Application of containment codes to LMFBRs in the united states

This paper describes the application of containment codes to predict the response of the fast reactor containment and the primary piping loops to HCDAs. Five sample problems are given to illustrate their applications. The first problem deals with the response of the primary containment to an HCDA. The second problem deals with the coolant flow in the reactor lower plenum. The third problem concerns sodium spillage and slug impact. The fourth problem deals with the response of a piping loop. The fifth problem analyzes the response of a reactor head closure. Application of codes in parametric studies and comparison of code predictions with experiments are also discussed.     

Analysis of LMFBR primary system response to an HCDA using a eulerian computer code

This paper describes the use of the Eulerian code to predict the response of the fast reactor containment and the primary piping loops to HCDAs, and to analyze the sodium-spillage and the bubble motion problems which cannot be analyzed with a Lagrangian code. The basic equations and numerical techniques used in the Eulerian computer code are described in detail. Four sample problems are given. The first problems are given. The first problem deals with the response of the primary containment to an HCDA. The second problem deals with the slug impact on the reactor head and the spillage of sodium through the opening holes. The third problem deals with the dynamics of an HCDA bubble. The fourth problem deals with the response of a piping loop.     

Prevention and mitigation measures to ensure containment integrity

The consequences of severe reactor accidents depend greatly on containment safety features and containment performance in retaining radioactive material. In most severe accident sequences, the ability of the containment boundary to maintain its integrity is determined by two factors: (1) the magnitude of the loads; and (2) the response to those loads of the containment structure and the penetrations through the containment boundary. Severe accident phenomenology and consequences has been the subject of worldwide intense research for more than 20 years. In this paper, the main threats to the containment integrity are reviewed, the state of knowledge and remaining uncertainties are briefly described, and the links with the relevant Community research activities are made. Finally, areas in which further research may be needed are listed.     

Structural response of fast breeder reactors to hypothetical core disruptive accidents

Liquid metal-cooled fast breeder reactors (LMFBRs) so far have been analyzed for the consequences on the plant and the environment for hypothetical core disruptive accidents (HCDAs). To provide the appropriate analytical tools for this effort, analysis and codes are currently under development in several countries. They combine the hydrodynamics and solid mechanics (and more recently the bubble dynamics) phenomena to gage stresses, strains, and deformations of the important components of the system, and the overall adequacy of the primary and secondary containments. The effort is partitioned into the structural analysis of (a) the core components, and (b) the primary system components beyond the core.The core mechanics effort covers the structural response of fuel pins, hexcans, fuel elements, and fuel element clusters to transient pressures and thermal loads. Two- and three-dimensional finite element codes are under development for these core components. The results of these analyses would permit evaluation of the adequacy of the heat removal process to continue following severe core component deformations. Also, these analyses are currently being combined with neutronics, for the core transition phase, to allow for the mass movements for realistic neutronic calculations.The primary system and containment program treats the structural response of the components beyond the core, starting with the core barrel. Combined hydrodynamics-solid mechanics codes provide transient stresses and strains and final deformations for components such as the reactor vessel, reactor cover, cover holddown bolts, as well as the pulses for which the primary piping system is to be analyzed. Both, Lagrangian and Eulerian two-dimensional codes are under development, which in their combined form provide greater accuracy and longer durations for the treatment of HCDAs. More recently the codes are being augmented with bubble migration capability pertaining to the latter stages of the HCDA, after slug impact. This step will permit treatment of the instabilities following slug impact, the ultimate reversal of the sodium slug with the rising bubble, the bubble break-up, and the calculations of sodium splillage and radioactive gases, if any, in the secondary containment. The extent of sodium spillage and sodium fires should be known for evaluation of the secondary containment. The mechanics of bubble migration are needed for radiological study of post-accident phenomena. More recently dynamic fracture mechanics considerations are being incorporated to remove arbitrary failure criteria imposed on components such as the core barrel and vessel.Most recent developments involve the adaptation of the 2-D Eulerian primary system code to the 2-D elastic-plastic treatment of the primary piping. The pulses are provided at the vessel primary piping interfaces of the inlet and outlet nozzles. The calculation includes the elbows and pressure drops along the components of the primary piping system. Pressures larger than the ones used as input at the inlet and outlet nozzles were observed. As expected, they occur far from the nozzles, in the pipe, where the pulses meet.Recent improvements to the primary containment codes include introduction of bending strength in materials, Lagrangian mesh regularization techniques, and treatment of energy absorbing materials for the slug impact. A further development involves the combination of a 2-D finite element code for the reactor cover with the 2-D finite-difference hydrodynamic code for continuous monitoring of stresses, strains, and deformations in the cover, as well as pressure changes in the hydrodynamic code. Substantial experimental effort is in progress in various countries on the response to energy releases of vessels and internals, piping systems, subassemblies, and subassembly clusters. These experimental results are being utilized for the verification or modification of the analyses and codes under development.     

