模块式高温气冷堆三维中子动力学计算分析

模块式高温气冷堆是国际上公认的安全性好、发电效率高、用途广的先进堆型。本文研究开发了三维圆柱几何堆芯多群中子动力学改进准静态方法模拟计算程序。对给定的模块式高温气冷堆堆芯物理模型进行了模拟计算。初始状态下,该程序计算结果与中子扩散程序CITATION吻合很好。动态情况下,模拟了堆芯反应性、堆芯相对功率以及堆内r,z网格上各群点中子注量率三维分布随时间的变化,计算结果与理论分析一致。     

Safety analysis of reactivity insertion accidents in a heavy water nuclear research reactor core using coupled 3D neutron kinetics thermal-hydraulic system code technique

Nuclear power plant Safety analysis using coupled 3D neutron kinetics/thermal-hydraulic codes technique is increasingly used nowadays. Actually, the use of this technique allows getting less conservatism and more realistic simulations of the physical phenomena. The challenge today is oriented toward the application of this technique to the operating conditions of nuclear research reactors. In the current study, a three-Dimensional Neutron Kinetics and best estimate Thermal-Hydraulic model based upon the coupled PARCS/RELAP5 codes has been developed and applied for a heavy water research reactor. The objective is to perform safety analysis related to design accidents of this reactor types. In the current study two positive reactivity insertion transients are considered, SCRAM protected and self-limiting power excursion cases. The results of the steady state calculations were compared with results obtained from conventional diffusion codes, while transient calculations were assessed using the point kinetic model of the RELAP5 code. Through this study, the applicability and the suitability of using the coupled code technique with respect to the classical models are emphasized and discussed.     

Release kinetics of tritium generated in lithium-enriched Li2+xTiO3 by thermal neutron irradiation

Tritium release kinetics for Li_2+xTiO₃, the lithium-enriched Li₂TiO₃, was investigated by isochronal and isothermal annealing experiments. Tritium release by isochronal annealing showed that the dominant release stage was found at around 600 K. An additional release stage at lower temperature side was appeared with increasing excess lithium, which was attributed to the release of tritium trapped in Li₄TiO₄ structure. The dominant release stage was considered to a release of tritium trapped by irradiation defects. Isothermal annealing experiments indicated that tritium releases were controlled by diffusion process. The diffusion coefficient of Li_2.0TiO₃ was one order of magnitude as large as those of Li_2.2TiO₃ and Li_2.4TiO₃, although their activation energies were almost the same. These results showed that rate-determining step was the diffusion process of tritium in Li₂TiO₃ structure for Li_2+xTiO₃ and excess lithium would make diffusion coefficient smaller. Simulation of tritium-TDS spectra for Li_2.0TiO₃ has clarified that the TDS spectrum forLi_2.0TiO₃ can be demonstrated by using Arrhenius diffusion parameters obtained by isothermal annealing experiment in the present study.     

一维特征线法中子输运程序开发及验证

为探讨两维/一维综合法堆芯分析方法,本文基于特征线法研制了一维中子输运程序--PEACH-1D.不同于通常的平源近似特征线方法,PEACH-1D可对子区的中子源项作线性近似;程序运用指数函数插值表和渐近源外推技术来加速计算过程.相关数值结果表明,PEACH-1D具有很高的计算精度和效率,线性源近似的特征线法具备处理较粗网格的能力,值得推广.     

Calculation of thermal neutron self-shielding correction factors for aqueous bulk sample prompt gamma neutron activation analysis using the MCNP code

In this work thermal neutron self-shielding in aqueous bulk samples containing neutron absorbing materials is studied using bulk sample prompt gamma neutron activation analysis (BSPGNAA) with the MCNP code. The code was used to perform three dimensional simulations of a neutron source, neutron detector and sample of various material compositions. The MCNP model was validated against experimental measurements of the neutron flux performed using a BF₃ detector. Simulations were performed to predict thermal neutron self-shielding in aqueous bulk samples containing neutron absorbing solutes. In practice, the MCNP calculations are combined with experimental measurements of the relative thermal neutron flux over the sample’s surface, with respect to a reference water sample, to derive the thermal neutron self-shielding within the sample. The proposed methodology can be used for the determination of the elemental concentration of unknown aqueous samples by BSPGNAA where knowledge of the average thermal neutron flux within the sample volume is required.     

Uncertainty and sensitivity analyses of the Kozloduy pump trip test using coupled thermal–hydraulic 3D kinetics code

The modeling of complex transients in nuclear power plants (NPP) remains a challenging topic for best estimate three-dimensional coupled code computational tools. This technique is, nowadays, extensively used since it allows decreasing conservatism in the calculation models and performs more realistic simulation and more precise consideration of multidimensional effects under complex transients in NPPs. Therefore, large international activities are in progress aiming to assess the capabilities of coupled codes and the new frontiers for the nuclear technology that could be opened by this technique. In the current paper, a contribution to the assessment and validation of coupled code technique through the Kozloduy VVER100 pump trip test is performed. For this purpose, the coupled RELAP5/3.3-PARCS/2.6 code is used. The code results were assessed against experimental data. Deviations between code predictions and measurements are mainly due to the used models for evaluating and modeling of the Doppler feedback effect. Further investigations through the use of two “antagonist” uncertainty GRS and the CIAU methods, were considered in order to evaluate and quantify the origin of the observed discrepancies. It was revealed on one hand that relative error quantification discrepancies exist between the two approaches, and further enhancements for both methods are needed.     

