A new method for rapid computation of transient fuel cycle material balances

This paper develops and demonstrates a fast, medium-fidelity coupled burnup, criticality and fuel cycle mass flow model. The burnup model requires pre-calculated parameters (neutron production and destruction rates, burnup, and isotopic transformation) that are functions of fluence and nuclide. These parameters are specific to a given reactor. It then uses mass-weighted superposition to recombine these parameters as needed and calculate criticality and maximum discharge burnup. This may then be folded in with a piecewise-linear reactivity model to simulate multi-batch cores. The resultant model is then applied to two fuel cycle scenarios: a recyclable uranium cycle and a fast burner reactor cycle. Various fuel cycle parameters are measured, such as isotopics at every stage of the respective fuel cycles. The dynamic and flexible nature of the model allows such fuel cycle data to be quickly recalculated for various initial conditions.     

The anisotropy of the high temperature deformation of Zircaloy-cladding tubes for nuclear fuel rods

The anisotropy of the high temperature deformation of Zircaloy-4 cladding tubes for nuclear fuel rods for pressurized water reactors has been investigated. The axial and tangential components of the deformation of internally pressurized tube samples during closed end creep rupture tests in air at 800°C have been measured. An axial contraction of the tube sample is observed. Using Hill's theory of plasticity the axial strain can be described by anisotropy coefficients which depend on the texture of the tube material. The anisotropy coefficients are quantitatively related to the orientation distribution of the basal poles in the radial/tangential plane of the tube sample. For the typical texture of Zircaloy cladding tubes of nuclear fuel rods for pressurized water reactors, an axial contraction has to be expected under the biaxial stress conditions applied.     

FRAP-T6: A computer code for the transient analysis of oxide fuel rods

The FRAP-T6 computer code was developed to model the transient performance of light water reactor fuel rods during reactor transients ranging from mild operational transients to large break loss-of-coolant accidents. The code models all of the thermal, structural, and chemical phenomena needed for the complete evaluation of light water reactor fuel rod performance. The code was developed using rigorous quality assurance procedures and a large assessment data base. The results of assessment show that the code accurately models the response of light water reactor fuel rods.     

The DIMCO system for mechanical design of fuel rods

This study presents the system developed at JEN for one- and two-dimensional mechanical analysis of nuclear fuels. The mathematical and numerical bases are described, as well as different models representing phenomena such as cracking, swelling, etc. Numerical results of several test cases are presented: (a) checking of one- and two-dimensional analyses, (b) applicability margin, (c) interaction of effects; and (d) influence of loading histories.     

A method for accurate computation of elastic and discrete inelastic scattering transfer matrices

A method for accurate computation of elastic and discrete inelastic scattering transfer matrices is described. In particular, a partition scheme for the source energy range that avoids integration over intervals containing points where the integrand has a discontinuous derivative is developed. Five-figure accurate numerical results are obtained for several test problems with the trama program, which incorporates the proposed method. A comparison with numerical results from existing processing codes is also presented.     

WTRLGD—A computer program for the transient analysis of waterlogged fuel rods under the RIA condition

A computer code WTRLGD has been developed to describe the transient internal pressure of a waterlogged fuel rod during power burst and also to predict the possibility of the rod failure in the mode of cladding rupture. The code predicts transient thermal behavior of the fuel rod on the basis of an assumption of axisymmetry, and thermal-hydraulic transients of the internal water on the basis of a homogeneous volume-junction model modified so as to involve the cladding deformation. Calculated transients of the rod pressure are in fairly good agreement with those measured in the NSRR experiments, simulating the fuel rod behavior under an RIA condition. The comparison between calculation and experiment verifies that the code is an effective tool for the prediction of the failure of a waterlogged fuel rod.     

Determination of within-rod caesium and europium isotopic distributions in high burnup fuel rods through computerised gamma-ray emission tomography

Computerised gamma-ray emission tomography has been applied to single PWR UO₂ fuel rods, with pellet averaged burnups of 52, 71, 91 and 126 GWd/t respectively, for the determination of ¹³⁴Cs, ¹³⁷Cs and ¹⁵⁴Eu internal radial distributions. State-of-the-art image reconstruction techniques, analytical and iterative, have been applied, evaluated and compared using test phantoms first and, in a second step, the actual measured data. Further, linear attenuation maps, previously derived on the same samples by means of gamma-ray transmission tomography, have been used to correct for density inhomogeneities. The final results have indicated large central depressions in the caesium distributions, but of varying extent from sample to sample. Particularly interesting is the case of the 126 GWd/t sample, showing a very deep central depression (periphery-to-centre ratios of ∼2.5 for ¹³⁷Cs and ∼3 for ¹³⁴Cs). In addition, a difference in the relative activity distributions of ¹³⁷Cs and ¹³⁴Cs has been observed for all the samples. In contrast, the europium shows an almost flat distribution.     

