PHEBUS FPT0～FPT2实验堆芯降级过程数值模拟

严重事故下,由于堆芯冷却剂丧失引起的堆芯裸露、过热和熔化过程对后期安全壳完整性、裂变产物行为等具有重要影响。法国辐射防护与核安全研究所主导的PHEBUS-FP研究项目旨在研究轻水堆严重事故下堆芯降级过程以及裂变产物行为。本文使用ATHLET-CD程序对PHEBUS-FP中的FPT0、FPT1和FPT2进行建模计算,主要分析堆芯过热,包壳氧化,堆内材料熔化、迁移及再定位过程。计算结果表明:不同蒸汽流量、不同加热功率将导致不同堆芯降级进程,在趋势上计算值与实验值吻合;模型的限制导致了部分计算值的偏差,本文讨论了包壳氧化与燃料再定位现象中的模型参数。     

燃料棒包壳氧化层溶解失效准则研究

为确定严重事故条件下燃料棒包壳温度达到金属锆的熔点后包壳氧化层的失效时间、再定位熔融物的成分以及氧化层失效对堆芯熔化进程的影响,本文基于熔融锆同时溶解UO_2和ZrO_2动力学模型及燃料棒包壳水侧氧化层的受力分析建立了氧化层在熔融锆中溶解失效的准则。以FPT-0实验结果验证后发现该失效准则可以较准确地预测包壳氧化层的溶解失效。为增加该准则在严重事故计算程序中的适用性,在燃料棒设计结构一定的条件下,进一步将该准则量化为温度的函数,分析表明包壳氧化程度和燃料棒温度上升速率是影响包壳氧化层失效温度的主要因素。利用该失效准则可以同时获得包壳氧化层失效后再定位的熔融物的质量及成分含量。     

应用MECLOR对严重事故实验Phebus FPT3的模拟分析

使用MECLOR1-8.6程序对严重事故实验Phebus FPT3进行了模拟分析。通过建模计算,得到了严重事故过程中燃料棒的行为,氢气的产生,裂变产物的释放、迁移和沉降及安全壳的热工水力响应等相关数据。计算值与实验值的对比分析表明,燃料棒的行为、氢气产生的时间和趋势及安全壳的热工水力响应与实验值吻合良好。由于相应程序模型的限制,最终产氢的总量及裂变产物相关的计算值与实验值有所不同。其中,计算的氢气总量较实验值偏小,计算的裂变产物释放量和在安全壳中的沉降量大多较实验值稍高。此外,还利用快速傅里叶变换方法（FFTBM）对整个建模计算进行了详细的定量化评估。     

UO_2－锆合金化学反应条件下的包壳水侧氧化动力学

在堆芯熔化情况下，ＵＯ２燃料与锆合金包壳之间的反应将导致一系列的新相生成及氧自ＵＯ２燃料向包壳合金的扩散，进而对包壳的水侧氧化动力学过程中的相间界面推移产生重要影响。本文推导建立了在这种条件下计算锆合金包壳水侧氧化动力学的方法。     

A Cladding Oxidation Model Based on Diffusion Equations

在轻水堆的严重事故中,堆芯中锆包壳与水蒸汽的氧化反应对事故进程的影响至关重要.当氧化时间较长,或者包壳表面水蒸汽流量很小时,基于实验结果的抛物线型公式存在着不足,影响对包壳失效、氢气产生量以及温度的预测.本研究从菲克定律出发,建立以扩散方程为出发点、温度范围宽广的包壳氧化模型.该模型的适用范围比抛物线型公式广,且不受抛物线型公式中假设氧化时间较短和堆芯中水蒸汽流量足够的限制,能很好地模拟长期氧化实验和水蒸汽流量很小时的氧化.另外,该模型能给出包壳中的氧原子分布,为精细模拟包壳脆化失效等现象以及发展先进的包壳失效准则提供条件.     

