Primary Pressure Boundary Research Articles

An FE-MCS based modeling method for event sequence development for the multi-module probabilistic risk assessment

The Multi-Unit Probabilistic Risk Assessment (MUPRA) has been a research hotspot since the Fukushima Daichi accident in 2011. Among the technical elements, event sequence modeling is the most challenging one in MUPRA implementation. During the PRA development for a 600 MW HTR-PM nuclear power plant (HTR-PM600) which consist of 6 reactor modules, we find that the complexity of the obtained event sequence model increases rapidly and massively with the number of reactors, therefore it is quite difficult to guarantee the modeling correctness. To address this issue, in this paper, we propose a multi-module event sequence modeling method based on Minimal Cut Sets at the level of Functional Event (FE-MCS). The proposed method is illustrated with two cases in this paper, namely, Loss of Coolant Accident caused by the Isolatable Large breach in primary pressure boundary (LLOCA-I) and Loss of vital Direct Current bus (LSDC).

Fretting wear assessment in PHWR pressure tubes using transient dynamic FEA and Archard’s wear model

In Pressurized Heavy Water Reactors (PHWRs), the fuel bundles are placed in a pressure tube (PT). These bundles rest on inside of pressure tubes through the bearing pads. The pressure tubes are part of the primary pressure boundary of the reactor, and are crucial from functionality and safety point of view. Any phenomena/ factors that may cause structural degradation of PT and may affect its structural integrity, requires thorough investigation. The fretting of pressure tube at bearing pad locations is one such phenomenon, wherein the local material removal due to fretting may lead to stress concentration and associated structural degradations. Therefore, a detailed assessment of the PT fretting damage incurred during its intended life is required. Present paper covers the premise and methodology based on Archard’s wear model for evaluation of the volumetric wear rate of pressure tube. A case study has been done to evaluate the fretting wear in pressure tubes of a large Indian PHWR. 3D transient computational fluid dynamic analysis and finite element analysis with contact modelled between the fuel bundle bearing pads and pressure tube are carried out to evaluate the contact force and sliding distance time histories. Based on the analyses results, the work rate and average saturated fret depth are evaluated. The assessment revealed insignificant concern owing to the fretting of pressure tube for the considered cases.

Estimation of pipe failure frequencies in the absence of operational experience data: A pilot study

Probabilistic failure metrics such as leak frequency and rupture frequency are commonly used to characterize piping reliability. The methodologies for calculating the failure metrics rely on a complex set of input parameters. Operating experience data and experimental data play an important role in informing the different input parameters. The paper describes results and conclusions of a coordinated research project to benchmark three different reliability models using a four-step procedure: reference case definition of relevance to advanced reactor designs, input parameter calibration, validation of results, and application of different methodologies upon completion of the calibration and validation steps. The reference case is a weld consisting of nickel-base alloy 152/52 and located within a primary pressure boundary of an advanced reactor. This alloy is a class of structural materials known to be highly resistant to stress corrosion cracking. Synergies between the different methods are noted and the importance of a multi-disciplinary approach to input parameter development is underscored. A key conclusion is that the three methods are equally suitable for estimating failure frequencies. In any specific application, a selection of the most practical or effective computational tool can be considered. The comparison of alternative models confirms and helps to gain confidence in the computed failure frequency estimates. The study was part of a coordinated research project organized by the International Atomic Energy Agency.

An innovation idea for SBLOCA mitigation strategy: Self-coagulation system in comparison with human hemostatic mechanism

The loss of coolant accident (LOCA), as one of the design basis accidents (DBAs), is a hypothetical accident that is usually considered in the design of nuclear power plant. LOCA is caused by small/large breaks in the reactor primary coolant and pressure boundary and may result in a loss of reactor coolant at a rate in excess of the reactor makeup system capability. Currently, the typical mitigation measures for LOCA are through the vessel high/low pressure coolant injection (HPCI/LPCI), core spray system, or reactor core isolation cooling (RCIC) system, to protect reactor from fuel melting and core degradation. However, is it possible to directly repair the small breaks passively? Inspired by the hemostasis and coagulation mechanism of human blood vessels, the authors are considering to develop a technology which would allow the reactor self-coagulation during the small break loss of coolant accident (SBLOCA). With the assistance of this innovative technology, the speed of losing coolant through reactor primary loop breaks could be significantly reduced or even the loss is stopped. With more and longer existing coolant in the core, the possibility of core degradation will also be significantly reduced. This innovative technology needs to be designed feasible and simple, which can be directly used into the existing nuclear power plants to improve the inherent safety of the cooling system, benefit the economy and safety of the nuclear power plant. In this paper, the authors compare the reactor primary loop of nuclear power plant with the blood circulation system of human beings from comprehensive and multi-angle perspectives, discuss the similarity, feasibility, and economy of nuclear power plant self-coagulation system based on the existing hemostasis and coagulation mechanism of human blood vessels. If possible, this technology will play a revolutionary positive role in alleviating DBAs related to SBLOCA.

