Secondary Coolant System Research Articles

Effect of organic film on magnetite deposition behavior of alloy 690 steam generator tubes in simulated PWR secondary system

The purpose of this study is to elucidate the effect of a film forming amine (FFA) on the magnetite deposition behavior of Alloy 690 tube in the secondary coolant system of pressurized water reactors. By the addition of octadecylamine (ODA), which is a FFA, an organic film with a petal-like form was evenly formed on the surface of the steam generator (SG) tube. Regardless of the addition of ODA, all simulated deposits were verified as magnetite and had a polyhedral or spherical shape. The amount of deposits on the film-formed SG tube was reduced by approximately 58% compared to the deposits on the as-received SG tube. The influence of the film-formed surface of the SG tube is discussed from the point of view of wettability, zeta potentials, and resistance of the electric double layer.

Evaluation of the Potential for Containment Bypass due to Steam Generator Tube Rupture in VVER-1000/V320 Reactor during Extended SBO sequence using SCDAP/RELAP5 code

A severe accident-induced of a Steam Generator (SG) tube releases radioactivity from the Reactor Coolant System (RCS) into the SG secondary coolant system from where it may escape to the environment through the pressure relief valves and an environmental release in this manner is called “Containment Bypass”. This study aims to evaluate the potential for “Containment Bypass” in VVER/V320 reactor during extended Station Blackout (SBO) scenarios that challenge the tubes by primarily involving a natural circulation of superheated steam inside the piping loop and then induce creep rupture tube failure. Assessments are made of SCDAP/RELAP5 code capabilities for predicting the plant behavior during an SBO event and estimates are made of the uncertainties associated with the SCDAP/RELAP5 predictions for key fluid and components condition and for the SG tube failure margins.

Artificial neural network for predicting nuclear power plant dynamic behaviors

A Nuclear Power Plant (NPP) is a complex dynamic system-of-systems with highly nonlinear behaviors. In order to control the plant operation under both normal and abnormal conditions, the different systems in NPPs (e.g., the reactor core components, primary and secondary coolant systems) are usually monitored continuously, resulting in very large amounts of data. This situation makes it possible to integrate relevant qualitative and quantitative knowledge with artificial intelligence techniques to provide faster and more accurate behavior predictions, leading to more rapid decisions, based on actual NPP operation data. Data-driven models (DDM) rely on artificial intelligence to learn autonomously based on patterns in data, and they represent alternatives to physics-based models that typically require significant computational resources and might not fully represent the actual operation conditions of an NPP. In this study, a feed-forward backpropagation artificial neural network (ANN) model was trained to simulate the interaction between the reactor core and the primary and secondary coolant systems in a pressurized water reactor. The transients used for model training included perturbations in reactivity, steam valve coefficient, reactor core inlet temperature, and steam generator inlet temperature. Uncertainties of the plant physical parameters and operating conditions were also incorporated in these transients. Eight training functions were adopted during the training stage to develop the most efficient network. The developed ANN model predictions were subsequently tested successfully considering different new transients. Overall, through prompt prediction of NPP behavior under different transients, the study aims at demonstrating the potential of artificial intelligence to empower rapid emergency response planning and risk mitigation strategies.

System dynamics simulation of the thermal dynamic processes in nuclear power plants

Abstract A nuclear power plant (NPP) is a highly complex system-of-systems as manifested through its internal systems interdependence. The negative impact of such interdependence was demonstrated through the 2011 Fukushima Daiichi nuclear disaster. As such, there is a critical need for new strategies to overcome the limitations of current risk assessment techniques (e.g. the use of static event and fault tree schemes), particularly through simulation of the nonlinear dynamic feedback mechanisms between the different NPP systems/components. As the first and key step towards developing an integrated NPP dynamic probabilistic risk assessment platform that can account for such feedback mechanisms, the current study adopts a system dynamics simulation approach to model the thermal dynamic processes in: the reactor core; the secondary coolant system; and the pressurized water reactor. The reactor core and secondary coolant system parameters used to develop system dynamics models are based on those of the Palo Verde Nuclear Generating Station. These three system dynamics models are subsequently validated, using results from published work, under different system perturbations including the change in reactivity, the steam valve coefficient, the primary coolant flow, and others. Moving forward, the developed system dynamics models can be integrated with other interacting processes within a NPP to form the basis of a dynamic system-level (systemic) risk assessment tool.