基于扩散方程的包壳氧化模型

在轻水堆的严重事故中,堆芯中锆包壳与水蒸汽的氧化反应对事故进程的影响至关重要。当氧化时间较长,或者包壳表面水蒸汽流量很小时,基于实验结果的抛物线型公式存在着不足,影响对包壳失效、氢气产生量以及温度的预测。本研究从菲克定律出发,建立以扩散方程为出发点、温度范围宽广的包壳氧化模型。该模型的适用范围比抛物线型公式广,且不受抛物线型公式中假设氧化时间较短和堆芯中水蒸汽流量足够的限制,能很好地模拟长期氧化实验和水蒸汽流量很小时的氧化。另外,该模型能给出包壳中的氧原子分布,为精细模拟包壳脆化失效等现象以及发展先进的包壳失效准则提供条件。     

基于替代材料的燃料棒熔化实验研究

为获得核反应堆燃料元件熔化以及熔融物扩展和消熔过程中的关键实验数据,本研究将典型压水堆中的燃料棒元件作为研究对象,在堆芯材料严重事故现象可视化研究实验装置FROMA上开展了低温条件下的燃料棒熔化实验。实验采用锌-铝的替代材料燃料棒,开展了单棒的熔化凝固可视化研究,获得了严重事故过程中燃料棒包壳的瞬态轴向温度分布特性以及熔融物扩展、迁移和再定位的动态过程。本研究基于实验数据对熔融物的流动、扩展和凝固、迁移等相关的物理现象和过程进行深入分析,为反应堆严重事故现象分析模型的开发提供了数据支持。     

基于MPS算法的2×2棒束结构熔化行为数值模拟

以典型压水堆燃料组件2×2棒束结构为研究对象,建立了含定位格架和不含定位格架的棒束三维模型,基于半隐式运动粒子（MPS）算法对严重事故背景下棒束结构的熔化行为进行了数值模拟,分析了定位格架对棒束熔化过程中流道堵塞进程的影响。结果表明:MPS算法能够较好地模拟棒束结构熔化行为,定位格架会加快堆芯的熔化进程和冷却流道的堵塞速度,本文研究结果有利于严重事故下堆芯熔化模型的优化改进。      

压水堆核电厂裂变产物源项计算方法研究

裂变产物作为一回路冷却剂中放射性核素的重要组成部分,在核电厂设计中具有非常重要的意义。文中对堆芯积存量计算模型、燃料包壳内裂变产物向一回路冷却剂释放模型、裂变产物在一回路中的平衡模型进行了分析与研究,并以典型压水堆核电厂为例进行了计算与验证,证实了本文中给出计算模型的合理性以及适用性,可供压水堆核电站裂变产物源项计算分析参考。     

快堆严重事故钠燃烧过程裂变产物释放模拟试验研究

钠燃烧过程产生的裂变产物及钠气溶胶迁移是快堆严重事故重要的源项之一。本研究对钠燃烧过程裂变产物随钠蒸汽和钠气溶胶迁移的行为进行分析,针对钠蒸发作用下裂变产物释放、钠燃烧作用下裂变产物释放以及气相空间气溶胶迁移行为分别提出了物理模型,并在确定计算方法的基础上通过CFD软件建模进行了仿真计算,最后通过开展小规模钠燃烧试验,获取了真实钠燃烧过程裂变产物沉降数据,对计算模型进行了修正和补充。试验数据与仿真计算结果表明,气溶胶迁移模型能够较好地表征裂变产物及钠气溶胶迁移行为,钠燃烧作用下裂变产物的释放系数为10-3时计算结果与试验结果较吻合。     

The fracture behaviors of non-irradiated zircaloy-4 fuel cladding with a pinhole under simulated LOCA conditions

In a case where a pinhole leak occurs in a fuel rod incidentally, it is possible that coolant enters the fuel rod through the pinhole. Since knowledge about the behavior of the fuel rod with a pinhole under LOCA conditions is limited, semi-integral quench tests were performed with non-irradiated zircaloy-4 fuel cladding tubes with a pinhole in order to investigate the difference in the fracture behaviors between normal and leaker fuels under LOCA conditions. Isothermal oxidation temperature and time ranged from 1100 to 1225 °C and 0 to 4200 seconds, respectively. Ballooning and rupture during the heat-up process did not occur in the case of test rods with a pinhole and initially injected water. Initially injected water affected the oxidation behavior of the inner surface of cladding during the test, and the fracture boundary of the test rod was dependent on not only the axial restrained condition during the test but also the existence of a pinhole and initially injected water. This tendency seemed to be related to the amount of oxidation of cladding inner surface caused by the steam which remained in or entered the test rod during the test.     

