Model for melt blockage (slug) relocation and physico-chemical interactions during core degradation under severe accident conditions

The model describing massive melt blockage (slug) relocation and physico-chemical interactions with steam and surrounding fuel rods of a bundle is developed on the base of the observations in the CORA tests. Mass exchange owing to slug oxidation and fuel rods dissolution is described by the previously developed 2D model for the molten pool oxidation. Heat fluxes in oxidising melt along with the oxidation heat effect at the melt relocation front are counterbalanced by the heat losses in the surrounding media and the fusion heat effect of the Zr claddings attacked by the melt. As a result, the slug relocation velocity is calculated from the heat flux matches at the melt propagation front (Stefan problem). A numerical module simulating the slug behaviour is developed by tight coupling of the heat and mass exchange modules. The new model demonstrates a reasonable capability to simulate the main features of the massive slug behaviour observed in the CORA-W1 test.     

Iodine-steel reactions under severe accident conditions in light-water reactors

Owing to large surface areas, the reaction of volatile molecular iodine (I₂) with steel surfaces in the containment may play an important role in predicting the source term to the environment. Both wall retention of iodine and conversion of volatile into non-volatile iodine compounds at steel surfaces have to be considered. Two types of laboratory experiment were carried out at Siemens (KWU) in order to investigate the reaction of I₂ at steel surfaces representative for German power plants.

• 1. 

  (1) For steel coupons submerged in an I2 solution at T = 50, 90 or 140 °C the reaction rate of the I2−I− conversion was determined. No iodine loading was observed on the steel in the aqueous phase tests. I2 reacts with the steel components (Fe, Cr or Ni) to form metal iodides on the surface which are all immediately dissolved in water under dissociation into the metal and the iodide ions. From these experiments, the I2−I− conversion rate constants over the temperature range 50–140 °C as well as the activation energy were determined. The measured data are suitable to be included in severe accident iodine codes such as IMPAIR.

• 2. 

  (2) Steel tubes were exposed to a steam-I2 flow under dry air at T = 120 °C and steam-condensing conditions at T = 120 and 160 °C. In dry air, I2, was retained on the steel surface and a deposition rate constant was measured. Under steam-condensing conditions there is an effective conversion of volatile I2 to non-volatile I− which is subsequently washed off from the steel surface. The I2−T− conversion rate constants suitable for modelling this process were determined. No temperature dependence was found in the range 120–160 °C.

     

Hydrogen distribution tests under severe accident conditions at the large-scale HDR-facility

The E11 tests series covered a combination of important issues dominating the physical phenomena controlling the hydrogen distribution mechanisms, namely: large-scale, multi-compartment, geometry with large-sized dome volume, high gas release rates, multiple steam and gas injection phases at different axial positions and examinations of the efficiencies of mitigative system features including the impact of external sprays at the top of the dome.The test series consisted of a total of eight different experiments covering all aspects of the H₂-distribution and potential mitigation features.A total of 700 sensors were applied during these experiments.The paper outlines experimental and computational results of tests E11.2 and E11.4 which were chosen for two computational PHDR-Benchmark Exercises in the context of blind posttest predictions with broad international participation applying the majority of known computer codes. In addition test E11.2 was selected as an open post-test, OECD International Standard Problem No. 29 (H. Karwat, Distribution of hydrogen within the HDR-containment under severe accident conditions — Task specifications (July 1990)) which is presently in progress.     

Effects of accident management strategy on the severe accident environmental conditions

