Transition from depressurization to long term cooling in AP600 scaled integral test facilities

A novel light water reactor design called the AP600 has been proposed by the Westinghouse Electric Corporation. In the evaluation of this plant’s behavior during a small break loss of coolant accident (LOCA), the crucial transition to low pressure, long-term cooling is marked by the injection of the gravitationally driven flow from the in-containment refueling water storage tank (IRWST). The onset of this injection is characterized by intermittency in the IRWST flow. This happens at a time when the reactor vessel reaches its minimum inventory. Therefore, it is important to understand and scale the behavior of the integral experimental test facilities during this portion of the transient. The explanation is that the periodic liquid drains and refills of the pressurizer are the reason for the intermittent behavior. The momentum balance for the surge line yields the nondimensional parameter controlling this process. Data from one of the three experimental facilities represent the phenomena well at the prototypical scale. The impact of the intermittent IRWST injection on the safe plant operation is assessed and its implications are successfully resolved. The oscillation is found to result from, in effect, excess water in the primary system and it is not of safety significance.     

Waste disposal aspects of the long term cooling characteristics of irradiated nuclear fuels

A study has been made of the long term cooling characteristics of nuclear fuels irradiated in commercial reactor designs of interest within the U.K. In the case of thermal reactors, Magnox, AGR, SGHWR, LWR and HTR systems fuelled with either natural or enriched uranium are considered, together with a fast reactor fuelled with plutonium derived either from the Magnox or the AGR programmes. Alternative uses for Magnox plutonium are considered by simulating a plutonium fuelled HTR thermal system and the development of a $\text{Th}{}^{233}\text{U}$ fuel cycle has been anticipated for both a fast reactor and an HTR.For each system the activities as a function of cooling time are considered on the assumption of U/Pu recovery from the fuel during reprocessing within a year of discharge from the reactor and for the alternative case of no U/Pu extraction. The reprocessing waste products associated with the various fuel cycles have then been compared both on the basis of decay heating and radiological hazard per GW(e) yr. Finally, recycling of transplutonium elements is also considered with a view to reducing the long term heating commitment from the higher actinides.     

压水堆释放源项快速估算程序开发

事故时向环境释放的源项是确定核电厂(NPP)应急响应水平和防护行动决策的重要依据。基于电厂工况估算源项是核电厂严重事故应急响应期间重要的应急评价内容之一。在国际原子能机构(IAEA)和美国核管会(NRC)的有关技术文档基础上,本文介绍了基于压水反应堆(PWR)工况进行事故释放源项估算的步骤和基础数据,并归纳了7种实用的事故释放源项估算方法。基于这些方法,开发了PWR事故时环境释放源项快速估算程序。该程序为不同估算方法提供4种释放途径:安全壳泄漏、安全壳旁通、蒸汽发生器传热管破裂(SGTR)和直接环境释放,除直接环境释放途径外,其他释放途径都估算了核素释放过程中的衰变、滞留、喷淋和过滤等减弱过程。对比发现,软件计算结果与美国核管会的RASCAL软件释放源项计算结果接近。     

Scaling methodology for a reduced-height reduced-pressure integral test facility to investigate direct vessel injection line break SBLOCA

A scaling methodology for a small-scale integral test facility was investigated in order to analyze thermal-hydraulic phenomena during a DVI (direct vessel injection) line SBLOCA (small break loss-of-coolant accident) in an APR1400 (advanced power reactor 1400 MWe) pressurized water reactor. The test facility SNUF (Seoul National University Facility) was utilized as a reduced-height and reduced-pressure integral test loop. To determine suitable test conditions for simulating the prototype in the SNUF experiment, the energy scaling methodology was propose to scale the coolant mass inventory and the thermal power for a reduced-pressure condition. The energy scaling methodology was validated with a system code (MARS) analysis for an ideally scaled-down SNUF model and that predicted a reasonable transient of pressure and coolant inventory when compared to the prototype model. For the actually constructed SNUF, the effect of scaling distortions in the test facility's thermal power and the loop geometry was analytically investigated. To overcome the limitation of the thermal power supply in the facility, the convective heat transfer between primary and secondary systems at the steam generator U-tubes was excluded and a modified power curve was applied for simulating the core decay heat. From the code analysis results for the actual SNUF model, the application of the modified power curve did not affect the major events occurring during the transient condition. The results revealed that the scaling distortion in the actual SNUF geometry also did not strongly disturb significant thermal-hydraulic phenomena such as the downcomer seal clearing. Thus, with an adoption of the energy scaling methodology, the thermal-hydraulic phenomena observed in the SNUF experiment can be properly utilized in a safety analysis for a DVI line break SBLOCA in the APR1400.     