Effect Of Notch Orientation On Fracture Behaviour Of Textured Zr-2.5Nb Material

Zr-2.5Nb pressure tubes are used as primary pressure boundary material in Indian Pressurized Heavy Water Reactors (PHWRs). Due to presence of strong tangential basal pole texture in the HCP α-phase of Zr-2.5Nb alloy, the mechanical properties of the pressure tube are different along axial and circumferential directions. As pressure tubes see higher hoop stress, most of the fracture studies have been carried out on pressure tubes having axial notch or crack orientation. However, in practice, flaws may be present in material in any orientation. Hence, fracture studies have been carried out on textured Zr-2.5Nb pressure tube to evaluate its fracture properties along axial and transverse directions. Single specimen J-test with Direct Current Potential Drop (DCPD) system was carried out from room temperature to 300°C using Disc compact tension (DCT) specimens. This paper compares the fracture toughness values in these two different directions of textured Zr-2.5Nb pressure tube.

Research on the Application of Fen for Environmentally Assisted Fatigue Evaluation of the Austenitic SS Pipe Under Combined Transient Loads

Accumulative test data indicate that the effects of the light water reactor (LWR) environment could cause the fatigue resistance of primary pressure boundary components materials to be significantly reduced. Environmentally assisted fatigue (EAF) is the abbreviation of the environmentally assisted fatigue. In 2007, Nuclear Regulatory Commission (NRC) issued RG. 1.207. It was updated in 2014. And, it requires that the effects of LWR environment on the fatigue life reduction of metal components should be considered for new design plants. And it suggests to use environmental correction factor, Fen, to account for EAF. NRC regulation (NUREG), NUREG/CR-6909 (NRC, 2013, “Effect of LWR Coolant Environments on the Fatigue Life of Reactor Materials,” U.S. Nuclear Regulatory Commission, Argonne, IL, Standard no. NUREG/CR-6909), presents the detail Fen calculation formula. Fen is a function of temperature, strain rate, dissolved oxygen level in water, and sulfur content of the steel. Accordingly, Fen calculation will present a comparatively conservative result. Depending on the experience of the primary pressure boundary piping transient operation, Fen varies during each transient. More uncertainty and confusion are raised during the application of the Fen method. The research work in this paper includes: first, the typical character of piping thermal transient is derived based on the existing experience. Second, small specimen EAF tests are conducted depending on the above derived combined loading characters. Then effort is taken to improve the application of the Fen method for the combined multitransient loading conditions. And the results are compared with those of the lowest instantaneous Fen method and equalization of the weighted Fen method. Finally, a designed test plan is presented.

Critical heat flux for downward-facing pool boiling on CANDU calandria tube surface

One accident scenario in CANDU reactors that can challenge the integrity of the primary pressure boundary is a loss of coolant accident, referred to as critical break LOCA, in which the pressure tube (PT) can undergo thermal creep strain deformation and contact its calandria tube (CT). In such case, rapid redistribution of stored heat from PT to CT, leads to a large spike in heat flux to the moderator which can cause bubble accumulation and dryout on the CT surface. A challenge to fuel channel integrity is posed if critical heat flux occurs on the surface of the CT and results in sustained film boiling. If the post-dryout temperature becomes sufficiently high then continued creep strain of the PT and CT may lead to fuel channel failure. In this study, a mechanistic model is developed to predict the critical heat flux variations along the downward facing outer surface of CT. The hydrodynamic model considers a liquid macrolayer beneath an elongated vapor slug on the surface. Local dryout is postulated to occur whenever the fresh liquid supply to the macrolayer is not sufficient to compensate for the liquid depletion. A boundary layer analysis is performed, treating the two phase motion as an external buoyancy driven flow. The model shows good agreement with the available experimental data and has been modified to take into account the effect of subcooling.

