Improvements to COMETHE III-J fuel rod modeling code

Current improvements to the COMETHE fuel performance code focus on pellet-clad axial interaction and Zircaloy cladding failure predictions. Slipping and sticking between pellets and clad as well as trapped stack are evaluated. The main conclusions are that slipping with friction concerns only local effects while axial PCMI is primarily dependent on pellet expansion with a strong ‘strain biaxiality’ effect dictated by the dishing. The notion of locking prior to radial PCMI is also introduced, which explains experimental features not previously understood. Benckmarking of the version of COMETHE against ramp tests has been initiated and will enable assessment of the code capability in Zircaloy clad failure predictions.     

Benchmarking a diffusion code for a cylindrical-core reactor

The distorted-buckling method, proposed by us previously, allows the benchmarking of a diffusion code by comparing it with an analytic model in either 2 or 3 dimensions. Here, the method is applied to the case of a cylindrical TRIGA-type reactor to compare the fluxes predicted by an analytic model of the core and reflector, to those predicted by the code CITATION. The match is everywhere excellent. ©     

A COMETHE version with transient capability

A version of the COMETHE code is under development to simulate transient situations. This paper focuses on some aspects of the transient heat transfer models. Initially the coupling between transient heat transfer and other thermomechanical models is discussed. An estimation of the thermal characteristic times shows that the cladding temperatures are often in quasi-steady state. In order to reduce the computing time, calculations are therefore switched from a transient to a quasi-static numerical procedure as soon as such a quasi-equilibrium is detected. The temperature calculation is performed by use of the Lebon-Lambermont restricted variational principle, with piecewise polynoms as trial functions. The method has been checked by comparison with some exact results and yields good agreement for transient as well as for quasi-static situations. This method therefore provides a valuable tool for the simulation of the transient behaviour of nuclear reactor fuel rods.     

General description and organization of COMETHE III-J

The COMETHE III-J code is a computer programme describing the thermal and mechanical behaviour of integral fuel rods. Its range of application is described and the physical phenomena which are modelled in the code are briefly reviewed. Some particular features of the calculation method are presented; they reduce the computation time, but allow simultaneously to take sophisticated models into account. Use of the code is then considered: the concept of recommended values and the material properties libraries make this very easy. Surrounding programmes allow a quick and economic interpretation of the output. The result is that COMETHE III-J is best fit for use by the utility industry: calibrated on PIE data, it can be used for review and checking of fuel design and for increasing plant load factor.     

Fuel behaviour models incorporated in COMETHE III-J

Some of the fuel behaviour models incorporated in the COMETHE III-J computer code are reviewed. The fuel swelling model is first described and each of its components is discussud. The fuel restructuring calculation takes equiaxed or columnar grain growth into account. Grain growth and gaseous swelling are coupled in a realistic way to the gas release model. One of the milestones of the COMETHE III-J code is the crack pattern calculation by means of the “pivot” concept. This model couples cracking with thermal expansion and three-dimensional plasticity effects. The effects of radial and axial restraints, coupled with fuel swelling or densification resulting from columnar grain growth, account for fuel relo cation and dish, crack or central hole filling. The power cycling effects are therefore naturally modelled and no additional relocation is required to explain the gap closure.     

NGFM程序对堆芯物理在线仿真结果的校验

将格林函数节块程序NGFM/TNGFM移植到windows系统下运行。对二维、三维基准题进行了验算,并对秦山二期反应堆仿真结果进行了校验,验证了将该程序应用到反应堆物理在线仿真系统的可行性。     

Benchmarking analysis of water hammer effects using RELAP5 code and development of RBMK-1500 reactor main circulation circuit model

Correct prediction of water hammer transients is of paramount importance for the safe operation of the plant. Therefore, verification of computer codes capability to simulate water hammer type transients is a very important issue at performing safety analyses for nuclear power plants. Verification of RELAP5/MOD3.3 code capability to simulate water hammer type transients employing the experimental investigations is presented. Experience gained from benchmarking analyses has been used at development of the detail RELAP5 code RBMK-1500 model for simulation of water hammer effects in reactor main circulation circuit. In RBMK-type reactors the water hammers can occur in cases of rapid check valve operation. The performed analysis using RELAP5 code RBMK-1500 model has shown that in general the maximum values of the pressure pulses due to water hammer do not exceed the permissible loads on the pipelines.     

