压水堆平衡堆芯钍铀燃料循环初步研究

建立WIMSD5-SN2-CYCLE3D和CASMO3-CYCLE3D物理分析系统作为钍铀燃料循环研究工具.以大亚湾第1机组压水堆为参考堆型,不改变反应堆栅元、组件和堆芯的结构与几何尺寸,设计出含36根钍棒、4.2#5U富集度的新型含钍组件,并对含钍组件和3.2%富集度的铀组件进行中子学计算和分析.模拟并分析了大亚湾压水堆12个月换料从初始循环到铀钚平衡循环的换料过程.再从平衡铀堆芯出发,逐步加入含钍组件代替铀组件,对铀钚平衡循环到钍铀平衡循环的换料过程进行了模拟与分析.计算结果表明:钍铀平衡循环比铀钚平衡循环每天节省裂变核素质量约18.4%,并减少了长寿命放射性核废料的产生.不利因素是使得循环长度减少90EFPD,缩短了换料周期,增加运行费用,并给燃料管理、安全控制以及乏燃料的处理带来困难.建议提高组件的235U富集度,在压水堆上进行钍利用研究.     

Optimization of core reload pattern for PARR-1 using evolutionary techniques

A procedure for searching optimum loading pattern to ensure safe and efficient reactor utilization was established for Pakistan Research Reactor-1 (PARR-1). A 6 × 5 core of PARR-1 consisting of 24 standard fuel elements, 5 control fuel elements and one flux trap was used in the study. The optimum loading schemes for maximum cycle length and maximum thermal flux in the flux trap were recommended using the developed methodology.     

An integrated expert system for optimum in core fuel management

An integrated expert system called Efficient and Safe Optimum In-core Fuel Management (ESOIFM Package) has been constructed to achieve an optimum in core fuel management and automate the process of data analysis. The Package combines the constructed mathematical models with the adopted artificial intelligence techniques. The paper gives a brief discussion of the ESOIFM Package modules, inputs and outputs. The Package was applied on the DALAT Nuclear Research Reactor (0.5 MW). Moreover, the data of DNRR have been used as a case study for testing and evaluation of ESOIFM Package. This paper shows the comparison between the ESOIFM Package burn-up results, the DNRR experimental burn-up data, and other DNRR Codes burn-up results. The results showed good agreement.     

A two-time step algorithm for seismic analysis of a PWR core

The present analysis considers the colliding vibration of a row of fuel assemblies under seismic excitation from the core plates. The differential equation model of each assembly is integrated separately, the impact forces due to the shocks between the spacer grids, being considered as external forces and reevaluated at each time step. The linear behaviour of the fuel assemblies for low amplitude vibrations allows a modal superposition.Two different time steps are considered in the analysis; the first one, related to the seismic excitation, applies to the fuel assemblies not involved in shocks; the second one, a submultiple of the first and related to the impact characteristics, is used for those assemblies involved in shocks.The resulting computer program, CLASH, is relatively economical to use; this allows such analysis becoming part of the standard design procedures.     

Thermal hydraulic design of a hydride-fueled inverted PWR core

An inverted PWR core design utilizing U(45%, w/o)ZrH_1.6 fuel (here referred to as U-ZrH_1.6) is proposed and its thermal hydraulic performance is compared to that of a standard rod bundle core design also fueled with U-ZrH_1.6. The inverted design features circular cooling channels surrounded by prisms of fuel. Hence the relative position of coolant and fuel is inverted with respect to the standard rod bundle design. Inverted core designs with and without twisted tape inserts, used to enhance critical heat flux, were analyzed. It was found that higher power and longer cycle length can be concurrently achieved by the inverted core with twisted tape relative to the optimal standard core, provided that higher core pressure drop can be accommodated. The optimal power of the inverted design with twisted tape is 6869 MW_t, which is 135% of the optimally powered standard design (5080 MW_t—determined herein). Uncertainties in this design regarding fuel and clad dimensions needed to accommodate mechanical loads and fuel swelling are presented. If mechanical and neutronic feasibility of these designs can be confirmed, these thermal assessments imply significant economic advantages for inverted core designs.     