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.     

ANL survey of LWR containment penetrations: A progress report

Argonne National Laboratory is currently working on specific tasks in a containment penetration integrity program funded by NRC and managed by the Sandia National Laboratories. The first of these tasks is called “Characterization of Existing Penetration Designs”. The objective of this task is to identify those penetrations in nuclear reactor containments which, because of historical data or expected behavior under accident loads, are believed to have a relatively high probability of developing leakage when subjected to temperatures and pressures well beyond the containment design basis values. The program focuses on large and operating penetrations — such as personnel airlocks, equipment hatches, and bellows seals — and excludes electrical penetration assemblies and valve penetrations. (Sandia is working on electrical penetrations and EG&G is studying valve penetration assemblies.) This task will determine which penetrations require detailed study to determine leakage characteristics, and will identify which types of penetrations may require specific model and/or large-scale testing to obtain such characteristics. The survey is concentrating on containments built primarily between 1970 and 1982, and includes a comprehensive sample involving not only all types of containment types and materials, but also includes work performed by a large number of A-E design firms. The survey includes a good sample of containment penetration fabrication vendors. About 40 containments have been completed mid-August 1984.     

Reliability-based assessment of integrity of bonded prestressed concrete containment structures

The main function of a nuclear containment structure is to prevent the leakage of radioactive materials from the reactor in the event of a serious failure in the process system. To maintain a high level of leak integrity, prestressed concrete is widely utilized in containment construction. In bonded prestressing systems, excessive prestressing losses caused by unexpected material deformations and degradation of tendons could result in the loss of leak integrity under an accident. To safeguard against this, the Canadian Standard, CSA N287.7 (1995), recommends periodic inspection and evaluation of prestressing systems of CANDU containments. As bonded tendons are not amenable to direct inspection, the evaluation is based on the testing of a set of beams with features identical to the containment. The paper presents a quantitative reliability-based approach to evaluate the containment integrity in terms of the condition of bonded prestressing systems. The proposed approach utilizes the results of lift-off, destructive, and flexural tests to update the probability distribution of prestressing force, and to revise the calculated reliability against through-wall cracking of containment elements. An acceptable criterion for the results of beam tests is established on the basis of maintaining adequate reliability throughout the service life of the containment.     

严重事故条件下安全壳响应模拟研究

压水堆核电厂发生严重事故期间,从主系统释放的蒸汽、氢气以及下封头失效后进入安全壳的堆芯熔融物均对安全壳的完整性构成威胁。以国内典型二代加压水堆为研究对象,采用MAAP程序进行安全壳响应分析。选取了两种典型的严重事故序列：热管段中破口叠加设备冷却水失效和再循环高压安注失效,堆芯因冷却不足升温熔化导致压力容器失效,熔融物与混凝土发生反应（MCCI）,安全壳超压失效;冷管段大破口叠加再循环失效,安全壳内蒸汽不断聚集,发生超压失效。通过对两种事故工况的分析,证实了再循环高压安注、安全壳喷淋这两种缓解措施对保证安全壳完整性的重要作用。     

Evaluation of plutonium oxide aerosol release from an LMFBR in a hypothetical accident

This work was undertaken to prepare a computer code for the hazard evaluation of plutonium oxide aerosol released to the atmosphere in the event of a hypothetical accident in a 50 MW(th) scale LMFBR. The reactor building structure consists of semi-double containments as follows: the primary containment has a large volume in comparison with the secondary annular containment in which a part is connected to the atmosphere through an emergency filter system. Sodium oxide aerosol containing PuO₂---UO₂ fuel, fission products and structural steel agglomerates quickly by coagulation due to its high concentration. Simultaneously, the aerosol concentration decreases due to settling, plating and thermophoresis. Using the present code, the amount of PuO₂ aerosol leakage to the atmosphere was evaluated.     

