秦山核电厂安全壳系统B、C类密封性试验

叙述了秦山核电厂安全壳系统B、C类密封性能试验概况,主要包括试验范围、泄漏率分配、试验结果和总体评价等。     

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.     

Pneumatic pressure tests of steel containment models — Recent developments

This paper describes the results of recent pneumatic pressure tests of steel containment models. These tests are part of the Containment Integrity Program whose objective is the qualification of methods for predicting containment response during severe accidents and extreme environments. Sandia National Laboratories is conducting this combined experimental and analytical program for the U.S. Nuclear Regulatory Commission (NRC). The long-range plans for the program include the following three containment loading conditions: static internal pressurization, dynamic internal pressurization, and seismic loadings. Steel, reinforced concrete, and prestressed concrete containment types are being considered.In the present experimental effort, models of steel containment structures are being subjected to static internal pressurization. The first set of models are about the size of hybrid-steel containments. Tests of these models are nearly finished. Testing of a large steel model, about of full size, will complete the static pressure experiments with steel models. Analysis of the models is paralleling the experimental effort.The Containment Integrity Program is being coordinated with other NRC programs on potential leakage of penetrations in containments. The results from all of the programs should provide a basis for predicting the structural and leakage behavior of containments during temperature and internal pressure loadings.     

Tension tests of concrete containment wall elements

Tension tests of concrete containment wall elements were conducted as part of a three-phase research program sponsored by the Electric Power Research Institute (EPRI). The objective of the EPRI experimental/analytical program is twofold. The first objective is to provide the utility industry with a test-verified analytical method for making realistic estimates of actual capacities of reinforced and prestressed concrete containments under internal over-pressurization from postulated degraded core accidents. The second objective is to determine qualitative and quantitative leak rate characteristics of typical containment cross-sections with and without penetrations. This paper covers the experimental portion the the EPRI program.The testing program for Phase 1 included eight large-scale specimens representing elements from the wall of a containment. Each specimen was 60-in (1525-mm) square, 24-in (610-mm) thick, and had full-size reinforcing bars. Six specimens were representative of prototypical reinforced concrete containment designs. The remaining two specimens represented prototypical prestressed containment designs.Various reinforcement configurations and loading arrangements resulted in data that permit comparisons of the effects of controlled variables on cracking and subsequent concrete/reinforcement/liner interaction in containment elements.Subtle differences, due to variations in reinforcement patterns and load applications among the eight specimens, are being used to benchmark the codes being developed in the analytical portion of the EPRI program.Phases 2 and 3 of the test program will examine leak rate characteristics and failure mechanisms at penetrations and structural discontinuities.     

On-power containment integrity monitoring in CANDU multi-unit stations

Based on a general review of current on-power test methods and experience in CANDU multi-unit containments, it is concluded that such tests make a significant contribution to plant safety. In particular continuous monitoring at low pressure differentials merits further development and more widespread application.Current on-power tests include individual component testing, quarterly reduced pressure tests (typically at −15 kPa (g)), and continuous pressure trend monitoring at normal operational pressure of −3 kPa(g). A continuous monitoring concept is outlined which consists of a periodically updated mass balance. Instrument error uncertainty for this technique was estimated to be on the order of a 1 cm hole. However systematic fluctuations (often attributable to physical causes) dominate the error analysis in on-power tests. With precautions on sampling interval, a moving regression may be used to generate a leakage rate time series such that the fluctuations can be bounded or eliminated.Experience to-date has indicated that most containment boundary impairments are detectable by component tests or continuous monitoring. On-power tests methods are capable of addressing a significant portion of the containment failure mode spectrum. Station risk assessment and regulatory testing requirements are identified as means by which these methods can be credited in demonstrating containment integrity.     

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.     

核电厂安全壳首次整体在役试验方案优化

安全壳整体试验是压水堆核电机组一项特大型、高风险、高难度的试验,通过模拟设计基准事故工况下安全壳内的峰值压力,在事故峰值压力平台下,进行安全壳整体泄漏率测量及各压力平台安全壳结构试验,以验证其密封和结构性能。安全壳整体试验是国家核安全局监管的一个重要见证点,试验结果直接决定是否能够启动反应堆发电。301大修安全壳整体试验是3号机组首次在役试验,本次试验汲取了秦山第二核电厂以往6次安全壳整体试验的经验和其他电厂的反馈,试验方案更加科学,试验的组织管理更为规范。文章对301大修安全壳整体试验的经验进行了论述和总结,希望对电厂以后的安全壳整体试验提供参考。     

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.     

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.     

Allowable stresses to limit shearing deformation in reinforced concrete containment design

An extensive program of the U.S. Nuclear Regulatory Commission (NRC) to study reinforced concrete containment wall behavior has been completed for orthogonal reinforcement. The transfer of shear caused by the action of seismic load has been studied sufficiently to recommend the seismic shear design and allowable shear stresses. However, the recommendations made in this paper are not the NRC position for the design.     