3D neutron diffusion computational code based on GFEM with unstructured tetrahedron elements: A comparative study for linear and quadratic approximations

In the present study, the comparison between the results obtained from the linear and quadratic approximations of the Galerkin Finite Element Method (GFEM) for neutronic reactor core calculation was reported. The sensitivity analysis of the calculated neutron multiplication factor, neutron flux and power distributions in the reactor core vs. the number of the unstructured tetrahedron elements and order of the considered shape function was performed. The cost of the performed calculation using linear and quadratic approximation was compared through the calculation of the FOM. The neutronic core calculation was performed for both rectangular and hexagonal geometries. Both the criticality and fixed source calculations were done using the developed GFEM-3D computational code. An acceptable accuracy with low computational cost is the main advantage of applying the unstructured tetrahedron elements. The generated unstructured tetrahedron elements with Gambit software were used in the GFEM-3D computational code via a developed interface. The criticality calculation was benchmarked against the valid data for IAEA-3D and VVER-1000 benchmark problems. Also, the neutron fixed source calculation was validated through the comparison with the similar computational code. The results show that the accuracy of the calculation for the both linear and quadratic approximations improves vs. the number of elements. Quadratic approximation gives acceptable results for almost all considered number of the elements, while the results obtained from the linear approximation have good accuracy for only high number of the elements.     

Improvement of HYCA3D code and experimental verification in rectangular geometry

Concerns about the local hydrogen behavior in a nuclear power plant (NPP) containment during severe accidents have increased with the 10CFR50.34(f) regulation after TMI accident. Consequently, investigations on the local hydrogen behaviors under severe accident conditions were required. An analytical model named HYCA3D was developed at Seoul National University (SNU) to predict the thermodynamics and the three dimensional behavior of a hydrogen/steam mixture, within a subdivided containment volume following hydrogen generation during a severe accident in NPPs. In this study, the HYCA3D code was improved with a steam condensation and spray model, and verified with hydrogen mixing experiments executed in a SNU rectangular mixing facility. Helium was used to simulate hydrogen in both the calculations and the experiments. The calculation results show good agreement with the experimental data.     

A methodology for the coupling of RAMONA-3B neutron kinetics and TRAC-BF1 thermal-hydraulics

The initial objective of this project was to directly couple the RAMONA and TRAC codes running on different PCs. The idea was to use the best part of each one and eliminate some of their limitations and widen the applicability of these codes to simulate different BWR and system components. It was required to try to minimize the amount of changes to present code subroutines and calculation procedures. If possible, just substitute values obtained in the parallel code. Preliminary results indicated that using a CHAN component of TRAC to model thermal-hydraulic phenomena for each neutronic channel modeled in RAMONA is rather difficult. Large amounts of CPU time consumption are obtained and lots of PCs would make this solution difficult, besides considerable large transients are introduced by the differences in thermal-hydraulic results of these codes. The substitution of the thermal-hydraulics of RAMONA, by the TRAC channel calculations, is possible but simulation of a null transient on both codes must be planed and a gradual change must be controlled by an additional supervisory subroutine. An indirect coupling of these codes, it is therefore proposed, in order to eliminate most of these limitations. In this indirect coupling, a thermal-hydraulic model of the average tube in a bundle and the global channel cooling fluid dynamics is programmed for each neutronic channel while the global reactor vessel and core is modeled by TRAC with just four channels and four rings. Results are more reliable, coupling is simpler and faster simulations are possible.     

Analysis of the Phebus FPTO containment thermal hydraulics with the Jericho and Trio-VF codes

This paper presents the analysis of the thermal hydraulic behavior of the containment, during the Phebus FPTO test performed on 2 December 1993, with the Jericho code which deals with the thermal hydraulics of containment in the severe accident field. This code is part of Escadre which is the French system of codes in charge of predicting severe accidents in pressurized-water reactors. After summarizing the relevant Jericho code characteristics and the preliminary assessment work for the Phebus conditions, we briefly describe the REPF 502 test facility and report the thermal hydraulic FPTO experimental protocol. Then, comparisons of the experiment and Jericho calculations are analyzed. Because the Jericho code assumes a well-mixed atmosphere, some additional three-dimensional calculations have been carried out in order to gain further insight into the convection flow patterns and qualify the well-mixed atmosphere assumption in the Phebus containment.     

