全厂断电时安全壳直接加热的概率研究

安全壳直接加热(DCH)是导致安全壳早期失效的潜在因素,本文应用基于风险导向的事故分析方法(ROAAM),对堆芯碎片中UO_2的质量和Zr的氧化份额的概率密度分布抽样,对安全壳直接加热模型TCE(Two-cell Equilibrium)编程,将抽样结果带人TCE模型中计算,得到安全壳压力峰值的累积概率分布和安全壳失效概率,研究压水堆全厂断电始发事故下轴封破口面积不同的情况对下封头失效后安全壳压力峰值的影响。其中TCE模型的输入数据由严重事故分析程序计算给出。     

三代压水堆核电厂DCH参数敏感性研究

安全壳直接加热（DCH）是导致安全壳早期超压的主要贡献之一,严重威胁安全壳完整性,并可能造成放射性物质早期大量不可控释放。本文以我国某三代压水堆为研究对象,首先基于风险导向的事故分析方法（ROAAM）,利用双隔间平衡（TCE）模型编写程序计算典型事故工况下的DCH载荷;其次结合安全壳失效概率曲线得出DCH现象造成的安全壳失效概率;最后对计算程序中不易得到的参数或经验值等不确定性较大的参数进行敏感性分析,归纳敏感性分析结果,找出敏感参数的不确定因素。结果表明：熔融物质量、堆腔几何设计、安全壳布置设计会直接影响DCH后果。     

A two-cell equilibrium model for predicting direct containment heating

This paper discusses two adiabatic equilibrium models. Assessment and validation of the separate effects (kinetic) models and the parameters (i.e. particle size) that control them are not required. The first, a single-cell equilibrium model, places a true upper bound on direct containment heating (DCH) loads. This upper bound, when compared with the entire DCH database, often far exceeds experiment observations by a margin too large to be useful in reactor analyses. The single-cell model is used as a conceptual seed for a two-cell model. A two-cell equilibrium (TCE) model is developed that captures the dominant mitigating features of containment compartmentalization and the noncoherence of the entrainment and blowdown processes. The existing DCH database has been used to extensively validate the TCE model. DCH loads are shown to be insensitive to physical scale and details of the subcompartment geometry. A simple model is developed to predict the coherence of debris dispersal and reactor coolant system blowdown. The coherence ratio is independent of physical scale and only weakly dependent on cavity design.     

Investigations into the physical phenomena and mechanisms that effect direct containment heating loads

Integral direct containment heating (DCH) experiment results are presented. The results are analyzed and discussed for the insights they have given into understanding the important physical phenomena and mechanisms that effect DCH loads to the containment. Particular attention is paid to (1) debris dispersal from the cavity and containment structure trapping, (2) hydrogen production and combustion, (3) the importance of difference in corium simulants used in integral DCH experiments and (4) corium debris quenching by flooded cavities. It is found that much has been learned about DCH phenomena that can be used for modeling and assessing potential containment loads.     

Closure of the direct containment heating issue for Zion

This paper, which was originally published in more detail (M.M. Pilch, M.D. Allen, D.L. Knudsen, D.W. Stamps and E.L. Tadios, Rep. NUREG/CR-6075, Supplement 1, 1994b (Sandia National Laboratories, Albuquerque, NM)), provides closure of the direct containment heating (DCH) issue for the Zion plant. It incorporates the comments and suggestions of the peer reviewers of NUREG/CR-6075 (M.M. Pilch, H. Yan, and T.G. Theofanous, Rep. NUREG/CR-6075, SAND93-1535, 1994a (Sandia National Laboratories, Albuquerque, NM)) and specifically includes assessments of four new splinter scenarios defined in working group meetings and modeling enhancements recommended by the working groups. In the four new scenarios, consistency of the initial conditions has been implemented by using insights from systems-level codes. was used to analyze three short-term station blackout cases with different leak rates. In all three cases, the hot leg or surge line failed well before the lower head and thus the primary system depressurized to a point where DCH was no longer considered a threat. However, these calculations were continued to lower head failure in order to gain insights that were useful in establishing the initial and boundary conditions. The most useful insights are that the reactor coolant system pressure is low at vessel breach, metallic blockages in the core region do not melt and relocate into the lower plenum, and melting of upper plenum steel is correlated with hot leg failure. The output was used as input to to assess the containment conditions at vessel breach. The containment-side conditions predicted by are similar to those originally specified in NUREG/CR-6075.The methodology originally developed in NUREG/CR-6075 (M.M. Pilch, H. Yan, and T.G. Theofanous, Rep. NUREG/CR-6075, SAND93-1535, 1994a (Sandia National Laboratories, Albuquerque, NM)) was used to analyze the new splinter scenarios. Some modeling enhancements in response to working group discussions were implemented for these analyses. The entrainment of hydrogen pre-existing in the atmosphere into a burning jet was examined more carefully. In addition, the impact of DCH-induced deflagrations on DCH loads was quantified. A new computational tool—the two-cell equilibrium—Latin hypercube sampling (TCE-LHS) code—was developed for this effort to perform Monte Carlo sampling of the scenario distributions. The TCE-LHS code was benchmarked against the original Scenario I calculations in NUREG/CR-6075 performed using the code, which is based on the method of discrete probability distributions. The results were in excellent agreement.The analyses of the new scenarios showed no intersection of the load distributions and the containment fragility curves, and thus the containment failure probability was negligible for each scenario. These supplemental analyses complete closure of the DCH issue for Zion.     

