Reactor Coolant Loop Research Articles

Vibration cause analysis and elimination of stagnant branch pipeline

Fluid acoustic structure coupling vibration is rare in nuclear power plant (NPP), especially in the stagnant branch of reactor coolant loop (RCL). This paper reports the phenomenon of fluid acoustic solid coupling vibration of one stagnant branch pipeline founding in the hot function test (HFT) of NPP. Field measurement shows that the vibration exceeds the limit value so it is necessary to determine the root cause of pipeline vibration and provide the modified layout. Based on the data of single point measurement, on-line monitoring, knocking test, acoustic modal calculation and modal analysis, the authors obtain the vibration characteristics and laws of pipelines and infers the acoustic solid coupling. The computational fluid dynamics (CFD) analysis and Strouhal number calculation, the inference of fluid acoustic coupling is identified. According to the above inference, the modified layout and the vibration prediction of the modified layout are proposed. After modification the HFT and the power up test was started. The vibration trend was consistent with the prediction, which verified that the fluid acoustic structure coupling was the root cause of the stagnant branch vibration. The amplitude of the modified the stagnant branch pipeline reduced in the whole process of temperature rise. The operation meets the limit requirements. The stagnant branch can operate normally.

The probability of the existence of defects that lead to the destruction of the pressure vessel without leak

Relevance. To ensure the safety of a nuclear power plant on the basis of the requirements of norms and rules in the field of the use of atomic energy for pipelines of the primary circuit of a nuclear reactor, the design should apply the leak before break concept. The main idea of the concept is to prevent a sudden rupture of the pipelines of the reactor coolant loop, and consists in substantiating the fact that the rupture is preceded by the formation of a stable through crack, which is detected by the provided leak control means. When substantiating the concept, it is assumed that break without leak is an impossible event. This article provides a method for determining the probability of a failure event without a leak. Purpose - estimate the probability of the existence of a defect that can lead to the destruction of the vessel or pressure pipeline without leakage, as well as the probability of failure without leakage for a known number of loading cycles. Methods. To systematize the data obtained by different methods of non-destructive testing, conservative assumptions were used to determine the area of detected defects. On the basis of the obtained defect sizes, the defect size regions were determined, which can determine the scenarios of crack growth. Using the methods of mathematical statistics, the probability of the existence of a defect, which can lead to failure without leakage, was determined. Based on the methods of the theory of reliability, a comparison of the obtained probability of destruction with the admissible value is carried out. Results. A method for processing non-destructive testing data based on an assessment of the area of detected defects has been developed to systematize the data obtained by different non-destructive testing methods. The criterion for the development of cracks according to the scheme leak before destruction is determined. A method has been developed for determining the probability of a defect that can lead to failure without leakage. An example of calculation based on feed water pipelines is considered.

Heat transport and chimney design in a typical MTR reactor during natural convection cooling regime

In the present work, three dimensional CFD analysis is presented for a typical MTR reactor coolant loop during natural convection cooling regime. Design optimization studies are carried out. A model consisting of the core, the chimney, and the cold legs is simulated. The research reactor core is simulated as a porous zone with a cosine shape heat distribution. The chimney dimensions are varied to optimize its design for best thermal-hydraulic performance under natural convection conditions. The study is carried out for chimney extension ratio in the range of 1.5 ≤ Lch/Lcore ≤ 7.2 and chimney opening (expansion) ratio in the range of 1 ≤ B/b ≤ 2.5. Contours plot of temperature at different values of chimney extension and opening ratios were illustrated. The Nusslet number, coolant mass flow rate, maximum reactor core temperature, coolant velocity and core pressure drop are correlated as a function of the extension and opening ratios. Comparison with the results of CONVEC and RELAP5 codes shows a good agreement. The recommended optimum design conditions are for Lch/Lcore greater than 3.5, and B/b = 1 for best thermal-hydraulic performance.