BETA experiments in verification of the Wechsl code: experimental results on the melt-concrete interaction

The BETA experiments are conducted to investigate the melt-concrete interaction in a large-scale melt facility using internally heated simulated core melts. The experimental findings are extrapolated to reactor accident conditions by means of computer codes verified experimentally.The experiments cover a wide range of temperatures and power rates typical of accident conditions. In high-temperature melts, fast downward erosion determines the cavity shape and the very high downward heat transfer causes the temperature of the melt to drop rapidly, even with high internal heating. Crust formation at the interface between the melt and the concrete during the low-temperature interaction allows the gases evolved by the concrete to percolate through the melt, thus establishing an effective gas driven mode of heat transfer. Measurements of gases and aerosols are reported and discussed for silicate and limestone types of concrete.     

SORA原型装置临界实验的基准化分析

临界装置实验数据的基准化分析是充实临界安全实验基准数据手册的必要条件。本文首先介绍了SORA(Sorgente Rapida Reactor)原型装置的活性区结构组成、临界实验等情况,然后对15个典型的临界实验数据进行了不确定度分析,其实验keff不确定度在0.002 3~0.002 7范围内,并进一步分析了对实验装置进行模型化处理的偏倚及其不确定度,最后得到了SORA原型装置基准模型的keff值及其不确定度。SORA模型的数值计算结果与实验基准化分析的keff值相比略低,其最大相对偏差小于1%。研究结果满足临界安全实验基准数据手册收录的要求。     

燃耗数据库基准检验方法研究

燃耗数据库基准检验方法对于研制高准确度的燃耗数据库至关重要。本文以TAKAHAMA 3压水堆辐照后检验实验中SF95样品的建模为例,研究了建模要素对燃耗计算的影响,确定了燃耗实验建模的方法,开展了燃耗信用制研究感兴趣的锕系和裂变产物核素积存量计算值与实验值的比对。比对结果显示,主锕系核素计算偏差小于2%,大部分次锕系核素偏差小于10%,大部分重要裂变产物核素偏差小于5%。本文还对125Sb积存量随燃耗深度变化规律进行了理论分析,确认了破坏性放化实验测量结果存在缺陷,并进一步获得了125Sb积存量的修正值,使计算偏差从接近170%下降到20%以内。本次研究表明,燃耗数据库基准检验研究不仅需发展适当的燃耗实验建模方法,还需对实验数据进行适当的评价。     

Supposed existence of Np4+ in a genuine dissolver solution from the results of extraction simulation by PARC-L code

With a genuine spent fuel solution (a dissolver solution), a laboratory-scale reprocessing experiment of an extraction–separation process was performed using mixer-settlers as extractors. In the experiment, n-butyraldehyde was utilized as a reducing reagent of Np^(VI)O₂²⁺ to Np^(V)O₂⁺ for the purpose to distinguish Np^(VI)O₂²⁺ from Np⁴⁺. From the Np concentration in the aqueous phase, Np would be extracted from the dissolver solution together with U and Pu. The scrutiny of Np behavior was performed utilizing 66 cases of calculation results by a Japan Atomic Energy Agency open extraction simulation code, the Program for Advanced Extraction with Radiation Effect Calculation–Lightened version. From the scrutiny, the authors found that the calculation result with 60% of Np⁴⁺ in the dissolver solution represented the best experimental extraction–separation behavior of Np. Therefore, it was supposed that the dissolver solution contained sufficient proportion of Np⁴⁺ to affect the extraction–separation behavior of Np.     