Thermal-Hydraulics in Uncovered Core of Light Water Reactor in Severe Core Damage Accident, (II)

The SEFDAN is a computer program to analyze the one-dimensional thermal-hydraulics of a partially uncovered core of a light water reactor in a severe degraded-cooling event. In order to verify the code and to obtain better understanding of the severe core damage process, SEFDAN has been applied to analyses of the thermal response of fuel rods in the Power Burst Facility Severe Fuel Damage Scoping Test. This paper presents the calculated results and discusses, based on the results, on the phenomena that are important for prediction of the thermal response of fuel rods to a severe accident under the partially un-covered core condition. The calculated results are in good agreement with the experimental results. Namely the dry-out time of each elevation and the temperature behavior in both the slow heat-up and rapid temperature excursion processes are well simulated. The analysis indicates that fuel cladding temperature of the upper part of the test bundle would have reached the melting point of ZrO₂ and fuel center line temperature would have reached the melting point of UO₂ during a rapid temperature excursion which was caused by rapid decreasing of the dry-out level and accelerated by zirconium-water reaction in the lower part.     

应用MELCOR对FPT1棒束失效过程的模拟分析

应用MELCOR1.8.6程序对严重事故试验PHEBUS-FPT1进行了模拟分析.通过对棒束毁损过程中涉及的燃料棒过热、锆水反应、裂变产物释放和迁移、燃料熔融坍塌等现象和机理的建模计算,得到的结果和趋势与试验测量值进行了比较分析.分析结果表明:计算得到的棒束失效过程中发生重要事件与试验值较吻合;表征严重事故过程的重要现象--锫水反应所产生的氢气趋势,计算值与试验值比较一致;棒束栅元单一控制体划分,会使得计算得到的燃料峰值温度等表征严重事故来临时间晚于试验值;用CORSOR-M模型预测得到的大部分裂变产物核素释放总量要低于试验测量值,并且该模型较高的估计了氧化热对Xe、Cs、I、Te等易挥发核素释放的影响.     

Core temperature transient in the early phase of core uncovering at TMI-2

This paper describes a best-estimate analysis of the initial core boil-down and heat-up transient at Three Mile Island Unit (2) on 28 March 1979. This transient began shortly after all reactor coolant pumps were secured (100 min after reactor trip) and was terminated by a period of sustained high pressure injection of emergency cooling water, starting at 202 min.

The analysis is primarily directed to understanding the progression of core damage, rather than providing a detailed characterization of the core end-state condition. The latter objective can be achieved only after vessel head removal and visual examination.

The thrust of the present effort has been to: (1) develop a core coolant mixture level (dry-out level) calculation which satisfies the boundary conditions implied by various instrument responses and system operational characteristics; (2) couple the level calculation with a core heat-up modelto simulate the accumulation of thermal damage in the exposed, upper regions of the core; (3) compare calculated gross damage to the core with measurements of hydrogen and fission product releases subsequent to the accident.

Results indicate that:

1. (i) Observed containment hydrogen levels were due to Zircaloy/stainless steel corrosion that occurred during the period of core uncovering between the de-activation of the loop A reactor coolant pump (100 min after trip) and sustained operation of the high pressure injection system 100 min later. Appreciable zircaloy oxidation probably commenced at 150 min after trip, and continued at a high rate until the sustained high pressure injection at 202 min caused a major core quench.

2. (ii) There was some potential for fuel liquefaction. Calculations imply that peak fuel temperatures did not exceed the UO₂ pellet melting temperature, but 30% of the fuel was exposed to temperatures where liquid U---Zr---O alloys could have formed.

3. (iii) A substantial fission product release was obtained from fuel over-heating; however, an apparent disparity between the expected fission product release by calculation and the high range of fission product estimates obtained from plant measurements suggests that a significant release fraction may have originated from powdered or rubbilized fuel during cooldown. Additional gas releases may have developed from hot spots which persisted after core quench.

4. (iv) Steam temperatures in the upper plenum, at the outlet nozzle elevation, were generally below 900°C (1650°F) although this value was probably exceeded for a few min during the partial fuel quench caused by activation of the loop 2B reactor coolant pump, at 174 min after trip. The metal-work in the upper plenum, above the upper tieplate did not experience appreciable heating.

