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.     

ARARAT — A computer code for thermal design of cooling towers

A computer program is presented for thermal and hydraulic design of cooling towers. Options have been provided for the evaluation of cooling tower size and performance curves by applying a basic physical model of heat and mass transfer.The solution is conducted by multiple iteration, in which iteration loops are mutually inclusive. Both film and spray-filled cooling towers are considered with either induced or natural air circulation.Numerical solutions are presented to a number of natural draft cooling towers which serve present nuclear or conventional power plants.     

Senhor-II — A computer code for loss-of-coolant accidents of pressure tube type reactors

An ATR LOCA analysis code, SENHOR-II, was developed which evaluates the loss-of-coolant accident in a reactor primary loop composed of parallel pressure tubes and downcomers connecting a steam drum to a lower header. The reactor system is divided into reservoirs and channels. The reservoirs are assumed to be saturated and equilibrated. The channels are treated one-dimensionally and their flows are assumed quasi-steady. The reservoir effect of piping, the heating up of fuel rods, the thermal capacity of structures, and the effects of steam separators and water level in the steam drum are considered. Calculated results are compared with the experimental results of the blowdown test performed with the mock-up test loop in -arai Engineering Center of PNC, and the adequacy of the calculation model and formulae is confirmed.     

COMPBRN IIIe—a computer code for probabilistic fire risk analysis

The COMPBRN code has been used extensively to predict deterministically the time-to-damage of critical components in nuclear power plant fire risk analyses. Because there is a significant amount of uncertainties in the input parameters used in room fire simulations, the assessment of the damage time of the specified components must be performed probabilistically. This paper presents an updated version of the code, called COMPBRN IIIe, which emphasizes the importance of parameter uncertainty propagation by incorporating capabilities to provide probability distributions for component damage times. COMPBRN IIIe eliminates several errors from its previous versions and incorporates a user-friendly environment to assist users in preparing input files. With these improvements, the code can significantly reduce the time and effort required in the performance of a probabilistic fire risk assessment. A compartment fire simulation is also provided to demonstrate the application of the code.     

Diana — A fast and high capacity computer code for interchannel coolant mixing in rod arrays

A mathematical model and digital computer program are presented for the subchannel thermal and hydraulic analysis of sodium-cooled fast reactor fuel assemblies. The newly developed FORTRAN-IV computer code ‘DIANA’ is much more useful than many other subchannel mixing analysis codes, especially for large size fuel assemblies which contain more than about 80 subchannels, and for assemblies undergoing swelling and thermal bowing which cause deformed coolant flow ducts, because of high computing speed, reduction of necessary core memory and accurate solution by momentum conservation. Numerical solutions are presented for a deformed rod bundle which contains 179 subchannels.     

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.     

高温气冷堆包覆燃料颗粒破损机制及失效模型

高温气冷堆的燃料元件的基本构成单元是全陶瓷型的包覆燃料颗粒,其性能决定了高温气冷堆的安全性。除了传统的辐照实验检测外,建立理论模型对其研究具有重要的意义。本文主要介绍了TRI-SO型包覆燃料颗粒的结构及破损机制,以及国外现有的几个主要模型的基本假设,计算原理和特点,通过对比几个模型的优缺点,提出今后研究的方向。     

SIMPLE-2: A computer code for calculation of steady-state thermal behavior of rod bundles with flow sweeping

A computer code has been developed for use in making single-phase thermal hydraulic calculations in rod bundle arrays with flow sweeping due to spiral wraps as the predominant crossflow mixing effect. This code, called SIMPLE-2, makes the assumption that the axial pressure gradient is identical for each subchannel over a given axial increment, and is unique in that no empirical coefficients must be specified for its use. Results from this code have been favorably compared with experimental data for both uniform and highly nonuniform power distributions. Typical calculations for various bundle sizes applicable to the LMFBR program are also included.     

SWING-R—A computer code to calculate the time history of multi-mass systems with consideration of non-linearities or dry friction

The computer code SWING-R calculates the time history of the positions, velocities and accelerations of multi-mass systems with up to ten degrees of freedom. In additon to the well-known linear mechanical system $\text{[M]{}\text{x}\text{?}\text{} + [C]{}\text{x}\text{dot}\text{} + [K]{x}={fnof;(t)}}$, non-linearities or dry friction can be considered. The calculation of the frictional process includes the transition from static to kinetic friction and vice versa. The matrix exponential method was used for integrating the matrix differential equation of motion.     

