5MW供热试验堆上空腔小破口失水事故模拟实验研究

本文给出了位于上空腔的中小尺寸管道破裂或安全阀意外开启引起的小破口失水事故实验研究。在实验中研究了系统压力、温度、空泡份额的变化和总失水量。结果表明总失水量约为初始装水量的百分之二十。这种事故对于清华大学核能研究所建造的低温供热堆是安全的。     

破口大小和二回路运行状态对核供热堆自然循环断流瞬态的影响

为研究一体化布置的核供热堆在发生破口失水事故中破口大小和从中间回路排出热量减少对断流过程的影响，选用不同的破口尺寸和不同的二回路工作状态，在５ＭＷ核供热堆热工水力模拟回路ＨＲＴＬ－５上进行了实验研究。稳态运行工况的系统压力为１．５ＭＰａ，在发生小破口失水事故后，加热功率维持为额定功率的５％以模拟剩余发热情况。实验研究并比较了不同条件下压力、温度、循环流量、液位和失水量等重要参数的变化。这些实验数据为核供热堆的安全分析提供了实验依据。     

小破口失水事故研究综述

对小破口失水事故（ＳＢＬＯＣＡ）及其研究状况进行了综述。描述了典型的压水堆和沸水堆小破口失水事故过程和破口位置、破口尺寸及反应堆冷却泵对失水过程的影响，对现有文献按实验和数值模拟两大类进行了归纳，给出了目前世界上用于小破口失水事故研究的主要设备，对小破口失水事故的研究进行了总结。     

核电站不同严重事故序列下的MCCI及其缓解措施计算分析

概述了MEDICS程序的主要机理和模型,介绍了利用MEDICS程序进行严重事故下堆芯熔融物与混凝土相互作用(MCCI)的计算方法,并给出了大亚湾核电站全厂断电、小破口失水事故、大破口失水事故等典型初因事故导致的严重事故下的MCCI及其缓解措施的计算分析结果.计算结果表明,在无缓解措施情况下,安全壳底板熔穿时间在10.08～13.4d范围内,H2的产生量在12760～13159kg范围内;顶部冷却是较好的MCCI缓解措施,能明显延长安全壳底板熔穿时间、降低H2和总不可凝气体释放量.     

丙烯基类单体-壳聚糖共聚物水凝胶的辐射制备及性能研究

以壳聚糖、丙烯酰胺（AM）和二甲基二烯丙基氯化铵（DMDACC）为原料,N,N-亚甲基双丙烯酰胺（MBA）为交联剂,经^60Co γ-射线辐照,制备了丙烯基类单体一壳聚糖共聚物水凝胶,并研究了MBA的用量对丙烯基类单体一壳聚糖共聚物水凝胶吸水性能和保水性能的影响。结果表明,在吸收剂量为2kGy的条件下,MBA用量较低时,随着MBA用量的提高,水凝胶的平衡吸水率反而下降;而MBA用量较高时,水凝胶的平衡吸水率受MBA用量的影响较小;凝胶吸水率随着溶胀时间的延长而增加;水凝胶在溶胀初期的溶胀动力学可用non-Fickian扩散定律来描述：随着交联剂MBA用量的逐渐增加,水凝胶的初始失水率逐渐增大,而最大失水率逐渐下降。     

低温堆上空腔失水事故模拟实验研究

叙述了位于低温堆上空腔位置的中小尺寸管道破裂引起的小破口失水事故研究。在核供热堆热工水力学实验系统ＨＲＴＬ－５上，对停堆后堆内有剩余功率的上空腔小破口失水事故进行了模拟实验，分析了小破口失水事故发生后，系统运行重要参数的变化，给出了上空腔小破口失水事故对低温安全性的影响。     

DVI管小破口失水事故实验研究

在模块化小型反应堆非能动安全系统综合模拟实验装置上进行了压力容器直接注入（DVI）管小破口失水事故实验,研究了DVI管小破口失水事故过程中的热工水力现象和非能动安全系统运行特性。研究结果表明：模块化小型反应堆DVI管小破口失水事故中,非能动安全系统可对堆芯进行注水,有效导出堆芯衰变热量,保护堆芯安全。     