Thermal damage to the fuel and consequential weakening and mechanical disruption of the core was essentially complete 230 min after turbine trip.     

Some insights into the role of axial gas flow in fuel rod behaviour during the LOCA based on Halden tests and calculations with the FALCON-PSI code

A model for axial gas flow in a fuel rod during the LOCA is integrated into the FRELAX model that deals with the thermal behaviour and fuel relocation in the fuel rods of the Halden LOCA test series. The first verification was carried out using the experimental data for the inner pressure during the gas outflow after cladding rupture in tests 3, 4 and 5. Furthermore, the modified FRELAX model is implicitly coupled to the FALCON fuel behaviour code.The analysis with the new methodology shows that the dynamics of axial gas-flow along the rod and through the cladding rupture can have a strong influence on the fuel rod behaviour. Specifically, a delayed axial gas redistribution during the heat-up phase of the LOCA can result in a drop of local pressure in the ballooned area, which is eventually able to affect the cladding burst. The results of the new model seem to be useful when analysing some of the Halden LOCA tests (showing considerable fuel relocation) and selected cases of LOCA in full-length fuel rods. While the short rods used in the Halden tests only show a very small effect of the delayed gas redistribution during the clad ballooning, such an effect is predicted to be significant in the full-scale rods - with a power peak located sufficiently away from the plenum - resulting in a considerable delay of the predicted moment of cladding rupture.     

LWR severe accident simulation: synthesis of the results and interpretation of the first Phebus FP experiment FPT0

The international Phebus Fission Product (FP) programme, initiated in 1988 and performed by the French Institut de Radioprotection et de Sûreté Nucléaire (IRSN), investigates key phenomena of severe water reactor accidents. Six in-pile experiments were planned. Four have been successfully performed in 1993, 1996, 1999 and 2000.The first experiment, called FPT0, used uranium dioxide fuel of 4.5% enrichment in-situ irradiated for 9 days to a burn-up of 230 MWd t⁻¹. It was designed to reach significant fuel melting and to study low pressure fission products release and transport through the primary cooling system including a non-condensing steam generator and into the containment vessel. As the first test of the programme, FPT0 was intended to demonstrate the adequacy of the new, complex Phebus facility to simulate the anticipated phenomena and was the first attempt in using the new experimental results for verifying codes.The scientific results from FPT0 were sufficiently challenging that they deserve to be documented and interpreted. Since some of them did not correspond to the predicted and pre-calculated behaviour, the post-test analysis and interpretation period was rather long. Three years later, the second experiment FPT1, rather similar in its boundary conditions but using a fuel burned in a reactor (23 GWd t⁻¹), confirmed certain FPT0 results, helping in their final interpretation and removing doubts about possible fundamental shortcomings of the Phebus facility. More detailed experimental results of the test are available in the final test report deliverable on CD-ROM support. It can be obtained upon request from IRSN.¹This report retraces the history of FPT0 and its general programme context, and briefly describes the layout of the facility, supporting separate effect tests and computational tools. It then presents the synthesis of the results and of the international understanding reached concerning their interpretation, with emphasis on fuel and fission product behaviour.Finally, conclusions are presented about the impact of FPT0 on severe accident modelling with implications on source term evaluation and on accident prevention and mitigation studies.     

Thematic network for a Phebus FPT1 international standard problem (THENPHEBISP)

