NRC. Mandatory multilevel safety goals are proposed to ensure that adequate safety and adequate treatment of uncertainties are provided. A case-study finally illustrates how this framework would operate. It is based on the Gas-cooled Fast Reactor (GFR) project developed at MIT. We show how PRA provides guidance for the design. Especially, PRA work makes designers consider otherwise overlooked uncertainties and find proper solutions. In a second phase, a simulation of the review by the regulator is conducted. Few new safety concerns are brought. The discussion shows that the proposed risk-based framework has been effective. However, it also highlights that improvements of PRA methodology and clarification over the treatment of key uncertainties are needed. 2005-12-01T00:00:00Z https://hdl.handle.net/1721.1/67673 Analysis Of Flow Instabilities In Supercritical Watercooled Nuclear Reactors Zhao, J.; Saha, P.; Kazimi, Mujid S. Near the supercritical thermodynamic point, coolant density is very sensitive to temperature which gives potential to several instabilities in the supercritical water-cooled nuclear reactors. The flow stability features of the U.S. reference Supercritical Water- Cooled Reactor (SCWR) have been investigated. Single channel stability features were studied by the decay ratio calculations for Density Wave Oscillations (DWO). The system response matrix was developed through perturbation and linearization of the conservation equations in the time domain. Then, the DWO decay ratio was calculated from the dominant eigenvalue of the system response matrix. It was found that the U. S. reference SCWR will satisfy the stability criterion at steady state if an inlet orifice coefficient was properly chosen. Simplified stability maps that define the onset of DWO instability have also been constructed based on a frequency domain method for both the single channel and the channel-to-channel DWO. At supercritical pressure, a three-region model consisting of heavy fluid region, heavy-light fluid mixture region and light fluid region has been used. New non-dimensional governing parameters, namely, the Expansion Number and the Pseudo-Subcooling Number have been identified. It has been found that the U.S. reference SCWR will be stable at full power operating condition with large margin. Although the SCWR operates in the supercritical pressure region at steady state, operation at subcritical pressure will occur during a sliding pressure startup process. At subcritical pressure, the stability maps have been developed based on the traditional Subcooling Number and Phase Change Number (also called as Zuber Number). The sensitivity of stability boundaries due to different two phase flow models has been studied. It has been found that the Homogenous-Nonequilibrium model (HNEM) yields more conservative results at high subcooling numbers while the Homogenous Equilibrium (HEM) model is more conservative at low subcooling numbers. Based on these stability maps, a stable sliding pressure startup procedure has been suggested for the reference SCWR design. 2004-09-01T00:00:00Z https://hdl.handle.net/1721.1/67672 Design of Compact Intermediate Heat Exchangers for Gas Cooled Fast Reactors Gezelius, K.; Driscoll, Michael J. ;; Hejzlar, Pavel Two aspects of an intermediate heat exchanger (IHX) for GFR service have been investigated: (1) the intrinsic characteristics of the proposed compact printed circuit heat exchanger (PCHE); and (2) a specific design optimizing economic and technical efficiency while coupling a supercritical CO[subscript 2] Brayton power cycle to a helium cooled fast reactor core. In particular, the wavy channel friction factor and the effective conduction thickness between channels were evaluated by simulations using state of the art software (Fluent[superscript TM]). To support the competitiveness of the PCHE, it was directly compared to other potential IHX candidates with respect to performance and size for identical operating conditions. All PCHE modeling conservatively assumed straight channels and was carried out using an MIT in-house code. The PCHEs designed specifically for the He/S-CO[subscript 2] cycle were designed to be deployed in a prestressed cast iron reactor vessel (PCIV) pod and to permit a net cycle efficiency of at least 40%. Optimization theory, sensitivity studies, and thermal-hydraulic constraints contributed to shaping the final design. The friction factor analysis showed that the correlations cited in the literature overestimate the value by approximately a factor of two. As regards the effective conduction thickness ratio, it was found to be around 0.6 for a 2.0 mm channel diameter. Since the value of the ratio employed in the MIT in-house code is 1.0, the results generated by the code should be conservative. Comparing the competing IHX types clearly illustrated the advantages of using a compact design, thus favoring PCHEs and plate-fin designs. A maximum net cycle efficiency of 40.9% was achieved for the proposed cycle utilizing a low-pressure-drop reference core. The cost and core volume of this 600 MWt PCHE design were estimated to be $2.4M and 16.4 m[superscript 3], respectively. The largest uncertainty associated with the computations is whether the PCIV pod provides sufficient space for deployment of the PCHE, a blower, and other ancillary equipment. However, studies of PCHEs based on zig-zag channels indicate that the compactness can be further enhanced by a factor of 2 to 3 thanks to the increased heat transfer capability of the saw-tooth channel geometry. More research is needed to verify this projection. 