非线性迭代半解析节块法在CANDU堆燃料管理中的应用

研究目前压水堆中常用的现代的非线性迭代节块法在CANDU堆的燃料管理中的应用 ,研发了非线性迭代半解析节块法燃料管理程序FMPHWR。通过基准题及对秦山三期CANDU堆的计算表明 :同目前采用有限差分和有限元方法的重水堆燃料管理程序相比 ,在相当的精度下 ,FMPHWR可以获得较高的计算效率。它完全可以用于CANDU堆燃料管理计算。     

新一代CANDU—9重水堆核电厂简介

ＣＡＮＤＵ－９是电功率为９００ＭＷ级的重水堆核电厂,其设计基于达灵顿和布鲁斯Ｂ多机组核电厂,并融入了一些最新的工程设计和研究成果,除了继续采用成熟的系统和部件外,在安全性,地可靠性和可维护性方面作了重要改进。ＣＡＮＤＵ－９综合考虑了安全审评和执照申请过程中发现的问题,产使其体现在安全设计理念中,特别是对慢化剂系统,端屏蔽冷却系统,系统和应急堆芯冷却系统进行了改进。     

在重水堆中用贫铀作为核燃料的应用研究

对以贫铀和钚组成ＭＯＸ核燃料替代ＣＡＮＤＵ堆中的天然铀的可能性进行了探讨、从而开一用于核能的途径。经过初步验算,得出了用＾２３５Ｕ含量为０．２５％的贫化铀浓缩残渣和钚组成ＭＯＸ核燃料可以替代ＣＡＮＤＵ堆中的天然铀来维持重水堆中的链式反就,达到核能利用的的目的。并展望了贫铀应用的前景。     

CANDU型核电站技术特点及其发展趋势

以中国秦山三期核电站的CANDU 6机组为例,介绍加拿大原子能公司原创开发的CANDU型核电站技术特点及其发展概况。讨论这种类型核电站的系统组成与压水堆核电站之间的相似性,CANDU堆芯结构的基本特点;着重指出这些特点对核电经济性、安全性、燃料可持续性的重要意义,以及可以不断向更新一代先进设计发展迈进的优势。     

Enhanced tritium production for fusion reactors via3He(n,p)3 H in the heavy water moderator of a CANDU reactor

There is an increasing requirement for tritium to supply the fuel needs of current experimental fusion devices and in the initial startup of future power generating reactors. Tritium is produced in heavy water reactors through deuterium activation, but the total production capacity of Canadian operated CANDUs will fall short of future demands, during the period before and for some time after self-sufficient reactors become available. Consequently, methods of enhancing tritium generating rates warrant investigation. Herein we provide the results of an inquiry into the feasibility of enhancing tritium production levels through the activation of helium-3 following its external addition to the heavy water moderator system of a hypothetical 500–600 MWe CANDU reactor. The approach adopted involves simulation of the temporal evolution of the tritium activities, originating from²H(n,)³H and³He(n, p)³H, as described by a simple first order kinetic model. The results suggest that the frequent addition of helium-3 to the moderator water will enhance tritium production inventories. The enhancement factor is highly dependent upon the rate at which helium-3 irretrievably escapes to the moderator cover gas. However, the direct activation of helium-3, contained in a closed loop such as the annulus gas system, for example, would be essentially complete within a few weeks without any significant loss.     

核安全关键软件的验证与确认技术

介绍了加拿大重水铀反应堆CANDU的安全关键软件的验证与确认(V&V)技术,说明了CANDU核反应堆停堆系统的脱扣计算机及其设计方法,详细描述了用于停堆脱扣计算机软件的确认和可靠性V&R测试的多功能测试平台,以及该测试平台在停堆脱扣计算机软件V&R测试中的应用.V&V技术已被成功地应用于各国CANDU核反应堆停堆系统的脱扣计算机设计中,如韩国的月城CANDU核反应堆、中国的秦山CANDU核反应堆、罗马尼亚Cernavda的2号CANDU核反应堆和加拿大Point Lepreau重建的CANDU核反应堆.随着计算机软硬件技术的发展,所描述的过程和工具在近期的项目中已得到了不断的改善.     

Comment on ‘A combined SIF and temperature model of delayed hydride cracking in zirconium materials’, by A.A. Shmakov et al.

Given the experimental facts related to delayed hydride cracking in zirconium alloys and the DHC models proposed so far, critical comments on “A combined SIF and temperature model of delayed hydride cracking in zirconium materials” by A.A. Shmakov were made, demonstrating that the authors’ DHC model is irrational.     

Adapting and applying SCDAP/RELAP5 to CANDU in-vessel retention studies

In CANDU reactors, the cool moderator surrounding the calandria tubes provides a potential heat sink following an accident initiator if the emergency coolant injection fails. However, in scenarios when a subsequent loss of all heat sinks occurs, the fuel channels fail and ultimately, the entire reactor core collapses and relocates into the bottom of calandria vessel (CV), which is externally cooled by shield-tank water. Previous studies using MAAP4-CANDU and ISAAC computer codes were found to investigate the long-term coolability of the CV in the late phase of core degradation in course of a severe accident. SCDAP/RELAP5 was applied in a previous work of the authors to the study of the in-vessel retention issue using the COUPLE models with user-defined slumping inside the 2D COUPLE mesh. This option allows for thermal and mechanical analyses of the reactor lower head avoiding the necessity to calculate the preceding course of core degradation during the accident. The former analyses used an equivalent spherically shaped CV while, for the present paper, calculations are performed with COUPLE routines modified to properly use the option for a horizontal pipe in plane geometry. The paper describes the modifications and the application of the resulted SCDAP/RELAPSIM/MOD3.4 code version to the study of the coolability of a CV starting with a dry debris bed. The vessel rupture time is compared to the ISAAC calculated value for a LOCA with loss of all heat sinks and no recovery actions. Parametric studies are performed in order to quantify the effect of several identified sources of uncertainty: boundary conditions of the vessel above debris, gap heat transfer coefficient and metallic fraction of zirconium inside the debris.     

CANDU重水反应堆钴调节棒组件结构设计

利用秦山三期CANDU重水反应堆生产⁶⁰Co放射源具有活度高、产量高、成本低等优点。CANDU重水反应堆原有的21个不锈钢调节棒组件改成同样数量和位置的钴调节棒组件后,在保持原来调节棒功能的条件下,利用⁵⁹Co吸收中子转变为⁶⁰Co,生产放射性钴源。本工作详细阐述了钴调节棒组件设计要求及结构设计过程中与各种设计接口之间的关系,并通过对设计的钴调节棒组件进行结构完整性分析、提插棒时间分析及跌落事故分析,论证了其在重水反应堆内运行的安全性。经反应堆成功运行经验证明,钴调节棒组件结构设计安全可靠。     

Conditioned optimization of fuel feed rate for a CANDU-600 MWe reactor

The paper solves an optimization problem concerning the conditioned minimization of fuel feed rate against a control vector which incorporates both zonal irradiation and the configuration of an adjusting rod system. One starts from a reference core with two irradiation zones of which the first contains 188 channels and for which all control devices and parasitic absorbers are taken into account. The fuel feed rate is minimized subject to four restrictions imposed, namely reactivity to equilibrium, maximum channel and bundle powers and reactivity of an adjusting rod system. The codes FMDP and LPROG are used to solve this problem.