Benchmark Calculations of a Radiation Heat Transfer for a CANDU Fuel Channel Analysis using the CFD Code

To justify the use of a commercial Computational Fluid Dynamics (CFD) code for a CANDU fuel channel analysis, especially for the radiation heat transfer dominant conditions, the CFX-10 code is tested against three benchmark problems which were used for the validation of a radiation heat transfer in the CANDU analysis code, a CATHENA. These three benchmark problems are representative of the CANDU fuel channel configurations from a simple geometry to a whole fuel channel geometry. For the solutions of the benchmark problems, the temperature or the net radiation heat flux boundary conditions are prescribed for each radiating surface to determine the radiation heat transfer rate or the surface temperature, respectively by using the network method.

The Discrete Transfer Model (DTM) is used for the CFX-10 radiation model and its calculation results are compared with the solutions of the benchmark problems. The CFX-10 results for the three benchmark problems are in close agreement with these solutions, so it is concluded that the CFX-10 with a DTM radiation model can be applied to the CANDU fuel channel analysis where a surface radiation heat transfer is a dominant mode of the heat transfer.     

CANDU6核电厂无过滤安全壳通风模式的研究

CANDU6核电厂早期设计未考虑严重事故对策,在严重事故下,CANDU6核电厂的安全壳容易失效。为了解决这一问题,本文研究了无过滤安全壳通风模式对CANDU6核电厂安全壳的影响。本文选取典型的全厂断电严重事故,利用重水蒸气回收系统作为无过滤安全壳通风的路径,初步研究了该通风模式下对安全壳完整性的保持和对裂变产物源项的滞留能力。研究表明:该通风模式可以有效保持安全壳的完整性,同时,对裂变产物源项也有一定的滞留能力。     

用RELAP5分析RD-14装置的破口实验

用RELAP5 /MOD3 .2程序模拟了在RD 1 4实验装置上进行的两个CANDU反应堆临界破口实验。对破口出现以后 ,冷却剂系统压力、堆芯压降和元件包壳温度的变化趋势进行了研究 ,计算结果和实验数据符合较好 ,表明用RELAP5程序模拟CANDU反应堆在LOCA事故后系统瞬变是可行的     

CANDU refuelling optimization by successive two-step mathematical programming – I: MILP models for refuelling channel selection

A CANDU refuelling optimization method based on successive two-step mathematical programming (MP) is developed. The first step is for the selection of weekly refuelling candidates, while the second step is for the determination of the detailed weekly refuelling scheme with the prescribed candidates. To search for the overall optimal solution with the aftereffects of refuellings being considered, a zero- dimensional core physics model and a three-dimensional core physics model are directly incorporated into their corresponding MP model. The formulated large-scale MP problems are solved by use of the powerful generic MP problem solver CPLEX® (commercial product of IBM ILOG). The potential for engineering application of the method is demonstrated by its on-site test application to Qinshan CANDU 6 reactor. Refuelling simulation against the operation history of 392 full power days demonstrates that in addition to the good regional power and core reactivity ripple control, higher average fuel discharge burnup of 176.54 MWh/kgU is achieved. No violation of reactor operating limits is found. Regarding the computational cost of the method, though for most tested cases the performance of the method satisfies the stringent on-site time requirement, some exceptional cases are observed. However this occasional speed problem is within a factor of two or three from the targeted requirement, which is realistically optimistic in the near future considering today’s rapid advancement in hardware and software.     

Utilization of nuclear waste plutonium and thorium mixed fuel in candu reactors

Spent nuclear fuel out of conventional light water reactors contains significant amount of even plutonium isotopes, so called reactor grade plutonium. Excellent neutron economy of Canada deuterium uranium (CANDU) reactors can further burn reactor grade plutonium, which has been used as a booster fissile fuel material in form of mixed ThO₂/PuO₂ fuel in a CANDU fuel bundle in order to assure reactor criticality. The paper investigates incineration of nuclear waste and the prospects of exploitation of rich world thorium reserves in CANDU reactors. In the present work, the criticality calculations have been performed with 3‐D geometrical modeling of a CANDU reactor, where the structure of all fuel rods and bundles is represented individually. In the course of time calculations, nuclear transformation and radioactive decay of all actinide elements as well as fission products are considered. Four different fuel compositions have been selected for investigations: ① 95% thoria (ThO₂) + 5% PuO₂, ② 90% ThO₂ + 10% PuO₂, ③ 85% ThO₂ + 15% PuO₂ and ④ 80% ThO₂ + 20% PuO₂. The latter is used for the purpose of denaturing the new ²³³U fuel with ²³⁸U. The behavior of the criticality k_∞ and the burnup values of the reactor have been pursued by full power operation for ~10 years. Among the investigated four modes, 90% ThO₂ + 10% PuO₂ seems a reasonable choice. This mixed fuel would continue make possible extensive exploitation of thorium resources with respect to reactor criticality. Reactor will run with the same fuel charge for ~7 years and allow a fuel burnup ~55 GWd/t. Copyright © 2016 John Wiley & Sons, Ltd.     

