Leak Rate Test Research Articles

Applicability of the Alternative Reagents to CFC112 for the Leak Rate Test of Iodine Removal Filters in PWR Nuclear Power Plants

Applicability of the Alternative Reagents to CFC112 for the Leak Rate Test of Iodine Removal Filters in PWR Nuclear Power Plants

Time-dependent leak behavior of flawed Alloy 600 tube specimens at constant pressure

Leak rate testing has been performed using Alloy 600 tube specimens with throughwall flaws. Some specimens have shown time-dependent leak behavior at constant pressure conditions. Fractographic characterization was performed to identify the time-dependent crack growth mechanism. The fracture surface of the specimens showed the typical features of ductile fracture, as well as the distinct crystallographic facets, typical of fatigue crack growth at low Δ K level. Structural vibration appears to have been caused by the oscillation of pressure, induced by a high-pressure pump used in a test facility, and by the water jet/tube structure interaction. Analyses of the leak behaviors and crack growth indicated that both the high-pressure pump and the water jet could significantly contribute to fatigue crack growth. To determine whether the fatigue crack growth during the leak testing can occur solely by the water jet effect, leak rate tests at constant pressure without the high-pressure pump need to be performed.

Penetration of gas into concrete during a leakage rate test of reactor containments and its significance for the drop in pressure

The objective of the project described in the paper was to develop a simulation model that describes transient air pressure distribution in concrete in order to see if the leakage rates obtained from the Containment Integrated Leakage Rate Tests can be explained by the transient air pressurization of concrete pores inside the steel liner. A partial differential equation was derived which describes transient air pressure distribution in concrete pores. The model was validated against experimental results. The simulation model shows that there are significant air fluxes into the concrete structures that can explain the pressure drop during a leakage test.

Modeling and analysis of aging behavior of concrete structures in nuclear power plants

The design rules for reactor containments and similar safety structures in nuclear power plants have been established decades ago based on their initial material and structural conditions, and design-based assumptions. However, as these plants begin to approach the end of their original design life and transition to the life-extension phase, analytical methods are needed to evaluate potential aging-related effects. In the case, of pre-stressed concrete containments, for example, the concrete is subject to long-term creep in the presence of time-varying loads due to tendon relaxation and possible re-tensioning, periodic leak-rate testing and service-and weather-dependent thermal cycling. Also, some PWR plants are facing the prospect of replacing steam generators, which require the cutting of large opening in the pre-stressed containment. Performing structural modification for such purposes should be guided by conducting detailed numerical simulations to avoid introducing new, or obscuring pre-existing, damage. This requires the use of qualified, well-validated analytical methodology with robust material-behavior representations of damage states, including the ability to model pre-existing cracks and other service-induced damage conditions. Such a methodology has been developed considering a wide range of structural behavior and service conditions. This paper describes the behavioral modeling attributes of this methodology and its application to the diagnostic evaluation of the effects of long-term service and environmental conditions on concrete infrastructures in nuclear power plants, including structural retrofitting and rehabilitation of reactor containments, and proof of performance for life-extension service. Reinforced concrete, despite its apparent simplicity, presents one of the most difficult problems in constitutive modeling and numerical simulation. The difficulties in constitutive modelling of concrete structures stem from the fact that cracks can form at relatively low stress levels. This leads to highly complex interaction between post-cracking behavioral regimes, which include: multidirectional cracking, crushing, shearing along rough crack surfaces and the consequential formation of rebarreacted compression normal to crack surfaces. One observable manifestation of this interaction is the potential for split cracking parallel to the compressive stress field imposed by the tendon in a pre-stressed containment, especially if the tendons are located close to the external surface of the containment wall where concrete confinement is small. Such split cracking could grow with time and may lead to wall delamination, with potentially serious consequences for the containment's design-basis function. The ageing-dependent constitutive properties of the concrete material considered are: tensile strength, post-cracking shear behavior, and compressive strength, with both, hardening effects due to the maturity of concrete compressive strength and softening effects due to environmental degradation, temperature, internal micro-cracking and past-loading-related macro cracks. A robust analytical treatment of these properties in a general three-dimensional stress-strain constitutive formulation that recognizes the interaction with the reinforcement is fundamental to the degree of fidelity of the predicted response to true physical behavior.

Enhancement of leak rate estimation model for corroded cracked thin tubes

During the last couple of decades, lots of researches on structural integrity assessment and leak rate estimation have been carried out to prevent unanticipated catastrophic failures of pressure retaining nuclear components. However, from the standpoint of leakage integrity, there are still some arguments for predicting the leak rate of cracked components due primarily to uncertainties attached to various parameters in flow models. The purpose of present work is to suggest a leak rate estimation method for thin tubes with artificial cracks. In this context, 23 leak rate tests are carried out for laboratory generated stress corrosion cracked tube specimens subjected to internal pressure. Engineering equations to calculate crack opening displacements are developed from detailed three-dimensional elastic-plastic finite element analyses and then a simplified practical model is proposed based on the equations as well as test data. Verification of the proposed method is done through comparing leak rates and it will enable more reliable design and/or operation of thin tubes.

