Analysis of hot leg natural circulation under station blackout severe accident

Under severe accidents, natural circulation flows are important to influence the accident progression and result in a pressurized water reactor （PWR）. In a station blackout accident with no recovery of steam generator （SG） auxiliary feedwater （TMLB＇ severe accident scenario）, the hot leg countercurrent natural circulation flow is analyzed by using a severe-accident code, to better understand its potential impacts on the creep-rupture timing among the surge line, the hot leg, and SG tubes. The results show that the natural circulation may delay the failure time of the hot leg. The recirculation ratio and the hot mixing factor are also calculated and discussed.     

Effect of water injection on hydrogen generation during severe accident in PWR

Effect of water injection on hydrogen generation during severe accident in a 1000 MWe pressurized water reactor was studied.The analyses were carried out with different water injection rates at different core damage stages.The core can be quenched and accident progression can be terminated by water injection at the time before cohesive core debris is formed at lower core region.Hydrogen generation rate decreases with water injection into the core at the peak core temperature of 1700 K,because the core is quenched and reflooded quickly.The water injection at the peak core temperature of 1900 K,the hydrogen generation rate increases at low injection rates of the water,as the core is quenched slowly and the core remains in uncovered condition at high temperatures for a longer time than the situation of high injection rate.At peak core temperature of 2100-2300 K,the Hydrogen generation rate increases by water injection because of the steam serving to the high temperature steam-starved core.Hydrogen generation rate increases significantly after water injection into the core at peak core temperature of 2500 K because of the steam serving to the relocating Zr-U-O mixture.Almost no hydrogen generation can be seen in base case after formation of the molten pool at the lower core region.However,hydrogen is generated if water is injected into the molten pool,because steam serves to the crust supporting the molten pool.Reactor coolant system (RCS) depressurization by opening power operated relief valves has important effect on hydrogen generation.Special attention should be paid to hydrogen generation enhancement caused by RCS depressurization.     

Models for the release from fuel of Ce, La, Sr, and Eu in accident conditions

During a hypothetical severe nuclear accident involving a pressurised water reactor, it is of primary importance to assess the potential radionuclide releases into the environment. With that view in mind, analytical models have been developed for the release kinetics from fuel of four low-volatile fission products (namely cerium, lanthanum, strontium, and europium) in severe accident conditions. The vaporisation of those fission products from fuel has been evaluated by means of thermodynamic equilibrium calculations. The effects of the fuel temperature and the oxygen potential on the chemical form and volatility of the fission products have been determined. Finally, these models have been integrated in the ASTEC and have been subsequently validated against annealing experiments with various oxidizing and reducing conditions.     

Study on severe accident mitigation measures for the development of PWR SAMG

In the development of the Severe Accident Management Guidelines （SAMG）, it is very important to choose the main severe accident sequences and verify their mitigation measures. In this article, Loss-of-Coolant Accident （LOCA）, Steam Generator Tube Rupture （SGTR）, Station Blackout （SBO）, and Anticipated Transients without Scram （ATWS） in PWR with 300 MWe are selected as the main severe accident sequences. The core damage progressions induced by the above-mentioned sequences are analyzed using SCDAP/RELAP5. To arrest the core damage progression and mitigate the consequences of severe accidents, the measures for the severe accident management （SAM） such as feed and bleed, and depressurizations are verified using the calculation. The results suggest that implementing feed and bleed and depressurization could be an effective way to arrest the severe accident sequences in PWR.     

