10MW高温气冷堆气体采样分析系统研究

10MW高温气冷堆的氦气净化系统由氧化铜床、分子筛床、低温吸附器等主要净化设备及其它辅助设备组成,气体采样分析系统由气相色谱仪,湿度计、红外分析仪组成。在投入HTR-10使用中,其湿度计和红外分析仪均能达到设计要求,实现了对反应堆一回路氦气中H2O,CO,CO2的连续监测。其气相色谱仪满足设计要求．实现了对反应堆一回路氦气中H2,O2、N2、CH4、CO、CO2的间歇取样分析。     

氦气杂质对中间换热器材料的影响初探

中间换热器是高温气冷堆氦气透平间接循环和高温工艺热应用的关键部件.中间换热器属于一回路压力边界,它将堆芯出口温度达900～1 000 ℃氦气的热量传递给二回路氦气,此外还承受一、二回路氦气压差,因此,目前能够用于中间换热器的耐热金属材料非常有限.高温气冷堆一、二回路氦气中含有H_2、H_2O、CO、CH_4等杂质,在高温下,氦气杂质对中间换热器材料的影响主要是氧化、碳化和脱碳,降低材料的机械性能,其影响不可忽视.对于中间换热器设计,现有规范的温度范围需扩展,氦气杂质对材料强度的影响也需考虑.     

高温气冷堆冷却剂杂质对高温合金材料高温性能影响的研究模型分析

高温气冷堆一回路冷却剂中含有的少量CO、H_2、H_2O、CH_4等杂质,这些杂质对高温堆蒸汽发生器用高温合金的高温性能有重要影响。国外在超高温运行工况下冷却剂杂质对高温合金材料性能的影响方面开展了大量研究,由于研究过程中对试验氦气中痕量的杂质含量控制十分困难,致使相关的研究成果分布比较分散,需要对相关的研究模型进行归纳和分析。高温镍铬合金中的铬是被氧化的主要合金元素,而保护性Cr_2O_3层的形成是合金是否被腐蚀的主要决定因素;铬的稳定相图模型和气体组成三元相图模型是两种被广泛应用的理论模型。本文对铬的这两种理论模型研究方法及其应用情况进行比较和分析。     

1OMW高温气冷实验堆氦气安全阀的设计与性能试验

10Mw高温气冷实验堆(HTR-10)一回路安全泄放系统安装了两台核一级氦气安全阀,对反应堆一回路进行超压保护,是保证HTR-10安全的重要设备之一.本文介绍了氦气安全阀的设计要求、结构特点及性能要求,并按相关规范要求对其性能进行了实验验证.结果表明,安全阀的性能满足设计要求.     

10MW高温气冷实验堆氦气安全阀的设计与性能试验

10MW高温气冷实验堆(HTR-10)一回路安全泄放系统安装了两台核一级氦气安全阀，对反应堆一回路进行超压保护，是保证HTR-10安全的重要设备之一。本文介绍了氦气安全阀的设计要求、结构特点及性能要求，并按相关规范要求对其性能进行了实验验证。结果表明，安全阀的性能满足设计要求。     

固态钍基熔盐堆中~(14)C的产生及释放探讨

熔盐堆作为第四代反应堆论坛推荐的6种候选堆型之一,具有输出温度高、能量密度高、无水冷却等特点。固态钍基熔盐堆(Thorium Molten Salt Reactor with Solid Fuel,TMSR-SF1)堆芯大部分结构材料为石墨,冷却剂杂质及石墨材料中的13C和杂质N、O易被活化产生14C。14C半衰期较长,同其他稳态核素12C、13C一样广泛参与各种复杂的生物循环,在反应堆中受到关注。TMSR-SF1中的14C广泛分布于冷却剂、堆芯石墨结构材料和燃料元件。本文采用输运燃耗耦合方法,应用SCALE6.1的TRITION控制模块对反应堆各区域的14C放射性活度进行计算分析,结果表明,反应堆在正常运行工况下一回路每年产生的14C放射性活度为0.34 TBq,满足现有的压水堆、重水堆管理限值要求。向环境释放的14C主要来自于一回路熔盐中N杂质的活化。     