The role of β-Zr in a Zr-2.5Nb alloy during aqueous corrosion: A multi-technique study

The Zr–2.5Nb alloy is used as pressure tubes in Canadian Deuterium Uranium (CANDU) nuclear reactors, and the typical starting microstructure consists of α-Zr grains elongated in both transverse and longitudinal directions and thin layers of partially decomposed β-Zr lying between the α-Zr grains. In this study, we have used state-of-the-art microscopy techniques to characterise the long-term thermally decomposed β phase in this alloy, and the oxide scale formed on them in a reactor coolant loop with the aim of understanding the mechanisms underpinning the thermal decomposition behaviour at service temperatures and exploring the role of the decomposed β-Zr phase in controlling the microstructure and microchemistry of the zirconium oxide, and hence its influence on the general corrosion resistance of the alloy. We observe that these β-Zr layers are heavily decomposed even after the short stress stage at 400 °C at the end of the manufacturing cycle, with a closely packed array of β-Nb precipitates forming in an α-Zr matrix. We have shown that the oxidation of these bands is significantly slower than the surrounding α-Zr matrix and that zirconium oxide grains are re-nucleated under each band. We conclude that it is the combination of the Nb-rich remnants of the original β-Zr layers arising from the hot extrusion and drawing stages and this new dense oxide that offers a significant barrier to the oxidation front (and also to the penetration of hydrogenic species), so the characteristic layered microstructure arising from the original manufacturing process is very important in determining the overall oxidation behaviour.

The Role of Β-Zr in Zr-2.5nb Alloys During Aqueous Corrosion: A Multi-Technique Study

Zr–2.5Nb alloy are used as pressure tubes in Canadian Deuterium Uranium (CANDU) nuclear reactors, and the typical starting microstructure consists of α-Zr grains elongated in both transverse and longitudinal directions and thin layers of partly decomposed β-Zr lying between the α-Zr grains. In this study, we have used state-of-the-art microscopy techniques to characterise the long-term thermally decomposed β phase in these alloys, and the oxide scale formed on them in a reactor coolant loop with the aim of understanding the mechanisms underpinning the thermal decomposition behaviour at service temperatures and exploring the role of the decomposed β-Zr phase in controlling the microstructure and microchemistry of the zirconium oxide, and hence its influence on the general corrosion resistance of the alloy. We observe that these β-Zr layers are heavily decomposed even after the short stress stage at 400°C at the end of the manufacturing cycle, with a closely packed array of β-Nb precipitates forming in an α-Zr matrix. We have shown that the oxidation of these bands is significantly slower than the surrounding α-Zr matrix and that zirconium oxide grains are re-nucleated under each band. We conclude that it is the combination of the Nb-rich remnants of the original β-Zr layers arising from the hot extrusion and drawing stages and this new dense oxide that offers a significant barrier to the oxidation front (and also to the penetration of hydrogenic species), so the characteristic layered microstructure arising from the original manufacturing process is very important in determining the overall oxidation behaviour.

Modeling and simulation of an integrated regenerative transcritical cycle with a small modular reactor

In recent years, small modular reactors (SMR) have gained substantial development advancement and world-wide acceptance. While various advanced power cycles (e.g. supercritical CO2 Brayton cycle) have been proposed for Gen IV reactors, the steam cycle remains the only cycle for light water reactors. In this work, a unique regenerative transcritical power cycle using small molecule organic fluids (methanol and ethanol) was modeled with the reactor coolant loop of a small modular light-water reactor. In addition to modeling the thermodynamic cycle and associated transport processes for prominent components, the solution algorithms of the combined cycle model using MATLAB/REFPROP are illustrated in this article. The simulation results show that, by leveraging the superior thermophysical properties around the critical points of the working fluids, the transcritical methanol/ethanol cycles have shown similar performance as the steam cycle in terms of thermal efficiency and power output. With further system optimization, the regenerative transcritical cycle is expected to offer increased power output from the small modular light-water reactor as an alternative energy conversion system.

ANALYSIS OF RESIDUAL STRESS FOR NARROW GAP WELDING USING FINITE ELEMENT METHOD

Reactor coolant loop (RCL) pipes circulating the heat generated in a nuclear power plant consist of so large diameter pipes that the installation of these pipes is one of the major construction processes. Conventionally, a shield metal arc welding (SMAW) process has been mainly used in RCL piping installations, which sometimes caused severe deformations, dislocation of main equipments and various other complications due to excessive heat input in welding processes. Hence, automation of the work of welding is required and narrow-gap welding (NGW) process is being reviewed for new nuclear power plants as an alternative method of welding. In this study, transient heat transfer and thermo-elastic-plastic analyses have been performed for the residual stress distribution on the narrow gap weldment of RCL by finite element method under various conditions including surface heat flux and temperature dependent thermo-physical properties.