Transmission computer microtomography for nondestructive monitoring of fuel and control rods in a reactor

A method of x-ray transmission computer microtomography has been developed for solving the problems of monitoring the quality of fuel elements and control rods in nuclear reactors: the geometric resolution of a defect is several microns. The solution of the problem of nondestructive monitoring of such objects has made it possible to perform a wide range of investigations of the technical characteristics of definite x-ray detectors, organize principles of scanning, based on the principles of laser interferometry and the design of data processing systems, that are different from those of conventional systems. The investigations performed have made it possible to implement computer-aided design for problem-oriented computer microtomographs and for the instrumentational implementation of an experimental variant of such a device. Investigations performed on specially fabricated test samples with calibrated defects have demonstrated that the approach to the design of such setups for nondestructive monitoring of objects for nuclear power generation is correct and that this is a promising direction, 8 figures, 9 references. All-Russia Scientific-Research Institute of Automatic Machine Engineering. Translated from Atomnaya énergiya, Vol. 88, No. 2, pp. 125–137, February, 2000.     

Effects of a central hole in a fuel tablet on the thermophysical characteristics of VVER fuel rods

Reactors Institute and Kurchatov Institute, Russian Scientific Center. Translated from Atomnaya énergiya, Vol. 80, No. 4, pp. 306–308, April, 1996.     

Evaluation of uranium thorium and plutonium thorium fuel cycles in a very high temperature hybrid system

In recent times, there is a renewed and additional interest in thorium because of its interesting benefits. Thorium fuel cycle is an attractive way to produce long term nuclear energy with low radiotoxicity waste. In addition, the transition to thorium could be done through the incineration of weapons grade plutonium or civilian plutonium. Th-based fuel cycles have intrinsic proliferation-resistance and thorium is 3–4 times more abundant than uranium. Therefore, thorium fuels can complement uranium fuels and ensure long term sustainability of nuclear power.In this paper, the main advantages of the use of fuel cycles based on uranium-thorium and plutonium-thorium fuel mixtures are evaluated in a hybrid system to reach the deep burn of the fuel. To reach this goal, the preliminary conceptual design of a hybrid system composed of a critical reactor and two Accelerated Driven Systems, of the type of very high temperature pebble-bed systems, moderated by graphite and cooled by gas, is analyzed.Uranium-thorium and plutonium-thorium once-through and two stages fuel cycles are evaluated. Several parameters describing fuel behaviour and minor actinide stockpile are compared for the analyzed cycles.     

Calculating the transient temperature field in a reactor tube and the thermoelastic stresses in the fuel element cladding

Approximate analytic methods are given for calculating the transient temperature field in the fuel elements and the coolant temperatures at any point along the reactor tube, as well as the transient thermoelastic stresses in the cladding of a cylindrical fuel element. The coolant temperature at the input to the tube is constant, and the coolant undergoes no changes in state of aggregation. The approximate methods are illustrated by examples.Results are given, for comparison, of accurate calculations of the same examples made with a rapid calculating machine.List of symbols time - r; z coordinates (radius, distance along tube) - r₁; r₂ internal and external radii of fuel element cladding respectively - H total active length of fuel element - a₁; ₁;c _{1 1} coefficients of temperature conductivity, heat conductivity, specific heat capacity and specific gravity of fissionable material respectively - a₂; ₂; C_p2; ₂ cladding parameters - a; ; c_p; coolant parameters - mean cladding radius - f:f₂ cross-sectional area of tube for coolant and cladding respectively - w coolant velocity - coefficient of heat release to coolant - t (r, ); (); () fuel temperature, mean temperature over cross section of cladding, and coolant temperature at pointz. along tube respectively - qv() specific volume of coolant at pointz - values averaged overz - quantities at the initial instant of time - ₃ delay time - _n time required for coolant to go from z=0 to the point in question     

The effect of a changing fuel solution composition on a transient in a fissile solution

This paper presents an extension to a point kinetics model of fissile solution undergoing a transient through the development and addition of correlations which describe neutronics and thermal parameters and physical models. These correlations allow relevant parameters to be modelled as a function of time as the composition of the solution changes over time due to the addition of material and the evaporation of water from the surface of the solution. This allows the simulation of two scenarios. In the first scenario a critical system eventually becomes subcritical through under-moderation as its water content evaporates. In the second scenario an under-moderated system becomes critical as water is added before becoming subcritical as it becomes over-moderated. The models and correlations used in this paper are relatively idealised and are limited to a particular geometry and fissile solution composition. However, the results produced appear physically plausible and demonstrate that simulation of these processes are important to the long term development of transients in fissile solutions and provide a qualitative indication of the types of behaviour that may result in such situations.     