Zirconium hydrides and Fe redistribution in Zr-2.5%Nb alloy under ion irradiation

Zr-2.5%Nb alloy is used to fabricate the pressure tubes of the CANDU reactor. The pressure tube is the primary pressure boundary for coolant in the CANDU design and is susceptible to delayed hydride cracking, reduction in fracture toughness upon hydride precipitation and potentially hydride blister formation. The morphology and nature of hydrides in Zr-2.5%Nb with 100 wppm hydrogen has been investigated using transmission electron microscopy. The effect of hydrides on heavy ion irradiation induced decomposition of the β phase has been reported. STEM-EDX mapping was employed to investigate the distribution of alloying elements. The results show that hydrides are present in the form of stacks of different sizes, with length scales from nano- to micro-meters.Heavy ion irradiation experiments at 250 °C on as-received and hydrided Zr-2.5%Nb alloy, show interesting effects of hydrogen on the irradiation induced redistribution of Fe. It was found that Fe is widely redistributed from the β phase into the α phase in the as-received material, however, the loss of Fe from the β phase and subsequent precipitation is retarded in the hydrided material. This preliminary work will further the current understanding of microstructural evolution of Zr based alloys in the presence of hydrogen.

Steam generator tube rupture simulation using extended finite element method

A steam generator (SG) is an important component of any pressurized water reactor. Steam generator tubes represent a primary pressure boundary whose integrity is vital to the safe operation of the reactor. SG tubes may rupture due to propagation of a crack created by mechanisms such as stress corrosion cracking, fatigue, etc. It is thus important to estimate the rupture pressures of cracked tubes for structural integrity evaluation of SGs. The objective of the present paper is to demonstrate the use of extended finite element method capability of commercially available ABAQUS software, to model SG tubes with preexisting flaws and to estimate their rupture pressures. For the purpose, elastic–plastic finite element models were developed for different SG tubes made from Alloy 600 material. The simulation results were compared with experimental results available from the steam generator tube integrity program (SGTIP) sponsored by the United States Nuclear Regulatory Commission (NRC) and conducted at Argonne National Laboratory (ANL). A reasonable correlation was found between extended finite element model results and experimental results.

Investigation of Burst Pressures in PWR Primary Pressure Boundary Components

In a reactor coolant system of a nuclear power plant (NPP), an overpressure protection system keeps pressure in the loop within 110% of design pressure. However if the system does not work properly, pressure in the loop could elevate hugely in a short time. It would be seriously disastrous if a weak point in the pressure boundary component bursts and releases radioactive material within the containment; and it may lead to a leak outside the containment. In this study, a gross deformation that leads to a burst of pressure boundary components was investigated. Major components in the primary pressure boundary that is structurally important were selected based on structural mechanics, then, they were used to study the burst pressure of components by finite element method (FEM) analysis and by number of closed forms of theoretical relations. The burst pressure was also used as a metric of design optimization. It revealed which component was the weakest and which component had the highest margin to bursting failure. This information is valuable in severe accident progression prediction. The burst pressures of APR-1400, AP1000 and VVER-1000 reactor coolant systems were evaluated and compared to give relative margins of safety.

Irradiation response of delta ferrite in as-cast and thermally aged cast stainless steel

To enable the life extension of Light Water Reactors (LWRs) beyond 60 years, it is critical to gain adequate knowledge for making conclusive predictions to assure the integrity of duplex stainless steel reactor components, e.g. primary pressure boundary and reactor vessel internal. Microstructural changes in the ferrite of thermally aged, neutron irradiated only, and neutron irradiated after being thermally aged cast austenitic stainless steels (CASS) were investigated using atom probe tomography. The thermal aging was performed at 400 °C for 10,000 h and the irradiation was conducted in the Halden reactor at ∼315 °C to 0.08 dpa (5.6 × 1019 n/cm2, E > 1 MeV). Low dose neutron irradiation at a dose rate of 5 × 10−9 dpa/s was found to induce spinodal decomposition in the ferrite of as-cast microstructure, and further to enhance the spinodal decomposition in the thermally aged cast alloys. Regarding the G-phase precipitates, the neutron irradiation dramatically increases the precipitate size, and alters the composition of the precipitates with increased, Mn, Ni, Si and Mo and reduced Fe and Cr contents. The results have shown that low dose neutron irradiation can further accelerate the degradation of ferrite in a duplex stainless steel at the LWR relevant condition.