The Study on Mechanical Properties in Narrow-Gap Welds for SA312 TP347

Conventionally, shielded-metal arc welding (SMAW) process has been applied to join pipes of reactor coolant loop, which caused defects and lot of loss in time and cost due to excessive heat input in joining section. Recently, narrow-gap welding (NGW) process was introduced to overcome the disadvantages of SMAW. However, the application of NGW to nuclear power plant is not yet commonly used, because safety of NGW process is not fully proven. In the present paper, welded coupons are made of stainless steel. They are manufactured under different processes; general welding (GW), and repair welding after GW. Performed are various mechanical tests to investigate microstructure, tensile strength and so on. It is verified that the mechanical properties of stainless steel are slightly changed after repair welding process. It is also found from stress corrosion cracking tests that the failure time of repair welding is shorter than that of general welding.

Transition to a new size of the sanitary-protective zone of a nuclear power plant

In our country, in accordance with world-wide practice, nuclear power plant safety is ensured by the systematic implementation of the concept of deeply esheloned protection based on the use of a system of physical barriers, including a fuel matrix, fuel-element cladding, boundary of the reactor coolant loop, airtight sealing of the reactor system, and radiation protection. In the Soviet Union, in contrast to other countries, an additional barrier was introduced in the construction and operation of nuclear power plants-a sanitary-protective zone separating the atomic power plant from housing developments.

Evaluation of Mechanical Properties for Narrow-Gap Weld Zone in Reactor Coolant Loop

Conventionally, shield metal arc welding (SMAW) process was applied to join pipes of RCL, which caused lot of loss in time and cost due to excessive heat input and defects in joining section. Recently, narrow-gap welding (NGW) process was introduced to overcome the disadvantages of SMAW. However, the application of NGW to nuclear power plant is not yet common because safety of NGW process is not proven. In present paper, the welded coupons are manufactured under different welding processes in carbon steel. Then, microstructure observations and various mechanical tests are performed. It is verified that the mechanical properties of carbon steel are greatly changed after repair welding process due to applied heat flux, and that the effect of post-welding heat treatment is beneficial.

Crud Removal Performance with Ion Exchange Resins in BWR Plants

It is needless to say that one of the most important roles of the condensate demineralizer in Japanese boiling water reactors (BWR) is to eliminate such impurities during accidental occurrence of sea water leakage from condensate cooling system. Ion exchange resins packed in condensate demineralizer have also been expected to decrease crud, or corrosion products (CP) in condensate water in order to finally reduce activated corrosion products (ACP) in the reactor coolant loop. It is perceived that crud removal ability of a condensate demineralizer has been improved year by year. And we call this phenomenon as “Aging Effect”. Typical property changes of aged cation exchange resin consisted of an increase of water retention capacity and a change of surface texture. Based on these findings, we formulated a new concept and developed new gel type ion exchange resins for the better crud removal. The results from column tests using actual condensate water for approximately three years and full scale test at an ac...

Behaviour of nonlinear supports on PWR coolant system during a postulated LOCA: Part 2: Effects of stiffness variations and cavity pressurisation

Part 2 of this two part paper deals with the variations in stiffness values of the supports and whip restraints significantly affected by uncertainties in evaluating concrete embedment stiffnesses and the effect of cavity pressurisation in the case of a Reactor Pressure Vessel (RPV) inlet or outlet nozzle break. Part 1 of the paper discussed the methodology of the LOCA analysis of the Reactor Coolant Loop (RCL) and a sensitivity study on the method of modelling the columns supporting the Reactor Coolant Pump and the Steam Generator. They were part of a series of sensitivity analyses performed to evaluate the influence of variations in the variables utilised as input to the primary coolant pipe dynamic structural analyses. The stiffness variations cause variations in loads received at the supports or restraints, thus influencing the design and the degree of accuracy to which it is necessary to define the stiffnesses. Cavity pressurisation is asymmetric pressurisation of the cavity between the RPV external surface and the primary shield wall. This generate external upthrust and lateral forces on the RPV thus increasing RPV movements and RCL support loadings. Different blockage situations impeding fluid pressure relief routes through the primary shield wall have been postulated due to the spalling of thermal insulation material. The loadings on the RCL supports and restraints as a result of increased RPV movements were analysed.