The THENPHEBISP 2-year thematic network started in December 2001, and was concerned with OECD/CSNI International Standard Problem 46, itself based on the Phebus FPT1 core degradation/source term experiment. The aim was to assess the capability of computer codes to model in an integrated way the physical processes taking place during a severe accident in a pressurised water reactor, from the initial stages of core degradation, the fission product transport through the primary circuit and the behaviour of the released fission products in the containment. ISP-46, coordinated by IRSN/DRS Cadarache, attracted 33 participating organisations, from 23 countries and international bodies, who submitted 47 base case calculations and 21 best-estimate calculations, using 15 different codes.The thermal behaviour of the fuel bundle and the hydrogen production were generally well captured, and good agreement for the core final state could be obtained with a suitable choice of bulk fuel relocation temperature, however this is unlikely to be representative of all plant studies so sensitivity calculations are needed with the modelling in its current state. Total volatile fission product release was simulated, but its kinetics, and the overall modelling of semi-volatile, low-volatile and structural material release (Ag/In/Cd, Sn) needs improvement. Overall retention in the circuit is well predicted, but calculations underestimate deposits in the upper plenum and overestimate those in the steam generator, also the volatility of some elements could be better predicted. Containment thermal hydraulics and depletion rate of aerosols are well calculated, but with difficulties related to partition amongst the deposition mechanisms. Calculation of iodine chemistry in the containment turned out to be more difficult. Its quality strongly depends of the calculation of release and transport in the integral codes. The major difficulties are related to the existence of gaseous iodine in the primary circuit and to the prediction of the amount of organic iodine in the gas phase. This paper summarises the results achieved and the implications for plant calculations.     

Analysis of MOX fuel behavior in reduced-moderation water reactor by fuel performance code FEMAXI-RM

To assess the feasibility of the 31% Pu-MOX fuel rod design of reduced-moderation water reactor (RMWR) in terms of thermal and mechanical behaviors, a single rod assumed to be irradiated in the core of RMWR up to 106 GWd/tHM has been analyzed by a fuel performance code FEMAXI-RM which is an extended version of FEMAXI-6 code. In the analysis, design specifications of fuel rod and irradiation conditions have been input, and available models of both MOX fuel and UO₂ fuel have been used as appropriate. The results are: fission gas release is several tens of percent, rod internal pressure does not exceed the coolant pressure, and the highest fuel center temperature is 2400 K, while cladding diameter increase caused by pellet swelling is within 1% strain. These predictions suggest that the MOX fuel rod integrity will be held during irradiation in RMWR, though actual behavior of MOX pellet swelling and cladding oxidation require to be investigated in detail.     

Iodine behaviour under LWR accident conditions: Lessons learnt from analyses of the first two Phebus FP tests

The International Phebus Fission Product programme, initiated in 1988 and performed by the French “Institut de Radioprotection et de Sûreté Nucléaire” (IRSN), investigates through a series of in-pile integral experiments, key phenomena involved in light water reactor (LWR) severe accidents. The tests cover fuel rod degradation and the behaviour of fission products released via the primary coolant circuit into the containment building.The results of the first two tests, called FPT0 and FPT1, carried out under low pressure, in a steam rich atmosphere and using fresh fuel for FPT0 and fuel burned in a reactor at 23 GWdt⁻¹ for FPT1, were immensely challenging, especially with regard to the iodine radiochemistry. Some of the most important observed phenomena with regard to the chemistry of iodine were indeed neither predicted nor pre-calculated, which clearly shows the interest and the need for carrying out integral experiments to study the complex phenomena governing fission product behaviour in a PWR in accident conditions. The three most unexpected results in the iodine behaviour related to early detection during fuel degradation of a weak but significant fraction of volatile iodine in the containment, the key role played by silver rapidly binding iodine to form insoluble AgI in the containment sump and the importance of painted surfaces in the containment atmosphere for the formation of a large quantity of volatile organic iodides.To support the Phebus test interpretation small-scale analytical experiments and computer code analyses were carried out. The former, helping towards a better understanding of overall iodine behaviour, were used to develop or improve models while the latter mainly aimed at identifying relevant key phenomena and at modelling weaknesses. Specific efforts were devoted to exploring the potential origins of the early-detected volatile iodine in the containment building. If a clear explanation has not yet been found, the non-equilibrium chemical processes favoured in the primary coolant circuit and the early radiolytic oxidation of iodides in the condensed water films are at present the most likely explanations. Models that were modified or developed and embodied in the computer codes for organic iodide formation/destruction in the gas phase and Ag–I reactions in the sump lead, in agreement with the Phebus findings respectively to greatly enhanced organic iodide formation kinetics and long term concentration in the containment atmosphere on one hand and, in the conditions of Phebus experiments, to significantly limited molecular iodine volatilisation from the sump in so far as silver was in excess compared to iodine, on the other hand. Organic iodides then quickly gain in importance and become the predominant volatile iodine species at long term.