2004-05-01T00:00:00Z https://hdl.handle.net/1721.1/67671 A Super Critical Carbon Dioxide Cycle for Next Generation Nuclear Reactors Dostal, Vaclav; Driscoll, Michael J.; Hejzlar, Pavel A systematic, detailed major component and system design evaluation and multiple-parameter optimization under practical constraints has been performed of the family of supercritical CO[subscript 2] Brayton power cycles for application to advanced nuclear reactors. The recompression cycle is shown to excel with respect to simplicity, compactness, cost and thermal efficiency. The main advantage of the supercritical CO[subscript 2] cycle is comparable efficiency with the helium Brayton cycle at significantly lower temperature (550ºC vs. 850ºC, but higher pressure (20 MPa vs. 8 MPa). The supercritical CO[subscript 2] cycle is well suited to any type of nuclear reactor with core outlet temperature above ~ 500ºC in either direct or indirect versions. By taking advantage of the abrupt property changes near the critical point of CO[subscript 2] the compression work can be reduced, which results in a significant efficiency improvement. However, a real gas cycle requires much more careful optimization than an ideal gas Brayton cycle. Previous investigations by earlier authors were systematized and refined in the present work to survey several different CO[subscript 2] cycle layouts. Inter-cooling, re-heating, re-compressing and pre-compressing were considered. The recompression cycle was found to yield the highest efficiency, while still retaining simplicity. Inter-cooling is not attractive for this type of cycle as it offers a very modest efficiency improvement. Re-heating has a better potential, but it is applicable only to indirect cycles. Economic analysis of the benefit of re-heating for the indirect cycle showed that using more than one stage of re-heat is economically unattractive. For the basic design, turbine inlet temperature was conservatively selected to be 550ºC and the compressor outlet pressure set at 20 MPa. For these operating conditions the direct cycle achieves 45.3% thermal efficiency and reduces the cost of the power plant by ~18% compared to a conventional Rankine steam cycle. The capital cost of the basic design compared to a helium Brayton cycle is about the same, but the supercritical CO[subscript 2] cycle operates at significantly lower temperature. The current reactor operating experience with CO[subscript 2] is up to 650ºC, which is used as the turbine inlet temperature of an advanced design. The thermal efficiency of the advanced design is close to 50% and the reactor system with the direct supercritical CO[subscript 2] cycle is ~24% less expensive than the steam indirect cycle and 7% less expensive than a helium direct Brayton cycle. It is expected in the future that high temperature materials will become available and a high performance design with turbine inlet temperatures of 700ºC will be possible. This high performance design achieves a thermal efficiency approaching 53%, which yields additional cost savings. The turbomachinery is highly compact and achieves efficiencies of more than 90%. For the 600 MWth/246 MWe power plant the turbine body is 1.2 m in diameter and 0.55 m long, which translates into an extremely high power density of 395 MWe/m3. The compressors are even more compact as they operate close to the critical point where the density of the fluid is higher than in the turbine. The power conversion unit that houses these components and the generator is 18 m tall and 7.6 m in diameter. Its power density (MWe/m3) is about ~ 46% higher than that of the helium GT-MHR (Gas Turbine Modular Helium Reactor). A by-pass control scheme is shown to be applicable to the supercritical CO[subscript 2] cycle and exhibits an almost linear efficiency decrease with power. The use of inventory control is difficult since it controls the cycle by changing the operating pressure, which changes the split of the flow between two compressors that work in parallel. The change is so significant that the compressors cannot cope with it. This is mainly because of the current cycle design with a single shaft synchronized with the grid, which was chosen in order to simplify the plant layout, the start-up procedure and eliminate the need for a start up motor. Multiple shaft layouts or compressors with adjustable blade geometry would be necessary to overcome this problem. Since these modifications would increase the capital cost of the system they are not pursued in the present work, which emphasizes base-load performance. 2004-03-01T00:00:00Z https://hdl.handle.net/1721.1/67670 Plant Design and Cost Assessment of Forced Circulation Lead-Bismuth Cooled Reactor with Conventional Power Conversion Cycles Dostal, Vaclav; Hejzlar, Pavel; Todreas, Neil E.; Kazimi, Mujid S. Cost of electricity is the key factor that determines competitiveness of a power plant. Thus the proper selection, design and optimization of the electric power generating cycle is of main importance. This report makes an assessment of power generation of the Actinide Burner Reactor (MABR). The reactor is a fast reactor cooled by lead bismuth eutectic. As a reference plant for capital cost evaluation, the Advanced Liquid Metal Reactor (ALMR) reactor was used based on its 1994 capital and busbar cost estimates. Two balance of plant schemes have been evaluated - a steam cycle and a helium cycle. For the steam cycle, the reference plant is the ALMR steam cycle and for the helium cycle the power generating side of the Modular High Temperature Gas-Cooled Reactor (MHTGR) was used. To identify the basic core design values, a hot channel analysis of the forced cooled core was performed. A scoping design study of the intermediate heat exchanger (IHX) for the helium cycle and the steam generator (SG) for the steam cycle was also carried out. Both were designed using the ALMR IHX as a base case in order to match the modularity criteria imposed on the reactor design and keep the MABR design as close to the reference plant as possible. The estimated cost of electricity for the helium cycle varies from 43.3 to 62.2 mills/kWhe, for the steam cycle from 30.5 to 33.3 mills/kWhe. These ranges in costs reflect the different thermal hydraulic cases. 2001-09-01T00:00:00Z https://hdl.handle.net/1721.1/67669 Conceptual Design of a Lead-Bismuth Cooled Fast Reactor with In-Vessel Direct-Contact Steam Generation Buongiorno, J.; Todreas, N. E; Kazimi, Mujid S.; Czerwinski, K. R. The feasibility of a lead-bismuth (Pb-Bi) cooled fast reactor that eliminates the need for steam generators and coolant pumps was explored. The working steam is generated by direct contact vaporization of water and liquid metal in the chimney above the core and then is sent to the turbine. The presence of a lighter fluid in the chimney drives the natural circulation of the Pb-Bi within the reactor pool. Three key technical issues were addressed: The maximum thermal power removable by direct contact heat transfer without violating the fuel, clad and vessel temperature limits; The consequences of Pb-Bi aerosol transport on the design and operation of the turbine; The release of radioactive polonium (a product of coolant activation) to the steam. Modeling of the multi-phase phenomena occurring in the chimney confirmed the effectiveness of the direct contact heat transfer mode within a well-defined design envelope for the reactor power, chimney height and steam superheat. A 1260MWth power is found possible for 10m chimney height and 25ºC superheat. The temperature of the low-nickel steel clad is maintained