Double-ended breaks in reactor primary piping

The US Nuclear Regulatory Commission (NRC) is reevaluating current design criteria for light water reactor plants that require postulation of a double-ended guillotine break (DEGB) in reactor coolant piping. In support of this reevaluation, the Lawrence Livermore National Laboratory has estimated the probability of occurrence of a DEGB, and has assessed the effect that earthquakes have on DEGB probability. This report describes a probabilistic evaluation of reactor coolant loop piping in pressurized water reactor (PWR) plants having nuclear steam supply systems designed by Westinghouse and by Combustion Engineering. Two causes of pipe break are considered: pipe fracture due to the growth of cracks at welded joints (“direct” DEGB), and pipe rupture indirectly caused by the seismically-induced failure of heavy component supports (“indirect” DEGB). The probability of direct DEGB was estimated using a probabilistic fracture mechanics model. The probability of indirect DEGB was estimated by convolving seismic hazard and support fragility. The results of this study indicate that the probability of a DEGB from either cause is very low for these plants, and that the NRC should therefore consider eliminating reactor coolant loop DEGB as a plant design basis in favor of more realistic criteria.

Simulation experiments of loose parts and abnormal vibration of core internals

The drifting and impact phenomena of loose parts in the reactor coolant loop and also the change of vibration characteristics of the core internals due to their looseness or defects were studied by means of simulation experiments employing reduced scale model. The model loop on the scale of one fifth consisted of a reactor vessel (R/V), core internals (C/I) and the water chamber of a steam generator (S/G) which were constructed to simulate the flow and vibration characteristics of the actual components according to similality law. The experiments were performed in two stages. In the first stage or a loose parts simulation test, simulated loose parts of different shapes scaled down to the model were thrown into the loop to study the drifting and impact characteristics by means of accelerometers installed on the surface of R/V and S/G. In the second stage, the core internals were intentionally subjected to abnormal conditions to examine changes in vibration characteristics of the core internals and study the possibility of monitoring abnormal vibration from outside R/V. On the basis of the model test results and data accumulated in actual plant operation, a loose parts monitoring system was developed that allows estimation of impact location and impact energy on-line real time.

Dynamic analysis of a primary reactor coolant system subjected to seismic and LOCA loads, including shield building interaction

This paper is intended to describe the development of a unique method of analysis for a primary reactor coolant piping system to evaluate its adequacy and the need for structural support system optimisation under hypothetical seismic and loss-of-coolant-accident loading conditions. An integrated approach is presented for a linear three-dimensional dynamic analysis of a typical 900 MWe PWR nuclear reactor coolant loop, accommodating dynamic coupling and interaction effects of shield building, including soil-foundation springs and damping parameters. Also included are major influences on sensitive equipment due to their variable support boundary conditions and rigid restraints. Due to the vital nature of reactor coolant systems, important dynamic design criteria have to be established in order to estimate natural frequencies, damping characteristics and overall system behaviour during LOCA and strong motion earthquakes as well as to optimise the structural design of various components.

Abstracts of contributed papers for annual meeting Professional Group on Nuclear Science October 6 and 7, 1954

The problem of designing a constant reactor outlet temperature control system for a thermal reactor power plant is considered. The dynamic control characteristics of the power plant are described in terns of (1) the response of reactivity to variations of reactor inlet coolant temperature, (2) the heat transfer characteristics of the reactor and boiler, and (3) the time lags of the reactor coolant loop. It is shown how the inherent characteristics of the power plant may be used to advantage by constant reactor outlet temperature control systems employing the reactor inlet and outlet temperatures as control variables. An important feature of these control systems is the computation of the reactor inlet temperature from the boiler outlet temperature or the steam plant variables using recently developed transport delay circuits to compensate for the boilerto-reactor coolant transport delay.