Comparison on Two Option Design of The RDE Cogeneration System

The Reaktor Daya Eksperimen (RDE) is the High Temperature Gas Cooled Reactor (HTGR) type that will be mainly used for research activities of nuclear energy utilization. The RDE thermal power is 10 MWt using Helium gas as a primary cooling system that applies indirect cycle and superheated steam resulted from the steam generator of secondary coolant system. In separate production of electricity, some energy is discarged as waste heat, but in the cogeneration, some of heat energy is removed to be used for heat applications. Regarding to safety aspect and to optimize the utilization of heat energy, the cogeneration system using the heat exchanger could be applied. There are two options to be stidied in cogeneration system which is the heat exchanger that installed in the steam line. The option-1 is installed in outlet line from the Steam Generator, the option-2 is installed in outlet line from turbine. Purpose of this paper is to compare the cogeneration system between the two options. The comparison is carried out by the ChemCAD software. The feed water for cogeneration usingmass flow rate from 0.6 kg/s until 2.0 kg/s at temperature of 28°C were studied. The results show that the heat removal on the option-2 provides high power of turbine for generating the electricity more than the option-1. This study is expected provide possibilities of extended studies in the area of process heat applications.

Micro-Galvanic Corrosion of Steam Generator Materials within Pores of Magnetite Flakes in Alkaline Solutions

In secondary coolant system of the pressurized water reactors, the reduced corrosion products such as metallic Cu and Pb particles were accumulated in the pores of the magnetite flakes and electrically contacted to the steam generator materials. The micro-galvanic corrosion behavior of steam generator materials (steam generator tube materials: Alloy 600 and Alloy 690, steam generator tube sheet materials: SA508 Gr.3) contacted to the corrosion products (magnetite, Cu, and Pb) was investigated in an alkaline solution. The steam generator materials considered in this study were all the anodic elements of the galvanic pair because their corrosion potentials were lower than those of the corrosion products. The corrosion rate of the steam generator materials was increased by the galvanic coupling with the each corrosion products, and was more accelerated with increasing the area ratio of the corrosion products to the steam generator materials. Among the corrosion products, Cu has the largest galvanic effect on steam generator materials in the pores when area ratio of cathode to anode is 10.

A new method for D2O to H2O leak detection and identification of leaky heat exchanger in PHWR by Cerenkov photon counting technique

In pressurized heavy water reactors, leaks from D2O primary coolant and moderator to H2O secondary coolant and other light-water systems in heat exchangers cannot be completely ruled out. High cost of D2O demands that its loss should be prevented to maximum extent possible. Traditionally D2O leak detection and identification of leaky heat exchanger is carried out by measurement of tritium activity in H2O. Since tritium emits low energy beta radiation, its concentration in H2O is measured by mixing it with liquid scintillation solution in a definite proportion in counting vial and counted in a Liquid Scintillation Analyzer. It is very sensitive method for leak detection, but identification of leaky heat exchanger is time consuming and may require low power operation or reactor shut down. In the new method, high energy beta emitting fission products, which emit Cherenkov photons in H2O, were used as the tracer. H2O was poured in 20 mL plastic vials without scintillator and counted on Liquid Scintillation Analyzer. D2O leak was identified by comparing the Cherenkov photon count rate with that of the blank. A discrimination ratio significantly higher than average Cherenkov photon count rate for all heat exchangers was used to identify the leaky one. The technique has advantageous over existing method of D2O leak detection, such as, (1) scintillation chemicals are not required (2) low power operation or reactor shut down is not required for identifying the leaky heat exchanger (3) no generation of radioactive chemical waste (4) on-power leak identification reduces generation of radioactive liquid waste.

Evaluation Of Operating Performance Of The Reactor Coolant System Of Rsg-Gas Using Chemcad6.1.4

<p>Evaluation of the operation of reactor main coolant system RSG-GAS is very important. This is related with the operation of the reactor that has been going on more than 27 years although the current reactor coolant system is still operating safely. The main coolant system RSG-GAS including both the primary and secondary cooling system. Evaluation can be conducted by calculating the coolant system performance by using the ChemCad6.1.4 software. Furthermore, this calculation result is compared with the most recent measurement in the cooling system operation. By this way it can be known the performance degradation in the Reactor Coolant System operation.The operating parameter data of primary and secondary cooling system are temperature, pressure and flow rate of coolant. The reactor operation at the power of 15 MW is used as a reference data.The calculation result shows that the temperature parameter between calculation by ChemCad and the measurement data qualitatively is no significant deviations. However in the secondary coolant system, the outlet pressure of both heat exchanger nozzles are 0.8bar and 1.2bar that indicate lower than the calculation results. It’s may be caused by the flow disturbances through the tube-side of heat exchangers. In generally the existing operating parameters on the reactor cooling system in terms of safety aspects still under safely condition.</p>

Modeling primary and secondary coolant of a nuclear power plant system with a unique framework (MCUF)

