Integrated Leak Rate Test Research Articles

Numerical simulation of gas diffusion and explosion limit of VOCs in containment integrated leak rate test

CILRT is an important work in the overhaul of nuclear power plants. VOCs from paints and cleaners used in the containment during the overhaul will migrate inside the containment, and the numerous rooms and compartments inside the containment make the air flow impeded, and the VOCs may be enriched to bring combustion risk. The complete fire risk analysis method includes volatilization kinetics of chemicals, mixture explosion limits, and numerical simulation of VOCs diffusion and migration. The volatilization rate at the lowest room temperature of 20°C is used as a conservative input, the complement of CAITA realistic model is used as the flow domain, and the geometry is appropriately simplified, and the steady ventilation process, the internal flow and the laws of diffusion and agglomeration of 12 kinds of VOCs gases in the 9-h pressure-holding process are investigated by Computational Fluid Dynamics. The simulation results of steady ventilation show that the VOCs gases are almost uniformly carried by the air to all spaces inside the containment, and the explosion risk is low. The concentration field of the steady ventilation process was taken as the initial boundary condition and unsteady state simulation of the 9-h holding pressure process was performed. The results show that the gas flow inside the containment basically stops after 3456 s of pressure holding. With the extension of time, the VOCs gases accumulate at the bottoms of the three lower layers of the containment under the effect of laminar diffusion and gravitational settling, where the total concentration of the mixed gases reaches the highest at the bottom of the lowest layer. The total concentration of the VOCs gas mixture was below the lower explosion limit throughout the 9-h pressure-holding process, and the explosion risk was low.

Development of new AE technology to detect and characterize leak on nuclear shielding building from inside during integrated leakage rate test

French Projects of Investment for the Future, called “Research for Nuclear Safety and Radiation Protection” have been initiated to further research on the causes, the management, the impact of the observed nuclear accidents and to propose and validate solutions to limit the risk and the consequences. In this context the "Non Destructive Evaluation of nuclear plants containment” project” (ENDE) with eight partners (six research institutes and two industrials) supported by the “National Agency of Research”, proposes a methodology for the Non Destructive Evaluation of the containment capacity. The two major functions to be assess are strength and leak tightness. Under the frame of the VeRCoRs project EDF has developed and build a containment mock-up on the 1/3 scale (diameter 20m, height 20m, wall thickness 40cm) which has been used for some partners of ENDE project. Mistras has been involved in characterisation of concrete cracking during pneumatic proof tests either in ENDE and VeRCoRs programs. Since some leaks occurs in the area of monitoring we use AE to characterized them. The mockup has been used to develop and characterized an instrumentation and technology for leak detection with a major challenge leak detection has to be performed from this inside of the containment building with all the contingences of operating of full scale operating nuclear plant. This paper describe the steps of development and validation of the methodology up to the application this measurement on site help the leak characterisation during pneumatic ten year Integrated Leakage Rate Test at 4,2 bar. EWGAE 35, Ljubljana, Slovenia, 13th – 16th Sep. www.ewg

Vibration Monitoring of Nuclear Power Plant Containment Buildings During the Integrated Leakage Rate Test for Structural Condition Assessment

Abstract This study investigates the use of the operational vibrations produced during the Integrated Leak Rate Test of nuclear power plant containment buildings for further informing on its mechanical condition. The experiment is performed on a 1:3-scale containment building mock-up. The results show that meaningful vibrations were generated during the pressurization test. Different features were extracted from the vibration signals and analyzed as a function of the internal pressure. Experimental modal analysis was performed and demonstrated that several frequency peaks generated during the pressurization cycle effectively corresponded to the eigenmodes of the containment building. The identified operational frequency modes exhibited remarkable hysteretic dependencies on the internal pressure. The latter was phenomenologically described through a simplified two-dimensional (2D) finite element model of the vessel. Besides, a surrogate statistical model based on the Principal Component Analysis of the vibration data was proposed as a baseline and so detect abnormal behavior. Then, different synthetic damage scenarios were created by subtlety altering the recorded signals and ultimately substantiate the capability of the statistical model to detect these odd signals. Finally, conclusions were drawn regarding the possibility of using mechanical vibrations for assisting in the licensing process of nuclear power plants and monitor the structural health condition of in-service containment buildings.

Evaluation of the thermal strain of an NPP containment structure during leakage rate tests

A strategy for isolating the thermal strain due to solar radiation from the total measured strain during the Integrated Leak Rate Test of a nuclear containment structure is presented in this paper. In the proposed method, a series of numerical models are developed to estimate the temperature change on the prestressed concrete containment structure and to predict the strain due to temperature and applied pressure. The effects of the solar radiation, heat exchange with internal and external air and heat conduction are considered in the proposed method. The developed method is validated by reproducing the total strain profiles measured from the leak rate tests conducted at the Point Lepreau CANDU®11CANDU (CANada Deuterium Uranium Reactor) is a registered trademark of the Atomic Energy of Canada Limited (AECL), used under exclusive license by Candu Energy Inc. 6 containment structure.

