Sensitivity Study of Fuel Cost in Extended Burnup BWR Core

A sensitivity study on the fuel cost of an extended burnup BWR core has been carried out on the basis of a realistic model of discharge burnup extension. Full power operating length in months in a refueling cycle and the number of refueling batches are chosen as independent variables in the model to describe extended burnup cores of various types. The reference core for the sensitivity study adopts 9-month full power operation and 4-batch refueling scheme. The difference in the plant cost between the extended burnup core and the reference core, which is referred to as plant capacity factor (PCF) credit, is estimated and combined with the fuel cost to calculate the fuel cost with PCF credit.

The results show that the fuel cost with PCF credit decreases for the extended burnup core with stretched operating length as the burnup extends in cases of constant non-operating length in a cycle, and that it may increase for the extended burnup core with decreased batch number in cases of constant plant capacity factor. It is also suggested that the cost minimum combination of the independent variables can be found to a given discharge burnup for the extended burnup core with decreased batch number in an intermediate case between these two extreme cases. Extended burnup cores with fixed batch number tend to have a lower natural uranium requirement, but larger separative work requirement.

The economic break-even condition for the extended burnup core with decreased batch number is discussed based on the fraction of fixed part in the non-operating length, which is insensitive to the cycle length stretch-out.     

核电站设备可靠性数据库的建立与应用

以大亚湾核电站为例,论述了核电站设备可靠性数据的采集与处理、可靠性参数的分析计算方法及可靠性数据库的建立与应用等。数据源涉及到设备的设计信息、运行信息、维修信息、定期试验记录等。在大亚湾核电站运行经验的基础上,形成了大亚湾核电站设备可靠性数据库,为深入地、客观地记录核电站各类设备的运行历史和现状、监控电站设备,特别是与安全相关设备的状态,提供了有效的工具;为加强设备的可靠性与可用性管理、确保电站的安全经济运行,提供了非常有实用价值的信息;同时还为核电站的安全管理、可靠性分析、概率安全评价、以可靠性为中心的维修及经济性管理等领域里的新技术在核电站的应用研究与开发,提供了必不可少的数据。     

Variations of a passive safety containment for a BWR with active and passive safety systems

The paper presents variations of a certain passive safety containment for a near future BWR. It is tentatively named Mark S containment in the paper. It uses the operating dome as the upper secondary containment vessel (USCV) to where the pressure of the primary containment vessel (PCV) can be released through the upper vent pipes. One of the merits of the Mark S containment is very low peak pressure at severe accidents without venting the containment atmosphere to the environment. Another merit is the capability to submerge the PCV and the reactor pressure vessel (RPV) above the core level by flooding water from the gravity-driven cooling system (GDCS) pool and the upper pool. The third merit is robustness against external events such as a large commercial airplane crash owing to the reinforced concrete USCV. The Mark S containment is applicable to a large reactor that generates 1830 MW electric power. The paper presents several examples of BWRs that use the Mark S containment. In those examples active safety systems and passive safety systems function independently and constitute in-depth hybrid safety (IDHS). The concept of the IDHS is also presented in the paper.     

Two types of a passive safety containment for a near future BWR with active and passive safety systems

The paper presents two types of a passive safety containment for a near future BWR. They are named Mark S and Mark X containment. One of their common merits is very low peak pressure at severe accidents without venting the containment atmosphere to the environment. The PCV pressure can be moderated within the design pressure. Another merit is the capability to submerge the PCV and the RPV above the core level. The third merit is robustness against external events such as a large commercial airplane crash. Both the containments have a passive cooling core catcher that has radial cooling channels. The Mark S containment is made of reinforced concrete and applicable to a large power BWR up to 1830 MWe. The Mark X containment has the steel secondary containment and can be cooled by natural circulation of outside air. It can accommodate a medium power BWR up to 1380 MWe. In both cases the plants have active and passive safety systems constituting in-depth hybrid safety (IDHS). The IDHS provides not only hardware diversity between active and passive safety systems but also more importantly diversity of the ultimate heat sinks between the atmosphere and the sea water. Although the plant concept discussed in the paper uses well-established technology, plant performance including economy is innovatively and evolutionally improved. Nothing is new in the hardware but everything is new in the performance.     

Recent Results in JT-60 Advanced Experiment JT-60 Team,JAERI (Presented by Naoyuki MIYA)

Major efforts in the recent JT-60 experiments have been concentrated on the improved confinement of plasmas with profile control and on the steady state operation study. Peaked density profiles were produced with the successive pellet injection. The energy confinement time was improved by 40% as large as that with the gas fuelled discharges. The fusion products n _e(0)τET _i(0)reached 1.2 × 10²⁰m^?3·s·keV, which was twice that of gas fuelled discharges. High-β_p, plasmas were obtained in low-I _p discharges with improved confinement and a high ion temperature T _i, (0) of 12 keV. The bootstrap current reached 80% of the total plasma current at β_p=3.2. The new concept of a steady-state tokamak power reactor has been proposed on the basis of this result. The maximum current drive efficiency ηCD of 3.4 × 10₁₉m^?2·MA/MW was obtained in the LH current drive experiments. Helium ash exhaust experiments using He-beam injection into H⁺ plasmas showed promising results for α-particle exhaust in a fusion power reactor.     

System-level hazard analysis using the sequence-tree method

A system-level PHA using the sequence-tree method is presented to perform safety-related digital I&C system SSA. The conventional PHA involves brainstorming among experts on various portions of the system to identify hazards through discussions. However, since the conventional PHA is not a systematic technique, the analysis results depend strongly on the experts’ subjective opinions. The quality of analysis cannot be appropriately controlled. Therefore, this study presents a system-level sequence tree based PHA, which can clarify the relationship among the major digital I&C systems. This sequence-tree-based technique has two major phases. The first phase adopts a table to analyze each event in SAR Chapter 15 for a specific safety-related I&C system, such as RPS. The second phase adopts a sequence tree to recognize the I&C systems involved in the event, the working of the safety-related systems and how the backup systems can be activated to mitigate the consequence if the primary safety systems fail. The defense-in-depth echelons, namely the Control echelon, Reactor trip echelon, ESFAS echelon and Monitoring and indicator echelon, are arranged to build the sequence-tree structure. All the related I&C systems, including the digital systems and the analog back-up systems, are allocated in their specific echelons. This system-centric sequence-tree analysis not only systematically identifies preliminary hazards, but also vulnerabilities in a nuclear power plant. Hence, an effective simplified D3 evaluation can also be conducted.     

Model extension and improvement for simulator-based software safety analysis

One of the major concerns when employing digital I&C system in nuclear power plant is digital system may introduce new failure mode, which differs with previous analog I&C system. Various techniques are under developing to analyze the hazard originated from software faults in digital systems. Preliminary hazard analysis, failure modes and effects analysis, and fault tree analysis are the most extensive used techniques. However, these techniques are static analysis methods, cannot perform dynamic analysis and the interactions among systems. This research utilizes “simulator/plant model testing” technique classified in (IEEE Std 7-4.3.2-2003, 2003. IEEE Standard for Digital Computers in Safety Systems of Nuclear Power Generating Stations) to identify hazards which might be induced by nuclear I&C software defects. The recirculation flow system, control rod system, feedwater system, steam line model, dynamic power-core flow map, and related control systems of PCTran–ABWR model were successfully extended and improved. The benchmark against ABWR SAR proves this modified model is capable to accomplish dynamic system level software safety analysis and better than the static methods. This improved plant simulation can then futher be applied to hazard analysis for operator/digital I&C interface interaction failure study, and the hardware-in-the-loop fault injection study.