Adaptive control of a PWR core power using neural networks

Reactor power control is important because of safety concerns and the call for regular and appropriate operation of nuclear power plants. It seems that the load-follow operation of these plants will be unavoidable in the future. Discrepancies between the real plant and the model used in controller design for load-follow operation encourage one to use auto-tuning and/or adaptive techniques. Neural network technology shows great promise for addressing many problems in non-model-based adaptive control methods. Also, there has been a great attention to inverse control especially in the neural and fuzzy control context. Fortunately, online adaptation eliminates some limitations of inverse control and its shortcomings for real world applications. We use a neural adaptive inverse controller to control the power of a PWR reactor. The stability of the system and convergence of the controller parameters are guaranteed during online adaptation phase provided the controller is near the plant’s real inverse after offline training period. The performance of the controller is verified using nonlinear simulations in diverse operating conditions.     

Reactivity feedback calculation of a conceptual TRISO fueled compact PWR core

Reactivity feedback coefficients have been calculated for a compact sized PWR core that utilizes carbon coated micro fuel particles instead of standard cylindrical fuel pellets with an inventive composition. A small amount of Pu-240 with 5 w/o has also been added in tristructural-isotropic (TRISO) fuel in place of U-238 for the reduction of excess reactivity. The values of fuel, moderator and void reactivity coefficients have been calculated at the middle of fuel cycle. All the reactivity coefficients were found negative which meet the design safety criteria. It was also observed that all reactivity feedback coefficients are interlinked and their effects are pronounced when coupled together.     

Gd burnable poison system for reactivity control of the first cycle of a PWR

The reactivity control of a PWR core may be performed by a system of burnable poison (BP) rods. In such a case, the soluble B system may be eliminated and the BP rods will be responsible for the excess reactivity provided for fuel depletion and fission products accumulation. A strong negative moderator temperature coefficient is a desirable safety feature, inherent to a poison-free moderator. The design objective of a PWR core controlled completely by a system of BP rods is achieved by utilization of Gd as the poison material and annular geometry of a BP rod. The proposed concept is tested as a retrofittable option for the current generation, as well as new PWR plants. A plausible incore fuel-management scheme is demonstrated, with planar power distribution, close to an acceptable range. The fuel-cycle penalty due to the residual poison content at EOC is relatively small.     

Emotional learning based intelligent controller for a PWR nuclear reactor core during load following operation

The design and evaluation of a novel approach to reactor core power control based on emotional learning is described. The controller includes a neuro-fuzzy system with power error and its derivative as inputs. A fuzzy critic evaluates the present situation, and provides the emotional signal (stress). The controller modifies its characteristics so that the critic’s stress is reduced. Simulation results show that the controller has good convergence and performance robustness characteristics over a wide range of operational parameters.     

Development of enriched Gd-155 and Gd-157 burnable poison designs for a PWR core

In this study, a genetic algorithm developed by the authors was applied to design the optimal enriched Gd-155 and Gd-157 burnable poisons in a reference PWR TMI-1 core. The CASMO-4/TABLES/SIMULATE-3 package calculated the neutronic performance of the enriched UO₂/Gd₂O₃ fuel pin configurations. These configurations included different fractions of neutron absorbing isotopes Gd-155 and Gd-157, and 100 w/o enriched Gd-155 designs. Fuel cost analysis was performed to evaluate the economical benefits of these optimized enriched gadolinium designs. The break-even point for unit Gd-155 enrichment cost was determined to be around ∼$30/gram-Gd-155 with current unit cost scenario. The projected savings were 3.13% in gross and 2.08% in net compared to total fuel cycle cost of a reference TMI-1 core loading, if all of the 68 feed assemblies would be replaced with the optimized designs.     

