Chemical cleaning of PWR steam generators with a low temperature process

Corrosion products generated in the steam, feedwater and condensate systems of a PWR will be transferred by the feedwater into the secondary side of the steam generators (SG). Up to several hundred kilograms of deposits may collect on the surfaces of the SG. These deposits not only reduce the efficiency of the SG by deterioration of the heat transfer, but also cause an acceleration of the corrosion of the SG tubes and blockage of support plate passages.The chemical removal of the corrosion products opens the possibility to eliminate causes of defects on heat transfer tubes by removing the corrosion products and trapped impurities.A low temperature cleaning process, based on the EPRI developed SGOG iron and copper solvents, was applied in November 1990 to remove the hard deposits from the tube sheets of a two-loop plant. The sludge containing over 60% copper was removed with the application of one iron removal step and several copper removal steps. Over 95% of the available sludge was removed. The corrosion of the unalloyed and low alloy materials was extremely low. The Incoloy 600 tubes showed no corrosion.In addition aspects of crevice cleaning at elevated temperatures are mentioned.     

高温气冷堆螺旋管式直流蒸汽发生器热工水力学

高温气冷堆蒸汽发生器具有一次侧氦气工质、二次侧直流、螺旋管结构、工作温度高等特点，其热工水力特性与传统压水堆自然循环蒸汽发生器存在很大区别。针对高温气冷堆蒸汽发生器的特点，对其基础热工水力及特有热工水力学问题进行了阐述，主要包括螺旋管内单相及两相流阻及换热计算、横掠螺旋管束流阻及换热计算、温度均匀性及两相流不稳定性等。同时介绍了清华大学核能与新能源技术研究院针对高温气冷堆蒸汽发生器热工设计、温度均匀性及两相流不稳定性等热工水力学问题所开发的一维稳态程序、一维瞬态程序、二维分析程序和方法，并对分析结果和结论进行了讨论。相关研究方法、程序和结论对其他相似参数螺旋管和直管式直流蒸汽发生器具有参考和借鉴意义。     

M310机组蒸汽发生器传热管5%破损率量化研究

M310核电机组《化学和放射化学技术规范》第3篇《放射化学规范》中规定,当“133Xe>92500 MBq/t或133Xe>37000 MBq/t和131I/133I>1.5”时,执行“如果至少1台蒸汽发生器（SG）传热管破损率超过5%,则应以50 MW/min 速率降负荷到NS/SG模式”指令。在实际运行过程中,由于无法判断传热管破损率是否超过5%,故无法确定是否执行“以50 MW/min 速率降负荷到NS/SG模式”的指令,因此,M310核电机组《放射化学规范》中该运行指令不具备可操作性。本文对国内外相关放射化学运行指令进行调研,充分理解“传热管破损率超过5%”的含义,并进行量化分析。通过对SG传热管在5、24、44 L/h泄漏率下的裂纹稳定性进行分析,发现在这些泄漏率下单根传热管都不会发生失稳断裂,进而给出了一个具有可操作性的建议,用指标“如果至少1台SG一次侧向二次侧的泄漏率超过5 L/h”,代替无法量化的“如果至少一台SG传热管破损率超过5%”,可保证核电厂运行的安全性和经济性。      

Predicting growth rate of wall thinning for severely degraded SG tubes using a statistical methodology

Based on the requirements of TSTF-449 or NEI 97-06, operational assessment (OA) should be performed to guarantee the steam generator (SG) tube integrity. OA is a forward looking evaluation of the SG tube conditions. One of main evaluations for the OA is to estimate the growth rate of tube degradation prior to the next SG tube inspection. Therefore, the majority of this paper is to predict the growth rate of wall thinning for the SG tubes by way of a statistical methodology. The wall thinning of degraded SG tubes predicted by the present model agrees well with the plant measured one. The relative errors between the predictions and measurements are less than 10%. In addition, the present model would over-predict the wall thinning in most cases, revealing that this methodology could provide a useful and conservative tool for the PWR plant staff to execute the OA for SG tubes.     