below 600ºC, which results in limited corrosion if tight control of the coolant oxygen concentration is adopted. Generation, transport and deposition of Pb-Bi aerosols were also modeled. It was found that the design of a chevron steam separator reduces the heavy liquid metal in the steam lines by about three orders of magnitude. Nevertheless, the residual Pb-Bi is predicted to cause embrittlement of the turbine blades. Four solutions to this problem were assessed: blade coating, employment of alternative materials, electrostatic precipitation and oxidation of the Pb-Bi droplets. An experimental campaign was conducted to investigate the polonium release from a hot Pb-Bi bath to a gas stream. The thermodynamics of the polonium hydride formation reaction (free-energy vs. temperature) as well as the vapor pressure of the lead-polonide were measured and then utilized to model the polonium transport in the reactor. It was found that the polonium concentration in the steam and on the surface of the power cycle components is significantly above the acceptable limits, which makes the very concept of a direct contact reactor open to question. 2001-03-01T00:00:00Z https://hdl.handle.net/1721.1/67668 Comparison Between Air and Helium for Use as Working Fluids in the Energy-Conversion Cycle of the MPBR Galen, T. A.; Wilson, D. G.; Kadak, A. C. A comparison between air and helium for use as working fluids in the energy-conversion cycle of the MPBR is presented. To date, helium has been selected in the MPBR indirect-cycle working reference design. Air open- and closed-cycle variants are considered in this thesis in order to identify relative advantages in cycle efficiency, component efficiency, size, and possible development work required for deployment. The results of this comparison indicate that the helium cycle results in the smallest-sized plant, uses well-established technology, has a high busbar efficiency, and thus best meets the design priorities of the MPBR. The open-cycle-air variant employs turbomachinery components with the greatest amount of industrial experience, the least amount of development work required, and a 6% advantage in busbar efficiency when compared with the helium cycle. However, it results in a plant roughly 5 times the size of the helium plant. The closed-air cycle has a 5% advantage in busbar efficiency over the helium plant, but results in a plant roughly 2.5 times the size of the helium plant and requires approximately the same amount of development work for near-term MPBR deployment. 2011-02-01T00:00:00Z https://hdl.handle.net/1721.1/67666 Plant Design and Cost Estimation of a Natural Circulation Lead-Bismuth Reactor with Helium Power Conversion Cycle Kim, D.; Todreas, N. E.; Kazimi, Mujid S.; Driscoll, M. J. The analysis of an indirect helium power conversion system with lead-bismuth natural circulation primary system has been performed. The work of this report is focused on 1) identifying the allowable design region for the important design parameters and 2) selecting the set of design parameter values for the helium secondary system which lead to the lowest electricity generating cost. An analysis was also performed to examine the capital cost of the ABR/GT and the sensitivity of the capital cost to key parameters. These capital cost estimation and sensitivity analyses were based on available cost estimates of the ALMR and a published HTGR/GT design. The following optimal design parameter values for the helium secondary system were established by this report. Pb-Bi in-tube design for the heat exchanger Triangular tube lattice in the heat exchanger Helium heat exchanger inlet temperature: 250 °C Helium heat exchanger outlet temperature: 500 °C Compression ratio: 3 The ABR/GT capital cost per unit electrical output with helium secondary system is about 36% above that of the steam secondary system case. Sensitivity analyses show about 10% reduction in cost is achieved by increasing the chimney height from 8 m to 15m, 22% cost reduction by increasing the capacity factor from 70-90% and 13% cost reduction by decreasing the construction time from 7 to 3 years. These cost reductions are comparable to those which can be achieved for the ABR with a steam secondary system. The increased cost for the helium versus the steam secondary side results principally from the thermal efficiency difference and the cost difference between steam cycle and helium cycle components. This report is restricted to the capital cost of the ABR/GT. A previous report has estimated the ABR fuel cycle cost. Future economic analysis will include the O&M costs and updated capital estimates based on comparison with the S-PRISM primary system. 2000-11-01T00:00:00Z https://hdl.handle.net/1721.1/67665 Modular Pebble Bed Reactor Kadak, Andrew C.; Ballinger, Ronald G.; Driscoll, Michael J.; Yip, Sidney; Wilson, David Gordon; No, Hee Cheon; Wang, Jing; MacLean, Heather; Galen, Tamara; Wang, Chunyun; Lebenhaft, Julian; Zhai, Tieliang; Petti, David A.; Terry, William K.; Gougar, Hans D.; Ougouag, Abderrafi M.; Oh, Chang H.; Moore, Richard L.; Miller, Gregory K.; Maki, John T.; Smolik, Galen R.; Varacalle, Dominic J. This project is developing a fundamental conceptual design for a gas-cooled, modular, pebble bed reactor. Key technology areas associated with this design are being investigated which intend to address issues concerning fuel performance, safety, core neutronics and proliferation resistance, economics and waste disposal. Research has been initiated in the following areas: • Improved fuel particle performance • Reactor physics • Economics • Proliferation resistance • Power conversion system modeling • Safety analysis • Regulatory and licensing strategy Recent accomplishments include: • Developed four conceptual models for fuel particle failures that are currently being evaluated by a series of ABAQUS analyses. Analytical fits to the results are being performed over a range of important parameters using statistical/factorial tools. The fits will be used in a Monte Carlo fuel performance code, which is under development. • A fracture mechanics approach has been used to develop a failure probability model for the fuel particle, which has resulted in significant improvement over earlier models. • Investigation of fuel particle physio-chemical behavior has been initiated which includes the development of a fission gas release model, particle temperature distributions, internal particle pressure, migration of fission products, and chemical attack of fuel particle layers. • A balance of plant, steady-state thermal hydraulics model