This paper presents a methodology utilizing an accident management strategy in order to determine accident environmental conditions to be used as inputs to equipment survivability assessments. In the case that there is a well-established accident management strategy for a specific nuclear power plant (NPP), an application of this tool can provide a technical rationale on equipment survivability assessment so that plant-specific and time-dependent accident environmental conditions could be practically and realistically defined in accordance with the equipment and instrumentation required for the accident management strategy or appropriate actions. For this work, three different tools are introduced; probabilistic safety assessment (PSA) outcomes, major accident management strategy actions, and accident environmental stages (AESs). In order to quantitatively investigate an applicability of accident management strategy on equipment survivability, the accident simulation for most likely scenario in Korean standard nuclear power plants (KSNPs) is performed with the MAAP4 code. The accident management guideline (AMG) actions such as the reactor coolant system (RCS) depressurization, water injection into the RCS, the containment pressure and temperature control, and hydrogen concentration control in containment are applied. The effects of these AMG actions on the accident environmental conditions are investigated by comparison to actions from previous normal accident simulation, especially focusing on equipment survivability assessment. As a result, the AMG-involved case shows the higher accident consequences along the accident environmental stages. This implies that plant-specific AMG actions need to be considered in order to determine accident environmental conditions in equipment survivability assessments.     

External cooling of a reactor vessel under severe accident conditions

The TMI-2 accident demonstrated that a significant quantity of molten core debris could drain into the lower plenum during a severe accident. For such conditions, the Individual Plant Examinations (IPEs) and severe accident management evaluations, consider the possibility that water could not be injected to the RCS. However, depending on the plant specific configuration and the accident sequence, water may be accumulated within the containment sufficient to submerge the lower head and part of the reactor vessel cylinder. This could provide external cooling of the RPV to prevent failure of the lower head and discharge of core debris into the containment.This paper evaluates the heat removal capabilities for external cooling of an insulated RPV in terms of (a) the water inflow through the insulation, (b) the two-phase heat removal in the gap between the insulation and the vessel and (c) the flow of steam through the insulation. These results show no significant limitation to heat removal from the bottom of the reactor vessel other than thermal conduction through the reactor vessel wall. Hence, external cooling is a possible means of preventing core debris from failing the reactor, which if successful, would eliminate the considerations of ex-vessel steam explosions, debris coolability, etc. and their uncertainties. Therefore, external cooling should be a major consideration in accident management evaluations and decision-making for current plants, as well as a possible design consideration for future plants.     

Leakage potential of LWR containment penetrations under severe accident conditions

This paper is an overview of a Sandia National Laboratories, Albuquerque (SNLA) study of the performance of mechanical penetrations in light-water reactor (LWR) containment buildings that are subjected to severe accident environments. The study is concerned with modes of failure as well as the magnitude of leakage. The following tests have been completed, are under way, or are planned: (a) seals and gaskets have been tested to register the effects of radiation aging, thermal aging, seal geometry, and squeeze on seal and gasket materials in severe accident environments; (b) the performance of a full-scale airlock will be evaluated at severe accident temperature and pressure levels; (c) personnel airlock and equipment hatch tests were made on a model of a steel containment building; and (d) tests of mechanical penetrations are planned as part of a test on a model of a reinforced concrete building. This program is part of an overall US Nuclear Regulatory Commission (USNRC) effort to evaluate the integrity of LWR containment buildings.     

The effects of debris bed's prototypical characteristics on corium coolability in a LWR severe accident

This paper is concerned with coolability assessment of a debris bed formed in fuel coolant interactions (FCIs) during a hypothetical severe accident in a light water reactor (LWR). The focus is placed the potential effect of the bed's prototypical characteristics on its coolability, in terms of (i) porosity range, (ii) multi-dimensionality, (iii) inhomogeneity, (iv) particle morphology, and (v) heat generation method (e.g. volumetric heating vs. local heaters). The analysis results indicate availability of substantial coolability margins compared to previous assessments based on models and experiments using an idealized bed configuration (e.g. 1D homogenous debris layer). Notably, high porosity (up to 70%) of debris beds, obtained in experiments and expected to be the case of prototypical debris beds, could increase the dryout heat flux by 100% and more, depending on particle size, compared with the dryout heat flux predicted for debris beds with traditionally assumed porosity of approximately 40%. Bed inhomogeneity represented by micro-channels in a mini bed is predicted to enhance the dryout heat flux by up to ∼50%, even if the micro-channels occupy only a small volume fraction (e.g., less than 4%) of the bed. The effect of coolant side ingress into a multidimensional bed is predicted to enhance the dryout heat flux by up to 40% for the beds analyzed.     

