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.     

URANUS — A computer programme for the thermal and mechanical analysis of the fuel rods in a nuclear reactor

The URANUS code, a digital computer programme for the thermal and mechanical analysis of integral fuel rods, is described. With this code the fuel rods found in the majority of power reactors can be analyzed. URANUS is built around a quasi two-dimensional analysis of fuel and cladding. The mechanical analysis can accommodate seven components of strain: elastic, time-independent plastic, creep and thermal strains, as well as strains due to swelling, cracking and densification. The heat generation and temperature distribution, cladding/fuel gap closure, pellet cracking and crack healing, fission-gas release, corrosion, O/M-distribution and plutonium redistribution are modelled. Geometric non-linearities (large displacements) are included; steady state or transient loading (pressure, temperature) is possible. In this paper special attention is paid to a theory for determining crack structures. The present status of the URANUS computer programme and a critical comparison with other fuel rod codes as well as sample analyses are given.     

FARST — A computer code for the evaluation of FBR fuel rod behavior under steady-state/transient conditions

FARST, a computer code for the evaluation of fuel rod thermal and mechanical behavior under steady-state/transient conditions has been developed. The code characteristics are summarized as follows:

1. (i) FARST evaluates the fuel rod behavior under the transient conditions. The code analyzes thermal and mechanical phenomena within a fuel rod, taking into account the temperature change in coolant surrounding the fuel rod.

2. (ii) Permanent strains such as plastic, creep and swelling strains as well as thermoelastic deformations can be analyzed by using the strain increment method.

3. (iii) Axial force and contact pressure which act on the fuel stack and cladding are analyzed based on the stick/slip conditions.

4. (iv) FARST used a pellet swelling model which depends on the contact pressure between pellet and cladding, and an empirical pellet relocation model, designated as “jump relocation model”.

The code was successfully applied to analyses of the fuel rod irradiation data from pulse reactor for nuclear safety research in Cadarache (CABRI) and pulse reactor for nuclear safety research in Japan Atomic Energy Research Institute (NSRR).The code was further applied to stress analysis of a 1000 MW class large FBR plant fuel rod during transient conditions. The steady-state model which was used so far gave the conservative results for cladding stress during overpower transient, but underestimated the results for cladding stress during a rapid temperature decrease of coolant.     

FEMAXI-IV: A computer code for the analysis of fuel rod behavior under transient conditions

A fuel rod behavior code FEMAXI-IV, presently under development, is an improved version of the FEMAXI-III code for the analysis of fuel rod behavior under transient conditions. To apply the FEMAXI-III code to transient conditions, the following additional models have been incorporated into the FEMAXI-III code: transient heat transfer model: axial gas mixing model; diffusion-type fission gas release model. This paper summarizes the above additional models, and the comparison of the FEMAXI-IV calculations with the experimental data.     

SAMURA—A fast computer code for reactor fuel pin performance modelling and analysis

A very fast integral numerical computer code for the modelling of transient and steady-state thermal and mechanical behaviour of Zircaloy-clad UO₂ fuel pins in water reactors has been developed. The computational technique which determines the stress and deformation state of the fuel pin is based upon an extremely efficient finite difference scheme, i.e. the non-linear terms in the constitutive equations which produce a non-linear system of equations have been linearised using a Taylor expansion technique coupled with a very sophisticated error minimization algorithm and then solved with great accuracy. An improved numerical method has also been developed for the fast and efficient solution of the transient heat conduction equation. In this way a very stable and economical one-dimensional code (with appropriate provisions made for its conversion to a quasi two-dimensional code) has been obtained. The physical processes included are thermo-elastic deformation, thermal and irradiation creep, plasticity, fission gas swelling and release, formation of cracks in the fuel, hot pressing, densification, pore migration and dish or central void filling. Here the mathematical basis of SAMURA is presented along with some preliminary calculations and benchmarkings. It is concluded that SAMURA is quite fast indeed, converges to accurate results and within the margins of the error criterion chosen has very reasonable computer demands. It is also stable under all conditions tested.     