Formation, characterisation and cooling of debris: Scenario discussion with emphasis on TMI-2

Regarding safety improvements for existing nuclear power plants, the TMI-2 accident is interesting because of the present commercial dominance of light water reactors (LWR). This accident demonstrated that the nuclear safety philosophy evolved over the years has to cover accident sequences involving massive core melt progression in order to develop reliable mitigation strategies for both, existing and advanced reactors. Although the TMI-2 core was reflooded, the results also appear applicable to the general melt progression phenomenology of most unrecovered (unreflooded) blocked core accident scenarios. Nevertheless, a large range in the initial conditions of core melt progression provides significant uncertainties in assessing the integrity of the lower head, the containment in severe reactor accidents, and the consequences of recovery actions in accident management, as well as core reflooding in particular. The probability of success of reflooding as an accident management strategy – in-vessel reflooding to terminate the accident and ex-vessel flooding to prevent reactor vessel melt-through – has to be assessed and discussed in detail.     

Testing of large-scale concrete containment structural elements

The tests described in this paper are part of an Electric Power Research Institute (EPRI) program (Research Project 2172-2) to provide a test-verified analytical method of estimating capacities of concrete reactor containment buildings under internal overpressurization from postulated degraded core accidents.Experimental study in Phase 2 of the investigation, on which this paper is based, includes tests of five large-scale specimens with steel liner plates representing structural elements of prestressed concrete containment buildings. Four square wall element specimens and one specimen representing the wall/basemat junction region were tested.This experimental work indicates that under internal overpressurization or other accident conditions, highly localized strains in the steel liner plate can result in liner tearing and subsequent containment leakage. These results support the theory of leak before break where liner tearing occurs in a controlled manner and leakage and depressurization occur rather than global failure.     

Inherent Prevention and Mitigation of Severe Accident Consequences in Sodium-Cooled Fast Reactors

Safety challenges for sodium-cooled fast reactors include maintaining core temperatures within design limits and assuring the geometry and integrity of the reactor core. Due to the high power density in the reactor core, heat removal requirements encourage the use of high-heat-transfer coolants such as liquid sodium. The variation of power across the core requires ducted assemblies to control fuel and coolant temperatures, which are also used to constrain core geometry. In a fast reactor, the fuel is not in the most neutronically reactive configuration during normal operation. Accidents leading to fuel melting, fuel pin failure, and fuel relocation can result in positive reactivity, increasing power, and possibly resulting in severe accident consequences including recriticalities that could threaten reactor and containment integrity. Inherent safety concepts, including favorable reactivity feedback, natural circulation cooling, and design choices resulting in favorable dispersive characteristics for failed fuel, can be used to increase the level of safety to the point where it is highly unlikely, or perhaps even not credible, for such severe accident consequences to occur.     

SNR 300 fast breeder reactor: steel containment - design, erection, testing

The Kalkar Nuclear Power Plant which is equipped with an 300 MW fast breeder reactor is being built by a Consortium mainly comprising German, Belgian and Dutch companies.The components of the fast breeder reactor are enclosed in a concrete containment which is designed to withstand severe external and internal loading.The concrete enclosure is surrounded by a steel containment which is designed to prevent the release of radioactivity following a postulated accident involving the nuclear components inside the concrete containment.The paper describes the solutions adopted for the different parts of the steel containment, the calculations verifying the suitability of the designs, the erection and the steel containment pressure and leak tests. The tests were performed with successful results in 1984.     

Natural Convection Sodium Boiling Experiments in 37-Pin Bundle Geometry

Decay heat removal capability under boiling condition was studied using an LMFBR fuel subassembly mockup loop. The sodium flow was driven by natural convection through the loop in which was installed a 37-pin bundle heated electrically over a length of 45 cm.

The heat flux furnished by the pins was increased stepwise, upon which the two-phase flow regime changed from bubble to slug flow and then to annular or annular mist flow. Dryout occurred even in slug flow regime, but only momentarily, and permanent dryout was not observed before establichment of annular flow. A suitable criterion for permanent dryout is considered to be 0.5 average exit sodium vapor quality. The results indicated that upon occurrence of sodium boiling, the coolability of fuel subassembly would be maintained by natural convection after reactor shutdown.     

