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[en] The economic incentives for low-cost electricity generation will continue to drive more plant owners to identify safe and reliable methods to increase the electrical power output of the current nuclear power plant fleet. A power uprate enables a nuclear power plant to increase its electrical output with low cost. However, power uprates brought new challenges to plant owners and operators. These include equipment damage or degraded performance, and unanticipated responses to plant conditions, etc. These problems have arisen mainly from using dated design and safety analysis tools and insufficient understanding of the full implications of the proposed power uprate or from insufficient attention to detail during the design and implementation phase. It is essential to demonstrate that all required safety margins have been properly retained and the existing safety level has been maintained or even increased, with consideration of all the conditions and parameters that have an influence on plant safety. The impact of the power uprate on plant life management for long term operation is also an important issue. Significant capital investments are required to extend the lifetime of an aging nuclear power plant. Power uprates can help the plant owner to recover the investment costs. However, plant aging issues may be aggravated by the power uprate due to plant conditions. More rigorous analyses, inspections and monitoring systems are required.
[en] The technological viability of sodium cooled fast reactors (SFR) has been established by various experimental and prototype (demonstration) reactors such as EBR-II, FFTF, Phoenix, JOYO, BN-600 etc. However, the economic competitiveness of SFR has not been proven yet. The perceived high cost premium of SFRs over LWRs has been the primary impediment to the commercial expansion of SFR technologies. In this paper, cost reduction options are discussed for advanced SFR designs. These include a hybrid loop-pool design to optimize the primary system, multiple reheat and intercooling helium Brayton cycle for the power conversion system and the potential for suppression of intermediate heat transport system. The design options for the fully passive decay heat removal systems are also thoroughly examined. These include direct reactor auxiliary cooling system (DRACS), reactor vessel auxiliary cooling system (RVACS) and the newly proposed pool reactor auxiliary cooling system (PRACS) in the context of the hybrid loop-pool design.
[en] One approach to address the United States Nuclear Power (NP) 2010 program for the advanced light water reactor (LWR) (Gen-III+) intermediate-term spent fuel disposal need is to reduce spent fuel storage volume while enhancing proliferation resistance. One proposed solution includes increasing burnup of the discharged spent fuel and mixing minor actinide (MA) transuranic nuclides (237Np and 241Am) in the high burnup fuel. Thus, we can reduce the spent fuel volume while increasing the proliferation resistance by increasing the isotopic ratio of 238Pu/Pu. For future advanced nuclear systems, MAs are viewed more as a resource to be recycled, and transmuted to less hazardous and possibly more useful forms, rather than simply disposed of as a waste stream in an expensive repository facility. MAs play a much larger part in the design of advanced systems and fuel cycles, not only as additional sources of useful energy, but also as direct contributors to the reactivity control of the systems into which they are incorporated. A typical boiling water reactor (BWR) fuel unit lattice cell model with UO2 fuel pins will be used to investigate the effectiveness of adding MAs (237Np and/or 241Am) to enhance proliferation resistance and improve fuel cycle performance for the intermediate-term goal of future nuclear energy systems. However, adding MAs will increase plutonium production in the discharged spent fuel. In this work, the Monte-Carlo coupling with ORIGEN-2.2 (MCWO) method was used to optimize the MA loading in the UO2 fuel such that the discharged spent fuel demonstrates enhanced proliferation resistance, while minimizing plutonium production. The axial averaged MA transmutation characteristics at different burnup were compared and their impact on neutronics criticality and the ratio of 238Pu/Pu discussed.
[en] An innovative hybrid loop-pool design for the sodium cooled fast reactor (SFR) has been recently proposed with the primary objective of achieving cost reduction and safety enhancement. With the hybrid loop-pool design, closed primary loops are immersed in a secondary buffer tank. This design takes advantage of features from conventional both pool and loop designs to further improve economics and safety. This paper will briefly introduce the hybrid loop-pool design concept and present the calculated thermal responses for unprotected (without reactor scram) loss of forced circulation (ULOF) transients using RELAP5-3D. The analyses examine both the inherent reactivity shutdown capability and decay heat removal performance by passive safety systems
[en] The suppression pool in a boiling water reactor (BWR) plant not only is the major heat sink within the containment system, but also provides the major emergency cooling water for the reactor core. In several accident scenarios, such as a loss-of-coolant accident and extended station blackout, thermal stratification tends to form in the pool after the initial rapid venting stage. Accurately predicting the pool stratification phenomenon is important because it affects the peak containment pressure; the pool temperature distribution also affects the NPSHa (available net positive suction head) and therefore the performance of the Emergency Core Cooling System and Reactor Core Isolation Cooling System pumps that draw cooling water back to the core. Current safety analysis codes use zero dimensional (0-D) lumped parameter models to calculate the energy and mass balance in the pool; therefore, they have large uncertainties in the prediction of scenarios in which stratification and mixing are important. While three-dimensional (3-D) computational fluid dynamics (CFD) methods can be used to analyze realistic 3-D configurations, these methods normally require very fine grid resolution to resolve thin substructures such as jets and wall boundaries, resulting in a long simulation time. For mixing in stably stratified large enclosures, the BMIX++ code (Berkeley mechanistic MIXing code in C++) has been developed to implement a highly efficient analysis method for stratification where the ambient fluid volume is represented by one-dimensional (1-D) transient partial differential equations and substructures (such as free or wall jets) are modeled with 1-D integral models. This allows very large reductions in computational effort compared to multi-dimensional CFD modeling. One heat-up experiment performed at the Finland POOLEX facility, which was designed to study phenomena relevant to Nordic design BWR suppression pool including thermal stratification and mixing, is used for validation. Comparisons between the BMIX++, GOTHIC, and CFD calculations against the POOLEX experimental data are discussed in detail
