一种节块多群精细注量率重构方法

在全堆粗网格节块解的基础上,经节块内精细注量率的重构来获得各燃料棒的功率是目前轻水堆堆芯分析计算中普遍采用的方法,然而,目前的精细注量率重构方法大都只适用于两群,无法进行多群重构。在韩国国立首尔大学提出的多群重构思想基础上,经进一步改进,研制出了多群重构方法,并开展了多个基准问题的检验。     

Multi-group pin power reconstruction method based on colorset form functions

A multi-group pin power reconstruction method that fully exploits nodal information obtained from global coarse mesh solution has been developed.It expands the intra-nodal flux distributions into nonseparable semi-analytic basis functions,and a colorset based form function generating method is proposed,which can accurately model the spectral interaction occurring at assembly interface.To demonstrate its accuracy and applicability to realistic problems,the new method is tested against two benchmark problems,including a mixed-oxide fuel problem.The results show that the new method is comparable in accuracy to fine-mesh methods.     

Three-dimensional pin power reconstruction for the axially heterogeneous region in BWR

A three-dimensional pin power reconstruction method was proposed and verified by applying to the axially heterogeneous region problem of the BWR core calculation. Because the production assembly calculation has been carried out by two-dimensional deterministic calculation methods like current coupling collision probability or the method of characteristics, it has been difficult to predict the detailed three-dimensional heterogeneous pin power distribution of the axially heterogeneous region. Consequently, only radial intranodal homogeneous power distributions have been considered, and axial intranodal homogeneous power distributions have not been considered in the estimation of linear-heat-generation-ratio at common BWR core calculations.     

On-line reconstruction of in-core power distribution by harmonics expansion method

Fixed in-core detectors are most suitable in real-time response to in-core power distributions in pressurized water reactors (PWRs). In this paper, a harmonics expansion method is used to reconstruct the in-core power distribution of a PWR on-line. In this method, the in-core power distribution is expanded by the harmonics of one reference case. The expansion coefficients are calculated using signals provided by fixed in-core detectors. To conserve computing time and improve reconstruction precision, a harmonics data library containing the harmonics of different reference cases is constructed. Upon reconstruction of the in-core power distribution on-line, the two closest reference cases are searched from the harmonics data library to produce expanded harmonics by interpolation. The Unit 1 reactor of DayaBay Nuclear Power Plant (DayaBay NPP) in China is considered for verification. The maximum relative error between the measurement and reconstruction results is less than 5.5%, and the computing time is about 0.53 s for a single reconstruction, indicating that this method is suitable for the on-line monitoring of PWRs.     

The behavior of reactor power and flux resulting from changes in core-coolant temperature for a miniature neutron source reactor

In this work, measurements were performed to verify the theoretical predictions of reactor power and flux parameters that result from changes in core inlet temperature (T_in) and the temperature difference between the coolant inlet and outlet (ΔT) in the Nigeria Research Reactor-1 (NIRR-1), which is a Miniature Neutron Source Reactor (MNSR). The measured data shows that there is a strong dependence of the reactor power on coolant temperature in agreement with the design of MNSR. The experimental parameters were found to be in good agreement with data obtained using a semi-empirical relationship between the reactor power, flux parameters, core inlet temperature, and the coolant temperature rise. The relationship was therefore used to predict the power level of NIRR-1 from its neutron flux parameters to which it has been found to be proportional. The variation of T_in and ΔT with the reactor power and flux was also investigated and the results obtained are hereby discussed.     

Pipe-anchor discontinuity analysis utilizing power series solutions, Bessel functions, and Fourier series

