Gas-cooled Fast Breeder Reactor Research Articles

THERMAL-HYDRAULIC ANALYSIS OF A HIGH-TEMPERATURE HELIUM-GAS-COOLED REACTOR UNDER VARIOUS STEADY-STATE OPERATING CONDITIONS

This paper focuses on thermal-hydraulic analysis, which plays a critical role in system efficiency and the selection of the optimal design of nuclear reactors. The analysis is done based on a one-dimensional computer code called MIGHT that performs a subchannel thermal-hydraulic analysis of a typical gas-cooled fast breeder reactor (GCFBR) cooled by helium (He). In steady-state operation, two typical channels, the hot and average channels, with the same flow rate and pressure drop were tested. Temperature distribution profiles and the heat flux were computed and compared for different types of power distribution. The effects of coolant mass flow rate and power level on the thermal-hydraulic performance of the tested GCFBR were investigated for cosine power profile. The results demonstrate that the lowest flow rate for the tested reactor to continue operation in the safe mode at the nominal operating power (2530 MWt) is 80% of the nominal flow rate (10 × 10<sup>6</sup> kg/h). The maximum cladding temperature stays within the suggested design limit of GCFRs (700-750°C) when the power is increased by 10% and 15%. The results revealed that temperature is more sensitive to changes in power level than mass flow rate. Data of GCFBR typical reactor were used as input data and for code validation. Good agreement between tested reactor data and MIGHT code calculation concerning the reactor proves the reliability of the methodology of analysis from a thermal-hydraulic perspective. The minor discrepancies could be explained by differences in the relevant physical parameters used in each method of calculation.

Preliminary Study of Gas Cooled Fast Breeder Reactor with Heterogen Percentage of Uranium–Plutonium Carbide based fuel and 300 MWt Power

A preliminary design study of GFR with helium gas-cooled has been performed. In this study used natural uranium and plutonium results LWR waste as fuel. Fuel with a small percentage of plutonium are arranged on the inside of the core area, and the fuel with a greater percentage set on the outside of the core area. The configuration of such fuel is deliberately set to increase breeding in this part of the central core and reduce the leakage of neutrons on the outer side of the core, in order to get long-lived reactor with a small reactivity. Configuration of fuel as it is also useful to generate a peak power reactors with relatively low in both the direction of axial or radial. Optimization has been done to fuel fraction 45.0% was found that the reactor may be operating in more than 10 year time with excess reactivity less than 1%.

Linear regression and sensitivity analysis in nuclear reactor design

The paper presents a general strategy applicable for sensitivity analysis (SA), and uncertainity quantification analysis (UA) of parameters related to a nuclear reactor design. This work also validates the use of linear regression (LR) for predictive analysis in a nuclear reactor design. The analysis helps to determine the parameters on which a LR model can be fit for predictive analysis. For those parameters, a regression surface is created based on trial data and predictions are made using this surface. A general strategy of SA to determine and identify the influential parameters those affect the operation of the reactor is mentioned. Identification of design parameters and validation of linearity assumption for the application of LR of reactor design based on a set of tests is performed. The testing methods used to determine the behavior of the parameters can be used as a general strategy for UA, and SA of nuclear reactor models, and thermal hydraulics calculations. A design of a gas cooled fast breeder reactor (GCFBR), with thermal–hydraulics, and energy transfer has been used for the demonstration of this method. MCNP6 is used to simulate the GCFBR design, and perform the necessary criticality calculations. Java is used to build and run input samples, and to extract data from the output files of MCNP6, and R is used to perform regression analysis and other multivariate variance, and analysis of the collinearity of data.

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Testing of small-scale cavity closure models in support of development of a 300 MW(e) Gas-Cooled fast breeder reactor

The Oak Ridge National Laboratory has been involved in design verification and support studies of the prestressed concrete reactor vessel (PCRV) for a 300 MW(e) Gas-Cooled Fast Reactor demonstration power plant. These studies have been related to the design and testing of small-scale models of the steam generator cavity and reactor core cavity closure plugs for the PCRV. The primary objective of these studies was to determine the structural response of the plugs both to normal operating pressure and to substantial overpressurization. Three closure plug models have been tested: a 1 15- scale full-thickness steam generator cavity closure plug, a 1 15- scale half-thickness steam generator cavity closure plug and a 1 20- scale reactor core cavity closure plug. Results obtained indicate that the closure plugs have a significant overload capacity; that is, in excess of 7·5 times the design pressure of the prototype.

