Long term performance of metal seals for transport and storage casks

Dual purpose casks for the transportation and storage of spent nuclear fuel and other radioactive materials require very high leak tightness of lid closure systems under accident conditions as well as in the long term to prevent activity release. For that purpose metal seals of specific types with an inner helical spring and outer metal liners are widely used and have shown their excellent performance if certain quality assurance requirements for fabrication and assembling are satisfied. Well defined surface roughness, clean and dry inert conditions are therefore essential. No seal failure in a loaded cask happened under these conditions until today. Nevertheless, the considered and licensed operation period is limited and all safety assessments have been performed and approved for this period of time which is 40 years in Germany so far. However, in the meantime longer storage periods might be necessary for the future and therefore additional material data will be required. BAM is involved in the qualification and evaluation procedures of those seals from the early beginning. Because long term tests are always time consuming BAM has early decided to perform additional tests with specific test seal configurations to gain a better understanding of the long term behaviour with regard to seal pressure force, leakage rate and useable resilience which is safety relevant mainly in case of accidental mechanical loads inside a storage facility or during a subsequent transport. Main test parameters are the material of the outer seal jacket (silver or aluminium) and the temperature. This paper presents the BAM test program including an innovative test mock-up and most recent test results. Based on these data extrapolation models to extended time periods are discussed, and also future plans to continue tests and to investigate seal behaviour for additional test parameters are explained.     

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.     

Severe accident analysis for a typical PWR using the MELCOR code

Using the MELCOR code, we simulated and analyzed a severe accident at a Chinese pressurized reactor 1000-MW (CPR1000) power plant caused by station blackout (SBO) with failure of the steam generator (SG) safety relief valve (SRV). The CPR1000 response and results for three different scenarios were analyzed: (i) seal leakage and an auxiliary feed water (AFW) supply; (ii) no seal leakage or AFW supply; and (iii) seal leakage but no AFW supply. The results for the three scenarios are compared with those for a simple SBO accident. According to our calculations, the SG SRV stuck in the open position would greatly accelerate the sequence for a severe accident. For an SBO accident with the SRV stuck open without seal leakage or an AFW supply, the pressure vessel would fail at 9576 s and the containment system would fail at 124,000 s. If AFW is supplied, pressure vessel failure would be delayed nearly 30000 s and containment failure would delay at least 50000 s. When seal leakage exists, pressure vessel failure is delayed about 50 s and containment failure time would delay about 30000 s. The results will be useful in gaining an insight into the detailed processes involved and establishing management guidelines for a CPR1000 severe accident.     

Natural circulation flow behavior at reduced inventory in a VVER geometry

This paper deals with the natural circulation flow characteristics of the VVER-440 geometry at reduced coolant inventory. Special emphasis is on the flow rate of the primary circuits during the two-phase flow regime. For studying two-phase natural circulation flow phenomena in a VVER geometry a series of cold leg small break loss-of-coolant accident (SBLOCA) tests was carried out in the PArallel Channel TEst Loop (PACTEL), a 1/305 volumetrically scaled model of a VVER-440 reactor. The tests were conducted with break areas ranging from 0.1 to 1.5 % of the scaled cold leg cross-sectional area of the reference reactor. A partial failure of the high-pressure injection system (HPIS) was assumed. The tests reveal a trend towards an increasing primary circuit mass flow rate with decreasing inventory. This contradicts the findings of earlier tests in multi-loop VVER geometry. With single-loop facilities, increased mass flow rates at reduced inventories have been reported before. The increase of the two-phase flow rate turns out to be a consequence of the combined effect of break size, pressure range and secondary side feed and bleed procedure. The physical phenomena of flow stagnation in the primary circuits, system pressurization, asymmetric loop flows, and loop seal clearing and refilling take place during the natural circulation cooling process from single-phase into two-phase and boiler–condenser modes. In addition, flow reversal in the undermost tubes of the horizontal steam generators (SG) is observed. These phenomena are discussed briefly while a general insight into the course of the tests is presented.     