Dynamic Characteristics Analysis of Steam Generator Tube of Nuclear Power Plant by Finite Elements

In nuclear power plants, the steam generator heat transfer tube is the weakest part of the primary circuit pressure boundary. Flow induced vibration is one of the main reasons for the failure of the heat transfer tube. In this paper, an ANSYS finite element software is used to carry out the modal analysis of the heat transfer tube, and to simulate the dynamic response of the heat transfer tube in the harmonic load based on the modal analysis.

The development of Module In-vessel degraded severe accident Analysis Code MIDAC and the relevant research for CPR1000 during the station blackout scenario

To meet the domestic demand of software autonomous scheduling, Xi'an Jiaotong University had developed MIDAC (a Module In-vessel degraded severe accident Analysis Code) which can not only analyze the processes of in-vessel severe accident, but also provide the related results of each one. This code was composed of five modules: the early behavior module, the core degradation module, the debris bed module, the melting materials IVR module and the connecting module. In this paper, the basic mathematic-physical models of those modules were briefly introduced. Then, the CPR1000 station blackout scenario severe accident was set as an example to calculate the primary system thermal-hydraulic transient, the core degradation and the debris behavior, as well as the creep rupture of heat structures in primary loop pressure boundary. At last, the accuracy of MIDAC was verified in partial comparison with SCDAP/RELAP5 in primary system thermal-hydraulic transient analyzing part.

Flashing Flow of Subcooled Liquid through Small Cracks

Abstract In Pressurized Water Reactors and in CANDU reactors steam generator (SG) tubes represent a major fraction of the reactor primary coolant pressure boundary. The ability to estimate the leak rates from the through wall cracks in the steam generator tube is important in terms of radiological source terms and overall operational management of steam generators as well as demonstration of the leak-before- break condition. In this study an experimental program and analysis methods were developed to measure and assess the choking flow rate of subcooled water through simulated steam generator tube crack geometries. Experiments were conducted on choking flow for various simulated crack geometries for vessel pressures up to 7 MPa with various subcoolings. Measurements were done on subcooled flashing flow rate through well defined simulated crack geometries with L/D

Transient pressure analysis for the reactor core and containment of a HTGR after a primary loop pressure boundary break accident

The pressure variances in the reactor core and containment of a High Temperature Gas-cooled Reactor (HTGR) after a primary loop pressure boundary break accident determine the structural integrity and safety of the reactor. Based on mass conservation, energy conservation and state equations, explicit formulae for the transient pressure and temperature variances in the pressure vessels were deduced, and a set of differential equations for the transient pressure and temperature variances in the containment were developed. Numerical simulation was also conducted to investigate the transient pressure and temperature variances in the pressure vessels and containment. The results show that energy transformation due to expansion work cannot be neglected. The maximum pressure in the containment could increase by 40 percent due to blockage caused by air in the containment. Detailed numerical simulations of the transient pressure and temperature variance in the reactor core flow passages were also conducted. The results show that the pressures acting on the reactor core and containment are below acceptable values.

Safety Analysis for Hot Gas Duct Vessel in HTR-PM

The hot gas duct vessel (HGDV) is an important part of the high-temperature reactor–pebble-bed module (HTR-PM) primary loop pressure boundary system. It connects the reactor pressure vessel (RPV) and steam generator pressure vessel. Because the dimensions of the HGDV are smaller than those of the other two vessels, it is often considered the weakest of the three vessels. Therefore, the safety of the HGDV has become one of the most important issues in the design of the HTR-PM. In the present paper, a comprehensive safety analysis of the HGDV in the HTR-PM was performed with an emphasis on the structural features. The designs of the HGDV and the support system of the primary loop pressure boundary are first described. A preliminary safety analysis of the HGDV, including the stress calculations and leak-before-break (LBB) analysis, is then presented. The results show that the stress levels of the HGDV under various accidents have a safety margin of at least 55.3% compared with the stress limits specified in American Society of Mechanical Engineers code, and the stress intensity factor at the postulated flaw is less than the critical stress intensity factor. The LBB analysis indicated that the leak detection system is capable of detecting leaks caused by a postulated through-thickness crack in the HGDV before it reaches the critical size. Although the preliminary analysis has proved the safety of the HGDV, the consequences of a hypothetical HGDV double-ended break accident were also studied to further investigate the safety features of the HTR-PM. Several mitigation measures were employed based on the original design. The structural integrity of the support system, the reactor internals, and the containment under double-ended break accident were evaluated. The results show that these main structures could maintain integrity under the HGDV double-ended break accident.