小型压水堆严重事故下一回路承压管道蠕变预测分析模型开发

一回路承压管道蠕变是压水堆核电厂严重事故重要现象之一。针对小型压水堆,本文基于SCDAP/RELAP5程序开发了严重事故分析模型,利用实验拟合方法得到了一回路主管道(SA321)、自然循环式蒸汽发生器传热管(00Cr25Ni35Al Ti)两种材料蠕变预测分析模型,改进了SCDAP/RELAP5程序蠕变预测分析功能模块,并通过假想事故序列验证了SA321、00Cr25Ni35Al Ti蠕变预测分析模型的合理性。为后续开展小型压水堆严重事故下一回路承压管道蠕变规律研究提供基础参考。     

小破口引发的严重事故工况及事故缓解的研究

利用MAAP4程序对方家山核电站进行建模,针对事故后果较为严重的小破口事件进行了计算分析,得到了假设事故下电厂系统的反应以及相应的严重事故现象.对事故中发生的DCH(安全壳直接加热)现象和安全壳失效以及裂变产物向环境的释放进行了分析.随后,本文根据相关的严重事故管理导则和该事故的特点,对缓解该事故的策略进行了研究和计算...     

First-principle studies of radioactive fission productions Cs/Sr/Ag/I adsorption on chrome-molybdenum steel in Chinese 200 MW HTR-PM

Chrome-molybdenum steel(2·1/4Cr1Mo) is one of the main products of steam generation.The adsorption behaviors of radioactive fission products on2·1/4Cr1Mo surface are critical in the analysis of HTR-PM.Here,the adsorption behavior of cesium,strontium,silver and iodine on 2·1/4Cr1Mo was investigated with first-principle calculations that the Ag and I atoms prefer to be adsorbed at the square hollow site of the face-centered cubic iron cell with a binding energy of about 1 and 3 eV,respectively.In contrast,Cs and Sr atoms are not adsorbed on the surface of the 2·1/4Cr1Mo.These results are again confirmed via analysis of charge density differences and the densities of state.Furthermore,the adsorption rates of these fission products show that only I and Ag have significant adsorption on the metal substrate.These adsorption results explain the amount of adsorbed radionuclides for an evaluation of nuclear safety in HTR-PM.These micro-pictures of the interaction between fission products and materials are a new and useful way to analyze the source term.     

Sensitivity analysis of fission product activity in primary coolant of typical PWRs

A model has been developed for static and dynamic activity analysis of the fission product activity (FPA) in primary coolant of typical pressurized water reactors (PWRs). It has been implemented in the FPCART based computer program FPCART-SA. For long steady power operation of reactor, the computed values of normalized static sensitivity have been compared with the corresponding values obtained by using the dynamic sensitivity analysis. The normalized sensitivity values for the reactor power (P), failed fuel fraction (D), coolant leakage rate (L), total mass of coolant (m) and the let-down flow rate (Q) have been calculated and the values: 1.0, 0.857, −2.0177 × 10⁻⁶, 2.349 × 10⁻⁴, −2.329 × 10⁻⁴ have been found correspondingly for Kr-88 with the dominant value of FPA as 0.273 μCi/g.     

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.     

严重事故下裂变产物气溶胶自然沉积现象研究

以600 MW压水堆核电厂为研究对象,在一体化安全分析模型的基础上建立重力沉降、扩散电泳、惯性碰撞和热电泳4种裂变产物气溶胶的自然沉积模型,选取典型的严重事故序列,分析严重事故下裂变产物气溶胶的自然沉积现象。将MELCOR程序的重力沉降模型植入本文的一体化分析模型,对重力沉降份额进行比较。研究表明,重力沉降对气溶胶沉积的贡献最大;本文采用的重力沉降模型比MELCOR程序重力沉降模型的沉降效应稍强。     

Efficacy assessment of independent severe accident mitigation measures in OPR1000 using MELCOR code

In-vessel and ex-vessel mitigation strategies have been revisited to improve the severe accident management (SAM) for operating nuclear power plants. Because independent mitigation measures tend to produce positive and adverse effects simultaneously, it is necessary to investigate the efficacy of individual measures by means of proper quantification. Thus, in the present study we investigated the overall efficacy of existing SA mitigation strategies prepared for the Optimized Power Reactor 1000 MWe (OPR1000) by means of MELCOR 1.8.6 code. The numerical evaluation showed that the Mitigation-01, feeding water into the steam generators, is the most effective among the other mitigations. In addition, Mitigation-02, reactor coolant system depressurization, could not mitigate the SA sufficiently when applied individually. Among the four ex-vessel mitigation strategies, execution of containment spray was effective in removing most of the aerosol fission product but also intensified hydrogen combustion by increasing the partial hydrogen pressure owing to steam condensation. Mitigation-07, operation of passive autocatalytic recombiners (PARs), could reduce the hydrogen concentration, though the catalytic reaction was predicted to increase the containment pressure. In conclusion, this study suggests that mitigation measures should be carefully selected, and that counteracting measures should be prepared to minimize potential adverse effects.     