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。     

反应堆冷却剂泵动压机械密封的工程开发与应用

反应堆冷却剂泵（以下简称主泵）轴密封由3级相同的动压机械密封串联组成，是主泵的心脏，其泄漏量直接决定主泵能否正常运行。本文提出了一种新型的挤压变形研磨法完成动压机械密封的制造，应用挤压变形工装和金属垫片使静环产生变形，在密封端面研磨出9个波形槽。功能实验表明，新型的机械密封在考核压力下的低压泄漏量满足主泵轴密封的设计要求；压力突变工况下的冲击考核实验表明，新型的动压机械密封摩擦副之间的液膜刚度未发生破坏，未出现密封失效。本文研发的动压机械密封在核电厂的运行状况与实验结果完全吻合，充分证明了该新型动压机械密封具有极高的工程应用可靠性。      

UPTF experiment: Flow phenomena during full-scale loop seal clearing of a PWR

To investigate the flow phenomena in the primary system of a pressurized water reactor (PWR) during a loss-of-coolant accident (LOCA) occurring with a small or intermediate break, experiments were performed at the full-scale Upper Plenum Test Facility (UPTF). Within the Transient and Accident Management (TRAM) program integral and separate effect tests were carried out to study loop seal clearing and to provide data for the further improvement of computer codes concerning the reactor safety analysis. This paper describes the UPTF tests that focus on the sequence of loop seal clearance in a four-loop operation for two different cold leg break sizes and the residual water levels, the flow patterns in, and the pressure drops across a single loop seal during the clearing. The UPTF results obtained from a single-loop seal operation are compared with experimental data and correlations available in the literature. Two correlations are proposed which allow the quantification of residual water levels in the loop seal under PWR conditions. It is shown that the steam–water test results gained from the full-scale UPTF with realistic PWR loop seal geometry differ from those obtained from the full or small-scale test facilities under air–water conditions. The UPTF experiments indicate the substantial need for steam-water test data from a full-scale facility with realistic PWR geometries in order to validate PWR LOCA thermal-hydraulic system codes to predict loop seal clearing correctly.     

First tests results for determination of seal life of EPDM O-rings at high temperature (determined by unique method)

Abstract

A recurrent concern in the design of packaging for the transportation of radioactive material is to determine the life of elastomeric O-rings at a high temperature. Following a precedent paper presenting the TN International method for the determination of ethylene propylene diene (EPDM) O-rings seal life, a first series of tests results have been obtained. These first results are used with the previously mentioned method to see that the EPDM seal life did follow an Arrhenius curve. This Arrhenian behaviour is verified below a certain 'threshold temperature' that seems to be the upper limit for the use of such elastomeric O-ring grades. With the 'seal life versus temperature' curve obtained from these tests results, the time–temperature profile of O-rings in casks can then be benchmarked with the seal resistance and its 'maximum permissible damage rate'.     

堆芯热电偶密封结构失效机理分析

核电厂堆芯热电偶密封结构属于主系统的压力边界,其可靠性对于核电厂主系统的严密性和安全性具有重要意义.文章以秦山核电站的堆芯热电偶密封结构为例,对目前国内压水堆核电厂堆芯热电偶密封结构中应用广泛的CONOSEAL和GRAYLOC的组合密封结构进行分析,对各种密封失效事件进行分析,并提出相应的改进措施.     

Comments on the probability of leakage in piping systems as used in PRA's

In risk analysis of power reactors the leakage or failure of piping structures has to be taken into account as a possible cause of loss of coolant.As part of the SMiRT post conference seminar on “PRA of NPP for External Events” the present practice of selecting pipe failure rates as initiating or related events for PRA's has been discussed.For pipe failures as initiating events an approach has been developed in the framework of the risk study for German PWR's. As compared to NUREG-1150 some significant differences are identified.For external events the effect of seismic induced loads on pipe failure has been a subject of considerable efforts in research. Several studies have demonstrated that for moderate siting conditions the effect of seismic induced loads on pipe failure rates of large diameter high pressure piping does not lead to significant contributions to the overall risk.The main subject for future research on pipe failure mechanisms is the detailed assessment of the influence of the water chemistry conditions.     