IAMBUS-1 — A digital computer code for the design, in-pile performance prediction and post-irradiation analysis of arbitrary fuel rods: Part I: Theory and modelling

IAMBUS-1^*, a digital computer code for the thermal and mechanical design, in-pile performance prediction and postirradiation analysis of arbitrary fuel rods, will be presented in two parts. Part I describes the theory and modelling and in Part II (to be published in a subsequent issue of Nuclear Engineering and Design) material behaviour will be discussed on a quantitative basis and some numerical results illustrating typical and diverse IAMBUS usage will be analysed.The multi-zone code IAMBUS is built around a sound but flexible mechanical analysis of fuel and cladding. A state of generalized plane strain approximates the cladding; the fuel is modelled by a state of plane stress, a state of generalized plane strain, or a combination of these two well-known stress—strain configurations depending on the macroscopic structure of the fuel prevailing. It is thus possible to follow closely the deformation of the fuel and cladding as these are subjected to varying (in part mutual) loads, beginning with a relatively loose, somewhat random assemblage of minute fuel fragments at BOL and progressing to a quasi compact continuum of fuel at EOL.Cladding analysis includes routines for plasticity, creep and swelling due to void nucleation and growth; in the fuel restructuring, plasticity, creep, swelling due to solid and gaseous fission products, fission-gas release and internal pressure build-up are modelled. Routines for friction and heat transfer between fuel and cladding are also incorporated. No strict temperature-dependent boundary is drawn between typically elastic and plastic behaviour, the multi-zone nature of the code models the gradual transition between these two types of material behaviour observed in practice with increasing temperature.Great care has been exercised in choosing numerical methods, since the most sophisticated/realistic modelling is of limited value if the effort expended in reaching a numerical solution becomes exorbitant. Multi-zone modelling lends itself readily to the method of finite differences. The finite difference equations are solved via the method of secants, modified to guarantee convergence for all IAMBUS functions in a feasible amount of computer time.     

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.     

高燃耗下M5合金包壳燃料棒稳态和瞬态性能分析

修改并验证了分析程序FEMAXI-IVM,增加了程序的适用范围。对采用M5合金包壳的FA300-4高性能燃料组件中的燃料棒在稳态和瞬态运行工况下的燃料性能进行了分析。结果表明,此种燃料棒在稳态和瞬态工况下都能保持其完整性,能保证反应堆的安全运行。     

SAP — A structural analysis program for linear systems

The current version of the computer program SAP for the static and dynamic analysis of linear structural systems is described. The analysis capabilities of the program, the finite element library, the numerical techniques used, the logical construction of the program and storage allocations are discussed. The main advantages of the program as a general purpose code become apparent. Results of analyses as comparisons with other existing solutions are given, and running times which demonstrate the efficiency of the program are included.     

Effects gap heat transfer on LWR fuel behavior during an RIA transient: In-pile experimental results with helium and xenon fillet rods

Gap heat transfer characteristics and their effects on LWR fuel behavior during an RIA have been studied through the in-pile experiment with UO₂ pellet fuel rods. The report describes the experimental results obtained in the NSRR tests in which PWR type test fuel rods of helium and xenon filled as the gap gas have been irradiated in the pulse reactor, NSRR, to simulate the prompt heat up of RIAs. The relation between the cladding temperature history and the gap heat transfer conditions, and the effects of gap gas composition on fuel behavior and on the fuel failure threshold are discussed based on the in-pile experimental data.     

NONSAP — A nonlinear structural analysis program

The current version of the computer program NONSAP for linear and nonlinear, static and dynamic finite element analysis is presented. The solution capabilities, the numerical techniques used, the finite element library, the logical construction of the program and storage allocations are discussed. The solutions of some sample problems considered during the development of the program are presented.     

