Statistical evaluation of pin gaps in fuel clusters for fugen

The computer program STGAP has been developed to estimate pin gaps in a fuel assembly for FUGEN. The program optionally computes the probable distribution of the pin gap between any adjacent pair of fuel pins, either at a desired location in an assembly or longitudinally averaged over the total effective length of a pin, based on the measured manufacturing and assembling tolerances in geometrical dimensions and mechanical properties of all the independent elements composing a fuel assembly. It also correlates the computed fuel gap distribution with the minimum critical heat flux ratio in the corresponding local subchannel. Sample calculations were performed for the probable distributions of the pin gaps between pairs of adjacent fuel pins in the outermost layer of a FUGEN fuel assembly using the program and satisfactory agreement was obtained with the corresponding measured distributions.     

Three-dimensional deflection analyses of wire-spaced fuel pin bundles under temperature and hydrodynamic force fields with irradiation effects

An analytical method of evaluating the effects of non-uniform thermal expansion, hydrodynamic force acting on the periphery of a fuel pin and thermal and irradiation-induced creep and swelling on the three-dimensional deflection modes of a fuel pin bundle, as well as the deviation in engineering hot channel factor, are presented.The analysis consists mainly of deriving an expression for a flexibility matrix in terms of the general solution of a beam deflection equation with an arbitrary number of loads under either fuel pin to fuel pin or fuel pin to wrapper tube contact condition. The resulting matrix equation is solved for loads corresponding to all contact points, in terms of which the deflection modes are given.The drag coefficient for a wire-wrapped fuel pin in flowing fluid was investigated experimentally. Several cases of sample calculation show that for a prototype LMFBR fuel subassembly consisting of 169 fuel pins, the engineering hot channel factor accounting for the three-dimensional fuel pin bundle deflection is around 1.023–1.035 at zero irradiation and further increases to 1.045 after 500 day irradiation. The maximum load due to pin contact and the order of pin bundle deflection increase according to the irradiation level.     

Pressure pulse on a subassembly wall due to gas release from a failed fuel pin

This paper presents a mathematical model to predict the pressure pulse on a subassembly of fuel pins due to rapid release of gas from a failed pin into liquid coolant between the pins. The subassembly is simulated by a rigid circular tube, and liquid flow inside the tube is assumed incompressible, inviscid, and irrotational. A gas bubble along the centerline of the subassembly is considered to be formed as a result of the gas release from the plenum, and a pressure pulse on the subassembly wall is a consequence of the liquid being accelerated by the gas bubble. It is assumed that the gas bubble grows spherically until it touches the subassembly wall, and then expands as a cylinder with hemispherical ends. This analysis is particularly applicable to the EBR-II reactor.     

The application of the simple fuel pin transient and melting model to thermal-hydraulics in LMFBR subassembly

A simple conduction model with phase change has been developed for the transient analysis of a reactor fuel pin based on lumped parameter techniques. The purpose of this analysis is to provide a simple useful tool to obtain the general information about fuel and clad leaning into the cooling transients and melting. Such a simple fuel and clad thermal transient model is particularly useful to multichannel analysis where conventional conduction computer codes require considerable computing time and storage space. At the present, this formulation is being employed for the analysis of sodium thermohydraulics, sodium voiding, and melting of cladding and fuel in a subassembly of a fast reactor core. A detailed analysis of the predicted coolant, fuel and cladding thermal transients leading into sodium voiding and fuel pin melting has been made in comparison with the results of various in-pile experiments and with the predictions from the existing more complicated codes.     

THIBO experiments — Thermohydraulically induced fuel pin oscillations in Na-cooled reactors

Unusual wear marks and damage to pins and spacer grids have been found in KNK II fuel element. For instance, there was a pronounced interaction not only between the spacers and the cladding, but also among the fuel pins outside the spacer grids. There is a theory postulating that the causes are not hydraulically induced vibrations, but powerful low-frequency pin oscillations created by special thermohydraulic and geometric conditions. This implies that the pin power, the sodium mass flow and velocity, the pin clearance within the spacer grid system, and the pin structure play the decisive roles.This phenomenon is so interesting that we are performing out-of-pile experiments in order to contribute to the verification of the theory and to learn something about the mechanism and the limiting conditions.The test section simulates a single fuel pin in a KNK II subassembly. There is an electric heater with dimensions of Mark II type pins. All spacers can be adjusted very precisely by micrometer screws. A large number of thermocouples indicate the sodium temperatures around the heater. An X-ray system allows part of the heater to be made visible.In 1988, we succeeded in making the fuel pin simulator oscillate with different parameters. The accompanying azimuthal temperature oscillations grew to more than 100 K and a period of more than 10 s. Maximum pin bending, as determined by X-raying, was nearly one millimeter. The reproducibility of the oscillations was remarkably good.     