反应堆一回路的化学去污技术介绍

一、前言随着反应堆运行时间的增长,一回路管道内壁的腐蚀层越来越厚,冷却水中的杂质活化后的产物越来越多地沉积在一回路的管壁上,燃料元件发生破损事故时裂变产物也会进入一回路。因此,一回路管道周围的辐射剂量将会升高,达到一定程度时,就会防碍反应堆正常运行和维修。降低辐射剂量主要有两个途径:一是抑制腐蚀,如水质管理     

Inconel 617合金在非纯氦气环境中的高温腐蚀行为研究

Inconel 617合金是高温气冷堆蒸汽发生器的候选材料,在反应堆超高温运行时可能会受到氦气中痕量杂质的腐蚀。为探究合金在高温堆环境中的腐蚀机理,本研究开展了Inconel 617合金在980℃的非纯氦气中的腐蚀实验,对气相以及腐蚀行为进行了分析。通过化学热力学和动力学计算,阐明了合金脱碳的机理,并建立了碳迁移判定模型和脱碳反应预测模型,与实验数据有良好的一致性。在此基础上,研究了预氧化和温度对脱碳反应的影响。研究结果表明,即使杂质含量极低,也会诱发相关的腐蚀行为。降低运行温度可以有效避免合金脱碳,但预氧化的抗脱碳效果不理想。因此,极低杂质含量并非高温堆一回路净化目标,应该根据模型预测和实验分析来选择更加合理的杂质控制方案。     

高温气冷堆氦气主流区中石墨粉尘运动特性初步分析

以高温气冷堆一回路氦气中石墨粉尘颗粒为研究对象,初步计算分析氦气主流区中单个石墨粉尘颗粒的各种受力.结果表明,各种受力中,气体阻力占绝对主导作用.在分析受力的基础上,推导出一回路中石墨粉尘颗粒的简化运动控制方程,并计算分析石墨粉尘颗粒速度随氦气流速的运动规律:当主流氦气流速度恒定时,石墨粉尘颗粒速度按照负指数的运动规律不断趋近于主流氦气速度；当主流氦气速度随时间发生线性变化时,石墨粉尘颗粒速度与主流氦气速度趋向于一个恒定的差值；在高温气冷堆氦气气轮机循环中,石墨粉尘颗粒从静止到流动时的加速度变化与反应堆功率调节的增减呈现同一性.     

高温气冷堆核电站安全级湿度仪研制

华能石岛湾高温气冷堆核电站与现役压水堆核电站一回路冷却剂不同,其采用氦气做为一回路冷却剂。氦气中水分含量过高可能导致安全事故,在运行时需要实时监控一回路冷却剂中的水分含量,当水分含量超过限值时发出停堆信号传送到保护系统。高温气冷堆湿度仪工作在高压环境下,输出信号与温度、湿度、压力等有关。通过试验压力、温度、湿度等物理量对湿度传感器输出信号的影响的分析研究,建立水分含量补偿模型,研制出基于纯硬件补偿电路的湿度仪样机并通过1E级鉴定试验验证。1E级湿度仪实现了实时监控一回路冷却剂中的水分含量,在水分含量超过限值时触发保护系统实现保护逻辑的目标,为反应堆安全运行提供了技术保障。     

Estimation of the Tritium Behavior in the Pebble Type Gas Cooled Reactor for Hydrogen Production

In Korea, a nuclear hydrogen program has been established to develop and demonstrate mass production system for hydrogen generation. The objective of this study is to establish the evaluation procedure for predicting the tritium behavior in the 300 MWth Pebble type gas cooled reactor which is the one of the candidate reactors for nuclear hydrogen development and demonstration plant. The tritium generated by the fission reaction can be leaked to the helium coolant from the coated ceramic particles and fuel elements. The annual total release rate of the tritium is estimated as 0.47% from the fuel kernel to the helium coolant by the numerical method. Tritium attributed by ⁶Li existing as impurities in the reflector can be released to the helium coolant by the diffusion process and the total annual release rate of the tritium is estimated as 5.3% through the reflector to the helium coolant. Based on the Siverts' law, tritium permeation from the primary coolant to the hydrogen production system is also evaluated and the result is calculated as 76?0.23 Bq/g-H₂ with respect to the PRF (Permeation Reduction Factor= 10?1000) in case of the normal operation of the 300 MWth Pebble type reactor.     