Remote Fabrication of DUPIC Fuel Pellets in a Hot Cell under Quality Assurance Program

The Korea Atomic Energy Research Institute (KAERI) has been developing the Direct Use of Spent Pressurized Water Reactor (PWR) Fuel in the CANada Deuterium Uranium (CANDU) Reactors (DUPIC) fuel fabrication technology since 1992, and the basic DUPIC fuel fabrication process was established in 2002. In order to demonstrate the robustness of the DUPIC fuel fabrication process through the irradiation test, it is important that a Quality Assurance (QA) program should be in place before a fabrication of the DUPIC fuel. Therefore, the Quality Assurance Manual (QM) for the DUPIC fuel was developed on the basis of the Canadian standard, CAN3-Z299.2-85. This manual describes the quality management system applicable to the activities performed for the DUPIC fuel fabrication at KAERI. In order to demonstrate the DUPIC fuel fabrication technology and produce qualified DUPIC fuel pellets, the process qualification tests were performed, which include three pre-qualification tests and three qualification tests. The characteristics of the DUPIC fuel pellets such as the sintered density, grain size, and surface roughness were measured and evaluated in accordance with the QA procedures. The optimum fabrication process of the DUPIC fuel pellet was also established based on the qualification results. Finally a production campaign was carried out to fabricate the DUPIC fuel pellets at a batch size of 1 kg following the QA program. As a result of the production campaign, qualified DUPIC fuel pellets were successfully produced and, therefore, the remote fuel fabrication technology of the DUPIC fuel pellet was demonstrated.     

Verification of Source Term Estimation Method against Measured Data for Decommissioning Waste from a CANDU Reactor

The method for the establishment of an equilibrium core model proposed in the previous paper and the source term calculation method proposed in this paper for the characterization of decommissioning waste were verified by comparing the nuclide inventory estimated by MCNP/ORIGEN2 simulations with the measured nuclide inventory according to a chemical assay in an irradiated pressure tube discharged from Wolsong Unit 1 in 1994. At first, the time-average pseudoequilibrium full-core model of Wolsong Unit 1 was developed on the basis of the previously proposed modeling method for the activation of in-core and ex-core structural components. Then, the application level of the neutron flux and cross section in the radionuclide buildup calculation were compromised. Fourteen major actinides and fission products were considered to represent the irradiated fuel condition, and a geometry simplification was also introduced in the burned full-core model for MCNP simulation. The assumption of a constant neutron flux and capture cross section as a function of the irradiation time was applied in the radionuclide buildup calculation in ORIGEN2. As a result, the values estimated from the analysis system agreed with the measured data within a difference range of 30%. Therefore, it was found that the MCNP/ORIGEN system and source term characterization method proposed can be viable to estimate the source terms of the decommissioning waste from a CANDU reactor.     

压水堆回收铀在重水堆应用示范验证试验

回收铀（RU）是一种重要的核能资源,随着核电发展和铀资源价格的上涨将更加受到重视,迄今为止国际上尚未很好地解决其有效利用问题。鉴于我国既有压水堆又有重水堆的现状,本文提出利用重水堆烧RU的设想,开发了一种与天然铀燃料中子学等效的由RU和贫铀（DU）混合而得的等效天然铀（NUE）燃料,并在秦山运行重水堆上开展随堆示范验证试验,以积累RU利用相关运行经验,为后续全堆应用提供了关键的技术支持。     

重水堆运行监督中的几个特殊问题探讨

秦山第三核电厂的CANDU-6型重水堆已经进入商业运行阶段,上海监督站在对重水堆的监督过程中遇到了一些新的问题,如重水泄漏与损失、燃料棒束包壳破损以及在电厂大修期间由启动仪表导致第一停堆系统误动作等.本文就这些特殊问题的处理原则进行探讨,并提出了自己的见解.     

CANDU堆应用RU的PWR/CANDU联合核燃料循环的研究

对压水堆乏燃料后处理回收铀（RU)在秦山三期CANDU堆中应用的可行性和经济性进行分析。使用ORIGEN2程序．对后处理回收铀在生产后放置不同时间后核素的成份和放射性活度进行了计算。证明RU燃料元件生产的放射性水平是可以接受的。使用DRAGON／DONJON程序对应用RU的秦山三期CANDU堆的时均堆芯和瞬时堆芯校验分析表明：采用简单的2燃耗区，2、4棒束的换料方案能满足最大通道功率、最大棒束功率限制。通过放射性分析和堆芯物理分析可以看出，秦山三期CANDU堆在不改变堆芯结构及运行模式的条件下，从天然铀(NU)燃料过渡到RU燃料是可行的。通过对秦山三期CANDU堆应用RU的经济性分析，可以看出PWR／CANDU联合核燃料循环的策略既可节约铀资源(23％)，提高燃料的能量输出(4l％)．又减少了废燃料的处置量(66％)．可大大降低核电成本。