使用FCI-CMFD软件分析岭澳二期核电厂直接安全壳加热事故

采用分析熔融物与冷却剂反应(FCI)的三维多相流数值计算软件MC3D,建立岭澳二期核电厂模型,对严重事故下可能发生的直接安全壳加热(DCH)现象进行了模拟和分析。为准确预测事故现象,本文结合全厂断电事故后期参数与岭澳二期核电厂核岛几何模型,模拟事故过程。计算得出了事故下安全壳内气体温度场、熔滴体积份额场、速度场及压力随时间的变化。结果表明：直接安全壳加热事故会在短时间内引起安全壳内压力和局部温度的迅速上升。     

Effect analysis of the intentional depressurization on fission product behavior during TMLB' severe accident

It has been found that the pressure in the reactor coolant system (RCS) remains high in some severe accident sequences at the time of reactor vessel failure, with the risk of causing direct containment heating (DCH).Intentional depressurization is an effective accident management strategy to prevent DCH or to mitigate its consequences. Fission product behavior is affected by intentional depressurization, especially for inert gas and volatile fission product. Because the pressurizer power-operated relief valves (PORVs) are latched open, fission product will transport into the containment directly. This may cause larger radiological consequences in containment before reactor vessel failure. Four cases are selected, including the TMLB' base case and the opening one, two and three pressurizer PORVs. The results show that inert gas transports into containment more quickly when opening one and two PORVs,but more slowly when opening three PORVs; more volatile fission product deposit in containment and less in reactor coolant system (RCS) for intentional depressurization cases. When opening one PORV, the phenomenon of revaporization is strong in the RCS.     

On the analysis of containment heat transfer following a LOCA

This paper is concerned with containment heat removal by steam condensation on passive heat sinks following a loss-of-coolant accident. A literature search and survey of technical publications including both experimental and analytical results that are relevant to the heat transfer mechanism inside the containment are summarized. Several small-scale experiments that relate to correlations used in licensing calculations as well as the large-scale Carolinas Virginia Tube Reactor Containment (CVTR) test data are discussed in detail. A theoretical model is then developed to predict the heat transfer rate. It is found that for the post-accident containment conditions, the heat transfer rate can be approximated by a closed form equation. The predictions compare favourably with various existing experimental data and offer logical explanations for the discrepancies among the data. It is observed that several crucial containment parameters are absent in the heat transfer correlations which are currently used for containment licensing calculations.     

Direct containment heating integral effects tests in geometries of European nuclear power plants

The DISCO test facility at Forschungszentrum Karlsruhe (FZK) has been used to perform experiments to investigate direct containment heating (DCH) effects during a severe accident in European nuclear power plants, comprising the EPR, the French 1300 MWe plant P’4, the VVER-1000 and the German Konvoi plant. A high-temperature iron–alumina melt is ejected by steam into scaled models of the respective reactor cavities and the containment vessel. Both heat transfer from dispersed melt and combustion of hydrogen lead to containment pressurization. The main experimental findings are presented and critical parameters are identified.The consequences of DCH are limited in reactors with no direct pathway between the cavity and the containment dome (closed pit). The situation is more severe for reactors which do have a direct pathway between the cavity and the containment (open pit). The experiments showed that substantial fractions of corium may be dispersed into the containment in such cases, if the pressure in the reactor coolant system is elevated at the time of RPV failure. Primary system pressures of 1 or 2 MPa are sufficient to lead to full scale DCH effects. Combustion of the hydrogen produced by oxidation as well as the hydrogen initially present appears to be the crucial phenomenon for containment pressurization.     

