Advanced Nuclear Power Technology Program (ANP) - Technical Reports

Vented Inverted Fuel Assembly Design for an SFR

Wed, 01 Jun 2011 00:00:00 GMT

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.

Effects of Surface Parameters on Boiling Heat Transfer Phenomena

Wed, 01 Jun 2011 00:00:00 GMT

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.

General Analysis of Breed-and-Burn Reactors and Limited-Separations Fuel Cycles

Tue, 01 Feb 2011 00:00:00 GMT

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.

Design of a Functionally Graded Composite for Service in High Temperature Lead and Lead-Bismuth Cooled Nuclear Reactors

Fri, 01 Oct 2010 00:00:00 GMT

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.

Conceptual Design of an Annular-Fueled Superheat Boiling Water Reactor

Fri, 01 Oct 2010 00:00:00 GMT

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.

Use of Response Surface for Evaluation of Functional Failure of Passive Safety System

Mon, 01 Mar 2010 00:00:00 GMT

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.

Thermal-Hydraulic Analysis of Innovative Fuel Configurations for the Sodium Fast Reactor

Sat, 01 Aug 2009 00:00:00 GMT

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.

Critical Heat Flux Enhancement via Surface Modification Using Colloidal Dispersions of Nanoparticles (Nanofluids)

Sun, 01 Jun 2008 00:00:00 GMT

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.

Methods for Comparative Assessment of Active and Passive Safety Systems

Fri, 01 Feb 2008 00:00:00 GMT

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.

Stability Analysis of Natural Circulation in BWRs at High Pressure Conditions

Mon, 01 Oct 2007 00:00:00 GMT

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.

Selection of Correlations and Look-Up Tables for Critical Heat Flux Prediction in the Generation IV "IRIS" Reactor

Thu, 01 Jun 2000 00:00:00 GMT

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.

Gas Heat Transfer in a Heated Vertical Channel under Deteriorated Turbulent Heat Transfer Regime

Sun, 01 Apr 2007 00:00:00 GMT

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.

Investigation of Pressure-Tube and Calandria-Tube Deformation Following a Single Channel Blockage Event in ACR-700

Tue, 01 Nov 2005 00:00:00 GMT

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.

Stability Analysis of Supercritical Water Cooled Reactors

Thu, 01 Sep 2005 00:00:00 GMT

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.

Using Risk-Based Regulations for Licensing Nuclear Power Plants: Case Study of the Gas-cooled Fast Reactor

Thu, 01 Dec 2005 00:00:00 GMT

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.

Analysis Of Flow Instabilities In Supercritical Watercooled Nuclear Reactors

Wed, 01 Sep 2004 00:00:00 GMT

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.

Design of Compact Intermediate Heat Exchangers for Gas Cooled Fast Reactors

Sat, 01 May 2004 00:00:00 GMT

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.

A Super Critical Carbon Dioxide Cycle for Next Generation Nuclear Reactors

Mon, 01 Mar 2004 00:00:00 GMT

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.

Plant Design and Cost Assessment of Forced Circulation Lead-Bismuth Cooled Reactor with Conventional Power Conversion Cycles

Sat, 01 Sep 2001 00:00:00 GMT

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.

Conceptual Design of a Lead-Bismuth Cooled Fast Reactor with In-Vessel Direct-Contact Steam Generation

Thu, 01 Mar 2001 00:00:00 GMT

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.

Comparison Between Air and Helium for Use as Working Fluids in the Energy-Conversion Cycle of the MPBR

Tue, 01 Feb 2011 00:00:00 GMT

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.

Plant Design and Cost Estimation of a Natural Circulation Lead-Bismuth Reactor with Helium Power Conversion Cycle

Wed, 01 Nov 2000 00:00:00 GMT

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.

Modular Pebble Bed Reactor

Sat, 01 Jul 2000 00:00:00 GMT

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).

Plant Design and Cost Estimation of a Natural Circulation Lead-Bismuth Reactor with Steam Power Conversion Cycle

Tue, 01 Aug 2000 00:00:00 GMT

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.

MCNP4B Modeling of Pebble-Bed Reactors

Mon, 15 Oct 2001 00:00:00 GMT

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.

Conceptual Reactor Physics Design of a Lead-Bismuth-Cooled Critical Actinide Burner

Tue, 01 Feb 2000 00:00:00 GMT

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.

A Semi-Passive Containment Cooling System Conceptual Design

Sun, 01 Feb 1998 00:00:00 GMT

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.

Risk-Informed, Performance-Based Regulatory Implications of Improved Emergency Diesel Generator Reliability

Thu, 01 Jan 1998 00:00:00 GMT

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.

Use of Performance Monitoring to Improve Reliability of Emergency Generators Diesel

Mon, 01 Dec 1997 00:00:00 GMT

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.

An Integrated Formal Approach for Developing High Quality Software for Safety-Critical Systems

Fri, 01 Sep 1995 00:00:00 GMT

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.

Conceptual Design of a Large, Passive Pressure-Tube Light Water Reactor

Wed, 01 Jun 1994 00:00:00 GMT

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.

Conceptual Design of a Large, Passive Pressure-Tube Light Water Reactor

Wed, 01 Jun 1994 00:00:00 GMT

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.

Effective Thermal Conductivity of Prismatic MHTGR Fuel

Sat, 30 Sep 1989 00:00:00 GMT

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.

Investigation of Pressure-Tube and Calandria-Tube Deformation Following a Single Channel Blockage Event in ACR-700

Tue, 01 Nov 2005 00:00:00 GMT

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.