Comparison of Measured and Calculated Data for NPP Krško CILR Test

Containment is the last barrier for release of radioactive materials in the case of an accident in the nuclear power plant (NPP). Its overall integrity is tested during a containment integrated leak rate test (CILRT) at the design pressure, at regular intervals. Due to applied risk based licensing, the test intervals can be increased up to once in 10 years and beyond. Taking that into account it is important to prepare the test properly and to use obtained results to assess the real status of the containment. The test can be used to verify existing containment calculation models. There is a potential benefit of verified computer models usage for the explanation of some test results, too. NPP Krško has performed CILRT during the plant outage in 2016. The paper presents a comparison between measured data and results calculated using a multivolume GOTHIC (Generation Of Thermal Hydraulic Information For Containment) model. The test scenario was reproduced using limited available data up to the end of the pressurization phase. The depressurization phase is calculated by the code and measured leakage rate is implemented in the model. Taking into account the necessary adjustments in the model, overall prediction of the measured results (in terms of pressure, temperature and humidity) is very good. In the last phase of the test some non-physical behavior is noticed (without influence on overall test results), probably caused by the combination of air redistribution within the containment and influence of heat transfer to plant systems that were in the operation during the test. GOTHIC model was used to check sensitivity of the predicted pressure (leak rate) to different heat inputs and to investigate the influence that operation of only one reactor containment fan cooler (RCFC) train during pressurization can have on the mixing of air within the containment. In addition, the influence of currently used weighting factors (weighting of measured temperature, relative humidity and pressure data) on the used test methodology is investigated. The possible non-conservative direction of the influence (currently used weighting factors are giving lower leakage rate) was demonstrated and a new set of weighting factors is proposed too.

중수로 원자로건물 총누설감시계통 시험 중지에 따른 리스크 영향 평가

Wolsong Unit 2,3&4 has been performing a containment integrity test during power operation. This test could impact to the safe operation during test. If an accident occurs during pressure dropping phase, reactor trip can be delayed because of the increased pressure difference which causes a time delay to reach the trip set-point. On the contrary, if an accident occurs during pressure increasing phase, reactor trip could be accelerated because the pressure difference to the trip set-point decrease. Point Lepreau nuclear power plant, which installed GCLMS (Gross Containment Leakage Monitoring System) in 1990, has discontinued the test since 1992 due to these adverse effects. Therefore, we evaluated the risk to obviate the GCLMS test based on PWR's ILRT (Integrated Leak Rate Test) extension methodologies. The results demonstrate that risk increase rate is not high in case of performing only ILRT test at every 5 years instead of doing GCLMS test at every 1.5 years. In addition, the result shows that GCLMS test can be removed on a risk-informed perspective since risk increasement is in acceptable area of regulatory acceptance criteria.

Structural Monitoring and Response Analysis of Prestressed Concrete Primary Containment of a Nuclear Power Plant

Before commissioning, Pressurised Heavy Water Reactor (PHWR) containments are tested for their integrity at proof pressure and Loss-of-Coolant Accident (LOCA) pressure. During operation, as per the current practice for Indian PHWR containments, in-service Integrated Leakage Rate Test (ILRT) is done once in two years at about one third of LOCA based test pressure. Kaiga Generating Station unit-2 (KGS-2) has been in operation for more than 11 years. For the first time, a periodic ILRT at LOCA pressure (0.86 kg/cm2 (g)) was conducted at KGS-2. During this ILRT, structural monitoring of pre-stressed concrete primary containment was carried out. The deflection and strains at critical locations in the IC were recorded during ILRT mainly to compare the same with the values obtained from analytical study and with those observed during the first ILRT. The strain and deflection measurements were taken using vibrating wire strain gauges (VWSG) and high precision dial gauges respectively connected to data-logger. This technical paper highlights the structural behaviour of KGS-2 primary containment after 11 years of operation.