Oxygen pump method for leak rate testing of SiO 2–B 2O 3–Al 2O 3–BaO–PbO 2–ZnO glass sealant for SOFC

A method for oxygen leak rate testing was presented based on oxygen pump technique. In this method, using the theory of oxygen pump and oxygen sensor, a formula for the calculation of leak rate was derived from Nernst equation, Faraday law and ideal gas law. In the gas leakage testing, a glass sealant based on SiO 2–B 2O 3–Al 2O 3–BaO–PbO 2–ZnO (named AF8) was used to seal two yttria stabilized zirconia (YSZ) plates together to form a small gas chamber, one plate worked as oxygen pump and the other worked as oxygen sensor. Constant current was applied on the oxygen pump to take out/into oxygen from the gas chamber resulting in its oxygen partial pressure change, the corresponding concentration voltage data were collected continuously for the calculation of the oxygen leak rate. With this method, the calculated leak rate of AF8 was 8.2 × 10 − 8 mbar l s − 1 cm − 1 at 800 °C which was smaller than the required value of 10 − 7 mbar l s − 1 cm − 1 .

In-Line Critical Leak Rate Testing of Vacuum-Sealed and Backfilled Resonating MEMS Devices

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Simplified processing of anode-supported thin film planar solid oxide fuel cells

The concept of anode-supported planar solid oxide fuel cells (SOFCs) involves different fabrication procedures and several sintering steps. This contribution is an attempt towards minimizing the number of sintering steps. By adopting an improved technique a pre-firing step was eliminated. Cells obtained following this co-firing technique showed relatively higher power density (∼1 A cm −2 at 0.7 V) at an operating temperature of 800 °C compared to the cells prepared by the conventional technique. In the present investigation, cells as large as 10 cm×10 cm were prepared and characterized following the new co-firing approach. Detailed microstructural, leak-rate testing (gastightness of the sintered electrolyte film) and electrochemical characterizations were carried out.

Decision Analysis and Its Application to the Frequency of Containment Integrated Leakage Rate Tests

For nuclear utilities to become competitive in a deregulated electricity market, costs must be reduced, safety must be maintained, and interested stakeholders must remain content with the decisions being made. One way to reduce costs is to reduce the frequency of preventive maintenance and testing. However, these changes must be weighed against their impact on safety and stakeholder relations. We present a methodology that allows the evaluation of decision options using a number of objectives that include safety, economics, and stakeholder relations. First, the candidate decision options are screened to make sure that they satisfy the relevant regulatory requirements. The remaining options are evaluated using multiattribute utility theory. The results of the formal analysis include a ranking of the options according to their desirability as well as the major reasons that explain this ranking. These results are submitted to a deliberative process in which the decision makers scrutinize the results to ensure that they are meaningful. During the deliberation, new decision options may be formulated based on the insights that the formal analysis provides, as happened in the case study of this paper. This case study deals with the reduction in frequency of the containment integrated leak rate test of a boiling water reactor.

Safety evaluation of rehabilitation of delaminated containment dome

Delamination of the inner (primary) containment dome of the Kaiga Atomic Power Project, Unit-1 (Kaiga-1) occurred during construction. Subsequent to the investigation of the delamination phenomenon, rehabilitation of the containment dome was initiated. A methodology for safety evaluation of rehabilitation of the containment dome was suitably devised based on the outcome of the investigation work. Emphasis was laid, during safety evaluation, on the safety of the existing ring beam, development of review basis for the re-engineered dome, design review and quality assurance in design, quality assurance of materials used, examination of constructibility using the new design, and safety review of construction activities of the re-engineered dome. The inner containment structure was accepted after the successful pre-commissioning proof (structural integrity) and leakage rate tests.

Polydimethylsioxane fluidic interconnects for microfluidic systems

This paper presents novel polydimethylsioxane (PDMS) based interconnects for microfluidic systems with a low dead volume. Through-hole type and /spl lceil/ type PDMS interconnects have been designed, fabricated, and tested for glass and plastic capillary tubing. Oxygen reactive ion etching and epoxy bonding methods are employed to bond PDMS interconnects to different substrate materials including silicon, glass, polymer and other thin film materials. Leakage pressure, leakage rate, and pull-out force are characterized for these interconnects. For reusable PDMS interconnects, the maximum leakage pressure reaches 510 kPa (75 psi) and the maximum pull-out force is about 800 mN. For nonreusable PDMS interconnects, the maximum leakage pressure is found to be 683 kPa (100 psi) and the maximum pull-out force is 2 N. For both types of PDMS interconnects, the leakage rate test demonstrates that the leakage is not detectable at a working pressure of 137 kPa (20 psi).