高温堆非纯氦气中预氧化对3种高温合金的腐蚀行为影响研究

高温气冷堆（简称高温堆）中,由于一回路冷却剂氦气中含有微量（ppm级）不纯杂质,其在高温环境中会对高温堆合金材料造成腐蚀,影响设备的性能。Inconel 617、Hastelloy X、Incoloy 800H是3种高温堆中间换热器及蒸汽发生器设备候选材料。研究表明,镍铬合金在高温下表面生成的富铬氧化层是防止合金在高温下发生严重腐蚀的重要因素。本文对3种合金在高温含杂质氦气中的腐蚀行为进行研究,探究预氧化对3种合金腐蚀行为的影响。并通过称重、扫描电镜、X射线能谱、电子探针显微分析仪以及碳硫分析仪对腐蚀结果进行分析。结果表明,3种合金均出现了不同的氧化和渗碳现象,预氧化对Hastelloy X合金抗腐蚀能力的提升不明显,对Inconel 617合金的抗氧化和渗碳能力有一定提升,对Incoloy 800H合金的抗渗碳腐蚀能力有一定提升。     

Thermal mixing characteristics of helium gas in high-temperature gas-cooled reactor, (I) thermal mixing behavior of helium gas in HTTR

The future high-temperature gas-cooled reactor (HTGR) is now designed in Japan Atomic Energy Agency. The reactor has many merging points of helium gas with different temperatures. It is needed to clear the thermal mixing characteristics of helium gas at the pipe in the HTGR from the viewpoint of structure integrity and temperature control. Previously, the reactor inlet coolant temperature was controlled lower than specific one in the high-temperature engineering test reactor (HTTR) due to lack of mixing of helium gas in the primary cooling system. Now, the control system is improved to use the calculated bulk temperature of reactor inlet helium gas. In this paper, thermal–hydraulic analysis on the primary cooling system of the HTTR was conducted to clarify the thermal mixing behavior of helium gas. As a result, it was confirmed that the thermal mixing behavior is mainly affected by the aspect ratio of annular flow path, and it is needed to consider the mixing characteristics of helium gas at the piping design of the HTGR.     

HCLL and HCPB coolant purification system: Design of the copper oxide bed

The coolant purification system (CPS) together with the tritium extraction system (TES) and helium cooling system (HCS) are the principal auxiliary circuits of helium-cooled-lithiium–lead (HCLL) and helium-cooled-pebble-bed (HCPB) test blanket modules (TBMs). To extract heat from TBMs, Helium is used as primary coolant. CPS is used to extract tritium from the helium primary circuit as well as to guarantee removal of impurities which could interact with structural material. The reference process proposed for CPS is composed of 3 main successive steps. Step 1 consists in oxidation of Q₂ and CO to Q₂O and CO₂ using a copper oxide bed (Q represents either: H, D or T). Step 2 is dedicated to the removal of water which is adsorbed together with CO₂ on molecular sieve bed. Step 3 will remove residual impurities using a heated getter.Based on the operating conditions of CPS (pressure, flowrate, temperature) and on an estimation of the impurities foreseen, this paper presents a design of the oxidising bed which fulfils all requirements in terms of efficiency and lifespan. The design is obtained using a phenomenological approach taking into account competition between oxidation of CO and Q₂ on the metal oxide. The model was implemented in matlab software. A column of 0.41 m large and 2 m long containing 480 kg of CuO is proposed to assure complete oxidation of Q₂ for 16 months long.     