Thermal-hydraulic and stress analysis of AP1000 reactor containment during LOCA in dry cooling mode

Some kinds of break in the reactor coolant system may cause the coolant to exit rapidly from the failure site,which leads to the loss of coolant accident (LOCA).In this paper,a stress analysis of an AP1000 reactor containment is performed in an LOCA,with the passive containment cooling system (PCCS) being available and not available for cooling the wall's containment.The variations in the mechanical properties of the wall's containment,including elastic modulus,strength,and stress,are analyzed using the ABAQUS code.A general two-phase model is applied for modeling thermal-hydraulic behavior inside the containment.Obtained pressure and temperature from thermal-hydraulic models are considered as boundary conditions of the ABAQUS code to obtain distributions of temperature and stress across steel shell of the containment in the accident.The results indicate that if the PCCS fails,the peak pressure inside the containment exceeds the design value.However,the stress would still be lower than the yield stress value,and no risk would threaten the integrity of the containment.     

CONTAIN-LMR程序中池式钠火事故分析计算模型的验证

CONTAIN-LMR是针对以液态钠为冷却剂的反应堆而开发的安全壳事故一体化分析程序。我国目前的CONTAIN-LMR程序版本为2000年左右从法国引进,还未进行过面向工程设计的系统性地程序开发和验证。本文主要针对CONTAIN-LMR程序中模拟池式钠火事故的分析模型进行详细分析,并采用国际上的池式钠火实验进行验证,实验验证结果表明CONTAIN-LMR程序可以较准确地模拟池式钠火事故造成的钠工艺间内的温度、压力升高及放射性钠气溶胶行为。本文的研究结果初步表明CONTAIN-LMR程序可用于钠冷快堆的钠火事故分析。     

Construction of a concrete containment model

This paper discusses the features and construction of a reinforced-concrete containment model that has been built at Sandia National Laboratories in Albuquerque, New Mexico. The model Light-Water-Reactor (LWR) containment building was designed and built to the American Society of Mechanical Engineers (ASME) code by United Engineers and Constructors, Inc. The containment model will be tested to failure to determine its response to static internal overpressurization. The results from testing the heavily instrumented containment will be used to assess the capability of analytical methods for predicting the performance of containments subject to severe accident loads as part of the US Nuclear Regulatory Commission's program on containment integrity.The scaled dimensions of the cylindrical wall and hemispherical dome are typical of a full-size containment. Features representative of a prototypical containment and included in the heavily reinforced model are equipment hatches, personnel airlocks, several small piping penetrations, and a thin steel liner attached to the concrete by headed studs.     

The passive ALWR approach to assuring containment integrity

Design requirements for the advanced light water reactor (ALWR) have been developed so as to provide high assurance of containment integrity even in the event of a severe accident. The containment integrity requirements are in the form of two design criteria, and associated methodology, which address containment severe accident performance and offsite dose and are specified in the ALWR utility requirements document (URD), a set of detailed design requirements for next generation plants in the US.The containment performance criterion, which is the main focus of this paper, specifies that plant design characteristics and features shall be provided to preclude core damage sequences which could bypass containment and to withstand core damage sequence loads. This containment performance capability, along with the associated dose mitigation capability, provides a technical basis for emergency planning change since there would not be the same need for rapid offsite emergency response that is called for under the existing US emergency planning basis.     

Accomplishments and challenges of the severe accident research

This paper briefly describes the progress of the severe accident research since 1980, in terms of the accomplishments made so far and the challenges that remain. Much has been accomplished: many important safety issues have been resolved and consensus is near on some others. However, some of the previously identified safety issues remain as challenges, while some new ones have arisen due to the shift in focus from containment to vessel integrity. New reactor designs have also created some new challenges. In general, the regulatory demands for new reactor designs are stricter, thereby requiring much greater attention to the safety issues concerned with the containment design of the new large reactors, and to the accident management procedures for mitigating the consequences of a severe accident. We apologize for not providing references to many fine investigations that contributed to the great progress made so far in the severe accident research.