On the failure modes of alternative containment designs following postulated core meltdown

The containment response to a postulated core meltdown accident in a PWR ice condenser containment, a BWR Mark III containment and a BWR non-inerted Mark I containment has been examined to see if the WASH-1400 containment failure mode judgement for the Surry large, dry containment and the Peach Bottom Mark I inerted-containment are likely to be appropriate for these alternative containment plant designs. For the PWR, the representative accident chosen for the analysis is a large cold leg break accompanied by a loss of all electric power while the BWR representative event chosen is a recirculation line break without adequate core cooling function. Two containment event paths are studied for each of these two cases, depending on whether or not containment vapor suppression function is assumed to be available. Both the core and the containment pressure and temperature response to the accident events are computed for the four time intervals which characterize (a) blowdown of the pipe break, (b) core melt, (c) vessel melt-through, and (d) containment foundation penetration. The calculations are based on a best estimate of the most probable sequence, but certain phenomena and events were followed down multiple tracks. These include the temperature of the non-condensibles escaping the ice condenser into the upper compartment, the performance of the pressure suppression system, the distribution of non-condensibles between compartments, and the degree and rate of combustion of hydrogen generated from metal-water reactions. For the PWR ice condenser case, results indicate that the containment would be breached by (i) steam overpressurization during the blowdown period (time less than 20 sec) if the ice condenser fails to perform its function, (ii) by overpressurization and thermal stress during the core melt period if 25% or more of the core zirconium reacts with water followed by hydrogen burning and, and (iii) by the overpressurization due to non-condensibles before containment floor penetration is completed. For the BWR Mark III case, similar conclusions can be drawn for the loss of vapor suppression, and for the hydrogen burning if the extent of zirconium-water reaction is more than 35% of the core inventory. If the hydrogen burning fails to materialize, the containment can retain its integrity until containment meltthrough provided the melting is confined to the reactor pedestal area. It appears that the non-inerted Mark I containment is not so vulnerable to overpressurization from hydrogen burning as the Mark III; however, acceptable temperatures may be exceeded.     

Evaluation of uncertainties in risk emanating from a number of severe accident issues

A study has been performed to estimate, for a particular pressurized water reactor, the uncertainty in risk associated with a number of key phenomenological issues. A second objective was to distinguish the individual importance of the various issues as contributors to the overall uncertainty in risk. The issues considered touched upon the areas of system behavior, containment loading, containment performance, and fission product source term behavior. It was found that the most important source of uncertainty for the plant in question (Surry) was direct containment heating (i.e., the transfer of heat from the core debris to the containment atmosphere when the debris is ejected at high pressure from the reactor vessel and dispersed throughout the atmosphere). Other significant issues included hydrogen burning, containment failure pressure, aerosol agglomeration uncertainties, the frequency of check valve failures leading to a loss-of-coolant accident (LOCA) outside containment, and the potential for having a LOCA induced by high temperatures in the reactor coolant system.     

Direct containment heating experiments for Vandellos and ASCo nuclear power plants

The ongoing IPE studies for the Vandellos and ASCo nuclear power plants require evaluation of accident phenomena that have been perceived to potentially challenge containment integrity including direct containment heating (DCH). Analyses and scaled experiments performed to date indicated that the lower containment structures play a substantial role in mitigating the extent of DCH given a high pressure melt ejection. Since the geometry is judged to be of major importance, linearly scaled experiments were conceived and conducted to evaluate the role of such structures in the Vandellos and ASCo specific configurations. The Vandellos test configuration with an initally dry cavity and significant exhaust area for the instrument tunnel resulted in the dispersal of a majority of the debris from the instrument tunnel into the lower compartment. The test of the ASCo configuration with an initially wet reactor cavity and limited exhaust area from the instrument tunnel exhibited the retention of the majority of the debris within the instrument tunnel and reactor cavity. The observed pressure responses in these scaled experiments for the seal table room, lower containment vessel, and upper containment vessel were all less than the containment design basis pressure. These test results contribute to the existing technical basis for concluding that direct containment heating would not represent a challenge to the integrity of these containments.     