An impact source localization technique for a nuclear power plant by using sensors of different types

In a nuclear power plant, a loose part monitoring system (LPMS) provides information on the location and the mass of a loosened or detached metal impacted onto the inner surface of the primary pressure boundary. Typically, accelerometers are mounted on the surface of a reactor vessel to localize the impact location caused by the impact of metallic substances on the reactor system. However, in some cases, the number of accelerometers is not sufficient to estimate the impact location precisely. In such a case, one of useful methods is to utilize other types of sensor that can measure the vibration of the reactor structure. For example, acoustic emission (AE) sensors are installed on the reactor structure to detect leakage or cracks on the primary pressure boundary. However, accelerometers and AE sensors have a different frequency range. The frequency of interest of AE sensors is higher than that of accelerometers. In this paper, we propose a method of impact source localization by using both accelerometer signals and AE signals, simultaneously. The main concept of impact location estimation is based on the arrival time difference of the impact stress wave between different sensor locations. However, it is difficult to find the arrival time difference between sensors, because the primary frequency ranges of accelerometers and AE sensors are different. To overcome the problem, we used phase delays of an envelope of impact signals. This is because the impact signals from the accelerometer and the AE sensor are similar in the whole shape (envelope). To verify the proposed method, we have performed experiments for a reactor mock-up model and a real nuclear power plant. The experimental results demonstrate that we can enhance the reliability and precision of the impact source localization. Therefore, if the proposed method is applied to a nuclear power plant, we can obtain the effect of additional installed sensors.

A Study of Air Ingress and Its Prevention in HTGR

A rupture of the primary piping in the helium-cooled and graphite-moderated high-temperature gas-cooled reactor (HTGR) represents a design-basis event that should not result in significant safety consequences. In such a loss-of-coolant event, the reactor would be shut down inherently, and the decay heat would be removed passively with the ultimate reactor temperature rise being less than the design limit. Still, an important concern for reactor safety continues to be graphite oxidation damage to the fuel and core should a major air ingress take place through the breached primary pressure boundary. Two major cases of air ingress are studied. The first case results from the rupture of a control rod or refuel access standpipe atop the reactor pressure vessel (RPV). To rule out the possibility of such a standpipe rupture, a design change is proposed in the vessel top structure. The feasibility of the modified vessel local structure is evaluated. The second case of air ingress results from the rupture of one or more main coolant pipes on the lower body of the RPV. Experiment and analysis are performed to understand the multiphased air ingress phenomena in the depressurized reactor. Accordingly, a new passive mechanism of sustained counter air diffusion is proposed and shown to be effective in preventing major air ingress through natural circulation in the reactor. The results of the present study are expected to enhance the HTGR safety and economics.

Application of continuous wavelet transform to the impact location estimation of the Loose Parts Monitoring System (LPMS)

The Loose Parts Monitoring System (LPMS) provides information on the location of a loosened or detached metal impacted onto the inner surface of a primary pressure boundary and on a mass or energy of the impact signal. For the conventional system, it is popular to employ the group delay among the sensors installed within 3 ft range from the impact source in a collection zone as stated in the Reg. Guide 1.133. However, there exist inherent errors for the estimation of the true location of the impact since it is highly dependent on the plant personnel's decision to determine the arrival time differences of the generated wave group among the neighbouring sensors. To overcome this problem in this study, the two-dimensional approach has been proposed and applied to effectively estimate the arrival time differences of the impact signal by using continuous wavelet transform which is one of the linear time-frequency analysis methods. The experiment has been implemented on the plate model and the steam generator's lower part in a real plant using steel balls. It is expected that the reliability of the location estimation could be enhanced when the proposed time-frequency method is introduced into the LPMS system.

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.