has been developed to represent all major components of a MPBR. Component models are being refined to accurately reflect transient performance. • A comparison between air and helium for use in the energy-conversion cycle of the MPBR has been completed and formed the basis of a master’s degree thesis. • Safety issues associated with air ingress are being evaluated. • Post shutdown, reactor heat removal characteristics are being evaluated by the Heating-7 code. • PEBBED, a fast deterministic neutronic code package suitable for numerous repetitive calculations has been developed. Use of the code has focused on scoping studies for MPBR design features and proliferation issues. Publication of an archival journal article covering this work is being prepared. • Detailed gas reactor physics calculations have also been performed with the MCNP and VSOP codes. Furthermore, studies on the proliferation resistance of the MPBR fuel cycle has been initiated using these code • Issues identified during the MPBR research has resulted in a NERI proposal dealing with turbo-machinery design being approved for funding beginning in FY01. Two other NERI proposals, dealing with the development of a burnup “meter” and modularization techniques, were also funded in which the MIT team will be a participant. • A South African MPBR fuel testing proposal is pending ($7.0M over nine years). 2000-07-01T00:00:00Z https://hdl.handle.net/1721.1/67664 Plant Design and Cost Estimation of a Natural Circulation Lead-Bismuth Reactor with Steam Power Conversion Cycle Kim, D.; Todreas, N. E.; Kazimi, Mujid S.; Driscoll, M. J. The analysis of an indirect steam power conversion system with lead-bismuth natural circulation primary system has been performed. The work of this report is focused on 1) identifying the allowable design region for the important design parameters, 2) selecting the set of design parameter values for the steam secondary system which leads to the lowest electricity generation cost and 3) comparing this approach to alternative fast systems. An analysis was performed to examine the capital cost of the ABR and the sensitivity of the capital cost to key design parameters: degree of superheat, secondary system pressure and reactor chimney height. These capital cost estimation and sensitivity analyses were based on the cost estimate of the ALMR report. The following optimal design parameter values for the steam secondary system were established by parameter studies presented in this report. - Pb-Bi in-tube design for the steam generator - Triangular tube lattice in the steam generator - Superheat in steam generator (30°C superheat) - Secondary pressure (70 bar) in the steam generator - No recirculation in the steam generator - Steam generator coolant inlet temperature. The ABR capital cost shows around 15% reduction compared to the ALMR. This is mainly due to the lower cost of the coolant systems due to elimination of the intermediate heat transport system and main coolant pump. Whether, the same ration of reduced cost can be expected in comparison to S-PRISM which is not known but is likely given that the same simplification apply. The ABR capital cost sensitivity analysis shows that the capital cost does not change with degree of superheat, increases with secondary system pressure and decreases with increased reactor chimney height. This report is restricted to the capital cost of the ABR. A previous report has estimated ABR fuel cycle cost. Future economic analysis will include the O&M costs and updated capital estimates based on comparison with the SPRISM. 2000-08-01T00:00:00Z https://hdl.handle.net/1721.1/67657 MCNP4B Modeling of Pebble-Bed Reactors Lebenhaft, Julian Robert The applicability of the Monte Carlo code MCNP4B to the neutronic modeling of pebble-bed reactors was investigated. A modeling methodology was developed based on an analysis of critical experiments carried out at the HTR-PROTEUS and ASTRA facilities, and the critical loading of the HTR-10 reactor. A body-centred cubic lattice of spheres with a specified packing fraction approximates the pebble bed, and exclusion zones offset the contribution of partial spheres generated by the geometry routines in MCNP4B at the core boundaries. The coated fuel particles are modeled in detail and are distributed over the fuelled region of the fuel sphere using a simple cubic lattice. This method predicted the critical core loading accurately in all cases. The calculation of control-rod worths in the more decoupled tall annular ASTRA core gave results within 10% compared to the reported experiments. An approximate method was also developed for the MCNP4B modeling of pebble-bed reactors with burnup. The nuclide densities of homogenized layers in the VSOP94 reactor model are transferred to the corresponding MCNP4B model with the lattice of spheres represented explicitly. The method was demonstrated on the PBMR equilibrium core, and used for a parallel study of burnup k∞ and isotopics on a single pebble. Finally, a study was carried out of the proliferation potential of a modular pebble-bed reactor for both normal and off-normal operation. VSOP94 analysis showed that spent fuel from pebble-bed reactors is proliferation resistant at high discharge burnup, because of its unfavourable plutonium isotopic composition and the need to divert ~157,000 pebbles to accumulate sufficient [superscript 239]Pu for a nuclear weapon. The isotopics of first-pass fuel pebbles are more favourable, but even more pebbles (~258,000) would be needed. However, a supercell MOCUP model was used to demonstrate that ~20,000 pebbles would be needed if loaded with depleted uranium. But the associated reactivity loss would necessitate a compensatory increase in core height of approximately 50 cm. Such a change in core loading, as well as the properties of the special pebbles, would be noticed in a safeguarded facility. 