Design studies of a typical PWR core using advanced computational tools and techniques

This study deals with the design and development of calculational techniques and evaluation of key neutronic parameters of a typical PWR core having a total reactor power of 2652 MW_t (890 MW_e). The PWR core consists of 157 fuel assemblies containing a total of ∼72 tons of uranium arranged vertically in a concentric square array within the core shroud. Each fuel assembly contains 264 UO₂ fuel pins with various enrichments (2.1, 2.6 and 3.1%), 24 control rods of Gd₂O₃ and one central water channel and all are arranged in a 17 × 17 array of matrix. Different computer codes including WIMS, TWOTRAN, CITATION and MCNP have been employed to develop a versatile and accurate reactor physics model of the PWR core. The computational methods, tools and techniques, customization of cross section libraries, various models for cells and super cells, and a lot of associated utilities have been standardized and established/validated for the overall core analysis. The analyses were performed in 3 steps: firstly for fuel pincells, then for the fuel assemblies and finally for the whole core. The WIMS and MCNP calculated infinite multiplication factors for fuel pincells having 2.1% enriched ²³⁵U were found to be 1.23393 and 1.23654, for 2.6% enrichment 1.28635 and 1.28887, and finally for 3.1% enrichment 1.32481 and 1.32812, respectively. For fuel assembly, WIMS and MCNP calculated infinite multiplication factors having 2.1% enrichment were found to be 1.24853 and 1.25445, for 2.6% enrichment 1.30372 and 1.30992, and for 3.1% enrichment 1.34424 and 1.35041, respectively. The effective multiplication factor calculated by CITATION, TWOTRAN and MCNP for whole core were found to be 1.25580, 1.25909 and 1.26382, respectively. The peak thermal neutron flux in the core calculated by MCNP was found to be 5.0298 × 10¹⁴ neutrons/cm² s and the average core power density was 17.1 kW/cm³. The calculated results from different codes were found to be very good agreement for different moderator conditions. The choice of computer codes like WIMSD, TWOTRAN, CITATION and MCNP which are being used in nuclear industry for many years were selected to identify and develop new capabilities needed to support PWR analysis. The ultimate goal of the validation of the computer codes for PWR applications is to acquire and reinforce the capability of these general purpose computer codes to perform the core design and optimization study.     

A prediction model of IASCC initiation stress for bolts in PWR core internals

A model of IASCC initiation stress for bolts of core internals in pressurized water reactors was developed considering differences in material property changes due to irradiation and material conditions. Assuming that IASCC initiation was controlled by grain boundary composition and yield strength, these values for each specimen of post-irradiation IASCC initiation tests were calculated by physical kinetic models considering dose rate, temperature, material composition and surface hardening. Then, correlations of grain boundary composition and yield strength with IASCC initiation stress were determined. The model predicted that the IASCC initiation stress became lower with dose and was lower for higher temperature, lower flux and higher surface hardening level.     

压水堆核燃料循环的经济分析程序介绍

所开发的压水堆核燃料循环分析计算机经济程序包括12个子程序,代表着压水堆整个核燃料循环各种不同的工艺过程。本程序能算出压水堆核电站核燃料循环中燃料费用对发电成本的影响,给出各工艺过程对燃料成本的敏感度分析,并尽可能给出燃料循环中燃料材料及服务的价格数据.     

压水堆核电站堆芯集中参数模型的微机仿真

阐述了PWR核电站堆芯的模型化问题,提适用于微机仿真的核电站堆芯的物理数学模型,将核电站堆芯分为三大块分别建立模型,中子动力学模块,反应性反馈模块,堆芯热力学模块,建立系统传递函数,运用MATLA仿真,得到良好结果。     