[en] This paper discusses the use of DRACS to Enhance HTGRs Passive Safety and Economy. One of the important requirements for Gen. IV High Temperature Gas Cooled Reactors (HTGR) is passive safety. Currently all the HTGR designs use Reactor Vessel Auxiliary Cooling System (RVACS) for passive decay heat removal. (1) The decay heat first is transferred to core barrel by conduction and radiation, and then to reactor vessel by thermal radiation and convection; finally the decay heat is transferred to natural circulated air or water systems. RVACS can be characterized as a surface based decay heat removal system. Similar concepts have been widely used in sodium cooled fast reactor (SFR) designs, advanced light water reactors like AP1000. The RVACS is especially suitable for smaller power reactors since small systems have relatively larger surface area. RVACS tends to be less expensive. However, it limits the largest achievable power level for modular HTGRs due to the mismatch between the reactor power (proportional to volume) and decay heat removal capability (proportional to surface). When the relative decay heat removal capability is reduced, the peak fuel temperature increases, even close to the design limit. Annual designs with internal reflector can mitigate this effect therefore further increase the power. Another way to increase power is to increase power density. However, it is also limited by the decay heat removal capability. Besides safety, HTGRs also need to be economical in order to compete with other reactor designs. The limit of decay heat removal capability set by using RVACS has affected the economy of HTGRs. Forsberg (2) pointed out other disadvantages of using RVACS such as conflicting functional requirements for the reactor vessel and scaling distortion for integral effect test of the system performance. A potential alternative solution is to use a volume based passive decay removal system, call Direct Reactor Auxiliary Cooling Systems (DRACS), to remove or mitigate the limitation on decay heat removal capability. DRACS has been widely used in SFR designs and in liquid salt cooled high temperature reactors. The containment cooling system in BWR is another example of volume based decay removal systems. DRACS composes of natural circulation loops with two sets of heat exchangers, one in reactor side and another is in environment side. DRACS has the benefits of increasing the power as needed (scalability) and modularity. This paper introduces the concept of using DRACS to enhance HTGRs passive safety and economy.
[en] The sodium intermediate heat transfer loop is used in existing sodium cooled fast reactor (SFR) plant design as a necessary safety measure to separate the radioactive primary loop sodium from the water of the steam Rankine power cycle. However, the intermediate heat transfer loop significantly increases the SFR plant cost and decreases the plant reliability due to the relatively high possibility of sodium leakage. A previous study shows that helium Brayton cycles with multiple reheat and intercooling for SFRs with reactor outlet temperature in the range of 510 C to 650 C can achieve thermal efficiencies comparable to or higher than steam cycles or recently proposed supercritical CO2 cycles. Use of inert helium as the power conversion working fluid provides major advantages over steam or CO2 by removing the requirement for safety systems to prevent and mitigate the sodium-water or sodium-CO2 reactions. A helium Brayton cycle power conversion system therefore makes the elimination of the intermediate heat transfer loop possible. This paper presents a pre-conceptual design of multiple reheat helium Brayton cycle for an advanced loop type SFR. This design widely refers the new horizontal shaft distributed PBMR helium power conversion design features. For a loop type SFR with reactor outlet temperature 550 C, the design achieves 42.4% thermal efficiency with favorable power density comparing with high temperature gas cooled reactors.
[en] Dispersion fuels represent a significant departure from typical ceramic fuels to address swelling and radiation damage in high burnup fuel. Such fuels use a manufacturing process in which fuel particles are encapsulated within a non-fuel matrix. Dispersion fuels have been studied since 1997 as part of an international effort to develop and test very high density fuel types for the Reduced Enrichment for Research and Test Reactors (RERTR) program. The Idaho National Laboratory is performing research in the development of an innovative dispersion fuel concept that will meet the challenges of transuranic (TRU) transmutation by providing an integral fission gas plenum within the fuel itself, to eliminate the swelling that accompanies the irradiation of TRU. In this process, a metal TRU vector produced in a separations process is atomized into solid microspheres. The dispersion fuel process overcoats the microspheres with a mixture of resin and hollow carbon microspheres to create a TRUC. The foam may then be heated and mixed with a metal power (e.g., Zr, Ti, or Si) and resin to form a matrix metal carbide, that may be compacted and extruded into fuel elements. In this paper, we perform reactor physics calculations for a core loaded with the conceptual fuel design. We will assume a 'typical' TRU vector and a reference matrix density. We will employ a fuel and core design based on the Advanced Burner Test Reactor (ABTR) design. Using the CSAS6 and TRITON modules of the SCALE system for preliminary scoping studies, we will demonstrate the feasibility of reactor operations. This paper will describe the results of these analyses.
[en] This paper investigates the use of enrichment and moderator zoning methods for optimizing the r-z power distribution within sodium cooled fast reactors. These methods allow overall greater fuel utilization in the core resulting in more fuel being irradiated near the maximum allowed thermal power. The peak-to-average power density was held to 1.18. This core design, in conjunction with a multiple-reheat Brayton power conversion system, has merit for producing an industrial level of electrical output (2400MWth, 1000MWe) from a relatively compact core size. The total core radius, including reflectors and shields, was held to 1.78m. Preliminary safety analysis suggests that positive reactivity insertion resulting from a leak between the sodium primary loop and helium power conversion system can be mitigated using simple gas-liquid centripetal separation strategies in the plant's primary loop.