Applicability of Chemical Cleaning Process to Steam Generator Secondary Side, (IV)

The application of chemical cleaning for dissolving and removing scale and sludge is being planned in the Japanese pressurized water reactor (PWR) plant in order to maintain high heat transfer performance and to prevent steam generator (SG) tube degradation. In this paper, the effectiveness of the Electric Power Research Institute (EPRI) and German Kraftwerk Union (KWU) processes on the integrity of structural materials other than SG tubes and the comprehensive applicability of chemical cleaning are discussed. The integrity of structural materials such as carbon steel, low-alloy steel and stainless steel was maintained after the EPRI and KWU processes. KWU chemical cleaning tailored for crevice cleaning has been studied to improve its cleaning effectiveness in crevices and to control the corrosion depth of structural materials less than the criterion for corrosion depth.     

Thermal-Hydraulic Study of Integrated Steam Generator in PWR

One of the safety aspects of innovative reactor concepts is the integration of steam generators (SGs) into the reactor vessel in the case of the pressurized water reactor (PWR). All of the reactor system components including the pressurizer are within the reactor vessel in the SG integrated PWR.

The simple heat transfer code was developed for the parametric study of the integrated SG. The code was compared to the once-through 19-tube SG experiment and the good agreement between the experimental results and the code predictions was obtained. The assessed code was used for the parametric study of the integrated once-through 16 m-straight-tube SG installed in the annular downcomer. The proposed integrated SG as a first attempt has approximately the same tube size and pitch as the present PWR and the SG primary and secondary sides in the present PWR is inverted in the integrated PWR. Based on the study, the reactor vessel size of the SG integrated PWR was calculated.     

钠冷快堆蒸汽发生器小钠水反应现象数值模拟

蒸汽发生器（SG）是钠冷快堆二回路主冷却系统的关键设备之一，其传热管破损会导致钠水反应事故，产生大量氢气、腐蚀性产物并放出热量，严重影响SG的安全运行。本文用FLUENT对小泄漏钠水反应区的瞬态现象进行数值模拟，计算得到泄漏孔径为0.2 mm时反应区最高温度可达1 564 K，最高温度随泄漏率的增加而升高，但保持在一定范围内，结果均与日本实验结果吻合，并且泄漏率会影响产物NaOH和H₂的扩散与分布。本文采用的数值模拟方法可用于小钠水反应现象分析，可得到不同泄漏率下小钠水反应能达到的最高温度、反应区任意位置的NaOH浓度和H₂浓度，以预测邻管损耗和失效时间，有利于进一步开展小钠水反应事故安全分析。     

Safety and availability of steam generator tubes affected by secondary side corrosion

The outside diameter stress corrosion cracking at tube support plates became the dominating ageing mechanism in steam generator tubes made of Inconel 600. A variety of maintenance approaches were developed and implemented world-wide to enable safe and reliable plant operation with affected tubes. Despite different philosophical and physical backgrounds involved, all applied approaches satisfy relevant regulatory requirements. The main goal followed in this paper is to quantify the degree of safety which is achieved through the implementation of selected maintenance approaches. A method is proposed which measures the operational safety and availability through three efficiency parameters: probability of steam generator tube rupture; predicted accidental leak rates through the defects in the tube bundle; and number of plugged tubes. An original probabilistic model quantifies the probability of tube rupture, while procedures available in literature were used to evaluate the accidental leak rates. A numerical example is based on data from the Kr ko NPP (PWR 623 MWe). The maintenance strategies analyzed are: (a) traditional defect depth (40%) plugging criterion; (b) alternate plugging criterion (bobbin coil voltage as defined by EPRI and US NRC); (c) combination of traditional and alternate plugging criteria; and (d) no plugging at all. Advantages of the defect specific approaches (b) and (c) over the traditional one (a) are clearly shown. The efficiency of the traditional approach (a) is shown to be comparable to the no plugging at all approach (d). Finally, a sensitivity analysis aimed at ranking of the input parameters is presented. Uncertain failure models are shown to be the major contributor to the scatter of obtained results.     