Over-pressure test on BARCOM pre-stressed concrete containment

Bhabha Atomic Research Centre (BARC), Trombay has organized an International Round Robin Analysis program to carry out the ultimate load capacity assessment of BARC Containment (BARCOM) test model. The test model located in BARC facilities Tarapur; is a 1:4 scale representation of 540MWe Pressurized Heavy Water Reactor (PHWR) pre-stressed concrete inner containment structure of Tarapur Atomic Power Station (TAPS) unit 3&4. There are a large number of sensors installed in BARCOM that include vibratory wire strain gauges of embedded and spot-welded type, surface mounted electrical resistance strain gauges, dial gauges, earth pressure cells, tilt meters and high resolution digital camera systems for structural response, crack monitoring and fracture parameter measurement to evaluate the local and global behavior of the containment test model.The model has been tested pneumatically during the low pressure tests (LPTs) followed by proof test (PT) and integrated leakage rate test (ILRT) during commissioning. Further the over pressure test (OPT) has been carried out to establish the failure mode of BARCOM Test-Model. The over-pressure test will be completed shortly to reach the functional failure of the test model. Pre-test evaluation of BARCOM was carried out with the results obtained from the registered international round robin participants in January 2009 followed by the post-test assessment in February 2011. The test results along with the various failure modes related to the structural members – concrete, rebars and tendons identified in terms of prescribed milestones are presented in this paper along with the comparison of the pre-test predictions submitted by the registered participants of the Round Robin Analysis for BARCOM test model.

Development of web-based Off-Site Consequence Analysis Program (OSCAP) for extending ILRT intervals and its application

For an offsite consequence analysis, MELCOR Accident Consequence Code System (MACCS) II code is widely used as a tool. In this study, the algorithm of web-based Off-Site Consequence Analysis Program (OSCAP) using the MACCS II code was developed for an integrated leak rate test (ILRT) interval extension and Level 3 probabilistic safety assessment (PSA), and verification and validation (V&V) of the program was performed. The main input data of the MACCS II code are meteorological data, population distribution data and source term data. However, it requires lots of time and efforts to generate the main input data for an offsite consequence analysis using the MACCS II code. For example, the meteorological data are collected from each nuclear power site in real time, but the formats of the raw data collected are different from each other as a site. To reduce efforts and time for risk assessments, the web-based OSCAP has an automatic processing module which converts the format of the raw data collected from each site in Korea to the input data format of the MACCS II code. The program also provides an automatic function of converting the latest population data from Statistics Korea, the National Statistical Office, to the population distribution input data format of the MACCS II code. In case of the source term data, the program includes the release fraction of each source term category resulting from Modular Accident Analysis Program (MAAP) code analysis and the core inventory data from ORIGEN code analysis. These analysis results of each plant in Korea are stored in a database module of the web-based OSCAP, so a user could select the defaulted source term data of each plant without handling source term input data. For V&V of the web-based OSCAP, the following analyses were performed. First of all, an offsite consequence analysis for reference plants was carried out using only the MACCS II code, and using the web-based OSCAP based on both methodologies of NUREG-1493 and NEI interim report. Secondly, the results of the offsite consequence analysis were compared with each other. Lastly, a sensitivity analysis was accomplished for the main input data of the meteorological data and the population distribution data using the web-based OSCAP. It is considered that the program can remarkably reduce man-power on performing a periodic risk assessment according to extending ILRT intervals, and Level 3 PSA. Also, the program can be applied to making accident management plan (AMP) and quantitative health objective (QHO), and to the field of public acceptance (PA).

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.

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.

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.

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.

The PWR integrated Leak Rate Test, a review of experiences and results

The paper reviews the Integrated Leak Rate Test (ILRT) as it has been carried out in the USA and as reported in papers in European countries. The test procedures are critically appraised and recommendations are given for modifications to them. The valves used in a PWR are identified as a main source of leaks and possibilities for improvement are discussed. The use of a part pressure test and its limitations are considered. A part pressure test cannot give the same assurance as a full pressure test but may be useful for the identification of gross leaks. Secondary effects such as weather and the use of Van de Waals equations are considered and are found to be not important for concrete containments.

高速実験炉「常陽」の格納容器全体漏えい率試験

Method of integrated leak rate test (ILRT) of reactor containment vessel for LMFBR is different from that of LWR, since the test is required to be conducted under operating condition of the reactor sodium coolant system, primary argon cover gas system and air conditioning system. So, it is a diffculty to obtain an adequate measurement accuracy under these conditions.The diffculty was solved by the past three ILRTs condcted in JOYO, and accuracy of the data was confirmed to be almost equal to that of LWR.From these, the adequacy of the test method and the test instrumentation system was demonstrated and the ILRT of LMFBR was established.

Engineering Innovations for Davis-Besse Containment

Design and construction innovations relative to size, construction sequence, and testing are presented to show how they provided maximum schedule flexibility to permit the containment vessel and shield building to be built on schedule despite serious delays caused by AEC licensing hearings. These innovations consisted of: (1) sufficient additional volume to preclude field stress relieving; (2) two-stage construction permitting simultaneous erection of the steel vessel and concrete of the vessel bottom head; (3) use of sufficient winter protection; and (4) elimination of the initial bare-vessel test with a single combined integrated leak rate test and structural integrity test conducted at the end of the overall plant construction.