Detailed Evaluation of RCS Boundary Rupture during High-Pressure Severe Accident Sequences

A depressurization possibility of the reactor coolant system (RCS) before a reactor vessel rupture during a high-pressure severe accident sequence has been evaluated for the consideration of direct containment heating (DCH) and containment bypass. A total loss of feed water (TLOFW) and a station blackout (SBO) of the advanced power reactor 1400 (APR1400) has been evaluated from an initiating event to a creep rupture of the RCS boundary by using the SCDAP/RELAP5 computer code. In addition, intentional depressurization of the RCS using power-operated safety relief valves (POSRVs) has been evaluated. The SCDAPRELAP5 results have shown that the pressurizer surge line broke before the reactor vessel rupture failure, but a containment bypass did not occur because steam generator U tubes did not break. The intentional depressurization of the RCS using POSRV was effective for the DCH prevention at a reactor vessel rupture.     

Modeling pressure suppression pool hydrodynamics in the ABWR containment

This paper focuses on the assessment of pressure suppression pool hydrodynamics in the advanced boiling water reactor (ABWR) containment under design-basis, loss-of-coolant accident (LOCA) conditions. The paper presents a mechanistic model for predicting various suppression pool hydrodynamics parameters. A phenomena identification and ranking table (PIRT) applicable to the ABWR containment pool hydrodynamics analysis is used as a basis for the development of the model. The highly ranked phenomena are represented by analytic equations or empirical correlations. The best estimate and several sensitivity calculations are performed for the ABWR containment using this model. Results of the sensitivity calculations are also presented that demonstrate the influence of key model parameters and assumptions on the pool hydrodynamics parameters. A comparison of model predictions to the results of the licensing analyses shows reasonable agreement. Comparison of the results of the proposed model to experimental data shows that the model predicted top vent clearance time, the pool swell height, and the bubble breakthrough elevation are within 10% of the data. The predicted pool surface velocity and the liquid slug thickness are within 30% of the measurements, which is considered adequate given the large uncertainties in the experimental measurements.     

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...     

Hydrogen management and overpressure protection of the containment for future boiling water reactors

New design and evaluation method for hydrogen management of containment atmosphere have been developed for application in the future boiling water reactor (BWR). These are intended as a part of consideration of severe accidents in the course of design so as to assure a high level of confidence that a large release of radioactivity to the environment that may result in unacceptable social consequences can reasonably be avoided. Emphasis on hydrogen management and protection against overpressure failure is based on the insights from probabilistic safety assessments (PSAs) that late phase overpressure (and associated leakage) and molten corium concrete reaction (MCCI) need attention to ensure that containment remains intact, in case energetic challenges to the containment such as DCH (direct containment heating) or FCI (fuel coolant interactions) are practically eliminated by design or resolved from risk standpoint of view. The authors studied the use of palladium-coated tantalum for hydrogen removal from containment atmosphere in order to avoid pressurization of the containment with small free volume by non-condensable gas and steam. Its effectiveness for ABWR (advanced boiling water reactor) containment was evaluated using laboratory test data. Although further experimental studies are necessary to confirm its effectiveness in real accident conditions, the design is a promising option and one that could be backfitted upon necessity to existing plants for which pressure retaining capability cannot be altered. Also new evaluation method for flammability control under severe accident conditions was developed. This method employes a realistic assessment of the amount of oxygen and hydrogen gases generated by radiolytic decomposition of water under severe accident conditions and their subsequent transport from water to containment atmosphere.     

压水堆核电厂高压熔堆严重事故序列分析

压水堆核电厂的高压熔堆事故覆盖了大部分的严重事故序列,且具有很大的潜在威胁。根据我国900MW压水堆核电厂的概率安全分析（PSA）结果选取了丧失厂外电、未能紧急停堆的预期瞬态、二回路管道破口、一回路小破口和蒸汽发生器传热管破裂5种典型的高压熔堆严重事故序列,并使用SCDAP/RELAP5程序对这些事故序列的进程和后果进行了计算分析。计算结果表明：5种典型高压熔堆事故序列可能导致高压熔喷和安全壳直接加热风险,可能引起安全壳早期失效,很有必要采取相应的一回路卸压措施。     