A simplified model of steam generators for digital simulation

This paper develops a simplified model of a PWR steam generator. A computer programme for the steady state operation was developed, which will be useful for the dynamic analysis for describing accident situations. The model incorporates all the various flow regimes and heat transfer regimes that are likely to be encountered by the secondary flow of the steam generator. The primary flow is considered as single phase compressed liquid. Given the heat transfer area, pitch and the size of the tubes the computer programme matches the total power generated within five percent accuracy. Detailed pressure and temperature distributions along the length of the preheater and evaporator are also computed.     

基于CFD的蒸汽发生器U型管内回流现象数值分析

反应堆一回路系统在自然循环条件下，蒸汽发生器（SG）部分U型管内可能会出现回流现象，利用计算流体动力学（CFD）方法，对某非能动三代反应堆蒸汽发生器U型管内流体的流动传热特性进行数值模拟分析。选取6组不同管长的U型管，对比分析U型管内单相流体的流动传热特性。基于数值仿真结果，得出6组U型管质量流量-进出口压降曲线，并?T分析了U型管长度和一次侧进口流体温度与二次侧壁面温度温差（?T）对流体回流的影响。研究结果表明，当?T一定时，随着进出口压降的降低，长管内更容易发生回流。当U型管长度一定时，?T越小越容易发生回流。      

Prediction of counter-current flow limitation at hot leg pipe during a small-break LOCA

The possibility of hot leg flooding during reflux condensation cooling after a small-break loss-of-coolant accident in a nuclear power plant is evaluated. The vapor and liquid velocities in hot leg and steam generator tubes are calculated during reflux condensation cooling with the accident scenarios of three typical break sizes, 0.13, 1.02 and 10.19% cold leg break. The effect of initial water level to counter-current flow limitation is taken into account. It is predicted that the hot leg flooding is precluded when all steam generators are available for heat removal. It is also shown that both hot leg flooding and SG flooding are possible under the operation of one steam generator. Therefore, it can be said that the occurrence of hot leg flooding under reflux condensation cooling is possible when the number of steam generators available for heat removal is limited.     

压水堆蒸汽发生器一、二次侧稳态流场耦合分析

蒸汽发生器（SG）在运行过程中主要面临流致振动所导致的传热管破裂事故,而流致振动分析需以SG内的三维两相流场作为输入条件。采用多孔介质模型,对SG二次侧流场进行求解,同时耦合一、二次侧换热,获得SG二次侧速度场、温度场、压力场及流动含气率分布,并获得传热管一维的一、二次侧流体温度和换热系数及传热管温度分布。由于一次侧向二次侧释热极不均匀,SG内流场分布及汽水分离器内的含气率分布极不均匀;汽水分离器内的最大、最小含气率分别为0.62和0.05,该参数可为汽水分离器负载设计提供依据。通过计算还获得弯管区速度分布,该分布可为传热管的流致振动磨损评估提供输入条件。     

传热管与支撑件随机碰撞实验研究

针对蒸汽发生器中传热管与支撑件的碰撞行为,对悬臂梁固定的传热管在不同支撑条件下开展了激振实验,获得了传热管均方根位移与接触率,分析了传热管与支撑件磨损功率的变化规律,并探究了传热管固有频率对振动特性的影响。结果表明,防振条支撑与波纹带支撑时传热管的法向均方根位移均随激振力增加逐渐放缓,而防振条支撑对应的切向位移呈线性增长。防振条支撑与波纹带支撑时的接触率均表现为随激振力增大趋于稳定,其中间隙对防振条支撑的接触率影响更明显。在以冲击为主导的激励方式下,激振力与磨损功率表现为明显的正相关。支撑间隙对磨损功率的影响相对复杂,防振条支撑下磨损功率在0.1 mm和0.25 mm间隙存在极值,而波纹带支撑磨损功率仅在0.2 mm间隙存在极值。传热管固有频率对振动响应结果的影响很小。     