Numerical simulation of a blanket cooling system for fusion reactor based on PWR conditions

The simulations of a blanket cooling system were presented to address the choice of cooling channel geometry and coolant input data which are related to blanket engineering implementation. This work was performed using computer aided design (CAD) and computational fluid dynamics (CFD) technology. Simulations were carried out for the blanket module with a size of 0.6 m × 0.45 m in toroidal plane, and the nuclear heat was applied on the cooling system at P_n (neutron wall load) of 5 MW/m². The structure factors and input data of hydraulics were investigated to explore the optimal parameters to match the PWR condition. It was found that the inlet velocity of first wall (FW) channel should be within the range of 2.48–3.34 m/s. As a result, the temperature rise (TR) of the coolant in the FW channel would be 24–25 K. This leads to the remaining space for TR within the range of 15 K in the piping circuits. It also indicated that the FW plays an important role in TR (reaches 60% of the whole cooling system) due to its high level of P_n and heat flux in the zones. It was predicted that the nuclear heat inside blanket module could be removed completely by the piping circuits with an acceptable pipe bore and the related input data. Finally, a possible design range of cooling parameters was proposed in view of engineering feasibility and blanket neutronics design.     

The contribution of non-wetting droplets to direct cooling of the fuel during PWR post-LOCA reflood

During the reflood of a Pressurised Water Reactor (PWR) following a loss of coolant accident, precursory cooling prior to the arrival of the rewetting front is of vital importance in limiting the rise in cladding temperature before rewet. This precursory cooling is achieved by a flow of superheated vapour, with entrained saturated drops, which evaporate into the vapour and act as a heat sink. In this paper we investigate a complementary mechanism; the direct cooling of the cladding by the drops themselves. Cladding temperatures are such that wetting by these droplets does not occur. On the contrary, droplets bounce off a vapour cushion formed during the ∼10 ms or so that they are in close proximity to the cladding. Using a combination of previous experimental correlations and recent CFD calculations, we estimate the rate of heat removal from the cladding surface as a result of the droplet impingement. Thus, we estimate the heat removed as a result of one impingement and estimate the total rate of heat removal by estimating the number droplets impinging on the cladding per unit surface area. The heat extracted by those droplets is found to be about 1/10 of the heat extracted by single-phase vapour under typical reflood conditions. Though there significant uncertainties in these estimates, it does seem that direct cooling by droplets, not generally incorporated in analyses of reflood, could actually be making a significant contribution to keeping cladding temperatures down to acceptable levels.     

Evaluation of ASME code flaw analysis procedure using the influence function method for application to PWR primary piping

This paper discusses stress intensity factor calculations and fatigue analysis for a PWR primary coolant piping system. The influence function method is applied to evaluate ASME Code Section XI Appendix A “analysis of flaw indication” for the application to a PWR primary piping. Results of the analysis are discussed in detail.     

A containment analysis for SBLOCA in the refurbished Wolsong-1 Nuclear Power Plant