One of the paradigmatic classes of problems that frequently arise in piping stress analysis discipline is the effect of local stresses created by supports and restraints attachments. Over the past 20 years, concerns have been identified by both regulatory agencies in the nuclear power industry and others in the process and chemicals industries concerning the effect of various stiff clamping arrangements on the expected life of the pipe and its various piping components. In many of the commonly utilized geometries and arrangements of pipe clamps, the elasticity problem becomes the axisymmetric stress and deformation determination in a hollow cylinder (pipe) subjected to the appropriate boundary conditions and respective loads per se. One of the geometries that serve as a pipe anchor is comprised of two pipe clamps that are bolted tightly to the pipe and affixed to a modified shoe-type arrangement. The shoe is employed for the purpose of providing an immovable base that can be easily attached either by bolting or welding to a structural steel pipe rack.Over the past 50 years, the computational tools available to the piping analyst have changed dramatically and thereby have caused the implementation of solutions to the basic problems of elasticity to change likewise. The need to obtain closed form elasticity solutions, however, has always been a driving force in engineering. The employment of symbolic calculus that is currently available through numerous software packages makes closed form solutions very economical. This paper briefly traces the solutions over the past 50 years to a variety of axisymmetric stress problems involving hollow circular cylinders employing a Fourier series representation. In the present example, a properly chosen Fourier series represent the mathematical simulation of the imposed axial displacements on the outside diametrical surface. A general solution technique is introduced for the axisymmetric discontinuity stresses resulting from an anchor restraint on a selected of pipe geometry. These solutions can be economically implemented on today's symbolic calculus software packages with no loss in solution accuracy when compared to often more expensive techniques such as the finite element method. Verification of the axisymmetric solution technique is illustrated by the comparison of results for the closed form solutions versus those approximated by the finite element technique. Extensions of the general axisymmetric solution technique to other geometries and applied loads are also discussed while the numerical and graphical results are tendered.     

System design and measurements of flux concentrator and its solid-state modulator for CEPC positron source

Positron sources are one of the most important components of the injector of a circular electron positron collector(CEPC).The CEPC is designed as an e+e-col-lid...     

Multigroup multi-layer models of neutron reflection and transmission for reactor transport calculations with anisotropic scattering

In this article, we extend the one-speed multi-layer models to neutron reflection and transmission developed in our earlier work (de Abreu, M.P., 2005. Multi-layer models to neutron reflection and transmission for whole-core transport calculations, Annals of Nuclear Energy 32, 215) to multigroup transport theory. We begin by considering a two-layer boundary region, and we develop for such a region discrete ordinates models to the diffuse reflection and transmission of neutrons for multigroup nuclear reactor core problems with anisotropic scattering. We perform numerical experiments to show that our models to neutron reflection and transmission can be used to replace efficiently and accurately two nonactive boundary layers in whole-core transport calculations. We conclude this article with an inductive extension of our two-layer results to a boundary region with an arbitrary number of layers.     

Study of relation between the gamma flux buildup factors and source geometry by M-C simulation

This paper is to achieve a gamma-ray source with the lowest rate of buildup factor, which is of great importance in medical, industrial and agricultural sciences.The flux buildup factor of gamma rays is calculated by the MCNP code for point, linear, surface and volume sources with shield layers of lead, iron and aluminum. The results show that for the high Z shielding material, the flux buildup factor of coaxial cylindrical sources is the lowest(1.6–2.3)of all sources, while for low Z shielding materials, the coaxial disk surface sources have smaller buildup factor(1.45–1.6).     

Nuclear reactor power and flux distribution fitting from a diffusion theory model and experimental data

Power and flux tracking in nuclear reactor cores are normally done using diffusion codes. These codes take into account geometric characteristics, compositions, operative conditions and eventually thermohydraulic feedbacks. However, due to operating requirements, in some cases these reactor are instrumented with in-core neutronic flux detectors. This information is usually used in order to verify instantaneous calculations, but it can also be used to fit the theoretical model and then to produce an improved solution. A scheme like this will fit into some extend with the deficiencies of the mathematical problem definition. In this work the integration between a theoretical model and experimental data is searched. It starts with a synthesis method called flux mapping and then its elements are studied. This is the minimization of a functional and selection and construction of expansion functions. For the first expansion function the solution of a diffusion code is chosen, while the other ones are built from Helmholtz equation solutions for a reactor having the same dimensions and boundary conditions. Flux mapping is then tested in a real reactor, a CANDU-600, for several reactivity device configurations. The set of measurement data available even from different physics principles made it suitable to perform the verifications. Comparison with experimental data in the zone at which flux detectors were located shows, for flux mapping, agreements of 2.4%, but 3.2% for the diffusion code used as standard for this reactor. Comparisons throughout the core shows agreements of 3.4 and 5.0%, respectively.     