2001-10-15T00:00:00Z https://hdl.handle.net/1721.1/67656 Conceptual Reactor Physics Design of a Lead-Bismuth-Cooled Critical Actinide Burner Hejzlar, Pavel; Driscoll, Michael J.; Kazimi, Mujid S. Destruction of actinides in accelerator-driven subcriticals and in stand-alone critical reactors is of contemporary interest as a means to reduce long-term high-level waste radiotoxicity. This topical report is focused on the neutronic design challenges of a pure critical actinide transmuter. The key objectives of the design were set to be (1) the attainment of a high actinide burning rate comparable to that of the ATW and (2) the attainment of plausible reactor physics characteristics so that the reactor safety performance is at least comparable to that of traditional fast breeder reactors. The proposed conceptual design is a Pb-Bi cooled 1800MWth-core with innovative streaming fuel assemblies that exhibits excellent reactivity performance upon coolant voiding, even for local voids in the core center. The core employs metallic, fertile-free fuel where the transuranics are dispersed in a zirconium matrix. The large reactivity excess at BOL is compensated by a system of double-entry control rods. The arrangement of top-entry and bottom-entry control rods in a staggered pattern allows the achievement of a very uniform axial power profile and a small reactivity change from CRD driveline expansion. Excellent void reactivity performance of the proposed design was demonstrated, together with other desirable features such as a very uniform power profile and tight neutronic coupling. A relatively long refueling interval of one and a half years is achieved using a two-batch refueling scheme. In terms of the TRU destruction rate per MWt per full power year the design is comparable to the accelerator-driven systems and other studied pure burner concepts based on sodium-cooled fast reactors. The effective delayed neutron fraction was found to be about 25% less than that of typical oxide-fueled fast reactors, making the requirements on reactor control performance more demanding. The Doppler coefficient was found to be negative with a magnitude appreciably lower than the typical values of oxide fuels in sodium-cooled reactors, but comparable to the values observed in IFR cores with metallic U-Pu-Zr fuels. The fuel thermal expansion coefficient is also negative, having a magnitude approximately equal to the Doppler coefficient. The proposed core can also incinerate long-lived fission products with an efficiency of about 2.6% of the initial Tc99 inventory per FPY – about the same as critical sodium-cooled pure burners under investigation elsewhere, but less than Tc99 incineration efficiency claimed for accelerator driven systems, like ATW. The strategy of mixing Tc99 uniformly in the fuel within the core at the expense of zirconium matrix was found to yield slightly better Tc transmutation efficiency than the use of designated fuel assemblies with zirconium hydride rods at the core periphery. Thermalized fuel assemblies are penalized by low neutron flux because of self shielding; in addition they increase capture to fission ratio in TRU nuclides in the adjacent fuel assemblies, worsening the TRU burning capability. The incineration of Tc99 in fast spectrum in the rods placed on the core periphery appears to be a more promising alternative than transmutation in thermalized fuel assemblies. 2000-02-01T00:00:00Z https://hdl.handle.net/1721.1/67654 A Semi-Passive Containment Cooling System Conceptual Design Liu, H.; Todreas, N. E.; Driscoll, M. J.; Byun, C. S.; Kim, Y. H.; Grodzinsky, M. The objective of this project was to investigate a passive containment cooling system (PCCS) for the double concrete containment of the Korean Next Generation Reactor (KNGR). Two conceptual PCCS designs: the thermosyphon loop and the internal evaporator-only (IEO) were studied. Based on their requirements, a number of full-scale single-tub experiments have been conducted to investigate the performance of the evaporator, the key component in both PCCS designs. The thermosyphon loop design consists of an evaporator (with integrated exit steam separator) and a condenser heat exchanger. The evaporator heat exchanger is located in the containment atmosphere; on its outside tube surfaces steam condensation in presence of noncondensable gases takes place. The condenser heat exchanger is placed in a large water pool located exterior to the containment building; its storage capability serves as the final heat sink. The numerical simulation in GOTHIC of this design shows that, depending on the water pool initial temperature, ten to fourteen thermosyphon loops are needed in order to keep the containment temperature and the total pressure below the design values for the design basis accident (60 psia) and three-to-five loops for the severe accident (120 psia). The IEO design is similar to the PCCS concept using internal condensers discussed earlier by KAIST. The difference between the IEO and the thermosyphon loop is that the steam exciting the evaporator is directly vented to atmosphere in the IEO rather than the exterior condenser in the thermosyphon loop design. The target of this system is to keep containment pressure below 8.3 bar (12 psia) in the absence of any other containment cooling following a severe accident, which postulates core damage and hydrogen combustion. A DBA scenario (LB LOCA, ECCS flow and no spray) and a severe accident scenario (LB LOCA without ECCS and containment spray flow, 100% Zr oxidation and complete hydrogen combustion), as used in KNGR safety analyses (similar to those in the standard safety analysis report for SYSTEM 80+) were modeled using the GOTHIC computer code. GOTHIC performance analysis of the IEO for the DBA condition shows that this concept can likely meet the design peak pressure of 60 psia, if 10 IEOs are used assuming that the separator water level is sufficiently low. However it is inherently difficult to meet the second design criterion of half of peak design pressure within 24 hours because the temperature difference between the containment and the IEO wall is low in the long term. Fir the severe accident, even with two IEOs, there is no problem in meeting the design criteria of 120 psia during the long-term period, with a generous margin. In addition the peak pressure is just 110 psia even assuming 100% zirconium oxidation and the complete burning of hydrogen. The fouling effect by aerosols on the IEO performance was calculated to be negligible. Judging from the above findings for the performance analysis involving DBAs and severe accidents, it is concluded that the IEO has considerable merit for severe accident mitigation and is worthy of further evaluation. A smooth tube, an axial-finned tube and a radial-finned tube have been tested to experimentally estimate the performance of the reference smooth evaporator tube and the enhancement factors, which may be achieved by finning smooth tubes. An empirical correlation has been developed for numerical analysis use. The Diffusion Layer Model (DLM) has been recommended for use beyond the range of this empirical correlation. Condensation in the presence of helium and tube bundle effects were also studied. Theoretical analysis and experimental results of the two finned tubes suggested an enhancement factor of 4 be used in GOTHIC simulation of the PCCS concept based on smooth tube modeling. 