Fuel Chemistry Impact on Vented Fuel Pin Design

Venting and pressure equalization of gas-cooled fast breeder reactor (GCFR) fuel rods can be maintained only as long as axial gas transport paths are available and operating within the fuel pin. Analysis of the chemistry expected within an irradiated GCFR fuel pin shows that axial gas transport paths may become plugged due to the migration of fission product cesium and the formation of low density cesium-urania or cesium-fuel compounds. Cesium transport in the fuel and blanket is controlled by oxygen concentrations and by temperature distribution. Analysis of the shorter Experimental Breeder Reactor II in-reactor fuel pins suggests cesium will be retained in the much longer fast test reactor pins. This will lead to minimal plugging of the gas transport paths.

Guided waves in a circular duct containing an assembly of circular cylinders

This paper provides an analytical scheme to calculate the admissible acoustic propagation modes of fluid in a circular duct containing an assembly of circular cylinders, as might occur in gas-cooled fast breeder reactors and advanced gas-cooled reactors. The duct wall and cylinders are assumed to be stationary, and their axes are assumed to be parallel to each other. The solution to the acoustic wave equation is expressed in a sum of the partial fluid velocity potentials associated with each rod co-ordinate and duct co-ordinate. The technique of transformation of cylindrical wave functions is then used to solve the boundary value problem. Two kinds of acoustic boundary conditions are considered, acoustically hard and acoustically soft, respectively.

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Measurement and Calculation of the Effectiveness of the Gas-Cooled Fast Breeder Reactor Grid-Plate Shield

An analysis of the Oak Ridge National Laboratory Tower Shielding Facility gas-cooled fast breeder reactor (GCFR) grid-plate shield design confirmation experiment was performed and inferences were m...

Fabrication of Grid-Spaced Bundle for the F-5 (X317) Irradiation Experiment

The fabrication of special components and hardware and the assembly of the grid-spaced bundle for the gas-cooled fast breeder reactor (GCFR) F-5 (X317) irradiation experiment now being irradiated in row 5 of the Experimental Breeder Reactor II are described. The F-5 experiment consists of 31 fueled rods and six coolant bypass tubes for the initial loading. The components fabricated and assembled by General Atomic Company are ribbed cladding, UO/sub 2/ blanket pellets (80% +- 2% TD), fission product traps, dosimeters, insulated flow bypass tubes, grid spacers, specially shaped axial spacer tubes, a thermal shield/orifice/flow mixer, a bundle attachment ring, hexagonal tubes with key slots for rod holddown, a rail assembly with attachment to the bottom subassembly shield, dummy fuel rods for flow testing, and assembly of the grid-spaced bundle including check-fit into the flow duct. A variety of fabrication methods, including powder pressing, sintering, outer diameter grinding, thread grinding, metal forming, machining, chemical milling, electrical discharge machining, and tungsten inert gas welding, were used to produce the components.

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Seismic response of gas-cooled fast breeder reactor core structural assembly via modal synthesis

An investigation has been conducted to determine theoretically the dynamic response of the GCFR core support structural assembly when subjected to boundary excitation from seismic disturbances. The system analyzed consists of a thick grid plate to which many core elements are vertically attached. The dynamic problem was solved by synthesizing component modes of two substructures and treating them as continuous subsystems. The investigation is of practical significance in the sense that the radial responses of the core elements in axisymmetric motions cause reactivity change of the core, and therefore an accurate assessment of the dynamic response of the system is important to the core and core support structure design. Numerical system modal data and time-history response results are presented.

Seismic model test of the GCFR core and core support structure

A 15% scale model was constructed to study the dynamic structural behavior of the GCFR (gas cooled fast breeder reactor) core support structure during seismic excitation. The model contains a perforated aluminum plate with a diameter of 20 in. and 265 model core elements constructed from 7/8 in.-diameter aluminum tubes. The proper frequency and mass ratios of the core elements and the perforated plate was ensured by placing steel inserts in the tubes. The natural frequencies, mode shapes and damping factors were individually measured for each of the components and for the complete system. Harmonic and simplified seismic forcing functions were applied to study the dynamic behavior of the core and its support structure. The test results were compared with both analytical and computer code results. Applying thick plate theory, the effective elastic modulus is 27% lower than that given in the ASME code. The resonant frequencies and the mode shapes of the “combined” core and grid plate assembly were also calculated. Applying thick plate theory to the analytical method, the two lowest frequencies were determined and the comparison with the test results shows differences od 3 and 6%.