Effectiveness and adverse effects of reactor coolant system depressurization strategy with various severe accident management guidance entry conditions for OPR1000

Severe accident analysis for Korean OPR1000 with MELCOR 1.8.6 was performed by adapting a mitigation strategy under different entry conditions of Severe Accident Management Guidance (SAMG). The analysis was focused on the effectiveness of the mitigation strategy and its adverse effects. Four core exit temperatures (CETs) were selected as SAMG entry conditions, and Small Break Loss of Coolant Accident (SBLOCA), Station Blackout (SBO), and Total Loss of Feed Water (TLOFW) were selected as postulated scenarios that may propagate into severe accidents. In order to delay reactor pressure vessel (RPV) failure, entering the SAMG when the CET reached 923 K, 923 K, and 753 K resulted in the best results for SBLOCA, SBO, and TLOFW scenarios, respectively. This implies that using event-based diagnosis for severe accidents may be more beneficial than using symptom-based diagnosis. There is no significant difference among selected SAMG entry conditions in light of the operator's available action time before the RPV failure. Potential vulnerability of the RPV due to hydrogen generation was analyzed to investigate the foreseeable adverse effects that act against the accident mitigation strategies. For the SBLOCA cases, mitigation cases generated more hydrogen than the base case. However, the amount of hydrogen generated was similar between the base and mitigation cases for SBO and TLOFW. Hydrogen concentrations of containment were less than 5% before RPV failure for most cases.     

稳压器卸压阀卸压效果研究

基于国际上模拟严重事故瞬态过程最详细的机理性程序SCDAP/RELAP5/MOD3.1,主要分析研究了核电站未紧急停堆的预期瞬变(ATWS)初因(失去主给水、失去厂外电和控制棒失控提升)叠加辅助给水失效导致的堆芯熔化严重事故进程,并验证阻止ATWS导致堆芯熔化进程的一次侧卸压缓解措施的充分性和有效性.计算分析结果显示,一列稳压器卸压阀不足以充分降低一回路压力,压力仍然停留在10MPa以上,存在很大高压熔堆的风险.增加一列卸压阀可把一回路压力降低到3MPa左右,安注系统得以投入,及时有效地阻止堆芯熔化进程,降低了高压熔堆风险.分析结果还显示高压安注系统的投入对一回路卸压具有重要影响.     

14.8 MeV中子诱发~(238)U裂变产物产额测量

裂变产物产额作为裂变过程的一个重要参数,其准确测量对有关裂变的很多方面都有重要意义。为了准确测量中子诱发~(238)U裂变产物产额,利用中国工程物理研究院PD-300加速器上的T(d,n)4He反应,产生14.8MeV的中子,诱发~(238)U裂变。辐照过程中,通过金硅面垒半导体探测器监测中子通量的变化。使用Al片作为监测片计算整个照射过程中样品的平均中子通量。辐照结束后,利用高纯锗(High-Purity Germanium,HPGe)探测器测得裂变产物特征γ射线计数,计算得到裂变产物的产额,使用MCNPX软件对中子的多次散射和自屏蔽效应进行修正,并通过计算得到样品和监测片的自吸收修正、中子通量波动因子。得到了95Zr、127Sb、140Ba、147Nd、131I、103Ru等长半衰期产物的累积产额值,并将结果与以前的文献值做了比对,研究结果有助于~(238)U裂变产物产额的分析和评价。     