1998-02-01T00:00:00Z https://hdl.handle.net/1721.1/67649 Risk-Informed, Performance-Based Regulatory Implications of Improved Emergency Diesel Generator Reliability Utton, S.; Golay, M. W. The Nuclear Regulatory Commission's (NRC) steady progress towards risk-informed performance-based regulation (RIPBR) prompted the practical application of this regulatory tool in order to demonstrate its potential benefits. This practical demonstration makes up one part of an Idaho National Engineering and Environmental Laboratory (INEEL) sponsored project entitled Integrated Models, Data Bases and Practices Needed for Performance-Based Safety Regulation. Project members selected the emergency diesel generator system as a candidate for assessment because of its high risk importance for core damage frequency (CDF) as well as for its failure to exhibit fulfillment of its current maintenance objectives. An analysis of current NRC maintenance and inspection requirements of the emergency diesel generators at the Millstone 3 nuclear power plant was performed by the project members. Maintenance and inspection items identified as unnecessary or harmful to the EDG qualified as candidates for removal from the current surveillance schedule. Expert testimony and comparisons with similar non-nuclear utility industries aided in the identification of candidate items. Calculations of the subsequent risk, reliability, safety, and economic implications revealed several benefits of the inspection alterations. The modified inspection provided improved backup power availability and defense in depth during the refueling outage. A sensitivity analysis performed on the EDG basic events affected by inspection alteration showed that a 50% reduction in these basic event failure rates would decrease the EDG system failure probability by 13.9%. The altered inspection also shortens the plant's refueling outage critical path therefore decreasing the risk of fuel damage and improving the risk profile of the plant outage. Transfer of the revised inspection to performance while the plant is operating at power resulted in identical refueling outage benefits. Performance of the inspection at power requires an increase in the allowed outage time (AOT) of the plant. The subsequent rise in core damage frequency due to the increased AOT is considered negligible. 1998-01-01T00:00:00Z https://hdl.handle.net/1721.1/67647 Use of Performance Monitoring to Improve Reliability of Emergency Generators Diesel Dulik, J. D.; Golay, M. W. Emergency diesel generators are one of the most important contributors to the core damage failure rate of nuclear power plants. Current required testing and maintenance procedures are excessively strict and expensive without any real justification. Probabilistic risk assessment is used to propose a monitoring system and Technical Specification changes to reduce EDG unavailability without jeopardizing safety, and to ease the excessive deterministic requirements. The EDG fault tree is analyzed to identify the critical failure modes of the EDG, the failure of service water pumps, the failure of EDG building ventilation dampers, and the failure of the EDG "supercomponent," which includes the fuel oil, lubricating oil, cooling water, and starting air systems. We use data from the nuclear industry and the U.S. Navy to identify the most significant EDG supercomponent failure modes, including system fluid leakages, instrumentation & controls failures, electrical power output failures, and the fuel system governors. The monitoring system proposed includes instrumentation for twenty-one of the 121 basic events in the fault tree, for a total of 94.9% of EDG failure contributions. The failure modes identified with industry data are monitored, as are diesel engine mechanical failures currently assessed with teardown inspections. With a 50% reduction in these twenty-one basic event failure rates, the EDG system failure rate is reduced by 41.6%, from 0.097 per year to 0.059 per year. With this reduced failure rate, we propose to extend the EDG surveillance interval from one month to twelve months, to lengthen the running tests from one hour to twenty-four hours, and to eliminate the tear-down inspections conducted during refueling outages. To fully assess the benefits of these proposed changes, the monitoring system should be installed on an EDG on a trial basis. The work reported here demonstrates the feasible gains which can be realized, and proposes, a method for evaluating the efficacy of the system as realized through experimentation. 1997-12-01T00:00:00Z https://hdl.handle.net/1721.1/67642 An Integrated Formal Approach for Developing High Quality Software for Safety-Critical Systems Ouyang, Meng; Golay, Michael W. This report presents the results of a study which devises an Integrated Formal Approach (IFA) for improving specifications of the designs of computer programs used in safety-critical systems. In this IFA, the formal specification techniques of a formal method — Development Before The Fact (DBTF) and its supporting tool — the OO1 Tool Suite, are used systematically to identify and remove various kinds of defects in software specifications. Defects usually exist in most computer programs developed using ad-hoc processes in which mathematical formality is not enforced in the program development effort. Five classes of defects are identified from program studies. The IFA here is designed in order to reduce the number of these defects more efficiently. The information produced from the application of the Approach is also used in a discussion of a conceptual process of updating one's knowledge of the quality of the tested specification. This IFA is then applied in two cases studies. On case is that for specifying the small and functionally simple Reactor Protection System (RPS) program. The other case that for specifying a larger sized, more complex program named the Signal Validation Algorithm (SVA) used in actual nuclear power plant safety systems. The results of the applications show that the IFA can quickly identify and remove any ambiguities and inconsistencies in using words and terms, and incompleteness in defining functions and operations in the specifications. The results also show that for a small program like the RPS, functional correctness can be achieved with very high confidence. For a larger program like the SVA, the IFA could efficiently help the system designers to identity there places where improvements of design in functional completeness and correctness should be made. In all, using this approach requires much less work force while producing larger benefits in obtaining a very reliable specification of the program. 