A small break leading to a loss of coolant accident (SBLOCA), being one of the topic accidents in the nuclear plant diagnosis in recent years, has been analysed and evaluated for the refurbished Wolsong-1 Nuclear Power Plant (NPP). The Industry Standard Toolset (IST) codes developed by CANDU (Canadian Deuterium Uranium reactor) Owners Group and updated models including design change parameters are applied newly to the event analyses. The GOTHIC code has been used for the containment thermal-hydraulic analysis of Wolsong-1. Also, the SMART-IST code fitted in the Iodine Chemistry (IMOD-2) model has been used to predict nuclide behavior within the containment considering various aspects. The IMOD-2 was incorporated into SMART-IST as a module dealing with chemical transformations and mass transfer of iodine species in containment. IMOD-2 model is very sensitive to paint and chemicals. The parametric studies for the IMOD-2 model are performed to decide the analysis value set. The iodine release amount increases as the paint thickness increases. But, the iodine release amount increases as the water pH (dousing and primary heat transport (PHT)) decreases. The developed containment analysis methodology and the results of SBLOCA without Emergency Coolant Injection (ECI) are presented herein. Under the most heat-up conditions, the radionuclide release from the failed fuel into the containment and subsequently to the environment is such that the radioactive doses to the public are below the acceptable limits.     

Thermal-hydraulic analysis on the effect of operator action in the loss of shutdown cooling system accident in a PWR

It is necessary to develop PSA methodology and integrated accident management technology during low power/shutdown operations. To develop this technology, thermal-hydraulic analysis is necessarily required to access the trend of plant process parameters and operator's grace time after initiation of the accident. In this study, the thermal-hydraulic behavior in the loss of shutdown cooling system accident during low power/shutdown operations at the Korean standard nuclear power plant was analyzed using the best-estimate thermal-hydraulic analysis code, MARS2.1. The effects of operator's action and initiation of accident mitigation system, such as safety injection and gravity feed on mitigation of the accident during shutdown operations are also analyzed.When steam generators are unavailable or vent paths with large cross-sectional area are open in the accident, the core damage occurs earlier than the cases of steam generators available or vent paths with small cross-sectional area. If an operator takes an action to mitigate the accident, the accident can be mitigated considerably. To mitigate the accident, high-pressure safety injection is more effective in POS4B and gravity feed is more effective in POS5. The results of this study can contribute to the plant safety improvement because those can provide the time for an operator to take an action to mitigate the accident by providing quantitative time of core damage. The results of this study can also provide information in developing operating procedure and accident management technology.     

压水堆吊篮下部防断支承组件流致振动分析

对秦山核电厂堆芯下腔流场、堆内下部防断支承组件振动特性及全组件的流致振动进行了分析,特别对旋涡脱落致振进行了定量分析.分析结果表明防断支承组件初始结构的整体转动振动的固有频率与旋涡脱落频率相差较大,发生大幅振动的可能性不大;只有当部分连接件松动,整体结构转动振动的固有频率下降时,才很有可能发生大幅振动.     

Experimental validation of the TASS/SMR code for an integral type pressurized water reactor

The transient and setpoint simulation small and medium reactor (TASS/SMR) code has been applied to perform the safety analysis and performance evaluation of an integral type pressurized water reactor. Till now, the code has only been verified by using simplified and analytical problems as well as a reliable system code due to the lack of available experimental data. Recently, several kinds of experiments have been performed by focusing on an identification of the heat transfer characteristics at a heat sink and source, and the thermal hydraulic characteristics and the natural circulation performance in an integral effect test facility. In this paper, the TASS/SMR code has been validated by using the experimental data obtained from a separate effect test facility by focusing on the heat transfer characteristics and an integral effect test facility by focusing on the thermal hydraulic characteristics and the natural circulation performance. According to the validation results of the TASS/SMR code against the separate effect test and the integral effect test, the code predicts the overall variation of the thermal hydraulic parameters well, including the system pressure, fluid temperature, mass flow rate, etc., and it is applicable for the safety analysis and performance evaluation of an integral type pressurized water reactor.     

先进非能动压水堆核电站燃料组件轴向燃耗分布研究

燃料组件的轴向燃耗分布以及末端效应是燃耗信任制技术应用中的难点。基于先进非能动压水堆核电站的运行模式及组件设计特点,结合可能的燃料管理策略,统计轴向两端使用低富集度抑制区的乏燃料组件,生成轴向燃耗包络线。以AP1000堆型的乏燃料贮存单元为例,通过分析统计,证明生成的轴向燃耗包络线用于临界安全分析是保守的。在此基础上,详细研究燃料组件顶部的低富集度抑制区对末端效应的贡献,并对燃料组件进行设计改进,减小至消除末端效应,为简化乏燃料组件相关的临界安全分析提供了一个方法。相关研究工作及成果,是先进非能动压水堆核电站乏燃料组件相关的设施设备的临界安全设计的基础,可为其他堆型的相应研究提供参考和借鉴。     