严重事故条件下压力容器完整性评价的研究进展

堆芯熔融物堆内滞留(In-Vessel Retention,IVR)是以AP1000为代表的第三代轻水反应堆严重事故管理的重要策略之一,也是严重事故条件下保证压力容器完整性(Reactor Vessel Integrity,RVI)的典型方法之一.该文综述了国外在严重事故条件下压力容器完整性试验研究和理论分析的现状,总...     

Features of a control blade degradation observed in situ during severe accident conditions in boiling water reactors

In the present paper, new results using in situ video are presented regarding boiling water reactor (BWR) control blade degradation up to 1750 K at the beginning of a nuclear severe accident. Energy-dispersive X-ray spectrometry (EDS) mapping indicated stratification of the absorber blade melt with formation of a chromium and boride-enriched layer. High-content-B- and C-containing material with increased melting temperature acted like a shielding and was found to prevent further relocation of control blade claddings. The interacted layers around the B₄C-granules prevented direct steam attack of residual B₄C. The results provide new insights for understanding of the absorber blade degradation mechanism under reducing conditions specific to Fukushima Dai-Ichi Unit 2 resulting from prolonged steam starvation.     

Prediction of structural integrity of steam generator tubes under severe accident conditions

Available models for predicting failure of flawed and unflawed steam generator tubes under normal operating and design-basis accident conditions are reviewed. These rate-independent flow stress models are inadequate for predicting failure of steam generator tubes under severe accident conditions because the temperature of the tubes during such accidents can reach as high as 800°C where creep effects become important. Therefore, a creep rupture model for predicting failure was developed and validated by tests on unflawed and flawed specimens containing axial and circumferential flaws and loaded by constant as well as ramped temperature and pressure loadings. Finally, tests were conducted using pressure and temperature histories that are calculated to occur during postulated severe accidents. In all cases, the creep rupture model predicted the failure temperature and time more accurately than the flow stress models.     

Behavior of a model VVER fuel assembly under the conditions of an unanticipated accident with cooldown by pouring water from above

Investigations of the behavior of model VVER-1000 fuel assemblies have been performed on the PARAMETR bench under conditions of an unanticipated accident with reserve water poured from above. The results of materials engineering investigations of a model fuel assembly, using optical and electron microscopy, as well as x-ray microspectral and x-ray structural analysis are presented. The results of these investigations are necessary for developing and verifying methods of computing improved evaluations for the purpose of modeling accidents with substantial core damage. __________ Translated from Atomnaya énergiya, Vol. 104, No. 5, pp. 276–279, May, 2008.     

Experiments to evaluate behavior of containment piping bellows under severe accident conditions

Bellows are an integral part of the containment pressure boundary in nuclear power plants. They are used at piping penetrations to allow relative movement between piping and the containment wall. In a severe accident they may be subjected to high pressure and temperature and a combination of axial and lateral deflections. A test program to determine the leak-tight capacity of containment penetration bellows is being conducted at Sandia National Laboratories, Albuquerque, NM. Several different bellows geometries representative of actual containment bellows are being subjected to extreme deflections along with pressure and temperature loads. The bellows geometries and loading conditions are described along with the testing apparatus and procedures. A total of 13 tests have been conducted. The tests showed that bellows are capable of withstanding relatively large deformations up to or near the point of full compression before developing leakage. The test data are presented and discussed.     

Methodology for developing channel disassembly criteria under severe accident conditions for PHWRs

This paper presents a methodology to develop a model for disassembly of the coolant channels in Pressurized Heavy Water Reactors under severe accident conditions. This model gives criteria to decide when under severe accident condition coolant channels will rupture due to deterioration in material properties at high temperatures and increase in load due to creep sag of channels above it and hence get disassembled. Presently available severe accident codes use simplistic and optimistic criteria based on a predefined temperature to predict failure of fuel channels and an explicit criterion for disassembly of the channel is not covered. The coolant channel disassembly model developed in this paper is based on modeling the sag and pile up of channels. A uniform temperature along the length of the channel is assumed. The disassembly of the channel is assumed when the total strain at any location exceeds the failure strain for a given temperature. A 3D failure surface which is a plot of time to failure, temperature of the calandria tube and load on the calandria tubes (on account of no of channels piled up) is developed. This failure surface can be used as an input to severe accident codes to predict the progress of the core disassembly. A set of failure surfaces is recommended to be used if metal–water reaction on the outer surface is to be accounted for loss in ductility due to metal water reaction. The temperature transient of the calandria tube for a severe accident obtained from system thermal hydraulic codes can be mapped onto the failure surface. The time at which the mapped transient crosses the failure surface gives the time at which the calandria tube is disassembled. This disassembly model is an engineered model which is much more realistic as compared to the current temperature based conservative model for predicting severe accident progression.     