A three-dimensional finite-element computer code for the analysis of complex structures

General computational procedures are described with emphasis on FORTRAN based ‘dynamic’ storage allocation and adaptability of specialized input routines for the analysis of complex structures, including pump casings and pipe tees and elbows. A brief discussion is given on the algorithms used in this code for the formation and solution of banded linear equatons. The basic finite element building blocks, including isoparametric tetrahedra, pentahedra, and hexahedra are described. The results of a complex structural analysis, along with sample computer output, is presented.     

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.     

DRAGON&DONJON程序在MSR中堆芯燃耗计算的适用性

DRAGONDONJON组件-堆芯"两步法"程序通过合理简化,理论可适用于任何堆芯与工况。使用蒙特卡罗方法 RMC(Reactor Monte Carlo code)、MCNP(Monte Carlo Neutron Particle transport code)程序验证DRADON程序是否能够承担快/热谱型熔盐堆(Molten Salt Reactor,MSR)焚烧TRU、Th U燃料燃耗计算。选出熔盐增殖堆(Molten Salt Breeder Reactor,MSBR)与熔盐锕系元素再循环和嬗变堆(Molten Salt Advanced Reactor Transmuter,MOSART)堆型进行比较,同时分别利用RMC程序验证DRAGON程序组件燃耗计算的准确性,利用MCNP程序验证DRAGON程序组件均匀化方法以及DONJON程序截面调用和程序全堆扩散的准确性。结果表明,组件燃耗计算中,TRU和Th U燃料满足燃耗计算要求;堆芯临界计算中,快/热谱堆芯计算误差均小于0.001。证明DRADON程序可以胜任快、热谱型MSR焚烧TRU、Th U燃料的物理计算任务。     

A comparison of the praise code and the paris code for the evaluation of the failure probability of crack-containing components

Two computer codes developed for the calculation of failure probabilities of crack-containing structures are compared with each other. The basic fracture mechanical, statistical, and numerical models used in the codes are described with special emphasis on probabilistic leak-before-break analysis. Sample problems taken from nuclear applications show that very small failure probabilities can be calculated with sufficient numerical accuracy.     

Subassembly experiments and a computer code to analyze the dynamic core deformation during local failure propagation

Safety investigations for LMFBRs have to consider local failure situations in one fuel element which may escalate to a hypothetical CDA. Such initiating events could produce high pressure pulses in a single subassembly which may expand and rupture the wrapper as well as load adjacent elements impulsively. The associated nonlinear dynamic core deformation problem is treated in this paper. In particular the multirow structural dynamics code CØRE-1 and underlying mechanical models are described. Each subassembly is simulated by an equivalent system of point masses and nonlinear coupling springs. The motion of the coolant layer between the elements is treated by an incompressible, non-stationary frictional flow model. In order to obtain realistic code input four types of static single subassembly deformation experiments are described which provided strongly nonlinear load deformation characteristics. Furthermore the transient pressure distribution within the core is obtained from a full scale explosion test. Finally code application is demonstrated and results are given of a transient analysis of the SNR 300 core.     

A Monte Carlo code for calculation of depolarization of beta-rays in materials

A Monte Carlo code is used to calculate the depolarization of electrons emitted by β-sources. A comparison of our results which the theory of Mühlschlegel yielded the same characteristic dependences of the depolarization on energy, nuclear charge and source thickness with our values being somewhat higher. This is clearly understood by the fact that our Monte Carlo calculations are not restricted to small angle scatterings and other approximations used by Mühlschlegel. The only limitation to our code results from depolarization contributions of inelastic processes which are not yet included.     

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.     

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.     

MIPAC, a computer code for fuel performance analysis by the finite element method

An axisymmetric finite element computer code named MIPAC has been developed for analysis of the mechanical interaction behaviour between a fuel pellet and cladding. This computer code can deal with elastoplasticity of the pellet and cladding materials, creep effects for the both materials, pellet-cladding and pellet-pellet contact problems, hot pressing effect of the fuel pellet, fuel pellet cracking, and the cracked pellet's stiffness. A cyclical boundary condition is introduced to deal with one pellet length instead of the full-size fuel rod. The contact problems are solved without a fictitious contact element. In the fuel pellet cracking model the crack opening and closing behaviour under arbitrary power changes can be treated by introducing five kinds of crack modes. Mismatch of irregular crack surfaces is taken into account in the evaluation of the cracked pellet's stiffness. Finally, calculated results are compared with experimental data to show validity of the computer code.     

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.