压水堆核电站常规运行工况下气载放射性物质释放源项的计算

本文介绍一个自行编制的用于计算压水堆核电站在常规运行工况下气载放射性物质向环境释放量的计算机程序ＭＧＡＬＥＳ。采用ＯＲＩＧＥＮ２程序，根据燃料元件的成份和燃耗情况计算堆芯的放射性核素谱；用放射性物质经堆芯向一回路迁移的逃脱率系数计算一回路冷却剂中的放射性核素浓度；再考虑核电站实际运行过程中一、二回路冷却剂的泄漏以及通风、除气等过程，计算其正常运行工况下气载放射性物质向环境的释放量。     

Experiment and Verification Test of the Once-through Steam Generator of the 10 MW High-temperature Gas-cooled Reactor Flow Stability of the Once-through Steam Generator

In January 2003, the 10MW High-temperature Gas-cooled Reactor (HTR-10) reached its full power for continuous operation of seventy-two hours in the Institute of Nuclear Energy Technology, Tsinghua University. The reactor operated smoothlyqbthe design parameters were successfully attained.

The once-through steam generator (SG) is one of key equipments of the HTR-10 reactor. The SG includes 30 modular heating helical tube assemblies. There are two thermal hydraulic requirements to be satisfied for the once-through steam generator: (1) enough heat transfer surface; (2) qualified steam can be produced under rated electrical generation power, and water-steam two phase flow un-stability can be avoided. In order to obtain the thermal hydraulic characteristics of the SG reliably, before design, a numerical code was developed for the design, and a full-scale test loop with two heating tubes as model was established, and series experiments had been carried out.

The purpose of this paper is to introduce the design of SG and researches on the stability of small bending radius helical coil-pipe used in HTR-10, for exempla, the effects of outlet steam pressure, inlet water sub-cooling degree, thermal power and inlet throttling degree. Up to now, the SG has experienced full power operation smoothly, and approvingly reached its original design requirements.

In the paper, some operational experimental data of the HTR-10 S.G have been presented.     

蒸汽发生器换热管流量分配及其对核主泵入口流场的影响

为研究蒸汽发生器（SG）换热管流量分配及其对反应堆冷却剂泵（RCP）入口流场的影响，进行了蒸汽发生器的缩尺模型冷态实验，并以实验获得的数据为SG下封头的入流条件，对SG下封头进行数值建模，并采用计算流体力学（CFD）方法对其进行了三维流场计算。结果表明:SG换热管存在较严重的流量分配不均，SG入口管会对其所正对部分的换热管的流量分配产生较大影响，使该部分流量增大，即形成高速区，而高速区周围会形成相对的低速区甚至回流区；在小流量时，换热管的沿程损失将对换热管的流量分配起主导作用；SG换热管内的不均匀流量分配会使SG出口管处的轴向速度更加紊乱，即在核主泵入口产生更加严重的入流畸变。      