A new procedure to generate resonance integral table with an explicit resonance interference for transport lattice codes

Resonance interference could not be considered explicitly in the conventional resonance treatment employing subgroup and direct resonance integral methods when using coarse energy group structure. This problem comes from the lack of information for the resonance shapes of resonant nuclides in the resonance interference formulas. As energy group boundaries get coarser, inaccuracy in estimating self-shielded cross sections with resonance interference gets bigger. A new method has been proposed to conserve the self-shielded cross sections for each group through an explicit consideration of resonance interference effect, which results in a good accuracy in predicting the multiplication factor. This method can be applicable to various mixing combinations of constituent resonant nuclides with resonance interference and can cover wide dilution range. The MERIT code has been used to generate resonance integral tables and to estimate resonance interference effects. And the 2-D transport lattice code KARMA has been used to perform sample calculations to see the effectiveness of the newly developed method. Sample calculations have been performed for single pins with various temperatures, ²³⁵U enrichments and dilution levels with the 47 and 190 energy group structures. The computational results show that this method is able to estimate self-shielded cross sections in each coarse energy group accurately for various temperatures and various geometry and composition configurations.     

Severe accident analysis for a typical PWR using the MELCOR code

Using the MELCOR code, we simulated and analyzed a severe accident at a Chinese pressurized reactor 1000-MW (CPR1000) power plant caused by station blackout (SBO) with failure of the steam generator (SG) safety relief valve (SRV). The CPR1000 response and results for three different scenarios were analyzed: (i) seal leakage and an auxiliary feed water (AFW) supply; (ii) no seal leakage or AFW supply; and (iii) seal leakage but no AFW supply. The results for the three scenarios are compared with those for a simple SBO accident. According to our calculations, the SG SRV stuck in the open position would greatly accelerate the sequence for a severe accident. For an SBO accident with the SRV stuck open without seal leakage or an AFW supply, the pressure vessel would fail at 9576 s and the containment system would fail at 124,000 s. If AFW is supplied, pressure vessel failure would be delayed nearly 30000 s and containment failure would delay at least 50000 s. When seal leakage exists, pressure vessel failure is delayed about 50 s and containment failure time would delay about 30000 s. The results will be useful in gaining an insight into the detailed processes involved and establishing management guidelines for a CPR1000 severe accident.     

A non linear model for the PWR fuel assembly seismic analysis

In seismic PWR core analysis, the non linear behavior of fuel assemblies has to be studied taking into account variations of stiffness and damping due to the slippage of rods in spacer grids. Based on the linear CEA’s assembly model, principally composed of two beams representing sets of rods and thimbles, a linear model is described. In the second part of this paper the slippage and loss of contact of rods on a grid relic is analyzed. Based on the Coulomb friction model and plasticity analogy, a grid model constituted of elastoplastic hinge and non linear springs is proposed to simulate the progressive slippage of rods inside the cells. In the third part, a non linear assembly model is designed and validated with experimental data. This model represented with success the increase of damping and decay of frequency when the magnitude of the dynamic loading increases.     

第三代压水堆核电站厂房辐射监测系统比较

以防城港核电一期CPR1000堆型、台山核电CEPR堆型和三门核电AP1000堆型为例, 比较分析了第三代压水堆核电站辐射监测系统的结构、功能及各自的特点, 对核电站厂房辐射监测系统的发展趋势做了分析讨论。     

国内压水堆核电厂安全壳喷淋试验的比较分析

以秦山第二核电厂1、2号机组和田湾核电厂1、2号机组为例,对国内已运行的压水堆核电厂在调试阶段进行安全壳喷淋试验时所采用的两种方式进行了比较分析,总结出两种喷淋试验方法各自的优缺点,可为国内新建核电机组的安全壳喷淋系统的调试试验提供借鉴。     

压水堆核电站的用钢量及要求

文中简要介绍了西德压水堆核电站机械设备的分级,用钢量及要求。