Modeling of WWER-440 fuel pin behavior at extended burn-up

Currently, there is an ongoing effort to increase fuel discharge burn-up of all LWRs fuel including WWERs as much as possible in order to decrease power production cost. Therefore, burn-up is expected to be increased from 60 to 70 MWd/kg U. The change in the fuel radial power distribution as a function of fuel burn-up can affect the radial fuel temperature distribution as well as the fuel microstructure in the fuel pellet rim. Both of these features, commonly termed the “rim effect.” High burn-up phenomena in WWER-440 UO₂ fuel pin, which are important for fission gas release (FGR) were modeled. The radial burn-up as a function of the pellet radius and enrichment has to be known to determine the local thermal conductivity.In this paper, the radial burn-up and fissile products distributions of WWER-440 UO₂ fuel pin were evaluated using MCNP4B and ORIGEN2 codes. The impact of the thermal conductivity on predicted FGR calculations is needed. For the analysis, a typical WWER-440 fuel pin and surrounding water moderator are considered in a hexagonal pin well. The thermal release and the athermal release from the pellet rim were modeled separately. The fraction of the rim structure and the excessive porosity in the rim structure in isothermal irradiation as a function of the fuel burn-up was predicted. A computer program; RIMSC-01, is developed to perform the required FGR calculations. Finally, the relevant phenomena and the corresponding models together with their validation are presented.     

Statistical evaluation of pin gaps in fuel clusters for fugen (II)

Fuel pin gaps of Fugen fuel assemblies deviate statistically from their nominal value due to manufacturing and assembling tolerances which influence the thermal and hydraulic characteristics of the reactor core. For assurance of the minimum fuel pin gap, an analytical method of evaluating the reliability of spacer gauge tests applied to selected fuel pin gaps arrayed within a Fugen fuel assembly is discussed where a computer program STGAP is utilized.Correlations among the thickness of a spacer gauge, the reliability of the test and the rate of rejecting fuel assemblies whose pin gaps all satisfy the minimum design criterion are discussed in connection with the optimum gauge thickness for a given realiability level of the test. Sample calculation shows that fuel subassemblies installed in a Fugen reactor core have the overall reliability level of 99.9954% at the beginning of fuel life.     

Accuracy evaluation of pin exposure calculations in current LWR core design codes

The deregulated utility environment and better utilization of fuel assemblies in nuclear power plants has allowed designers to burn fuel assemblies to maximum allowable exposures. Any uncertainties, associated with the technical approach and numerical methods used to perform pin exposure calculations may cause either peak power exposure to exceed the Nuclear Regulatory Commission (NRC) exposure limit or lead to excessive conservatism and thus inefficient fuel utilization. In this work, a Monte Carlo based coupled depletion code (MCNP5/ORIGEN-S) is utilized to provide reference solutions in order to assess the accuracy of pin power and pin exposure reconstruction methods in the current commercial and licensed three-dimensional (3D) nodal Light Water Reactor (LWR) core design codes. The developed at the Pennsylvania State University (PSU) MCNP5/ORIGEN-S coupled depletion code system was validated using measured data from the PSU TRIGA research reactor critical experiments.     

COBRA-3M: A modified version of COBRA for analyzing thermal-hydraulics in small pin bundles

COBRA-3M, a modified version of COBRA computer code, is most suitable for the analysis of thermal-hydraulics in small pin bundles commonly used in in-reactor or out-of-reactor experiments. It includes detailed thermal models for the fuel pins and duct walls. It can handle nonuniform power distribution across the bundle and/or within a fuel pin. Temperature dependence of material properties and fuel-cladding gap conductance can be treated. Heat generation in the duct walls and the effect of heat loss to the surroundings can also be simulated. COBRA-3M has been used extensively in the design and analysis of TREAT and SLSF experiments.     

A method of evaluating local turbulent diffusivity of heat in a gridded fuel subassembly

In evaluating the turbulent diffusivity of heat associated with the coolant flow past a grid spacer within an FBR fuel subassembly, a heat diffusion technique is usually employed. However, measurement of subchannel bulk coolant temperature using thermocouples usually involves difficulty due to a steep and non-linear temperature gradient in the subchannels adjacent to a heater pin.A series solution of the heat conduction equation for the coolant flow in subchannels past a grid spacer and a heated section of a dummy fuel pin was derived under a slug flow approximation where the boundary conditions on dummy fuel pins were satisfied by means of the point-matching technique. The solution may be utilized in analyzing the turbulent diffusivity of heat within subchannel coolant flow as a function of distance from a grid spacer based on the measured temperature distribution on the wall of dummy fuel pins, which may be obtained without affecting the subchannel coolant temperature.In an illustrative example, the turbulent diffusivity of heat was most exaggerated at about 50 mm beyond a grid spacer and was approximately five times larger than the corresponding diffusivity without a grid spacer.     