Semi-analytical modeling of a PWR lower head failure under severe accident conditions using an axisymetrical shell theory

The USNRC/SNL OLHF program was carried out within the framework of an OECD project. This program consisted of four one-fifth scale experiments of a reactor pressure vessel (RPV) lower head failure (LHF) under well controlled internal pressure and large throughwall temperature differentials; the objectives were to characterize the mode, timing and size of a possible PWR lower head failure in the event of a core meltdown accident. These experiments should also lead to a better understanding of the mechanical behavior of the reactor vessel lower head, which is of importance both in severe accident assessment and the definition of accident mitigation strategies. A well-characterized failure of the lower head is of prime importance for the evaluation of the quantity of core material that can escape into the containment, since this defines the initial conditions for all ex-vessel events. A large quantity of escaping corium may lead to direct heating of the containment or ex-vessel steam explosion. These are important issues due to their potential to cause early containment failure. The experiments also provide data for model development and validation. For our part, as one of the program partners, a 2D semi-analytical model has been developed and used to simulate these experiments. The aim of this effort is to develop a simplified but well predicting code that can be then implemented in European integral severe accident computer codes (ASTEC, ICARE/CATHARE). This paper presents the detailed mathematical formulation of this simplified method which is used to interpret the experimental results. The axi-symmetric shell theory under internal pressure proposed by Timoshenko has been utilised. The solution to the equilibrium equations is presented, with particular attention to the Rabotnov analytical formula. The radius and the polar angle of the deformed structure have been written as analytical expressions in order to take the large displacements and large strains into account using our mathematical formulation. The Norton type creep law and the Kachanov damage law have been used. Several failure criteria were used in the calculations and their effect on the numerical results is discussed. This 2D semi-analytical model gives very satisfactory results when compared, with the experimental and numerical results that were presented recently in the Benchmark calculations based on the first test of the OLHF program. The performance of this model is also illustrated by its capacity to accurately simulate the deformation of the lower head, including the variation of wall thickness.     

Experimental investigations of pressure and temperature loads on a containment after a loss-of-coolant accident

For the design of an LWR containment one of the important conditions to be considered is the rapid rise of internal pressure and temperature caused by a loss-of-coolant accident (LOCA) of the primary cooling system. The phenomena occurring within a containment during a LOCA are currently investigated through experiments with a model containment. The experimental results are compared with the results of model calculations to improve the calculational methods.An experimental facility was built, consisting of a primary coolant circuit and a special model containment. The model containment, built in conventional reinforced concrete, has a diameter of 12 m, a height of 12.5 m, a capacity of 580 m³ and is designed for an internal pressure of 6 bar. The interior is divided by concrete walls and removable partitions into several compartments, which are interconnected through openings with adjustable cross sections. By exchanging the removable partitions it is possible to modify the interior of the containment and to simulate different containment shapes. For the first experiments a PWR configuration with nine compartments has been installed. The model scales of the compartment volumes and the overflow areas are about 1:64 compared to the 1200 MW PWR plant Biblis A.Up to now the test facility has been used for four trial runs and nine PWR LOCA experiments with single- and double-ended pipe ruptures of 100 mm dia. in a steam generator compartment and in the nozzle compartment. The initial conditions of the pressurized water in the coolant circuit before rupture were 140 bar and 290°C. About 0.1 sec after the rupture the flow rate at the site of rupture reaches its maximum of about 400 kg/sec (single-ended rupture) and 800 kg/sec (double-ended rupture). From the compartment where the rupture takes place a water-steam-air mixture streams through openings into the other compartments of the containment. Differential pressures between the compartments were measured with maximums of up to a few bar 0.15–0.5 sec after rupture, depending on the positions of rooms and transducers.Approximately 30–40 sec after rupture the blowdown has finished and the pressure in the containment has reached about 4–5 bar. The maximum pressure in a model containment is lower and the decrease of the pressure by condensation is faster than in a full-scale containment, due to the greater ratio of inner surface area to volume of a model containment. During blowdown the temperature of the containment atmosphere rises to about 150°C. Several minutes later the temperature of the concrete walls has increased non-uniformly causing considerable stress in the walls. Approximately 30 min after rupture measurements on the outside of the outer containment wall show a temperature-caused strain of about 30–60% of the maximum pressure-caused strain. A comparison between experiments and calculations shows discrepancies indicating the need for further development of calculational methods.