Pu recycling in a full Th-MOX PWR core. Part I: Steady state analysis

Current practice of Pu recycling in existing Light Water Reactors (LWRs) in the form of U-Pu mixed oxide fuel (MOX) is not efficient due to continuous Pu production from U-238. The use of Th-Pu mixed oxide (TOX) fuel will considerably improve Pu consumption rates because virtually no new Pu is generated from thorium. In this study, the feasibility of Pu recycling in a typical pressurized water reactor (PWR) fully loaded with TOX fuel is investigated.Detailed 3-dimensional 100% TOX and 100% MOX PWR core designs are developed. The full MOX core is considered for comparison purposes. The design stages included determination of Pu loading required to achieve 18-month fuel cycle assuming three-batch fuel management scheme, selection of poison materials, development of the core loading pattern, optimization of burnable poison loadings, evaluation of critical boron concentration requirements, estimation of reactivity coefficients, core kinetic parameters, and shutdown margin.The performance of the MOX and TOX cores under steady-state condition and during selected reactivity initiated accidents (RIAs) is compared with that of the actual uranium oxide (UOX) PWR core.Part I of this paper describes the full TOX and MOX PWR core designs and reports the results of steady state analysis. The TOX core requires a slightly higher initial Pu loading than the MOX core to achieve the target fuel cycle length. However, the TOX core exhibits superior Pu incineration capabilities.The significantly degraded worth of control materials in Pu cores is partially addressed by the use of enriched soluble boron and B₄C as a control rod absorbing material. Wet annular burnable absorber (WABA) rods are used to flatten radial power distribution. The temperature reactivity coefficients of the TOX core were found to be always negative. The TOX core has a slightly reduced, as compared to UOX core, but still sufficient shutdown margin.In the TOX core β_eff is smaller by about a factor of two in comparison to the UOX core and even lower than that of the MOX core. The combination of small β_eff and reduced control materials worth may potentially deteriorate the performance under RIA conditions and requires an additional examination. The behavior of the considered cores during the most limiting RIAs, such as rod ejection, main steam line break, and boron dilution, is further investigated and reported in Part II of the paper.     

Evaluation of fluid flow in the lower core of a PWR with Code_Saturne

In order to better understand the stresses to which fuel rods are subjected, we need to improve our knowledge of the fluid flow inside the core. The first spacer grid region is of particular interest, as fuel rod fretting has sometimes been observed at that level. Entry conditions depend on the geometry of the lower core plate and of the assembly nozzles. Distribution of flow in the downcomer and lower plenum is also a factor. A series of calculations are thus run with the incompressible Navier–Stokes solver, Code_Saturne, using a classical RANS turbulence model. The first calculations involve a global geometry, including part of the cold legs, downcomer, lower plenum, and lower core of a pressurized water reactor. The level of detail includes most obstacles below the core. The lower core plate and the fuel rod assemblies above it cannot be well represented within a practical mesh size, so that a head loss model is used. Different types of assemblies can be represented through different head loss coefficients. We make full use of Code_Saturne’s non-conforming mesh possibilities to represent a complex geometry, being careful to retain a good mesh quality. Steady-state or near steady-state results are obtained, which may be used as realistic entry conditions for full core calculations at assembly width resolution, and beyond those, mechanical strain calculations. We are especially interested in more detailed flow conditions and in the lower core area, so as in the future to quantify vibrational input. This requires a much higher resolution, which is limited to a scale of a few assemblies for practical reasons. At this scale, most of the features of the fuel rods, nozzles, and guide tubes are represented, though the geometry of the spacer grids is still much simplified, and details such as debris-trapping grids are ignored. Different meshes are used for different fuel types. For the moment, a constant velocity upstream of the lower core plate is used as an inlet condition. We have also built a small lower fuel rod assembly mock-up (1/5 scale 7 × 7 tube, 3 × 3 assemblies) with which we plan to obtain detailed flow information, and better qualify the use of our CFD codes with regards to this type of application.     

Comparison of two models for a pebble bed modular reactor core coupled to a Brayton cycle

The pebble bed modular reactor (PBMR) plant is a promising concept for inherently safe nuclear power generation. This paper presents two dynamic models for the core of a high temperature reactor (HTR) power plant with a helium gas turbine. Both the PBMR and its power conversion unit (PCU) based on a three-shaft, closed cycle, recuperative, inter-cooled Brayton cycle have been modeled with the network simulation code Flownex.One model utilizes a core simulation already incorporated in the Flownex software package, and the other a core simulation based on multi-dimensional neutronics and thermal-hydraulics. The reactor core modeled in Flownex is a simplified model, based on a zero-dimensional point-kinetics approach, whereas the other model represents a state-of-the-art approach for the solution of the neutron diffusion equations coupled to a thermal-hydraulic part describing realistic fuel temperatures during fast transients. Both reactor models were integrated into a complete cycle, which includes a PCU modeled in Flownex.Flownex is a thermal-hydraulic network analysis code that can calculate both steady-state and transient flows. An interesting feature of the code is its ability to allow the integration of an external program into Flownex by means of a so called memory map file.The total plant models are compared with each other by calculating representative transient cases demonstrating that the coupling with external models works sufficiently. To demonstrate the features of the external program a hypothetical fast increase of reactivity was simulated.