Analysis of temperature noise propagation in a large fuel subassembly of FBR

We have developed a method to calculate the three-dimensional distribution of root-mean-square (RMS) values of temperature noise in the single phase flow in a fast reactor fuel subassembly with a local flow blockage. Employed are the subchannel method in a pin bundle region and the finite difference method in the region downstream of the bundle. We have compared the calculated RMS values of temperature noise with experimental data for a sodium loop test using a wire-spacered 91-pin-bundle fuel sub-assembly with a local blockage. We have investigated the possibility of detection of the blockage by temperature noise by taking into account the influence of structures in the upper part of the subassembly.     

Sensitivity analysis of fuel pin failure performance under slow-ramp type transient overpower condition by using a fuel performance analysis code FEMAXI-FBR

The FEMAXI-FBR is a fuel performance analysis code and has been developed as one module of core disruptive evaluation system, the ASTERIA-FBR. The FEMAXI-FBR has reproduced the failure pin behavior during slow transient overpower. The axial location of pin failure affects the power and reactivity behavior during core disruptive accident, and failure model of which pin failure occurs at upper part of pin is used by reflecting the results of the CABRI-2 test. By using the FEMAXI-FBR, sensitivity analysis of uncertainty of design parameters such as irradiation conditions and fuel fabrication tolerances was performed to clarify the effect on axial location of pin failure during slow transient overpower. The sensitivity analysis showed that the uncertainty of design parameters does not affect the failure location. It suggests that the failure model with which locations of failure occur at upper part of pin can be adopted for core disruptive calculation by taking into consideration of design uncertainties.     

Three-dimensional local coolant flow velocity field under the influence of wire-spacers in an FBR fuel subassembly

In order to evaluate the coolant cross-flow rate among subchannels of a wire-spaced FBR fuel subassembly, the three-dimensional Navier-Stokes equation was solved numerically for a detailed flow velocity distribution within several connected subchannels inside a subassembly, where consideration was focussed on setting up iteratively an approriate velocity field on boundary interfaces enclosing the subchannels under consideration as the boundary condition. Such subchannels may include a peripheral fuel pin.Some of the numerical results obtained are as follows: (1) In an annular channel coaxed with a wire-spaced fuel pin, the maximum azimuthal velocity component does not appear just at the upstream side of a wire spacer but it appears at the leading angle of 180° to the upstream side with respect to the wire-wrapping phase in a fully developed flow region apart from the entrance. (2) In a wire-spaced fuel pin bundle, the transverse velocity increases steeply in the vicinity of the upstream side of a wire-spacer, while it increases gradually with the development of wakes in the downstream side of a wire-spacer. (3) At the peripheral gaps the swirl flow is induced in the wire-wrapping direction along the inner surface of a wrapper tube and its circumferential evolution predicted in the present analysis is in good agreement with experimental data obtained by a MIT group.     

Thermoplastic analysis of a sphere-pac fuel pin

This analysis is concerned with the thermomechanical response of a sphere-pac fuel pin. The fuel is modeled as an elastic-plastic continuum governed by the Mohr-Coulomb yield criterion and associated flow rule under plane strain conditions. Yielding is found to initiate at the outer edge of the fuel, and a plastic zone progresses inward. When the fuel has completely yielded, there exist three distinct plastic zones corresponding to stress states on different facets of the yield surface. Closed form expressions for the displacement and stresses in each of the plastic zones are presented. Numerical results illustrating the variation of displacement and stress with radial position are given.     

Development of safety assessment methodology on fuel element failure propagation in SFR and its application to Monju

Probabilistic and deterministic safety assessments and experimental studies on local fault (LF) propagation in sodium-cooled fast reactors (SFRs) have been performed in many countries because LFs have been historically considered as one of the possible causes of severe accidents. Adventitious fuel pin failures have been considered to be the most dominant initiators of LFs in these probabilistic assessments because of its high frequency of occurrence during reactor operation and possibility of subsequent pin-to-pin failure propagation. Four possible mechanisms of fuel element failure propagation from adventitious fuel pin failure (FEFPA) were identified from a state-of-the-art review of open papers. All the mechanisms for FEFPA analysis including thermal, mechanical and chemical propagation are modeled into a safety assessment code which is applicable to arbitrary SFRs by developing some needed but missing methods. Furthermore, an assessment on FEFPA of Japanese prototype fast breeder reactor (Monju) was performed using this methodology. It was clarified that FEFPA was highly unlikely and limited at most within one subassembly in Monju owing to its redundant and diverse detection and shutdown systems for FEFPA even assuming the propagation. These results also suggested future possibility of run-beyond-cladding-breach operation which would enhance the economic efficiency in Monju.     

