Fuel Pin Assemblies Research Articles

Adaptive burnup stepsize selection using control theory for 2-D lattice depletion simulations

A control theory approach is adopted to determine the temporal discretization during two-dimensional lattice physics depletion simulations. Two primary applications of automated and adaptive stepsize control are identified: (i) the presence of strong absorbers such as gadolinium, where the accurate burnout of the isotopes requires a depletion stepsize smaller than typically required, and (ii) high fidelity multiphysics simulations, e.g. loosely coupled physics, where the coupled physics are nonlinear in time and stepsize changes may be necessary to obtain an accurate coupled solution. A conventional predictor–corrector method is used to address the nonlinearity of the nuclide transmutation and neutron flux. An adaptive stepsize method is developed based on monitoring the one-group scalar neutron flux at both the predictor and corrector steps to approximate the convergence residual of the nonlinear solution. A user-specified tolerance on the L2 relative error norm of the scalar neutron flux is utilized by the stepsize controller. Controllers that include integral, proportional, and/or derivative components are investigated and parameterized using Latin hypercube sampling of the controller input parameters. Three distinct fuel loadings of pressurized water reactor 17 × 17 fuel pin assemblies are considered, including no burnable absorbers, Integral Fuel Burnable Absorber, and gadolinium fuel pins. The required depletion stepsizes, as predicted throughout the cycle by the controller, are compared with a very small stepsize (0.01 MW d/kgHM) reference solution and a solution obtained by a typical rule of thumb depletion stepsize sequence.

Passive measurements of mixed-oxide fuel for nuclear nonproliferation

We present new results on passive measurements and simulations of mixed-oxide fuel-pin assemblies. Potential tools for mixed-oxide fuel pin characterization are discussed for future nuclear-nonproliferation applications. Four EJ-309 liquid scintillation detectors coupled with an accurate pulse timing and digital, offline and optimized pulse-shape discrimination method were used. Measurement analysis included pulse-height distributions to distinguish between purely fission neutron sources and alpha-n plus fission neutrons sources. Time-dependent cross-correlation functions were analyzed to measure the fission neutron contribution to the measured sample's neutron source. The use of Monte Carlo particle transport code MCNPX-PoliMi is discussed in conjunction with the measurements.

Investigation on heavy liquid metal cooling of ADS fuel pin assemblies

In the framework of accelerator driven sub-critical reactor systems heavy liquid metals are considered as coolant for the reactor core and the spallation target. In particular lead or lead bismuth eutectic (LBE) exhibit efficient heat removal properties and high production rate of neutrons. However, the excellent heat conductivity of LBE-flows expressed by a low molecular Prandtl number of the order 10 −2 requires improved modeling of the turbulent heat transfer. Although various models for thermal hydraulics of LBE flows are existing, validated heat transfer correlations for ADS-relevant conditions are still missing. In order to validate the sub-channel codes and computational fluid dynamics codes used to design fuel assemblies, the comparison with experimental data is inevitable. Therefore, an experimental program composed of three major experiments, a single electrically heated rod, a 19-pin hexagonal water rod bundle and a LBE rod bundle, has been initiated at the Karlsruhe Liquid metal Laboratory (KALLA) of the Karlsruhe Institute of Technology, in order to quantify and separate the individual phenomena occurring in the momentum and energy transfer of a fuel assembly.

Developments in nondestructive reactor fuel composition analysis

Methods and equipments are described that have been developed at the Radium Institute between 1973 and 1998 for nondestructive analysis of fissile materials present in spent and unirradiated reactor fuel. The methods are based mainly on recording the neutrons and γ rays emitted by the fuel-pin assemblies for various types of reactor. Sometimes, methods are used for recording the induced neutron emission on irradiation with neutrons from isotope sources.

Optimizing fuel-pin assembly location in a research reactor

Optimizing fuel-pin assembly location in a research reactor

Forecasting total energy production from the cores of pressurized-water reactors

The authors approach the problem of forecasting the total energy production from a PWR core by constructing a statistical model for reduced-scale numerical testing of fuel pin assemblies similar in design to those to be used in the core. They assume that test conditions are identical with operating conditions and that the power is produced uniformly over the volume of the test assembly. The monitoring feature for pin-sheath failure is the increased specific radioactivity of the coolant and for this reason the failure criterion for pin assemblies is taken as a preset limiting value of coolant radioactivity. Energy production in the core before this event occurs is taken as the basic characteristic of the working life. Another assumption for the model is that the power distribution is independent of time and coincides with the thermal neutron flux distribution. The authors also assume that coolant radioactivity is determined mainly by short-lived isotopes and that contributions to the total activity from individual defects are the same if the power densities are equal in the formation zones. The last assumption of the model is that at each instant the contribution to coolant activity from a previously formed and developing defect is identified withmore » the equivalent contribution from the corresponding number of elementary defects. Possible escape of fuel from the pins and contamination of the loop are not considered.« less

The P3 approximation for gases and vacuum

It is shown that the P/sub 3/ approximation gives good results for cells having cylindrical vacuum zones surrounded by the medium. Figures show neutron-flux distribution in a five-zone cell, and neutron-flux distribution in a 17-zone cell. The P/sub 3/ approximation for vacuum enables one to calculate a multizone cylindrical fuel-pin assembly containing cavities with little error by comparison with the FCP method from the MK program. Calculations on clusters with gas regions from WIMS and CLUST programs agree well, but the agree well, but the agreement deteriorates on going to blacker rods, which is evidently because the CLUST program does not incorporate the azimuthal field dependence.

Determining the critical power for the fuel-pin assemblies in the RBMK-1500

The authors attempt to increase the accuracy in determining the actual and critical values for the limiting parameters for each core state in order to improve the reliability. One of the basic parameters determining the reliability of the fuel-pin assemblies in the RBMK is the safety factor (k /SUB s/ ) with respect to the limiting permissible power conditional on the heat-transfer crisis. To determine the distribution of k /SUB s/ over the reactor, one needs to know the critical power for each channel.

Accelerating heat and mass transfer in fast-reactor fuel-pin assemblies with oppositely wound wire windings

To equalize the temperature over the cross section of a fastreactor fuel-pin assembly, it is promising to use oppositely wound wire windings on the fuel pins which accelerate the heat and mass transfer by comparison with ordinary (unidirectional) windings. This article presents results from thermohydraulic studies for two forms of such winding: the alternation of smooth and finned fuel pins with oppositely directed windings on adjacent pins and also oppositely directed fins in adjacent rows. The results are compared with experimental data for the currently used pin spacing with ordinary wire windings. Sodium was used as the coolant.

Coolant mixing in a fuel pin assembly utilizing helical wire wrap spacers

Coolant mixing in a helically wire-wrapped, 217 pin, full-scale model of an LMFBR fuel assembly is presented. Salt-injection-conductivity techniques were employed, using water as the coolant. Over four hundred probes were concentrated in a one-sixth sector of the assembly to exhaustively measure the salt distribution. The basic results are presented for use in the verification of thermal hydraulic computer codes and wire-wrap mixing models. Interior channel gross mixing rates were about 9%/cm (effectively 3% per cm per gap); this was more than three times the natural turbulence level. There existed only a small peripheral (swirl) flow around the bundle; its direction averaged only slightly greater than half helical wire wrap angle. The results were reduced into an effective turbulent mixing model, and a wire-wrap forced diversion cross flow model is also presented.