基于RELAP5与MELCOR联合分析方法的压水堆严重事故研究

针对严重事故的模拟研究,本文提出结合热工水力系统程序和严重事故一体化程序的分析方法,以典型三环路传统压水堆为对象,分别采用RELAP5和MELCOR程序建立模型,分析在全厂断电叠加汽动辅助给水泵失效事故下系统的瞬态响应。为了尽可能地利用RELAP5计算早期热工水力响应,同时保证严重事故计算结果的准确性,以MELCOR锆合金氧化模型开始工作温度的下限,即包壳温度达到1 100 K作为程序衔接准则并利用RELAP5的大编辑功能,提取所需计算结果导入MELCOR输入卡作为初始参数继续模拟。计算结果表明,数据连接过程整体保持了连续性,两种方法计算得出的主冷却剂系统压力、堆芯和稳压器水位、燃料包壳温度等参数的数值以及堆芯传热恶化和压力容器失效等现象的时序存在不同程度的差异,例如堆芯熔毁时间延后了约538 s。由于采用了RELAP5计算严重事故前的系统暂态响应,联合分析方法的计算结果比单独使用MELCOR分析的结果更加准确,该方法可以提高传统严重事故分析的可靠性。     

基于RELAP5与MELCOR联合分析方法的压水堆严重事故研究

针对严重事故的模拟研究,本文提出结合热工水力系统程序和严重事故一体化程序的分析方法,以典型三环路传统压水堆为对象,分别采用RELAP5和MELCOR程序建立模型,分析在全厂断电叠加汽动辅助给水泵失效事故下系统的瞬态响应。为了尽可能地利用RELAP5计算早期热工水力响应,同时保证严重事故计算结果的准确性,以MELCOR锆合金氧化模型开始工作温度的下限,即包壳温度达到1 100 K作为程序衔接准则并利用RELAP5的大编辑功能,提取所需计算结果导入MELCOR输入卡作为初始参数继续模拟。计算结果表明,数据连接过程整体保持了连续性,两种方法计算得出的主冷却剂系统压力、堆芯和稳压器水位、燃料包壳温度等参数的数值以及堆芯传热恶化和压力容器失效等现象的时序存在不同程度的差异,例如堆芯熔毁时间延后了约538 s。由于采用了RELAP5计算严重事故前的系统暂态响应,联合分析方法的计算结果比单独使用MELCOR分析的结果更加准确,该方法可以提高传统严重事故分析的可靠性。     

喷淋模式对氢气爆炸影响的初步研究

采用MELCOR程序比较了大亚湾核电站在全厂断电事故下．恢复供电后不同喷淋模式对事故进程的影响。结果显示,采用较短的喷淋持续时间和适宜的喷淋投入时间,能较明显地避免氢燃或降低氢燃的强度,从而延迟安全壳内压力到达限值的时间。     

MELCOR/MACCS simulation of the TMI-2 severe accident and initial recovery phases, off-site fission product release and consequences

MELCOR has become the preferred code of the Swiss nuclear industry and of PSI for severe accident analysis, on account of its integrated systems-level approach and validation against experiments and more detailed codes, while MACCS is commonly used by safety authorities for independent assessment of off-site consequences, in particular health effects. The present work arises out of a programme to assess MELCOR independently using empirical data consistent with the recommendations of the OECD/CSNI validation matrix for core degradation codes. The MELCOR 1.8.5RD calculations are based on a model for phases 1 and 2 provided by the code developers but with a simplified thermal hydraulic noding in certain regions and the inclusion of a simple representation of the fission product release and transport pathways. The model has also been extended to simulate phases 3, 4, and the continuing initial period of core recovery and stabilisation. These calculations are a first attempt to demonstrate a MELCOR–MACCS capability to simulate the whole plant accident sequence beyond phase 4, including the containment response and off-site consequences arising from fission product release from the containment. Emphasis is placed on the overall accident evolution and whole plant response, rather than the detailed behaviour. Results are compared with observed and deduced data for the major accident signatures and rough estimates for exposure based on off-site monitoring. The results provide a good basis for the NPP analysis foreseen.