BEAF - a computer program for analysis of light water reactor fuel rod behavior

BEAF - a computer program for analysis of light water reactor fuel rod behavior was developed. The BEAF code, which is appropriate for on-line prediction of fuel rod behavior, can analyze fuel rod thermal and mechanical behaviors using the axisymmetric, plane strain approximation and finite difference method to realize a fast running time.In the mechanical analysis, a new cracked pellet compliance model is introduced, in which pellet cracking and crack healing, pellet initial relocation, modified elastic moduli of a cracked fuel pellet, and stress dependent hot pressing are considered. Adding to those capabilities, fission gas flow and diffusion in the fuel-clad gap are analyzed to take into account the slow fission gas dilution effect on the gap conductance during power ramp.     

Mathematical modeling and analysis of fuel rods

This paper presents a comprehensive method of analysis for nuclear fuel rods by first developing appropriate constitutive relations for the fuel and the clad materials that take into account elastic-plastic behavior, primary as well as secondary creep, mechanical densification (hot pressing), fuel swelling, fuel cracking and healing, and fuel redistribution. The geometric problem is treated on the basis of a two-dimensional, axisymmetric, and plane model within the framework of the finite element method. General power histories of the transient, cyclic, and steady-state types are considered. Analysis results uncovering the importance of certain deformation processes, which heretofore have been neglected, and comparisons with experiments are given.     

FIPMIGR — A computer program for studying migration of fission products in a temperature gradient

FIPMIGR is a computer program for studying migration of fission products in a fuel pin. Migration in a temperature gradient and in a concentration gradient is considered. The geometry is cylindrical with migration only in the radial direction.As an example the diffusion of Ba is calculated and compared with experimental results. The migration of Ba is well described using the diffusion constant for Ba in BaO and a heat of transport of −100 kJ mol⁻¹. The great sensitivity of the theoretical prediction to temperature is clearly demonstrated. Both theory and experiments show that there is a temperature or power above which migration becomes clearly visible. The critical temperature is about 1700 K and the power level in the S176 experiments [2] was then about 40 kW m⁻¹.     

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.     

A mathematical model for investigation of the rewetting of fuel rods

The computer model ZETHYF simulating the reflood phase after a loss-of-coolant accident with emphasis on the investigation of coolant channel is described. The thermal behaviour of the fuel rod is modeled based on a detailed representation of the heat transfer mechanisms and a moving mesh around the quench front. The flow conditions in the coolant channel are simulated as a one-dimensional transient one- or two-phase flow.     

Validation of frey for the safety analysis of LWR fuel using transient fuel rod experiments

The analysis and comparison of severe light water reactor transient experiments are presented from the FREY verification and validation effort. The purpose of this study was to validate the predictive capabilities of the code for severe transient analysis. The FREY code, developed under the sponsorship of the Electric Power Research Institute, uses a two-dimensional finite-element computational method for the thermomechanical analysis of LWR fuel rods under steady state and transient conditions. A total of 10 test fuel rods from experimental programs conducted in both the Power Burst Facility and the Transient Reactor Test Facility have been used in this study. The fuel rods were selected from the following test programs: Power Coolant Mismatch Tests, PCM-2 and PCM-4: Reactivity Initiated Accident Test, RIA 1–2; Loss-of-Coolant Accident Test, LOC-3; First Fuel Rod Failure Test, FRF-1; and Irradiation Effects Test, IE-3. The test programs used in this study cover a large range of code applications for severe transient analysis. The methods used to model the fuel, cladding, and coolant geometry are discussed in addition to experimental data comparisons. The results of the PCM-2, RIA 1–2, and FRF-1 analyses are presented to highlight the full two-dimensional modeling capabilities of FREY and to compare the thermal and mechanical measurements with FREY's prediction. The comparisons show good general agreement, with a tendency for FREY to overpredict the peak cladding surface temperature for a few cases where strong three-dimensional effects have been identified.