Assessment of a DNB-type theoretical critical heat flux model for rod bundles with non-uniform axial power shapes

A general critical heat flux (CHF) prediction method with a wide applicable range and reasonable accuracy is essential to the thermal-hydraulic design and safety analysis at the conceptual design stage for a new pressurized water reactor (PWR). In this study, the Korea Advanced Institute of Science and Technology (KAIST) liquid sub-layer dryout CHF prediction model for Departure from Nucleate Boiling (DNB) region has been implemented in a sub-channel analysis code, and investigated for the method's possible use in a rod bundle environment with various non-uniform axial power shapes. The KAIST model showed comparable prediction capability to Lin's method for bottom-, center-, and top-peaked heat flux shapes. The KAIST model, without any correction factors or empirical constants, turned out to be suitable to fulfill the needs for a basis of a general CHF prediction method as compared to Lin's method and Westinghouse-3 (W-3) correlation.     

核电厂一回路源项和排放源项框架体系研究

我国核电机组堆型众多,来源广泛,这些引进堆型的源项在我国应用中还存在一些问题。源项设计是否合理,直接影响到排放源项的准确性和环境影响评价源项的合理性。本文通过分析不同堆型源项在我国应用中存在的问题,研究如何构建我国核电厂通用的一回路源项和排放源项框架体系,为解决国内核电厂源项计算中长期存在的问题,也为我国华龙一号和CAP1400堆型的源项计算提供技术基础。     

Calculation of reactivity and power factors depending on the external source location

We used the neutron diffusion equation with external neutron sources, in cartesian geometry and the two groups of energy, to verify the influence of external neutron source locations in the calculation of reactivity and power factors. To this end, the Coarse Mesh Finite Difference (CMFD) method was applied to the adjoint flux calculation and to simplify reactivity calculation in PWR type reactor, using the output of the Nodal Expansion Method (NEM). Different locations on the two-dimensional plane, as well as different types of fuel elements in the reactor core were used in the present study.     

Measurements of critical heat flux in CANDU 37-element bundle with a steep variation in radial power profile

An experiment has been performed to obtain dryout power measurements with a 37-element bundle string simulating radial power profiles of high-enriched fuel (which contains slightly higher enrichment than the CANDU fuel and is not the same as the traditional highly enriched uranium) and natural-uranium fuel. The electrically heated bundle string was cooled with Refrigerant-134a and installed vertically inside the test station. Occurrences of CHF were detected using sliding thermocouples installed inside each element. Measurements showed that the dryout power for the high-enriched fuel bundle is lower than that for the natural-uranium fuel bundle by 26%, on average. Similarly, the corresponding critical heat flux (CHF) values for the high-enriched fuel power profile are lower than those for the natural-uranium fuel power profile by about 48%. A correlation previously proposed for the effect of radial power profile on CHF underpredicts considerably the CHF for the high-enriched fuel power profile. The correlation has been revised to improve the prediction accuracy. The revised correlation represents closely the available database. It exhibits a correct asymptotic trend and hence can be extended to bundles with more severe variation in radial power profile than those covered in the current experiment.     

A modular heavy-water reactor as a source of energy in atomic thermal power stations and atomic electric power stations

Institute of Theoretical and Experimental Physics. Translated from Atomnaya Énergiya, Vol. 70, No. 5, pp. 319–322, May, 1991.     

压水堆核电厂源项控制实践与改进

本文分析了福清核电厂1号机组停堆沉积源项调查发现的一回路管道内壁⁵⁸Co和⁶⁰Co表面活度水平、剂量率贡献以及随机组运行时间发生的变化情况,并介绍了压水堆核电厂活化腐蚀产物的形成、沉积及存在形式。通过分析201大修主泵停运对氧化运行效果及蒸汽发生器(SG)下封头辐射水平的影响,结合酸性氧化环境下腐蚀产物溶解度变化的特点,提出改进主泵停运时机以提高氧化运行效果的建议。另外,还分析了阀门密封面维修导致向一回路系统引入含钴金属颗粒对机组源项的影响,建议严格控制阀门维修过程以减少⁵⁹Co进入一回路系统。