Pre-service tightness tests of HTR-10 primary loop pressure boundary system

The main design and operating parameters for the 10 MW high temperature gas-cooled reactor (HTR-10) primary loop pressure boundary system are introduced in this paper. The component installations and the pneumatic and tightness test are also described, including the objectives and methods. The leakage rate test results are analyzed to show that the results meet the design requirements and have enough safety redundancy.

ICONE11-36186 INTEGRATED LEAK RATE TESTING OF CONTAINMENTS-A PERSPECTIVE

Appendix J of 10 CFR 50,A Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors, @ sets the testing requirements for preoperational and periodic verification of the leak-tight integrity of the primary reactor containment, including systems and components which penetrate containment of light water-cooled power reactors, and establishes the acceptance criteria for such tests. The purposes of the tests are to assure that leakage through the primary reactor containment and systems and components penetrating primary containment would not exceed allowable leakage rate values as specified in the plants=technical specifications. The current Appendix J requirements provide two options for performing the tests. Option A (fully deterministic) requires that after the preoperational leakage rate tests, a set of three integrated leak rate tests (ILRT-termed as Type A tests) shall be performed, at approximately equal intervals during each 10-year service period. Option B does not provide a quantitative requirement for scheduling the periodic Type A tests. However, NEI 94-01, A Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50,Appendix J, @ and NRC Regulatory Guide (RG) 1.163, A Performance-Based Containment Leak Test Program, @ provide guidelines for determining the frequencies of preoperational and periodic leak rate tests using a performance based approach. The Option B requirements and NEI 94-01 guidance are based on NUREG-1493, A Performance Based Containment Leak-Test Program. @ Based on the information provided by the operating data and their risk significance, the study calculated the change in risk (in person-rem) to public for the 15 alternatives considered in the study. However, recognizing the non-sensitivity of risk to change in Type A leak rate testing frequencies, as depicted in NUREG-1493,a number of licensees are proposing changes to their Type A test frequencies using NRC risk-informed guidance in RG 1.174, A An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis. The paper discusses certain key deterministic aspects of the risk-informed decision for the plant-specific changes in the ILRT frequencies.

JV AWARDED OMAN GAS CONTRACT

The objective of this study is to investigate the long-term performance of a typical CANDU®1 containment structure. A three-dimensional nonlinear finite element model was built to realistically evaluate the performance of the structure under service load as well as a hypothetical beyond-design level internal pressure. Consideration is given to the time-dependent effects, such as shrinkage, creep, and relaxation of prestressing tendons, over a 60-year timeframe. In addition, the sensitivity of the response of the containment structure against support condition, internal temperature profile and temporary construction openings was also investigated. The accuracy of the numerical model was validated against structural measurements made during a routine leak rate test. The analysis results show that the containment structure would develop a ductile mechanism if the internal pressure significantly exceeded the design pressure. The pressure-deformation relationship of the structure is sensitive to the considered time-dependent parameters.

New plasma display panel packaging technology using electrostatic bonding method

In this work, we have developed a plasma display panel (PDP) packaging technology, replacing the tip off tube with an indirect glass-to-glass electrostatic bond with intermediate layers of indium tin oxide and amorphous silicon. The glass-to-glass bonding for the packaging was performed at a low temperature of 180 °C while applying a bias of 250 Vdc in an environment of three mixed gases: He–Ne(27%)–Xe(3%). The 3.6 in. color alternating current-PDP with 8 mm thickness was successfully fabricated, with full white color brightness at a firing voltage of 190 V. Additionally, the luminance during the sustained period was 900–1000 cd/m2. To investigate the capability of applying the glass bonding technology, the hermetic sealing test of the packaged panel was examined by a spinning rotor gauge for time variation. The leak rate test was performed with a helium detector. Since there is very low leakage through the bonded interface during 160 h, we can propose that it is highly possible to apply this technology to any vacuum packaged device.

Performance-oriented and risk-based regulation for containment testing

In August 1992, the US Nuclear Regulatory Commission (NRC) began a major initiative to develop requirements for containment testing that are less prescriptive and more performance oriented and risk based than current requirements. This action was a result of public comments and several studies that concluded that the economic burden of certain present containment testing requirements was not commensurate with their safety benefits. The rule-making included considering relaxing the allowable containment leakage rate, increasing the interval for the containment integrated leak rate test, and establishing intervals for the containment local leak rate tests on the basis of the performance of containment isolation valves and penetrations. A study has been conducted to provide technical information to establish the preformance criterion for containment tests, i.e. the allowable leakage rate, commensurate with its significance to total public risk. For the study, the results used were from a comprehensive study conducted by the NRC (NUREG-1150, `Severe accident risks: an assessment for five U.S. nuclear power plants') to examine the sensitivity of containment leakage to public risk. Risk was found to be insensitive to a containment leakage rate up to levels of about 100% volume per day for certain types of containments. Probabilistic risk assssment methods have also been developed to establish risk-based intervals for containment tests on the basis of experience. Evaluations show that increasing the interval for the integrated containment leakage test from three times to once every 10 years would have an insignificant impact on public risk. Analyses of operational experience data for local leak rate tests show that performance-based testing (valves and penetrations that preform well are tested less frequently) is feasible with a marginal impact on safety. These technical studies have been used to develop efficient (cost-effective) requirements for containment tests.