加速器热中子照相装置CCD芯片屏蔽的模拟计算

建立了研究加速器中子源热中子照相装置CCD芯片屏蔽效果的蒙特卡罗模拟方法,对γ与中子吸收剂量率的模拟计算结果与实验相符.进行了基于9Be(d,n)反应的热中子照相装置屏蔽系统的优化设计,在复杂几何条件下用蒙特卡罗模拟分别计算了CCD芯片在中子、γ混合场中的吸收剂量率和快中子注量率,对CCD相机在辐射场中安全性能进行了评估.     

Synergism of the method of characteristic, R-functions and diffusion solution for accurate representation of 3D neutron interactions in research reactors using the AGENT code system

This paper summarizes the theoretical and numerical aspects of the AGENT code methodology accurately applied for detailed three-dimensional (3D) multigroup steady-state modeling of neutron interactions in complex heterogeneous reactor domains. For the first time we show the fine-mesh neutron scalar flux distribution in Purdue research reactor (that was built over forty years ago).     

Elastic and viscoplastic pellets fragmentation modeling using an axisymmetrical 1D finite element code

During first rise to power in Power Water Reactor, fuel pellets crack because of thermal expansion. The phenomena of pellet cracking and fragments relocation have a major influence on rod behaviour and especially on the cladding behaviour in the case of pellet–cladding interaction.This article presents the modeling used to take into account the fragmented state of the pellet in the EDF fuel rod thermo-mechanical code, CYRANO3^®. The aim is to simulate more realistic stress and strain fields in the pellet.The investigated method consists in adding parameters in the 1D finite elements calculations in order to integrate the multi-dimensional fragmentation effects in the axisymmetrical 1D code CYRANO3^®. These parameters modify the material behaviour by describing the fuel as an anisotropic damaged material. The modeling accounts for the opening and closing of radial pellet cracks. It has been implemented in the code for elastic and viscoplastic fuel behaviours.     

Performance of the improved version of the Monte Carlo A3MCNP code in the design of cask shielding

Abstract

A³MCNP (automatic adjoint accelerated MCNP) is a revised version of the MCNP Monte Carlo code that automatically prepares variance reduction parameters for the CADIS (consistent adjoint driven importance sampling) methodology. Using a deterministic 'importance' (or adjoint) function, CADIS performs source and radiation particle transport biasing within the weight-window technique. The current version of A³MCNP uses the 3-D Sn transport TORT code to determine a 3-D importance function distribution. Based on simulation of several real-life problems, it is demonstrated that A³MCNP provides precise calculation results with a remarkably short computation time by using the proper and objective variance reduction parameters. However, since the first version of A³MCNP provided only a point source configuration option for large-scale shielding problems, such as spent fuel transport casks, a large amount of memory may be necessary to store enough points to properly represent the source. Hence, we have developed an improved version of A³MCNP (referred to as A³MCNPV) which has a volumetric source configuration option. This paper describes the successful use of A³MCNPV in solving the cask neutron and gamma-ray shielding problem.     

Study of delayed neutron decay curves from thermal neutron induced fission of U and Pu

The results of measurements of the delayed neutron decay curves obtained from the thermal neutron induced fission of ²³⁵U and ²³⁹Pu are presented. The data were obtained by the periodical irradiation method on the pulsed reactor IBR-2 during the time interval 5 to 730 msec following irradiation. A comparison of these newly measured decay curves with the curves calculated using several standard delayed neutron sets was then performed. Based on these measurements, a new 7-group delayed neutron model is proposed.     

Ion beam irradiation of nanostructures - A 3D Monte Carlo simulation code

A computer code for the simulation of ion beam irradiation of nanostructures has been developed. The code simulates the transport of energetic ions through matter by means of a Monte Carlo algorithm, similar to the often-used TRIM code (Ziegler et al. (1985) [1]). New effects occur compared to bulk, when irradiating nanostructures, which are of the same size as the ion range or the damage cascade. To account for these effects, the target in our code does not consist of layers like in TRIM but can be defined as an arbitrary 3-dimensional structure. This allows to obtain more accurate 3D distributions of implanted ions and implantation damage for nanostructures, which cannot be described by a stack of layers. We demonstrate the functionality of the code by comparing simulations with ion beam implantation into nanowires.     

Analysis and evaluation of thermal and resonance neutron activation data

A generalized least-squares technique has been applied to produce a consistent set of thermal and epithermal neutron activation data for the following 20 nuclides produced by neutron capture: ⁴⁶Sc, ⁵¹Ti, ⁵¹Cr, ⁵²V, ⁵⁶Mn, ⁵⁹Fe, ⁶⁰Co, ⁷⁵Se, ⁸⁶Rb, ⁹⁵Zr, ⁹⁷Zr, ¹²⁴Sb, ¹³¹Ba, ¹³⁴Cs, ¹⁴⁰La, ¹⁴¹Ce, ¹⁶⁰Tb, ¹⁸¹Hf, ¹⁸²Ta and ¹⁹⁸Au. The technique combines available information on nuclear data from the literature with measured activation data irradiated in nine reactor positions in Germany, the U.K. and Japan.Most of the solution nuclear data showed a distinct improvement, some by a large amount. These have been compared with the most recent evaluation.