1995-09-01T00:00:00Z https://hdl.handle.net/1721.1/67641 Conceptual Design of a Large, Passive Pressure-Tube Light Water Reactor Hejzlar, P.; Todreas, N. E.; Driscoll, M. J. A design for a large, passive, light water reactor has been developed. The proposed concept is a pressure tube reactor of similar design to CANDU reactors, but differing in three key aspects. First, a solid SiC-coated graphite fuel matrix is used in place of pin-rod bundles to enable the dissipation of decay heat from the fuel in the absence of primary coolant. Second, the heavy water coolant in the pressure tubes is replaced by light water, which serves also as the moderator. Finally, the calandria is connected to a light water heat sink. The cover gas keeps the light water out of the calandria during normal operation, which during loss of coolant or loss of heat sink accidents it allows passive calandria flooding. Calandria flooding also provides redundant and diverse reactor shutdown. The entire primary system is enclosed in a robust, free standing cylindrical steel containment cooled solely by buoyancy-induced air flow, and surrounded by a concrete shield building. It is show that the proposed reactor can survive loss of coolant accidents without scram and without replenishing primary coolant inventory, while the safe temperature limits on the fuel and pressure tube are not exceeded. It can cope with station blackout and anticipated transients without scram — the major traditional contributors to core damage frequency — without sustaining core damage. The fuel elements can operate under post-CHF conditions even at full power, without exceeding fuel design limits. The heterogeneous arrangement of the fuel and moderator ensures a negative void coefficient under all circumstances. Although light water is used as both coolant and moderator, the reactor exhibits high neutron thermalization and a large prompt neutron lifetime, similar to D[subscript 2]O moderated cores. Moreover, the extremely large neutron migration length results in a strongly coupled core with a flat thermal flux profile, and inherent stability against xenon spatial oscillations. 1994-06-01T00:00:00Z https://hdl.handle.net/1721.1/67640 Conceptual Design of a Large, Passive Pressure-Tube Light Water Reactor Hejzlar, P.; Todreas, N. E.; Driscoll, M. J. A design for a large, passive, light water reactor has been developed. The proposed concept is a pressure tube reactor of similar design to CANDU reactors, but differing in three key aspects. First, a solid Sic-coated graphite fuel matrix is used in place of pin-rod bundles to enable the dissipation of decay heat from the fuel in the absence of primary coolant. Second, the heavy water coolant in the pressure tubes is replaced by light water, which serves also as the moderator. Finally, the calandria tank, surrounded by a graphite reflector, contains a low pressure gas instead of heavy water moderator, and the normally-voided calandria is connected to a light water heat sink. The cover gas keeps the light water out of the calandria during normal operation, while during loss of coolant or loss of heat sink accidents it allows passive calandria flooding. Calandria flooding also provides redundant and diverse reactor shutdown. The entire primary system is enclosed in a robust, free standing cylindrical steel containment cooled solely by buoyancy-induced air flow, and surrounded by a concrete shield building. It is shown that the proposed reactor can survive loss of coolant accidents without scram and without replenishing primary coolant inventory, while the safe temperature limits on the fuel and pressure tube are not exceeded. It can cope with station blackout and anticipated transients without scram - the major traditional contributors to core damage frequency - without sustaining core damage. The fuel elements can operate under post-CHF conditions even at full power, without exceeding fuel design limits. The heterogeneous arrangement of the fuel and moderator ensures a negative void coefficient under all circumstances. Although light water is used as both coolant and moderator, the reactor exhibits high neutron thermalization and a large prompt neutron lifetime, similar to DgO moderated cores. Moreover, the extremely large neutron migration length results in a strongly coupled core with a flat thermal flux profile, and inherent stability against xenon spatial oscillations. 1994-06-01T00:00:00Z https://hdl.handle.net/1721.1/67639 Effective Thermal Conductivity of Prismatic MHTGR Fuel Han, J. C.; Driscoll, M. J.; Todreas, N. E. The Reactor Cavity Cooling System (RCCS) is an essential passive safety feature of the Modular High Temperature Gas-Cooled Reactor (MHTGR). Its function is to assure the protection of both public safety and owner investment. As shown schematically in Figure 1.1, the system relies upon all three of the classic modes of heat transfer: conduction through the graphic core dominates energy transport to the reactor vessel, from which radiation is the principal mechanism for heat transfer to the riser tubes, inside which natural convection transfers heat to ambient air which provides the ultimate heat sink. The latter two steps have been the subject of past and on-going analyses at MIT in the support of the MHTGR program, as document in references 1 through 10. In this report we focus attention on the in-vessel aspects of this sequence. 1989-09-30T00:00:00Z https://hdl.handle.net/1721.1/67638 Investigation of Pressure-Tube and Calandria-Tube Deformation Following a Single Channel Blockage Event in ACR-700 Gerardi, Craig Douglas; Buongiorno, Jacopo The ACR-700 is an advanced pressure-tube (PT) reactor being developed by Atomic Energy of Canada Limited (AECL). As in conventional CANDU reactors, the PTs are horizontal. Each PT is surrounded by a calandria tube (CT), and the gap in between is filled with carbon dioxide gas. The space between the CTs is filled with the heavy-water moderator. One postulated accident scenario for ACR-700 is the complete coolant flow blockage of a single PT. The flow is not monitored within each individual PT, thus during the early stages of this accident the reactor remains at full power and full pressure, resulting in rapid coolant boil-off and fuel overheating. Melting of the Zircaloy (Zry) components of the fuel bundle (cladding, end plates and end caps) can occur, with relocation of some molten material