Study on the structural analysis of the containment and the hydrogen detonation problem under severe accident conditions

A brief review of the possible failure modes of a reactor containment is made. And, in light of recent research achievements, the threat of loss of structural integrity of the containment due to gradual overpressure and hydrogen detonation is emphasized.With regard to the methods for assessing the ultimate capacity of a prestressed concrete containment, the simplified engineering method proposed by Bechtel Inc. is verified by comparing the results calculated by the Bechtel method with that obtained by more sophisticated methods and experimental investigations. The comparison demonstrates that the accuracy of the Bechtel method is quite satisfactory.As an extension of the idea of probabilistic evaluation of the occurrence of local hydrogen detonation in a containment, that was suggested by experts from Sandia Laboratory, a new approach based on the use of fuzzy mathematics is proposed. The fuzzy evaluation approach provides a flexible mathematical framework for systematical collection and processing of experimental data, and is a simple and quite effective tool for treating the problem of assessing the possibility of local hydrogen detonation within a containment.     

Dissolution behavior of core structure materials by molten corium in boiling water reactor plants during severe accidents

For the decommissioning of the Fukushima Daiichi Nuclear Power Plant, it is necessary to consider the access route to the fuel debris for its removal, which can be determined by knowing the corruption situation of the core support structure. To predict the damage condition of reactor vessel, dissolution behavior of the core structure material should be understood. In this study, the dissolution behavior of core structure materials (stainless steel) by molten metallic corium (stainless steel + B₄C) originated from control rod and its cladding was investigated. As a result of immersion experiment, it was found that there were two types of dissolution mode in this system: (1) chemical dissolution by eutectic reaction between Fe and B and (2) physical dissolution caused by the grains falling off from solid steel due to infiltration of molten metal. Moreover, on the basis of kinetic analysis, it was considered that the chemical dissolution in this system was slow. Therefore, the dissolution is considered to mainly occur through the mechanism that physical dissolution precedes chemical dissolution.     

Finite element analyses of a BWR vessel and penetration under severe accident conditions

This study assesses the two-dimensional thermal response of a BWR vessel and drain line penetration to three types of debris bed; primarily metallic, primarily ceramic and metallic and ceramic layered, with sensitivity studies for the most severe case. Structural finite element analysis evaluates vessel elastic, plastic and creep response for two cases which bound the thermal challenge to the vessel.Thermal analysis results indicate that drain line failure does not occur for the case when metallic debris relocates to the lower head; structural analysis predicts that the vessel also remains intact for this case. In cases where ceramic debris relocates to the lower head, drain line temperatures peak near values where failure may occur within several minutes; whereas vessel failure is not predicted for 3.5 to 4.0 hours. Sensitivity studies indicate that large porosity debris or high heat removal rates from the vessel and drain line outer surfaces can preclude failure temperatures from occurring.     

Bottom nozzle failure mechanism of water reactor pressure vessel under severe accident conditions