氦气试验回路中的氦净化

为了减少结构材料的腐蚀并验证高温气冷实验堆中的氦净化工艺，在氦气试验回路中设置了氦净化系统并进行了试验。试验结果表明，净化流量为５０ｍ＾３／ｈ的主要由分子筛和深活性炭床组成的氦净化系统，能把氦中２００００ｃｍ＾３／ｍ＾３的化学杂质净化到７６ｃｍ＾３／ｍ＾３以下。所采用的氦中痕量杂质分析测量技术达到１０＾－１ｃｍ＾３／ｍ＾３精度。     

Radionuclides in primary coolant of a fluoride salt-cooled high-temperature reactor during normal operation

The release of fission products from coated particle fuel to primary coolant,as well as the activation of coolant and impurities,were analysed for a fluoride saltcooled high-temperature reactor (FHR) system,and the activity of radionuclides accumulated in the coolant during normal operation was calculated.The release rate (release fraction per unit time) of fission products was calculated with STACY code,which is modelled mainly based on the Fick's law,while the activation of coolant and impurities was calculated with SCALE code.The accumulation of radionuclides in the coolant has been calculated with a simplified model,which is generally a time integration considering the generation and decay of radionuclides.The results show that activation products are the dominant gamma source in the primary coolant system during normal operation of the FHR while fission products become the dominant source after shutdown.In operation condition,health-impacts related nuclides such as 3H,and 14C originate from the activation of lithium and coolant impurities including carbon,nitrogen,and oxygen.According to the calculated effective cross sections of neutron activation,6Li and 14N are the dominant 3H production source and 14C production source,respectively.Considering the high production rate,3H and 14C should be treated before being released to the environment.     

Radionuclides in primary coolant of a fluoride salt-cooled high-temperature reactor during normal operation

The release of fission products from coated particle fuel to primary coolant,as well as the activation of coolant and impurities,were analysed for a fluoride salt-cooled high-temperature reactor (FHR) system,and the activity of radionuclides accumulated in the coolant during normal operation was calculated.The release rate (release fraction per unit time) of fission products was calculated with STACY code,which is modelled mainly based on the Fick's law,while the activation of coolant and impurities was calculated with SCALE code.The accumulation of radionuclides in the coolant has been calculated with a simplified model,which is generally a time integration considering the generation and decay of radionuclides.The results show that activation products are the dominant gamma source in the primary coolant system during normal operation of the FHR while fission products become the dominant source after shutdown.In operation condition,health-impacts related nuclides such as 3H,and 14C originate from the activation of lithium and coolant impurities including carbon,nitrogen,and oxygen.According to the calculated effective cross sections of neutron activation,6Li and 14N are the dominant 3H production source and 14C production source,respectively.Considering the high production rate,3H and 14C should be treated before being released to the environment.     

The Fort St. Vrain high temperature gas-cooled reactor: X. Core temperature fluctuations

In October 1977, during the rise to power test program, the Fort St. Vrain high temperature gas-cooled reactor experienced the first of 37 fluctuation events involving primary coolant outlet temperature, nuclear detector signals, steam generator module gas inlet temperature and steam generator module main and reheat steam temperatures. In a 3 year investigation it was determined that the apparent cause of the fluctuations was movements of core components accompanied by periodic changes in bypass flows and crossflows of primary coolant helium. Installation of region constraint devices has eliminated fluctuations, but a single small primary coolant helium core outlet temperature redistribution is experienced routinely during rise to power.     

Investigation of the impurities and fission products in the AVR coolant gas at an average hot-gas temperature of 950°C

Since 27 February, 1974, the AVR pebble bed reactor has been producing gas at an average temperature of 950°C. Therefore it is possible for the first time to gain experience in high temperature reactor operations and experiments with such a high temperature level. This is of particular interest with regard to efforts using high temperature reactors for production of nuclear process heat. This paper reports briefly on the preparations for a temperature increase and on the first experimental results obtained with a hot-gas temperature of 950°C. Measured data are given on the behaviour of inactive gaseous impurities, on the increase of fission gas activities, and on the increase of concentrations of solid fission products in the helium coolant gas. While the activities of the fission gases showed an insignificant increase in the coolant gas, considerable increase of activity was measured for solid fission products, especially for Ag isotopes. However, activities released from fuel elements are low so that there are no operational or safety problems.