to the bottom of the PT. The hot spot caused by the molten Zry/PT interaction may cause PT/CT failure due to localized plastic strains. Failure of the PT/CT results in depressurization of the primary system, which triggers a reactor scram, after which the decay heat is removed via reflooding, thus PT/CT rupture effectively terminates the accident. Clearly, prediction of the time scale and conditions under which PT/CT failure occurs is of great importance for this accident. We analyzed the following key phenomena occurring after the blockage: (a) Coolant boil-off (b) Cladding heat-up and melting (c) Dripping of molten Zircaloy (Zry) from the fuel pin (d) Thermal interaction between the molten Zry and the PT (e) Localized bulging of the PT (f) Interaction of the bulged PT with the CT Simple one-dimensional models were adequate to describe (a), (b) and (c), while the three-dimensional nature of (d), (e) and (f) required the use of more sophisticated models including a finite-element description of the thermal transients within the PT and the CT, implemented with the code COSMOSM. The main findings of the study are as follows: (1) Most coolant boils off within 3 s of accident initiation. (2) Depending on the magnitude of radiation heat transfer between adjacent fuel pins, the cladding of the hot fuel pin in the blocked PT reaches the melting point of Zry in 7 to 10 s after accident initiation. (3) Inception of melting of the UO2 fuel pellets is not expected for at least another 7 s after 2Zry melting. (4) Several effects could theoretically prevent molten Zry dripping from the fuel pins, including Zry/UO2 interaction and Zry oxidation. However, it was concluded that because of the very high heat-up rate typical of the flow blockage accident sequence, holdup of molten Zry would not occur. Experimental verification of this conclusion is recommended. (5) Once the molten Zry relocates to the bottom of the PT, a hot spot is created that causes the PT to bulge out radially under the effect of the reactor pressure. The PT may come in contact with the CT, which heats up, bulges and eventually fails. The inception and speed of the PT/CT bulging and ultimately the likelihood of failure depend strongly on the postulated mass of molten Zry in contact with the PT, and on the value of the thermal resistance at the Zry/PT interface. It was found that a Zry mass =/< 10 g will not cause PT/CT failure regardless of the contact resistance effect. On the other hand, a mass of 100 g would be sufficient to cause PT/CT failure even in the presence of a thick 0.2 mm oxide layer at the interface. The characteristic time scales for this 100-g case are as follows: PT bulging starts within 3 s of Zry/PT contact - PT makes contact with the CT in another 2 s - CT bulging starts in less than 1 s - CT failure occurs within another 5 s. Thus, the duration of the PT/CT deformation transient is 11 s, which gives a total duration of the accident (from PT blockage to PT/CT failure) of 18 to 21 s. 2005-11-01T00:00:00Z https://hdl.handle.net/1721.1/67634 The Martian Surface Reactor: An Advanced Nuclear Power Station for Manned Extraterrestrial Exploration Bushman, A.; Carpenter, D. M.; Ellis, T. S.; Gallagher, S. P.; Hershcovitch, M. D.; Hine, M. C.; Johnson, E. D.; Kane, S. C.; Presley, M. R.; Roach, A. H.; Shaikh, S.; Short, M. P.; Stawicki, M. A. As part of the 22.033/22.33 Nuclear Systems Design project, this group designed a 100 kW[subscript e] Martian/Lunar surface reactor system to work for 5 EFPY in support of extraterrestrial human exploration efforts. The reactor design was optimized over the following criteria: small mass and size, controllability, launchability/accident safety, and high reliability. The Martian Surface Reactor was comprised of four main systems: the core, power conversion system, radiator and shielding. The core produces 1.2 MW[subscript th] and operates in a fast spectrum. Li heat pipes cool the core and couple to the power conversion system. The heat pipes compliment the chosen pintype fuel geometry arranged in a tri-cusp configuration. The reactor fuel is UN (33.1w/o enriched), the cladding and structural materials in core are Re, and a Hf vessel encases the core. The reflector is Zr[subscript 3]Si[subscript 2], chosen for its high albedo. Control is achieved by rotating drums, using a TaB[subscript 2] shutter material. Under a wide range of postulated accident scenarios, this core remains sub-critical and poses minimal environmental hazards. The power conversion system consists of three parts: a power conversion unit, a transmission system and a heat exchanger. The power conversion unit is a series of cesium thermionic cells, each one wrapped around a core heat pipe. The thermionic emitter is Re at 1800 K, and the collector is molybdenum at 950 K. These units, operating at 10[superscript +]% efficiency, produce 125 kW[subscript e] DC and transmit 100 kW[subscript e] AC. The power transmission system includes 25 separate DC-to-AC converters, transformers to step up the transmission voltage, and 25 km of 22 gauge copper wire for actual electricity transmission. The remaining 900 kWth then gets transmitted to the heat pipes of the radiator via an annular heat pipe heat exchanger that fits over the thermionics. This power conversion system was designed with much redundancy and high safety margins; the highest percent power loss due to a single point failure is 4%. The radiator is a series of potassium heat pipes with carbon-carbon fins attached. For each core heat pipe there is one radiator heat pipe. The series of heat pipe/fin combinations form a conical shell around the reactor. There is only a 10 degree temperature drop between the heat exchanger and radiator surface, making the radiating temperature 940 K. In the radiator, the maximum cooling loss due to a single point failure is less than 1%. The shielding system is a bi-layer shadow shield that covers an 80º arc of the core. The inner layer of the shield is a boron carbide neutron shield; the outer layer is a tungsten gamma shield. The tungsten shield is coated with SiC to prevent oxidation in the Martian atmosphere. At a distance of 11 meters from the reactor, on the shielded side, the radiation dose falls to an acceptable 2 mrem/hr; on the unshielded side, an exclusion zone extends to 14 m from the core. The shield is movable to protect crew no matter the initial orientation of the core. When combined together, the four systems comprise the MSR. The system is roughly conical, 4.8 m in diameter and 3 m tall. The total mass of the reactor is 6.5 MT. 2004-12-01T00:00:00Z