氚气流发生器饱和系数测定

简述了用称重法对氚气流发生器饱和系数的测定,给出了不同流速等条件下失水量的修正及饱和系数的测量结果,并对测量结果不确定度进行了评定。     

脉冲堆失水事故计算程序的开发和应用

由于西安脉冲堆的特点，致使国际上通用的瞬时堆芯裸露模型不能使用。中国核动力研究设计院建立了反映西安脉冲堆失水事故机理和过程的真实真芯裸露模型，开发了相应的计算机程序，用于分析和评价西安脉冲堆的安全特性。分析结果表明，真实堆芯裸露模型具有广泛的实用性，可用于计算全部侧面破口和底部破口的失水事故。在破口直径相同的条件下，西安脉冲堆侧面破口失水事故后果比底部破口失水事故严重。在目前的设计条件下，即使发生失水事故，西安脉冲堆也能满足安全准则的要求。     

严重事故下氢气风险及氢气控制系统的初步分析

采用一体化严重事故分析工具,对600MWe压水堆核电厂严重事故下氢气风险及拟定的氢气控制系统进行分析。结果表明：相对于小破口失水始发事故和全厂断电始发事故工况,大破口失水始发严重事故堆芯快速熔化,在考虑100%锆 水反应产氢量的条件下,大破口失水始发事故氢气风险较大,有可能发生氢气快速燃烧;在氢气控制系统作用下,发生大破口失水始发严重事故时,安全壳内平均氢气浓度和隔间内氢气浓度低于10%,未达到氢气快速燃烧和爆炸的条件,满足美国联邦法规10CFR中关于氢气控制和风险分析的准则,认为该氢气控制系统是可行、有效的。     

Evaluation of precipitates used in strainer head loss testing. Part I. Chemically generated precipitates

The purpose of the current program was to evaluate the properties of chemical precipitates proposed by industry that have been used in sump strainer head loss testing. Specific precipitates that were evaluated included aluminum oxyhydroxide (AlOOH) and sodium aluminum silicate (SAS) prepared according to the procedures in WCAP-16530-NP, along with precipitates formed from injecting chemicals into the test loop according to the procedure used by one sump strainer test vendor for U.S. pressurized water reactors. The settling rates of the surrogate precipitates are strongly dependent on their particle size and are reasonably consistent with those expected from Stokes’ Law or colloid aggregation models. Head loss tests showed that AlOOH and SAS surrogates are quite effective in increasing the head loss across a perforated pump inlet strainer that has an accumulated fibrous debris bed. The characteristics of aluminum hydroxide precipitate using sodium aluminate were dependent on whether it was formed in high-purity or ordinary tap water and whether excess silicate was present or not.     

Neutronic evaluation of coating and cladding materials for accident tolerant fuels

In severe accident conditions with loss of active cooling in the core, zirconium alloys, used as fuel cladding materials for current light water reactors (LWR), undergo a rapid oxidation by high temperature steam with consequent hydrogen generation. Novel fuel technologies, named accident tolerant fuels (ATF), seek to improve the endurance of severe accident conditions in LWRs by eliminating or at least mitigating such detrimental steam-cladding interaction. Most ATF concepts are expected to work within the design framework of current and future light water reactors, and for that reason they must match or exceed the performance of conventional fuel in normal conditions. This study analyzed the neutronic performance of ATF when employed in both pressurized and boiling water reactors. Two concepts were evaluated: (1) coating the exterior of zirconium-alloy cladding with thin ceramics to limit the zirconium available for reaction with high-temperature steam; (2) replacing zirconium alloys with alternative materials possessing slower oxidation kinetics and reduced hydrogen production. Findings show that ceramic coatings should remain 10–30 μm thick to limit the neutronic penalty. Alternative cladding materials, with the exception of SiC, enhance neutron loss compared to zirconium-alloys. An extensive parametric analysis concluded that reference performance metrics can be met by employing 300-μm or less thick cladding or increasing fuel enrichment by up to 1.74% depending on material and geometry.     

Vessel and Channel Fast Reactors Cooled by Boiling Water or Water with Supercritical Parameters

The possibility is examined of developing a vessel-type fast reactor cooled by water with supercritical parameters (BR-VSP) and a channel-type fast plutonium–water reactor cooled by boiling water, using the RBMK scheme, or water with supercritical parameters for burning weapons or power-production plutonium (RBMK-Pu).A reactor with construction similar to that of an RBMK reactor but without the graphite moderator and zirconium can be used for burning plutonium. Removing the graphite masonry while retaining a 25-cm spacing in the square lattice gives a large free volume in the neutron field for holding, for example, cobalt. A neutron-physical substantiation is given for the serviceability of such a reactor with the maximum burnup of removed fuel 10% h.a. and the possibility of obtaining a negative reactivity effect in the case of water loss.     