Within the study and design of a nuclear power plant extensive system modeling is necessary to determine how the reactor is going to perform in any given situation, not only in the normal performance of the reactor but also transients including anticipated transients without scram (ATWS) and hypothetical accidents. Primary and other loops in multiple coolant loop systems in nuclear power plants can be modeled in nuclear system codes, such as Reactor Excursion and Leak Analysis Program (RELAP5), COBRA-TRAC, TRACE, and ATHLET, for the interaction and feedback effects between the coolant loops. In the recent years, new components, such as turbine, coolant pump, advanced compact heat exchangers, valves, sensors, instrumentation and control systems, are being designed to use in the coolant systems. However, current system codes are not capable of integrating the computational model of the whole nuclear power plant system to an experimental apparatus. Therefore, this article proposes a new coupling between a system code and Laboratory Virtual Instrument Engineering Workbench (LABVIEW) in a unique framework (called as Modeling of Coolant Systems with a Unique Framework -MCUF-) that allows integrating an experimental apparatus or a facility to a whole power plant model. MCUF allows online or interactive data exchange by using the powerful and flexible tools of LABVIEW. This article demonstrates how coupling between the primary and secondary coolant system of a typical Pressurized Water Reactor (PWR) can be performed by using MCUF. The primary and secondary sides of the PWR are modeled with RELAP5 and LabVIEW computer simulators respectively. The coupling between RELAP5 and LabVIEW has been executed with steady state and transients, in this case a loss of coolant accident (LOCA) for a four loop PWR. The results of the coupling have been compared with the typical RELAP5 results which significantly depend on “hard coded” data for second coolant loop. Code-to-code benchmark results demonstrate that this unique coupling framework has a good agreement with the RELAP5 code results with “hard coded” data for the secondary coolant system.

Evaluation of radioactivity and gamma spectra in the secondary coolant system of PWR

“Source Term” is the fundamental data used to evaluate the environmental impact of radioactive releases during normal operation. This paper presents a general investigation on the computational model of radiation source-term for the secondary coolant system of the pressurized water reactor (PWR). Research is carried out on the radionuclide migration, adsorption, retention, decay. Accordingly, mathematic models are described on the basis of the mechanism of radionuclides generation and disappearance in the system. Based on the implementation of these models, the corresponding function modules were developed and tested, which completes the source-term program previously developed. The nuclear reaction data of nuclides was extracted from the evaluated nuclear data library-ENDF/B-VII.1. The results obtained from preliminary verification between this work and ANSI/ANS-18, 1999 supported the models, indicating that the models could be used in the secondary coolant system for radiation shielding, accident prevention, environmental assessment and nuclear facility decommission.

Conducting water chemistry of the secondary coolant circuit of VVER-based nuclear power plant units constructed without using copper containing alloys

The secondary coolant circuit water chemistry with metering amines began to be put in use in Russia in 2005, and all nuclear power plant units equipped with VVER-1000 reactors have been shifted to operate with this water chemistry for the past seven years. Owing to the use of water chemistry with metering amines, the amount of products from corrosion of structural materials entering into the volume of steam generators has been reduced, and the flow-accelerated corrosion rate of pipelines and equipment has been slowed down. The article presents data on conducting water chemistry in nuclear power plant units with VVER-1000 reactors for the secondary coolant system equipment made without using copper-containing alloys. Statistical data are presented on conducting ammonia-morpholine and ammonia-ethanolamine water chemistries in new-generation operating power units with VVER-1000 reactors with an increased level of pH. The values of cooling water leaks in turbine condensers the tube system of which is made of stainless steel or titanium alloy are given.

Water chemistry used in the secondary coolant circuit of unit 3 at the rovno nuclear power station involving correction treatment of working medium with lithium hydroxide and ethanolamine

The all-volatile water chemistry used in the secondary coolant circuit involving correction treatment of the steam generator’s boiler water with lithium hydroxide and the ethanolamine water chemistry are analyzed from the viewpoint of their effect on the erosion-corrosion wear of equipment used in the secondary coolant system and damageability of heat-transfer tubes used in PGV-1000M steam generators.

Overview of safety and environmental issues for inertial fusion energy

This paper summarizes safety and environmental issues of Inertial Fusion Energy (IFE): inventories, effluents, maintenance, accident safety, waste management, and recycling. The fusion confinement approach among inertial and magnetic options affects how the fusion reaction is maintained and which materials surround the reaction chamber. The target fill technology has a major impact on the target factory tritium inventory. IFE fusion reaction chambers usually employ some means to protect the first structural wall from fusion pulses. This protective fluid or granular bed also moderates and absorbs most neutrons before they reach the first structural wall. Although the protective fluid activates, most candidate fluids have low activation hazard. Hands-on maintenance seems practical for the driver, target factory, and secondary coolant systems; remote maintenance is likely required for the reaction chamber, primary coolant, and vacuum exhaust cleanup systems. The driver and fuel target facility are well separated from the main reaction chamber.