Performance of containment penetrations under severe accident loadings

The paper provides a summary of efforts to date to better understand the leakage behavior of containment penetrations when subjected to severe accident conditions. The research activities discussed herein are a part of the Containment Integrity Programs, which are managed by Sandia National Laboratories for the U.S. Nuclear Regulatory Commission. Past containment penetration research topics, which are briefly described, include testing of typical compression seals and gaskets, electrical penetration assemblies, and a personnel airlock, as well as an investigation of leakage due to ovalization of penetration sleeves. The primary focus of the paper is on recent or ongoing research programs on the behavior of inflatable seals, bellows, and of pressure unseating equipment hatches.     

Synopsis of the results from overpressurizing a reinforced concrete containment model

Sandia National Laboratories completed the testing of a 1:6-scale containment building for a light water reactor in July 1987. Results from this and other containment model testing are being used by the US Nuclear Regulatory Commission to benchmark analytical techniques. The validated techniques can then be used to predict the behavior of actual nuclear power plant containments to a variety of hypothesized severe accidents.The most recent containment building tested was made of reinforced concrete and had many of the features found in full-size containments. Testing consistent of a structural integrity test, and integrated leak rate test, and concluded with an overpressurization test of the structure. Highlights of the results from the overpressurization of the containment model are presented.     

Accident progression within the secondary containment boundary of large LMFBRs

This study deals with the sodium spillage phenomenon as it relates to accident energetics and containment integrity. Sodium spillage has been identified as an important issue for large LMFBRs because of the large inventory of sodium present and the potential for energetic accidents. Energetic core-disruptive events leading to slug impact could open leak paths in the reactor cover and vent sodium into the secondary containment. Sodium fires in the containment building could lead to pressurization and thermal stressing of the surrounding structure and jeopardize containment integrity. The potential consequences of such a scenario have prompted the development of analytical tools to quantify the spillage process.One of the primary concerns in assessing the integrity of secondary containment is the amount and velocity of sodium which may be ejected from the primary vessel. A parametric study has been performed, the purpose of which was to study the sensitivity of sodium spillage to accident energetics. Treatment of the spillage process was accomplished with the ICECO code employing a quasi-Eulerian method. A 1000 MWe reactor, with prescribed leak paths, is modelled and analyzed during the slug impact phase. Leak paths are assumed to exist as annular penetrations in the reactor cover and as a gap at the vessel-head junction. The behavior of sodium spillage is investigated under conditions of different accident energetics, various opening sizes, and multiple leak paths, with both stationary and moving reactor covers. The relative influence of short and long term spillage is also addressed.During the transient period immediately following slug impact it was found that spillage from annular penetrations in the reactor cover is only weakly sensitive to changes in slug velocity. The same conclusion applies to spillage from a fixed gap at the vessel-head junction. Significant sensitivity of spillage to accident energetics was seen only in cases of spillage from the vessel-head junction when the reactor cover was movable. The influence of slug impact on the motion of the reactor cover leads to the conclusion that sodium spillage is most sensitive to accident energetics inasmuch as it affects the size of the leak path.     

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.     

Experimental Testing on a Simulated Flask Lid Seal System to Confirm Containment Measuring Procedures

Abstract

British Nuclear Fuels Ltd (BNFL) operates a large number of package types for the transport of irradiated nuclear fuels from customers' utilities to the Sellafield reprocessing facility in Cumbria. All fuel loading operations are carried out under water and consequently package lid sealing systems are saturated. All BNFL package types employ a double seal system on the lid which must be tested before despatch, to confirm containment integrity. The normal test procedure involves drying the interspace between the seals with compressed air before applying a gas pressure and measuring the pressure drop; the reliability of this procedues depends upon the seals being dry. In order to demonstrate the reliability of BNFL's containment testing methods, and to develop operational procedures that ensure acceptable dryness is achieved, an experimental test rig was designed and manufactured. Closely based on a typical package lid seal arrangemens, the test rig allowed leakage paths to be introduced by fine wires fitted across the seal faces. BNFL conducted a series of tests to investigate how the measured leak rate was influenced by the presence of water. Existing drying procedures were evaluated, and shown not to be fully effective in removing all moisture. New drying procedures were subsequently developed, which are totally efficient in drying the test inter-space and ensure that accurate containment measurements can be undertaken.     

秦山核电厂安全壳密封性能试验的仪表测量系统

简要介绍秦山核电厂反应堆安全壳密封性能试验所采用的仪表测量系统及其性能。     

Comparison of analytical predictions and experimental results for a 1:8-scale steel containment model pressurized to failure

The analyses used to predict the behavior of a 1:8-scale model of a steel LWR containment building to static overpressurization are described and results are presented. Finite strain, large displacement, and nonlinear material properties were accounted for using finite element methods. Three-dimensional models were needed to analyze the penetrations, which included operable equipment hatches, personnel lock representations, and a constrained pipe. It was concluded that the scale model would fail due to leakage caused by large deformations of the equipment hatch sleeves.