Fusion neutron test facility requirements for interactive effects in structural and high-heat-flux components

A relevant design data base is needed for structural components in near-term and commercial fusion devices. A high-flux, high-fluence fusion neutron test facility is required for testing the failure mechanisms and lifetime-limiting features for first wall, blanket, and high-heat-flux components. We describe here the key aspects of the fusion environment which influence the response of structural and high-heat-flux components. In addition to test capabilities for fundamental radiation-effects phenomena, e.g., swelling, creep, embrittlement, and hardening, it is shown that the facility must provide an adequate range of conditions for accelerated tests to study the limitations on component lifetime due to the interaction between such fundamental phenomena. In high-heat-flux components, testing of the failure mechanisms of duplex structures is shown to require maintenance of an appropriate temperature gradient in the 14-MeV neutron field. Thermal stresses are shown to result in component failure, particularly when the degradation in the thermal conductivity and mechanical properties by irradiation are considered. Several factors are discussed for assessment of the failure modes of the first wall and blanket structures. These are displacement-damage dose and dose rate, the amount of helium gas generated, the magnitude of irradiation and thermal creep, prototypical temperature and temperature-gradient distributions, module geometry, and external mechanical constraints.     

研究堆低压电气贯穿件的密封性能

电气贯穿件的密封性能直接关系着整个反应堆的安全性能。为了使电缆贯通式的电气贯穿件能满足研究堆设计文件的密封性能要求,首先分析了贯穿件筒体和电缆密封的难点,针对遇到的难点,分别采取了法兰挤压硅橡胶板和电缆芯线灌胶密封的方法,对贯穿件的缝隙进行密封。利用氦质谱仪器,对低压电气贯穿件样机和电缆进行了密封检测实验,实验结果表明,电缆贯通式的低压电气贯穿件密封结构能满足研究堆的密封要求。     

"In Situ" Radiation Tolerance Tests of MOS RAMs

To design the 4Mbit Scientific Data Store for the Faint Object Camera of the NASA/ESA space telescope, MOS RAM radiation hardness has been tested extensively. The radiation tolerance of 36 MOS RAM types has been evaluated by "in situ" testing. For comparison, one I2L type was added. The principal aim was the evaluation of failure doses. Beyond that, the impact of technology, storage mechanism, integration density and manufacturer on failure dose and failure mechanism has been studied. The tests have been performed with different test patterns, to ensure the recognition of specific failure types. Various irradiation sources have been used : X-ray tube, Co60-?-source and 1.5 MeV electron accelerator. The irradiation procedure and the functional test equipment are described in this paper. The results are presented and discussed. The influence of technological parameters is less pronounced than expected. Highly integrated memories are therefore preferable for those applications, where weight and shielding problems are involved.     

Integrated MELPROG/TRAC analyses of a PWR station blackout

A complete, coupled, mechanistic analysis of the entire reactor coolant system during a station blackout accident (TMLB') has been completed using the MELPROG/TRAC code. The analysis includes the failure of the seal on all coolant pumps at 100 min into the accident; in all other respects the case is identical to a previous station blackout calculation. Both cases started at accident initiation and continued through boiloff of the water, failure of the control and fuel rods, oxidation of the zircaloy and the formation of U---Zr---O eutectics, failure of the vessel internal structures due to melting and loading, massive core disruption, and subsequent vessel failure. The two cases reached significantly different end conditions. The basic TMLB' resulted in a high pressure (15 MPa) vessel failure approximately 4 h after accident initiation. The addition of a 12.5-mm hole in each pump seal caused the water in the loop seal to clear and resulted in a significantly lower pressure (0.27 MPa) at vessel failure, which occurred almost 10 h after accident initiation. Therefore, high pressure melt ejection (HPME) and the potential for subsequent direct containment heating (DCH) were predicted not to occur in the TMLB' accident scenario with pump seal failure.     

Testing of Elastomer Seals using Small-Size Rigs

Abstract

This paper looks at the use of small size seal leakage test rigs to demonstrate the compliance of full size container seals against the IAEA Transport Regulation's limits for activity release for normal transport and accident conditions. The detailed requirements of the regulations are discussed and it is concluded that an appropriate test programme to meet these requirements, using only small size test rigs, can normally be set up and carried out on a relatively short time scale. It is important that any small test rigs should be designed to represent the relevant features of the seal arrangement and the overall test programme should cover all of the conditions, specified by the regulations, for the type, classification and contents of the container under consideration. The parameters of elastomer O-rings, which affect their sealing ability, are considered and those which are amenable to small scale testing or have to be modelled at full size are identified. Generally, the seals used in leakage tests have to be modelled with a full size cross-section but can have a reduced peripheral length.     