An approximate method for calculating the temperature distribution in gas-filled cracked fuel elements

Treating radial cracks by an idealized geometry in fuel elements containing a thermal source gradient, the temperature distribution throughout a cracked fuel element, its gaseous bonding gap, and its surrounding can is found by an approximate series expansion method. This approximation allows the cracked fuel zone to be described by a single series expansion rather than by a multizone approach and allows approximate satisfaction of the boundary conditions at the interfaces of the cracked fuel zone and both at the non-cracked fuel zone and at the can. The method allows for the temperature dependence of the fuel and that of the gas mixture in the cracks and bonding gap, and takes into account the effect of crack openings on gap conductance at the fuel-can interface. The temperature distribution, its mean and variance resulting from the stochastic distribution of radial crack locations, are found. The case of hairline radial cracks, the mean locations of which are azimuthally uniformly distributed, is used to illustrate the importance of cracks on the can temperature distribution and it is shown that their presence may lead to central fuel zone melting problems.     

超临界水堆堆芯新型燃料组件设计分析

超临界水堆(SCWR)作为六种第四代未来堆型中唯一的水冷反应堆,具有良好的经济性与技术延续性.本文采用最新开发的热工-物理耦合计算程序对超临界水堆方形燃料组件进行稳态热工与中子物理耦合分析,提出一种新型的超临界水堆堆芯燃料组件设计.现有单排组件设计与新型双排燃料组件设计的计算结果表明,双排组件具有功率径向分布均匀,包壳...     

Verification of a mechanistic strain rate equation for transient fuel modelling

The behaviour of CANDU-PHW fuel elements in a transient is very dependent upon the development of sheath strain during the transient. So that uncertainties in predictions that usually involve extrapolations from the data base are kept small, a sheath strain rate equation that reflects the physical processes that are involved in strain has been developed. Recently completed verification tests reveal, that the average error of predictions by this model is indeed small, but the standard deviation is very large. It is shown that variations in structure and dimensions, permitted by the manufacturing tolerances for the specimens, and uncertainties in the experimental measurements, can account for this scatter. Some identified deformation mechanisms are not yet well enough quantified to be included in the model, and their omission could be the reason for the average trend to overpredict slightly. It is concluded that the model is reliable for probabilistic predictions of fuel behaviour.     

Using a relative weighting method for shim rods in a BN-600 reactor

I. V. Kurchatov Beloyarsk Nuclear Plant. Translated from Atomnaya Énergiya, Vol. 70, No. 1, pp. 49–50, January, 1991.     

Design of coated fuel particles for a hybrid fusion-fission system

Designs have been developed for coated ThO₂ fuel particles to be used in a hybrid fusion-fission system that could be operated without reprocessing. The fresh fertile fuel particle would first be cycled through the blanket of a fusion reactor to breed ²³³U, which would then be ‘burned’ in a thermal fission reactor. The depleted fuel would then be refreshed in a second pass through the fusion reactor, and the process above repeated as many times as feasible. Designs of coated particles for up to three cycles through the hybrid system of reactors have been developed. The outer structural layer for these particles is made from vapor-deposited silicon carbide, because of its remarkable dimensional stability under fast neutron irradiation, and an inner layer of porous pyrocarbon is used to accommodate the buildup of gaseous reaction products inside the particle. The production of gaseous emission products from the interaction of high-energy fusion neutrons with coating materials and with the oxygen in the kernel contributes significantly to pressure vessel stresses in these coatings, whereas gaseous fission products alone dominate in conventional thermal reactors. The most stringent design for the three-cycle particle is identical in fuel loading to the reference fertile particle for an HTGR, which would constitute an ideal hybrid partner for the fusion reactor. Consideration is also given to coated-particle designs for the containment of the bred tritium used to fuel the D-T fusion reactor.     

Modeling of gap conductance for LWR fuel rods applied in the BISON code

ABSTRACT

A new gap conductance model is proposed in this study as a combination of Toptan’s model and the Ross-Stoute model. A variance-based sensitivity analysis is performed to understand how simulation results depend on all input parameters of the proposed model. Additionally, new modeling options (e.g. fill gas thermal conductivity, temperature jump distance, thermal accommodation coefficient, etc.) are added into the nuclear fuel performance code, BISON. The need for further investigation of the gap heat transfer between fuel and cladding in BISON motivated this study to evaluate its impact on the code’s predictions. New gap conductance modeling is proposed. A series of integral-effects validation tests is performed: (1) to demonstrate the impact of the proposed model on the code’s fuel temperature predictions at the beginning of life and through the reactor’s life; (2) to ensure that the proposed model is capable of accurately modeling gap heat transfer characteristics in real-world problems; and (3) to investigate the impact of the estimation of fission gas release on the fuel temperature predictions with the proposed model. The results indicate that the proposed gap conductance model improves BISON’s predictions.