CFD investigations of helical wire-wrap fuel pin bundle and its comparison with straight wire bundle

Computational Fluid Dynamics (CFD) investigations of a fast reactor fuel pin bundle wrapped with helical and straight spacer wires have been carried out and the advantages of using helical spacer wire have been assessed. The flow and temperature distributions in the fuel pin bundle are obtained by solving the statistically averaged 3-Dimensional conservation equations of mass, momentum and energy along with high Reynolds number k-ε turbulence model using a customized CFD code CFDEXPERT. It is seen that due to the helical wire-wrap spacer, the coolant sodium not only flows in axial direction in the fuel pin bundle but also in a transverse direction. This transverse flow enhances mixing of coolant among the sub channels and due to this, the friction factor and heat transfer coefficient of the coolant increase. Estimation of friction factor, Nusselt number, sodium temperature uniformity at the outlet of the bundle and clad hot spot factor which are measures of the extent of coolant mixing and non-homogeneity in heat transfer coefficient around fuel pin are paid critical attention. It is seen that the friction factor and Nusselt number are higher (by 25% and 15% respectively) for the helical wire wrap pin bundle compared to straight wire bundle. It is seen that for 217 fuel pin bundle the maximum clad temperature is 750 K for straight wire case and the same for helical wire is 720 K due to the presence of transverse flow. The maximum temperature occurs at the location of the gap between pin and wire. The ΔT between the bulk sodium in the central sub-channel and peripheral sub-channel is 30 K for straight wire and the same for helical wire is 18 K due to the presence of secondary transverse flow which makes the outlet temperature more uniform. The hotspot factor and the hot channel factors predicted by CFD simulation are 10% lower than that used in conventional safety analysis indicating the conservatism in the safety analysis.     

Boiling and dryout behind local blockages in sodium cooled rod bundles

Much attention has been given in LMFBR safety analysis to cooling disturbances caused by local blockages within a fuel subassembly. Such blockages are generally considered to be more probable in gridded fuel pin clusters which present the possibility for solid particles in the coolant to be trapped at grids to form a radially extending flow obstruction. The temperature distribution produced in the region of impaired cooling has been studied in water and sodium experiments in pin bundles of various sizes. The experimental work at KfK on local cooling disturbances culminated in two local blockage experiments in the KNS sodium loop simulating LMFBR fuel elements with a 49% central and a 21% corner blockage. In the frame of this work pin cooling in the wake of the blockage was investigated in single-phase conditions, in boiling conditions up to dryout and in conditions simulating gas release from failed pins. The general aims of the studies were to demonstrate that the consequences of a local blockage do not lead to rapid propagation of damage within a pin bundle and to obtain data for validation of theoretical models.     

Post-irradiation examination of a fuel pin using a microscopic X-ray system: Measurement of carbon deposition and pin metrology

The paper presents some interesting aspects associated with X-ray imaging and its potential application in the nuclear industry. The feasibility of using X-ray technology for the post-irradiation examination of a fuel pin has been explored, more specifically pin metrology and carbon deposition measurement. The non-active sample was specially designed to mimic the structure of an AGR fuel pin whilst a carbon based material was applied to the mock up fuel rod in order to mimic carbon deposition. Short duration low energy (50 kV) 2D digital radiography was employed and provided encouraging results (with respect to carbon deposition thickness and structure measurements) for the mock up fuel pin with a spatial resolution of around 10 μm. Obtaining quantitative data from the resultant images is the principal added value associated with X-ray imaging. A higher intensity X-ray beam (90 kV) was also used in conjunction with the low energy set-up to produce a clear picture of the cladding as well as the interface between the lead (Pb mimics the uranium oxide) and stainless steel cladding. Spent fuel metrology and routine radiography are two additional tasks that X-ray imaging could perform for the post-irradiation examination programme. Therefore, when compared to other techniques developed to deliver information on one particular parameter, X-ray imaging offers the possibility to extract useful information on a range of parameters.     

The pin power distribution in theVVER-1000 mock-up on the LR-0 research reactor

The pin power density distribution in reactor is an important quantity, necessary for the adequate assessment of fuel conditions and of core structures and pressure vessel radiation embrittlement as well.The paper shows the detailed comparison of calculated and experimentally determined pin by pin power distribution. To verify the reliability of measured data used for comparison with calculated data, the symmetrically located pins were measured. The calculations have been done with deterministic and Monte Carlo approach. The effect of different data libraries used for calculations are discussed as well.     

Probabilistic evaluation of core-internal coolant flow distribution with correlation among fuel subassemblies

The effect of local and global uncertainties in properties and tolerances on the probabilistic variation in coolant flow rate of the hottest channel in a prototype FBR core is discussed in terms of the coolant flow correlation among all the core-internal coolant channels. Sample calculations of the probabilistic deviations in subassembly flow rate for a prototype FBR were about 30% smaller than those conventionally calculated.