Steam Generator Secondary pH during a Steam Generator Tube Rupture

The Nuclear Regulatory Commission requires utilizes to determine the response of a pressurized water reactor to a steam generator tube rupture (SGTR) as part of the safety analysis for the plant. The SGTR analysis includes assumptions regarding the partitioning of iodine between liquid and vapor in steam generator secondary. Experimental studies have determined that the partitioning of iodine in water is very sensitive to the pH. Based on this experimental evidence, the NRC requested the INEL to perform an analytical assessment of secondary coolant system (SCS) pH during an SGTR. Design basis thermal and hydraulic calculations were used together with industry standard chemistry guidelines to determine the SCS chemical concentrations during an SGTR. These were used as input to the Facility for Analysis of Chemical Thermodynamics computer system to calculate the equilibrium pH in the SCS at various discrete time during an SGTR. The results of this analysis indicate that the SCS pH decreases from the initial value of 8.8 to approximately 6.5 by the end of the transient, independent of PWR design.

Measurements of Radioiodine Species in Samples of Pressurized Water Reactor Coolant

Samples of primary coolant at two pressurized water reactors in the United States have been collected and analyzed to determine the fraction of the radioiodine present in volatile forms. The volatile species, I2 and organic iodides, would be available for prompt release following a steam generator tube rupture (SGTR) accident, which overpressurizes the secondary coolant system and causes venting to the atmosphere. Coolant samples were collected at full power, during power reduction at the start of an outage, and up to 48 h after shutdown. Radioiodine concentrations spiked as power was decreased, and an increase in the volatile species fraction to ∼20% was seen within 2 h of shutdown. A second peak of 30 to 40% volatile species occurred at ∼1.5 days after shutdown, but is not significant for SGTR accidents.

The Fort St. Vrain high temperature gas-cooled reactor

Abstract The Fort St. Vrain primary and secondary coolant systems have given satisfactory performance during the rise to power test program with the tests being terminated at the current maximum allowable thermal reactor power of 70% of rated. Because of a regenerative heat problem in the steam generators, rated conditions of 1000°F main and hot reheat steam temperatures, predicted to occur at 25% power, were not reached until 68%. The regenerative heat problem also forced “overblowing” of the core with primary coolant helum which resulted in higher fuel temperatures than predicted, lower core primary coolant outlet temperatures and higher core primary coolant inlet temperatures. Data suggest that all parameters will be at rated conditions at 80–100% power. A small steam generator tubing leak was detected by the primary coolant moisture monitors of the plant protective system. It was located by covergas techniques and repaired by plugging the leaking feedwater and steam subheaders external to the reactor.

Fundamental Experiment of Potassium Heat Exchanger Using Principle of Heat Pipe

Abstract In order to provide compact and reliable sodium equipments including a steam generator, performance tests are conducted with a potassium heat exchanger, which is featured by the separate construction of primary and secondary coolant systems. A small amount of potassium plays a role as an intermediate media of heat transportation between these two coolant systems. Heat is transfered by evaporation and condensation of potassium on the surfaces of the primary and the secondary coolant pipings, respectively. The tests are performed in the temperature range of 200-300°C and the maximum heat transfer reaches 1.3 kW (heat transfer rate at the primary heating source: 8.6 W/cm2 at 300°C). The experimental results are analyzed by using Langmuir's and Schrage's equations and close agreement between experiment and theory is obtained.

On estimated modes of failure of nuclear power plants by potential earthquakes

This paper deals with the estimated ‘modes of failure’ of nuclear power plants during future violent earthquakes. The authors have been surveying the damage to industrial plants caused by several violent earthquakes since 1960. Some of them have already been reported in English, but here the authors try to rearrange them from the viewpoint of ‘modes of failures’ of nuclear power plant buildings, equipment, vessels and piping. The authors categorize the mechanisms of failure as follows: (i) damaged by the dynamic effect of acceleration waves, (ii) by resonance in displacement waves, (iii) by the static effect of seismic force, (iv) by external force from attached piping and others, or forced deformation, and (v) by liquefaction of soil. The authors try to determine the modes of failure of the following items in a matrix form of the mechanisms: (i) the reactor building, (ii) steel containment vessel, (iii) auxiliary building, (iv) reactor vessel, (v) core internals, (vi) primary and secondary coolant system, (vii) emergency power supply system, (viii) emergency gas treatment system and stack, (ix) fuel cooling pond and fuel rack, (x) refuel machine crane, (xi) auxiliary system and component, (xii) turbine and its pedestal, and (xiii) main power system and control instrumentation. They also examine them from another point of view, i.e. in ‘the classification of the important factor’ of items for their aseismic design.