Study on creep-fatigue failure prediction methods for type 304 stainless steel

Creep-fatigue failure is one of the principal failure modes to be avoided in elevated-temperature components of liquid metal fast breeder reactor (LMFBR) plants. To prevent this failure during the plant life with sufficient confidence, accurate and reliable methods should be employed for evaluating creep-fatigue endurance. A number of creep-fatigue tests have been conduced to establish a reliable creep-fatigue design methodology applicable to LMFBR plants in the last two decades but the conditions of these tests are generally far from those expected in actual plants. For the purpose of studying the characteristics of various creep-fatigue life prediction methods in conditions closer to actual plant conditions, the authors initiated creep and creep-fatigue tests for type 304 austenitic stainless steel with a special emphasis on tests with longer durations than past tests. Interim results are summarized in this paper. Two representative life prediction methods, linear damage fraction rule and ductility exhaustion method, were then applied to these test conditions. It was found that both methods can predict the failure lives with reasonable accuracy. Some comparisons were made regarding the characteristics of these two methods.     

The 'Anti-Evidence' Approach to Tamper Detection

Abstract

Tamper-indicating seals are widely used for the transport, packaging and storage of nuclear material. Most existing seals operate under the same basic principle: once the seal is opened, information that the seal has indeed been unsealed is stored in or on the seal until such time as the seal can be inspected. This stored evidence of tampering, however, is often easy to hide or erase. A better approach, discussed theoretically in this paper, is to store information when the seal is first installed that tampering has not been detected; this anti-evidence is then erased once the seal is opened. Such anti-evidence seals may provide better security. They also have a number of potentially useful attributes, including an intrinsic check on whether the seal was actually inspected for evidence of tampering ('anti-gundecking').     

石墨垫片密封泄漏率计算方法研究

为完善核级主设备密封分析及设计方法,基于稳压器人孔密封结构建立了密封数值分析模型,对石墨垫片密封接触应力进行了分析研究;结合平行圆板流动模型和多孔介质渗流模型建立了石墨垫片密封质量泄漏率理论预测模型;基于理论预测模型计算了设计工况、试验工况和启动瞬态工况下的质量泄漏率,对主要影响参数进行了分析和讨论。研究结果表明,石墨垫片密封接触应力沿周向分布较为均匀,而石墨环沿径向的中间区域接触应力值略低于石墨环两侧;在温度和压力上升瞬态中,密封接触应力随时间呈现出下降的规律,密封质量泄漏率与接触应力呈负相关,增大密封接触应力可以降低质量泄漏率,但降低效率逐渐减小,减小粗糙度可以显著降低质量泄漏率。本文分析方法可为核级主设备密封泄漏率分析和紧密度评价提供重要参考。     

Analytical studies on local damage to reinforced concrete structures under impact loading by discrete element method

This paper proposes a new analytical approach for assessing local damage to reinforced concrete structures subjected to impact load, by applying the discrete element method (DEM). It first outlines the basis concept and analytical formulation of the DEM. Next, it discusses the results of simulation analyses of concrete material tests, uni-axial compression tests and tensile splitting tests conducted to determine appropriate analytical parameters such as material constants, failure criteria and strength increase factors depending on strain rate. Finally, the adaptability of the DEM to local damage to reinforced concrete structures impacted by rigid and deformable missiles is verified through simulation analyses of various types of impact tests. Furthermore, the various impact response characteristics and failure mechanisms, such as impact forces, penetration behavior, reduction in missile velocity and energy transfer process, which are difficult to obtain experimentally, are analytically evaluated by the DEM.     

Technical investigation on small water leakage incident occurrence in mercury target of J-PARC

Japan Proton Accelerator Research Complex experienced failures of two mercury targets, which were Target #5 and #7, in 2015 when the facility was operating with a proton beam power of 500 kW. The failures involved coolant water leak from the water shroud. In this paper, we investigate the root cause of the Target #5 failure. The results of the visual inspections, mockup tests, and analytical evaluations suggested that the water leak was caused by the possible combination of two incidents. One was the diffusion bonding failure due to the large thermal stress induced by welding of the bolt head during the fabrication process, and the other was the thermal fatigue failure of the seal weld due to the repetitive beam shutdown during beam operation. Though the investigation into the root cause of the Target #7 failure is still going on, these target failures point to the importance of eliminating initial defects and the need to secure the rigidity and stability of welded structures. The next mercury target, Target #8, was fabricated with an improved design and fabrication process to reduce the possibility of similar failures. The beam operation of this mercury target is planned to be started in October 2017.