A review of the design and construction of the Sizewell B pressurized water reactor building containment structures

The design and construction of the Sizewell B primary containment has resulted in a structure different in certain aspects from that envisaged at the time when the decision was made for Sizewell to be based on the SNUPPS design which originated in the USA. The successful structural overpressure test (SOT) and integrated leak rate test (ILRT), performed at the end of 1993, marked the completion of approximately 5 1 2 years of civil construction activity on the containments. With the construction of the structure complete, it is a suitable time to review the design of the structure and to ascertain whether significant improvements can be gained, in terms of cost and construction programme, for further projects based on the Sizewell design.

Results of reliability test program on light water reactor piping

The Japan Atomic Energy Research Institute has conducted a piping reliability test program to demonstrate the safety and reliability of light water reactor primary piping. In this program, pipe fatigue test, leak-before-break (LBB) verification test and pipe rupture test were carried out to examine the integrity of piping, to verify the LBB and to demonstrate the effectiveness of protective measures against jet impingement and pipe whip loads under a pipe rupture event. In the pipe fatigue test, a procedure to predict the fatigue crack growth was developed, and the integrity of piping during the plant service life was evaluated. In the LBB verification test, the pipe fracture test and the leak rate test were performed to verify the LBB in the primary piping. In the pipe rupture test, the influence of jet impingement on the target disk and the deformation behavior of whipping pipe and restraint were investigated. Using the test results, the jet impingement behavior and the effectiveness of pipe whip restraint were demonstrated.

Containment penetrations — Flexible metallic bellows: Testing, safety, life extension issues

The performance and long-term operational integrity of containment systems and components is being challenged as many of the world's nuclear plants progress into the second half of their design lives. As time in service increases, so does the likelihood of component failure due to long term degradation. By observing trends in containment degradation, potential weaknesses can be anticipated and corrected, minimizing interruptions in operations, increasing safety and improving the plant's life extension outlook. One such trend, affecting containment penetration flexible metallic bellows that are subject to cyclic loading, is beginning to appear in some BWR plants. Although the overall performance of flexible metallic bellows penetrations has appeared to be acceptable for approximately 20 years, aging and degradation of these components has been recently identified. Corrosion and fatigue mechanisms have lead to cracking and subsequent leakage through the stainless steel bellows elements. Although periodic testing in accordance with 10CFR50, Appendix J has been performed, fabrication features of some expansion joint assemblies has made it difficult to collect reliable leakage rate data. In fact, recent observations suggest that it is not possible to perform valid Local Leak Rate Testing (LLRT) of certain two-ply bellows elements. This paper presents a description of efforts made to characterize the nature and significance of the bellows degradation observed. Field pressure testing and laboratory examinations are discussed, as well as evaluations to determine leak areas and leakage rates. Crack growth from both corrosion and fatigue mechanisms are examined, and methods for predicting remaining useful life are discussed. Repair and replacement methods will be summarized as well as testing to demonstrate replacement effectiveness.

A review on operating experiences of containment isolation systems in Korea

Wide experience, on the order of 50 reactor-years, in operating containment isolation systems has been accumulated in Korea since 1978, necessitating a review of the operating data to confirm the integrity of containments and to establish the future direction to the containment test program. The objectives of the present work are to collect, consolidate and assess the operational data relevant to containment isolation systems, and then to identify types of isolation valve failures, dominant leakage paths and factors affecting integrated leakage rate test. Leakage is observed to be the most frequent type of isolation valve failure. The causes of the valve leakage are packing leakage, seat damage due to foreign material, and misadjustment of torque switches. Malfunction and deterioration are also observed to be frequent failure modes. General leakage trends show that more careful surveillance during preoperational test can reduce the containment leakage. Significant leakage paths are found to be through air locks and large valves of six-inch or greater nominal diameter, so that weighted surveillance and inspection on these dominant leakage paths can considerably reduce the containment leakage. Atmospheric stabilization is found to be the most important factor to obtain the reliable integrated leakage rate. In order to achieve a well stabilized atmosphere, temperature and flow rate of compressed air should be kept constant and it is preferable not to operate fan coolers during pressurizing the containment for test. These results are expected to be a guidance for improving the operation of containment system.