Fission products and secondary gamma-ray spectra in a plutonium-fueled fast reactor

Abstract The fission products' gamma-ray and gamma-ray energy source spectra for a gas-cooled fast breeder reactor (GCFR) are calculated for different times after shutdown by modifying the RIBD computer code. The secondary gamma-ray energy source spectrum in the core of a GCFR, from fission, inelastic scattering, and capture reactions, is calculated using a typical GCFR neutron spectrum. The computer code LAPHANO is used to generate the multigroup (n, xγ) neutron-coupled gamma-ray transfer matrix. The weak dependence of capture and inelastic gamma ray source spectrum on the neutron flux spectrum has been noted. The fission products and secondary gamma-ray source spectra obtained can be used to calculate heat generation and refueling shielding requirements, etc.

Status of helium-cooled nuclear power systems

Helium-cooled nuclear power systems offer a great potential for electricity generation when their long-term economic, environmental, conservation and energy self-sufficiency features are examined. The high-temperature gas-cooled reactor (HTGR) has the unique capability of providing high-temperature steam for electric power and process heat uses and/or high-temperature heat for endothermic chemical reactions. A variation of the standard steam cycle HTGR is one in which the helium coolant flows directly from the core to one or more closed cycle gas turbines. The effective use of nuclear fuel resources for electric power and nuclear process heat will be greatly enhanced by the gas-cooled fast breeder reactor (GCFR) currently being developed. A GCFR using thorium in the radial blanket could generate sufficient U-233 to supply the fuel for three HTGRs, or enough plutonium from a depleted uranium blanket to fuel a breeder economy expanding at about 10% per year. The feasibility of utilizing helium to cool a fusion reactor has been included in most research studies on thermonuclear fusion and is also discussed in this paper. This paper summarizes the status of helium-cooled nuclear energy systems as a basis for assessing their prospects.

Gas-cooled fast breeder reactors

The need for breeder reactors, breeder performance and the gas-cooled reactor are outlined. The performance of the gas-cooled reactor is compared with that of the liquid metal fast breeder reactor

Nuclear Heating in a Control Element of a Gas-Cooled Fast Breeder Reactor

For the design of control elements for a gas-cooled fast breeder reactor, accurate heat generation rates in the element must be known. The nuclear heating rates in all sections of the control element and the contribution of all radiation sources to the distribution of nuclear heating in the element are evaluated. The emphasis is placed on the direct neutron heating of the B4C control rod and on gamma-ray transport and energy deposition in the stainless-steel cladding and guide tube. The calculations include the helium and tritium production within the B4C control rod.

Pressure Drop of Spacer Grids in Smooth and Roughened Rod Bundles

Experiments have been performed on the pressure drop of a spacer grid in rod bundles of 12 rods. Both smooth and rough rod bundles were used in the experiments. The artificial roughness cut into the outer surface of the rods was chosen so as to meet the actual design of a gas-cooled fast breeder reactor. The results of this investigation for a Reynolds number range between Re = 4 x 10/sup 3/ and 7 x 10/sup 4/ indicate that the pressure drop at the spacer grids is higher in a roughened rod bundle than in a smooth rod bundle.

Analysis of the Venting System of the Gas-Cooled Fast Breeder Reactor Fuel Element for Various Operating Conditions

The fuel element concept of the gas-cooled fast breeder reactor (GCFR) is based on vented fuel pins to equalize pressure differences between the fission gas inside the fuel pin and the coolant. The fission products escaping from the fuel, mainly noble gases, are collected and swept separately from the primary coolant by a helium stream into a purification plant. Calculations were performed to estimate the activity release during normal operation, transient, and accident conditions for a 1000-MW(e) GCFR designed by Kraftwerk Union. The results show that during normal operation, only 0.8% of the total equilibrium noble gas activity in the core will be released into the purification plant. The most severe case for the activity release is a depressurization accident followed by the release of the whole fission gas inventory in the interstitial gas volume of the fuel pins of ∼5.3 × 107 Ci (2.0 EBq). To adsorb this amount of fission gases in the low-temperature charcoal beds of the purification plant, a temporary refrigeration load of ∼173 kW is necessary. Using a purification plant with a refrigeration capacity of ∼50 kW and an equivalent storage of liquid nitrogen for auxiliary purposes, no significant extrapolation from the designed high-temperature gas-cooled reactor purification plants is necessary.

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