Most of past studies devoted to the creep rupture of a nuclear reactor pressure vessel (RPV) lower head under severe accident conditions, have focused on global deformation and rupture modes. Limited efforts were made on local failure modes associated with penetration nozzles as a part of TMI-2 vessel investigation project (TMI-2 VIP) in 1990s. However, it was based on an excessively simplified shear deformation model. In the present study, the mode of nozzle failure has been investigated using data and nozzle materials from Sandia National Laboratory's lower head failure experiment (SNL-LHF). Crack-like separations were revealed at the nozzle weld metal to RPV interfaces indicating the importance of normal stress component rather than the shear stress in the creep rupture. Creep rupture tests were conducted for nozzle and weld metal materials, respectively, at various temperature and stress levels. Stress distribution in the nozzle region is calculated using elastic–viscoplastic finite element analysis (FEA) using the measured properties. Calculation results are compared with earlier results based on the pure shear model of TMI-2 VIP. It is concluded from both LHF-4 nozzle examination and FEA that normal stress at the nozzle/lower head interface is the dominant driving force for the local failure. From the FEA for the nozzle weld attached in RPV, it is shown that nozzle welds failure occur by displacement controlled fracture of nozzle hole not by load controlled fracture of internal pressure. Considering these characteristics of nozzle weld failure, new concept of nozzle failure time prediction is proposed.     

SMART behavior under over-pressurizing accident conditions

SMART (System-integrated Modular Advanced ReacTor) is an integral reactor of 330 MW capacity with passive safety features under development in Korea. The design is developed by combining the firmly-established commercial reactor technologies with new and advanced technologies such as industry proven KOFA (Korea Optimized Fuel Assembly) based nuclear fuels, self-pressurizing pressurizer, helically coiled once-through steam generators, and new control concepts. The design of SMART focuses on enhancing the safety and reliability of the reactor by employing inherent safety features such as low core power density, elimination of large break loss of coolant accident, etc. In addition, in order to prevent the progression of emergency situations into accidents, the SMART is provided with a number of engineered safety features such as Passive Residual Heat Removal System, Passive Emergency Core Cooling System, Safeguard Vessel, and Passive Containment Over-Pressure Protection System. This paper presents an overview of the SMART design, characteristics of it’s safety systems, and results of over-pressure accident analyses. The results of the accident analyses show that the SMART provides the inherent over-pressure protection capability for design basis accidents without actuation of any protection devices such as safety valves, rupture disks, etc.     

Failure strains and proposed limit strains for an reactor pressure vessel under severe accident conditions

The local failure strains of essential design elements of a reactor vessel are investigated. The size influence of the structure is of special interest. Typical severe accident conditions including elevated temperatures and dynamic loads are considered.The main part of work consists of test families with specimens under uniaxial and biaxial load. Within one test family the specimen geometry and the load conditions are similar, but the size is varied up to reactor dimensions. Special attention is given to geometries with a hole or a notch causing non-uniform stress and strain distributions typical for the reactor vessel. A key problem is to determine the local failure strain. Here suitable methods had to be developed including the so-called “vanishing gap method”, and the “forging die method”. They are based on post-test geometrical measurements of the fracture surfaces and reconstructions of the related strain fields using finite element models.The results indicate that stresses versus dimensionless deformations are approximately size independent up to failure for specimens of similar geometry under similar load conditions. Local failure strains could be determined. The values are rather high and size dependent. Statistical evaluation allow the proposal of limit strains which are also size dependent. If these limit strains are not exceeded, the structures will not fracture.     

Stepwise integral scaling method for severe accident analysis and its application to corium dispersion in direct containment heating

Accident sequences which lead to severe core damage and to possible radioactive fission products into the environment have a very low probability. However, the interest in this area increased significantly due to the occurrence of the small break loss-of-coolant accident at TM1–2 which led to partial core damage, and of the Chernobyl accident in the former USSR which led to extensive core disassembly and significant release of fission products over several countries. In particular, the latter accident raised the international concern over the potential consequences of severe accidents in nuclear reactor systems. One of the significant shortcomings in the analyses of severe accidents is the lack of well-established and reliable scaling criteria for various multiphase flow phenomena. However, the scaling criteria are essential to the severe accident, because the full scale tests are basically impossible to perform. They are required for (1) designing scaled down or simulation experiments, (2) evaluating data and extrapolating the data to prototypic conditions, and (3) developing correctly scaled physical models and correlations. In view of this, a new scaling method is developed for the analysis of severe accidents. Its approach is quite different from the conventional methods. In order to demonstrate its applicability, this new stepwise integral scaling method has been applied to the analysis of the corium dispersion problem in the direct containment heating.