JENDL-4.0 Benchmarking for Fission Reactor Applications

Benchmark testing for the newly developed Japanese evaluated nuclear data library JENDL-4.0 is carried out by using a huge amount of integral data. Benchmark calculations are performed with a continuous-energy Monte Carlo code and with the deterministic procedure, which has been developed for fast reactor analyses in Japan. Through the present benchmark testing using a wide range of benchmark data, significant improvement in the performance of JENDL-4.0 for fission reactor applications is clearly demonstrated in comparison with the former library JENDL-3.3. Much more accurate and reliable prediction for neutronic parameters for both thermal and fast reactors becomes possible by using the library JENDL-4.0.     

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.     

New features of the Monte-Carlo code MOCADI

MOCADI, the Monte-Carlo code for tracking of ions in ion-optical systems with non-Liouvillian elements, has been extended. Accurate atomic and nuclear interactions are taken into account when ions penetrate gaseous and solid matter placed within the ion-optical system. The new features of MOCADI are described in this article with practical examples which demonstrate the new possibilities, such as new event-generators for targets and spontaneous nuclear decay, the option of atomic-charge state fluctuation in matter, loops for multi-turn ion-optical systems and a graphical user interface for easier operating and control of the program. Experiments for investigation of nuclear structure and reactions with ions circulating in a storage ring can now be ideally studied with MOCADI.     

Experimental validation of the thermal-hydraulic code SACATRI

A sub-channel analysis steady state thermal-hydraulic code (SACATRI) was developed for the Moroccan TRIGA MARK II research reactor. The main objective of the thermal-hydraulic study of the whole reactor core is to evaluate the main safety parameters of the reactor core, and to ensure that they are within the safety limits for any operating conditions. The thermal-hydraulic model used in SACATRI is based on four partial differential equations that describe the conservation of mass, energy and momentum. In order to assess the thermal-hydraulic mathematical model of SACATRI, the present paper focuses on the quantification of the physical model accuracy to judge if the code is capable to represent the thermal-hydraulic behaviour of the reactor core with sufficient accuracy. The methodology adopted is based on the comparison between responses from SACATRI computational model and experimentally measured responses performed on the IPR-R1 TRIGA research reactor. The results showed good agreement between SACATRI predictions and the experimental measurements where the discrepancies observed (simulation-experiment) are less than 6%.     

基于六角形节块法程序的三维时空动力学程序开发

处理三维中子时空动力学问题时,在时间的处理上采用了改进型准静态方法,将总的中子通量密度分解成由时间相关的幅度函数和由时间空间能量相关的形状函数,幅度函数通过三阶多项式插值方法求解,而形状函数则通过节块法程序NAS程序系统的稳态扩散或者输运模块求解,以此为基础,开发了三维时空动力学程序NAS-K.并给出了与两个国际基准例题(无反馈弹棒事故和简单的绝热多普勒反馈弹棒事故)计算结果的比较,结果符合很好,初步验证了本程序的正确性.     

The verification basis of the PM-ALPHA code

An overall verification approach for the PM-ALPHA code is presented and implemented. The approach consists of a stepwise testing procedure focused principally on the multifield aspects of the premixing phenomenon. Breakup is treated empirically, but it is shown that, through reasonable choices of the breakup parameters, consistent interpretations of existing integral premixing experiments can be obtained. The present capability is deemed adequate for bounding energetics evaluations.     

Validation of the MCNP5 computer code

The MCNP5 computer code with the ENDF/B6 neutron data library is validated for problems which are of current importance at the Russian Federal Nuclear Center — All-Russia Research Institute of Technical Physics. Comparative calculations performed with the MCNP5 code and its preceding version MCNP4c are identical within the limits of computational error. This confirms that the MCNP5 code can be used instead of the previous versions. __________ Translated from Atomnaya énergiya, Vol. 101, No. 2, pp. 112–116, August, 2006.