An aging assessment on ex-service Alloy 800 steam generator tubing

To support SG life extension and plant life management, an aging assessment was performed on a number of ex-service Alloy 800 steam generator (SG) tubes removed from three CANDU^®1 stations with service life spanning from 2 to 27 years. Laboratory tests and examinations were carried out to investigate the potential aging mechanisms of SG tubing. High-temperature electrochemical experiments were performed under simulated SG secondary side crevice chemistry conditions to determine the corrosion susceptibility of the ex-service tubing; metallurgical examinations were carried out to check the chemical compositions, grain size, and hardness of the ex-service tubing materials and secondary ion mass spectrometry (SIMS) analysis was performed to assess the potential surface chromium depletion and grain boundary segregation of the ex-service tubing. Based on the results from the assessment, no increase in the corrosion susceptibility or changes in metallurgical properties of the ex-service tubes resulting from aging were observed. SIMS top-down profiles did not detect any aging-related surface chromium depletion on any of the ex-service tubes. However, SIMS imaging performed on the polished cross-sections of the ex-service tubes observed boron precipitation at the grain boundaries. Since no archived tubing with the same heat number as that of the ex-service tubing is available for comparison, whether the boron precipitation at grain boundaries is attributed to aging through SG operation is not conclusive and needs further clarification. In addition, the impact of this boron precipitation on the integrity of Alloy 800 SG tubing needs further investigation.     

Experiments on the Heat Transfer and Natural Circulation Characteristics of the Passive Residual Heat Removal System for an Advanced Integral Type Reactor

Experiments on the heat transfer characteristics and natural circulation performance of the passive residual heat removal system (PRHRS) for the SMART-P have been performed by using the high temperature/high pressure thermal-hydraulic test facility (VISTA). The VISTA facility consists of the primary loop, the secondary loop, the PRHRS loop, and the auxiliary systems to simulate the SMART-P, a pilot plant of the SMART. The primary loop is composed of the steam generator (SG) primary side, a simulated core, a main coolant pump, and the loop piping, and the PRHRS loop consists of the SG secondary side, a PRHRS heat exchanger, and the loop piping. The natural circulation performance of the PRHRS, the heat transfer characteristics of the PRHRS heat exchangers and the emergency cooldown tank (ECT), and the thermal-hydraulic behavior of the primary loop are intensively investigated. The experimental results show that the coolant flows steadily into the PRHRS loop and that the heat transfers through the PRHRS heat exchanger and the emergency cooldown tank are sufficient enough to enable a natural circulation of the coolant. The results also show that the core decay heat can be sufficiently removed from the primary loop with an operation of the PRHRS.     

Thermal hydraulic behaviors during loss of RHR system at mid-loop operation of Chinese 300 MWe PWR NPP

Thermal hydraulic behavior of nuclear power plant (NPP) is analyzed by using mechanistic computer code for loss of residual heat removal (RHR) system during mid-loop operation of Chinese 300 MWe two-loop pressurized water reactor is presented. In the absence of recovery of RHR or other accident management measures, the reactor core will be uncovered for a long term resulting in core heat-up, degradation and relocation to the lower plenum. The effectiveness of available mitigate measures, such as safety injection system, gravity feed from refueling water storage tank (RWST) and steam generator (SG) reflux-condensation, are investigated. Coolant injection is highly effective in halting the accident progression and make the core recovered. The cooling capability of SG reflux-condensation has a relationship with different availabilities of steam generators and decay heat power. 6 days after shutdown, 2SG operation can keep the water level at mid-line of hot leg. 12 days after shutdown, both 2SG operation and 1SG operation can keep the water level at mid-line of hot leg. The analyses also indicate that the cooling mechanism of safety injection system is more effective than gravity feed from RWST and SG reflux-condensation. Through confirming the success criteria of SG reflux-condensation, time windows can be devided. Then, event trees for loss of RHR system under mid-loop operation are built with considering the analysis results and abnormal procedure.     

TA16传热管的断裂韧度测试方法研究

近年来，以蒸汽发生器传热管为代表的小尺寸管材的断裂韧度评价方法受到了日益关注。本文设计了一种可用于TA16传热管断裂韧性测试的含径向裂纹C形试样，基于弹塑性有限元分析获得试样的应力强度因子K和J积分的计算式。采用规则化法完成了TA16传热管的断裂韧度试验，试验结果表明，不同试样得到的J阻力曲线和条件启裂韧度J_Q的分散性均较小，J_Q均值为32.875 MPa•mm，标准差为1.377 MPa•mm。