Manual for ENERGY I, II, III computer programs

The ENERGY codes for predicting coolant temperature distributions in LMFBR were wrapped fuel and blanket assemblies are described. The mathematical models, data input, code listings, and sample problems are presented. (JWR

Surveillance strategy for an extended operating cycle in commercial nuclear reactors

The impetus for improved economic performance of commercial nuclear power plants can be partially satisfied by increasing plant capacity factors through operating cycle extension. One aspect of an operating cycle extension effort is the modification of plant surveillance programs to complete required regulatory and investment protection surveillance activities within the extended planned outage schedule. The goal of this paper is to introduce a general strategy for existing power plants to transition their surveillance programs to an extended operating cycle up to 48 months in length, and to test the feasibility of this strategy through the complete analysis of the surveillance programs at operating BWR and PWR case study plants. The reconciliation of surveillances at these plants demonstrates that surveillance performance will not preclude 48 month operating cycles. Those surveillance activities that could not be resolved to an extended cycle are identified for further study. Finally, a number of general issues are presented that should be considered before implementing a cycle extension effort

Final report on MIT contract on fuel-coolant interactions

The work of this contract has yielded results in three areas which can be summarized as follows. The existing body of molten metal/water interaction data appears consistent with the criteria of the spontaneous nucleation theory. The observed fragmentation of UO/sub 2/ can be satisfactorily predicted using the proposed thermal stress initiated fracture model. The work potential of a hypothetical fuel vapor source acting on a sodium slug can be reduced by about a factor of two by considering entrainment due to Taylor instabilities at the vapor-liquid interface

COOLANT MIXING IN SODIUM COOLED FAST REACTOR FUEL BUNDLES.

Radial subchannel temperature imbalances can be quite severe in typical fast reactor designs of tight pitch to diameter ratios. Consideration of mixing between interconnected subchannels is required to avoid the significant design penalties which would otherwise would be imposed. The available experimental data and analytic methods for predicting coolant mixing are presented and their deficiencies relative to sodium cooled fast reactor application discussed. Sodium conduction effects, subchannel area changes and forced mixing effects due to grid and wire wrap pin spacing techniques are the key fuel assembly characteristics which must be considered in conducting experiments and formulating analytic methods for fast reactor applications

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

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.United States. Dept. of Energ

Modeling and design of a reload PWR core for a 48-month fuel cycle

The objective of this research was to use state-of-the-art nuclear and fuel performance packages to evaluate the feasibility and costs of a 48 calendar month core in existing pressurized water reactor (PWR) designs, considering the full range of practical design and economic considerations. The driving force behind this research is the desire to make nuclear power more economically competitive with fossil fuel options by expanding the scope for achievement of higher capacity factors. Using CASMO/SIMULATE, a core design with fuel enriched to 7{sup w}/{sub o} U{sup 235} for a single batch loaded, 48-month fuel cycle has been developed. This core achieves an ultra-long cycle length without exceeding current fuel burnup limits. The design uses two different types of burnable poisons. Gadolinium in the form of gadolinium oxide (Gd{sub 2}O{sub 3}) mixed with the UO{sub 2} of selected pins is sued to hold down initial reactivity and to control flux peaking throughout the life of the core. A zirconium di-boride (ZrB{sub 2}) integral fuel burnable absorber (IFBA) coating on the Gd{sub 2}O{sub 3}-UO{sub 2} fuel pellets is added to reduce the critical soluble boron concentration in the reactor coolant to within acceptable limits. Fuel performance issues of concern to this design are also outlined and areas which will require further research are highlighted

A Methodology for Characterizing Representativeness Uncertainty in Performance Indicator Measurements of Power Generating Systems

In this work, a general methodology and innovative framework to characterize and quantify representativeness uncertainty of performance indicator measurements of power generation systems is proposed. The representativeness uncertainty refers to the difference between a measurement value of a performance indicator quantity and its reference true value. It arises from the inherent variability of the quantity being measured. The main objectives of the methodology are to characterize and reduce the representativeness uncertainty by adopting numerical simulation in combination with experimental data and to improve the physical description of the measurement. The methodology is applied to an industrial case study for demonstration. The case study involves a computational fluid dynamics (CFD) simulation of an orifice plate-based mass flow rate measurement, using a commercially available package. Using the insight obtained from the CFD simulation, the representativeness uncertainty in mass flow rate measurement is quantified and the associated random uncertainties are comprehensively accounted for. Both parametric and nonparametric implementations of the methodology are illustrated. The case study also illustrates how the methodology is used to quantitatively test the level of statistical significance of the CFD simulation result after accounting for the relevant uncertainties