Development of safety injection flow map associated with target depressurization for effective severe accident management of OPR1000

If any severe accident occurs, application of the Severe Accident Management Guidance (SAMG) is initiated by the Technical Support Center (TSC). In order to provide advisory information to the TSC, required safety injection flow rate for maintaining the coolability of the reactor core has been suggested in terms of the depressurization pressure. In this study, mechanistic development of the safety injection flow map was performed by post-processing the core exit temperature (CET) data from MELCOR simulation. In addition, effect of oxidation during the core degradation was incorporated by including simulation data of core water level decrease rate. Using the CET increase rate and core water level decrease rate, safety injection flow maps required for removing the decay and oxidation heat and finally for maintaining the coolability of the reactor core were developed. Three initiating events of small break loss of coolant accidents without safety injection, station black out, and total loss of feed water were considered. Reactor coolant system depressurization pressure targeting the suggested injection flow achievable with one or two high pressure safety injections was included in the map. This study contributes on improving the current SAMG by providing more practical and mechanistic information to manage the severe accidents.     

Radiological consequences of a depressurization accident combined with water ingress in an HTR module-200

The principles of the reaction between Cs and steam or water with respect to the primary circuit of an HTR and results of corresponding experiments are presented. Based on this for the most important pressure loss accidents combined with water ingress of an HTR-200 the radiological consequences for the environment are investigated in detail. The results are presented and discussed.     

Enrichment and Volume Reduction of Tritiated Water Using Electrolysis Cell Having Hydrogen Permeable Cathode

The electrolysis rate and the separation factor for hydrogen isotopes are measured using the electrolysis cell having the hydrogen permeable cathode. As the hydrogen gas without the vapor of electrolyte is obtained by this method, decrease of the apparent separation factor by mixing with vapor can be avoided. It is also observed in this study that enrichment and volume reduction of tritiated water using the bipolar electrode electrolysis cell is effective because it gives small loss of tritium from the cell during volume reduction. The separation factor obtained in this study indicates that attachment of two or three sub-cells is enough for volume reduction of tritiated water.     

核电厂安全壳地坑滤网堵塞问题的机理分析

破口失水事故工况下,大量碎片可能随着泄漏的冷却剂和喷淋液迁移到安全壳地坑滤网处,并逐渐堆积形成碎片床,不断增大流体通过滤网的阻力,降低应急堆芯冷却系统或安全壳喷淋系统泵的净正吸入压头裕量并导致堆芯、安全壳丧失冷却,从而威胁核电厂安全。本文对核电厂发生假想破口失水事故后碎片的产生、迁移,以及在安全壳地坑滤网处堆积成碎片床,并造成地坑滤网堵塞的机理进行分析说明。     

Prediction of temperature and water level in a spent fuel pit during loss of all AC power supplies

A prediction method for water temperature in a spent fuel pit of a pressurized water reactor (PWR) has been developed to calculate the increase in water temperature during the shutdown of cooling systems. In this study, the prediction method was extended to calculate the water level in a spent fuel pit during loss of all AC power supplies, and predicted results were compared with measured values of spent fuel pools in the Fukushima Daiichi Nuclear Power Station. The calculations gave reasonable results, but overestimated the decreasing rate of the water level and the water temperature. This indicated that decay heat was overestimated and evaporation heat transfer from the water surface was underestimated. Results of calculations with 80% decay heat and 155% (Unit 4 pool) or 230% (Unit 2 pool) evaporation heat flux were in good agreement with measured values. The data-fitted evaporation heat fluxes agreed rather well with the evaporation heat transfer correlation proposed by Fujii et al.     

Stopping cross sections of liquid water for MeV energy protons

Cross sections for the stopping of swift protons in liquid water have been measured for the first time by using a liquid water jet target of 50 μm in diameter. The energy loss spectra of incident 2.0 MeV protons were measured at scattering angles of 5.0-50 mrad. Experimental energy loss spectra have been successfully reproduced by Monte Carlo simulation calculations (GEANT4.9.1.p02 toolkit) by taking account of multiple scattering of projectile ions inside the liquid water target. The present stopping cross sections are found to be considerably smaller than other standard stopping power data, revealing e.g. about 11% deviation from those of SRIM2003.     

西安脉冲堆失水事故后堆顶平台散射γ剂量率计算

计算了西安脉冲堆发生失水事故后，堆芯发射的γ射拇因失去堆水池屏蔽层，直接穿透空气被顶部天花板反散射后在堆水澉平台产生的辐射剂量率。计算结果与ＴＲＩＧＡ－２堆结果相符合，是确定事故后工作人员在平台可停留时间的基础，同时为适当地制定事故处理措施提供了参考。