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.     

Effects of interaction between molten zircaloy and irradiated MOX fuel on the fission product release behavior

As a first step for obtaining experimental data on the effects of high-temperature chemical interaction on fission product release behavior, we focused on the dissolution of irradiated uranium plutonium mixed oxide (MOX) fuel by molten zircaloy (Zry) and carried out a heating test under the reducing atmosphere. Pieces of an irradiated MOX fuel pellet and cladding were subjected to the heating test at 2373 K for five minutes. The fractional release rate of cesium (specifically ¹³⁷Cs) was monitored during the test and its release behavior was evaluated. The observation of microstructures and measurements of elemental distribution in the heated specimen were also performed. We demonstrated experimentally that the fuel dissolution by molten Zry accelerated the release of Cs from the fuel pellets.     

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.     

Chemical reaction kinetics dataset of Cs-I-B-Mo-O-H system for evaluation of fission product chemistry under LWR severe accident conditions

In order to improve LWR source term under severe accident conditions, the first version of a fission product chemistry database named ‘ECUME’ was developed. The ECUME is intended to include several datasets of major chemical reactions and their effective kinetic constants for representative severe accident sequences. It is expected that the ECUME can serve as a fundamental basis from which fission product chemical models can be elaborated for use in the severe accident analysis codes. The implemented chemical reactions in the first version were those for representative gas species in Cs-I-B-Mo-O-H system from 300 to 3000 K. The chemical reaction kinetic constants were evaluated from either literature data or calculated values using ab-initio calculations. The sample chemical reaction calculation using the presently constructed dataset showed meaningful kinetics effects at 1000 K. Comparison of the chemical equilibrium compositions by using the dataset with those by chemical equilibrium calculations has shown rather good consistency for the representative Cs-I-B-Mo-O-H species. From these results, it was concluded that the present dataset should be useful to evaluate fission product chemistry in Cs-I-B-Mo-O-H system under LWR severe accident conditions, where kinetics effects should be considered.     

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.     

MAAP5 simulation of the PWR severe accident induced by pressurizer safety valve stuck-open accident

The Modular Accident Analysis Program version 5 (MAAP5) is a computer code that can simulate the response of light water reactor power plants during severe accident sequences. The present work aims to simulate the severe accident of a typical Chinese pressure water reactor (PWR) with MAAP5. The pressurizer safety valve stuck-open accident is essentially a small break loss-of-coolant accident (SBLOCA), which becomes one of the major concerns on core melt initiating events of the PWR. Six cases with different assumptions in the pressurizer (PZR) safety valves (SVs) stuck-open accident stuck open accident were analyzed for comparison. The results of first three cases show that the severe accident sequence is correlated with the number of the stuck open valve. The primary system depressurized faster in a more SVs stuck open case, and the consequences in which is hence slighter. The remaining 3 cases along with the case 2 were then analyzed to study the effect of operator intervention to the accident. The results show that the auxiliary feed water (AFW) is effective to delay the core degradation and hence delayed the finally system recovery. The high pressure injection (HPI) operation and manually opening the steam generator (SG) SVs are effective to mitigate this kind of severe accident. The results are meaningful and significant for comprehending the detailed process of PWR severe accident, which is the basic standard for establishing the severe accident management guidelines.     

Simulation of the small modular reactor severe accident scenario response to SBO using MELCOR code

Advanced small modular reactors (SMRs) use different design in the systems, structures, components from large reactors for achieving a high level of safety and reliability. In present work, the SMRs severe accident caused by the station blackout (SBO) was modeled and analyzed using MELCOR code, and the simulation of the accident scenario response to SBO was conducted. Based on the steady state calculation, which agrees well with designed values, we introduced the SBO accident for transient calculation. First, the case of the SBO accident without the passive core cooling system (PXS) was calculated. The progression and scenario in the reactor pressure vessel (RPV) and the containment were simulated and analyzed, including the transient response, cooling capacity and thermal-hydraulic characteristics and so on. The station black-out transient in the SMR can be simulated accurately, and the main failure model in the accident process can be concluded. Then three other cases of the SBO accident with different passive safety systems (core makeup tank (CMT), accumulator (ACC), passive residual heat removal system heat exchanger (PRHR HX), automatic depressurization system (ADS)) of the PXS were calculated respectively, and the results for different passive safety systems were compared. The passive core cooling system can not only provide water to the primary coolant system, but also take away the reactor decay residual heat. So in a station black-out transient, we can get more time for restoring AC power, and effectively prevent the accidents such as Fukushima.     

Influence of chemical speciation in reactor cooling system on pH of suppression pool during BWR severe accident

The influences of chemical speciation for Cs–I–Te–Mo–Sn–B–C–O–H system, simulating a state in the reactor cooling system (RCS) of BWR, on pH of the suppression chamber (S/C) water pool were analytically investigated with PHREEQC code. Major conditions were chosen on the basis of the outputs from a BWR severe accident analysis by THALES2 code and chemical thermodynamic analysis with VICTORIA2.0 code. The chemical thermodynamic analysis showed that the chemical speciation of important volatile FPs, Cs and I was strongly influenced by Mo and B₄C control material. As a consequence, pH of the S/C water pool was predicted to range from approximately 6 to 10, depending on the fraction of volatile FPs transported from the RCS to the S/C water pool and the H₂/H₂O ratio associated with the oxygen potential. It was implied that the formation of volatile I species such as I₂ in the S/C water pool was larger by three orders at the lowest pH than that at the highest pH.     

严重事故下开启双层安全壳环形空间通风过滤系统对缓解放射性向环境释放影响研究

核电站发生严重事故后,安全壳能包容从堆芯释放出的裂变产物,防止向环境的大量释放,但即使在安全壳完好的情况下,仍然会存在一定量泄漏。目前国际上的三代核电机型,大多采用双层安全壳的设计,对裂变产物具有一定的包容、滞留和过滤作用。本文基于我国自主设计的第三代核电机组,结合双层安全壳的设计特点和特定源项分析,对严重事故下双层安全壳之间的环形空间及其通风过滤系统对缓解裂变产物向环境释放的作用进行了定量分析,结果显示双层安全壳及环形空间通风过滤系统能够显著降低放射性气溶胶对环境的释放,对惰性气体也有一定的延缓排放作用。     

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。     

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.     

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.