Pressurized Water Reactor System Research Articles

Pressurizer Modeling: Using Different Thermodynamic Models and Comparing Results with RELAP Code Results

In pressurized water nuclear power plants, pressurizer is applied to control the primary loop’s pressure during transient processes. There are different thermodynamic models, which calculate the pressurizer’s tank pressure and water/vapor thermodynamic properties in different times. In this article, single-region and two-region models have been employed to simulation of WWER-1000 type pressurizer. Here, the obtained pressure from both models has been compared with RELAP5/Mod3 code findings. The RELAP is a thermal-hydraulic code, which has been developed and accepted throughout the world for simulating loop flows, for example pressurized water reactor (PWR) systems together with their related scaled systems. Finally, the pressure results have been compared and indicate that the presented models accuracy, are acceptable. The differences between models and RELAP code results are justifiable.

Advanced progressive real coded genetic algorithm for nuclear system availability optimization through preventive maintenance scheduling

One of the main goals of reliability-centered maintenance programs is to find optimal maintenance strategies. Availability of emergency systems through preventive maintenance scheduling has an important role in Probabilistic Safety Analysis (PSA). Preventive maintenances can be replaced with reloading time maintenance and reduce unavailability of the system, especially in the last days of reload cycle. Of course, the emergency core cooling systems connect to circuit in accidents and control the intensity of the unavailability. For reducing unavailability of standby emergency systems, preventive maintenance scheduling method is offered. In this paper, an Advanced Progressive Real Coded Genetic Algorithm (APRCGA) is applied to optimize the availability of standby systems with preventive maintenance scheduling. Using APRCGA in nuclear maintenance systems and considering appropriate objective functions, the most suitable conditions are obtained. The preventive maintenance scheduling keeps unavailability of systems within safe and reliable conditions. In order to demonstrate the effectiveness of the proposed method, it is applied for two nuclear power plant emergency systems. First, for the nuclear emergency core cooling system of a two-loop pressurized water reactor, most available maintenance scheduling is presented. In the second case, for a four-loop pressurized water reactor, the maintenance scheduling for emergency core cooling system is proposed. The results are compared with those obtained by some standard maintenance policies, and previously published papers.

PFM APPLICATION FOR THE PWSCC INTEGRITY OF Ni-BASE ALLOY WELDS-DEVELOPMENT AND APPLICATION OF PINEP-PWSCC

Often, probabilistic fracture mechanics (PFM) approaches have been adopted to quantify the failure probabilities of Ni-base alloy components, especially due to primary water stress corrosion cracking (PWSCC), in a primary piping system of pressurized water reactors. In this paper, the key features of an advanced PFM code, PINEP-PWSCC (Probabilistic INtegrity Evaluation for nuclear Piping-PWSCC) for such purpose, are described. In developing the code, we adopted most recent research results and advanced models in calculation modules such as PWSCC crack initiation and growth models, a performance-based probability of detection (POD) model for Ni-base alloy welds, and so on. To verify the code, the failure probabilities for various Alloy 182 welds locations were evaluated and compared with field experience and other PFM codes. Finally, the effects of pre-existing crack, weld repair, and POD models on failure probability were evaluated to demonstrate the applicability of PINEP-PWSCC.

Verification and Validation of a thermal stratification experiment CFD simulation

Abstract Thermal stratification and striping are observed in many piping systems including those of nuclear power plants. Periodic occurrences of these thermal transients lead to fatigue and may induce undesirable failures and deformations to the piping. The Thermal Hydraulic Laboratory of the Centro de Desenvolvimento da Tecnologia Nuclear/Comissao Nacional de Energia Nuclear (CDTN/CNEN) conducts an experimental and numerical project simulating the thermal stratified flows in piping systems of pressurized water reactors (PWR) to obtain some understanding on these phenomena. Experiments were carried out in a test section simulating the steam generator injection nozzle of a PWR. A numerical simulation of one experiment was performed with the commercial finite volume Computational Fluid Dynamic code CFX 13.0. A vertical symmetry plane along the pipe was adopted to reduce the geometry in one half, reducing mesh element size and minimizing processing time. The RANS two equations RNG k–ɛ turbulence model with scalable wall function and the full buoyancy model were used in the simulation. In order to properly evaluate the numerical model it was performed a Verification and Validation (V&V) process. Numerical uncertainties due to mesh refinement and time step were evaluated. This validation process showed the great importance of a proper quantitative evaluation of numerical results. In past studies qualitative evaluations were considered enough and numerical results like the one presented here could be considered satisfactory for the prediction of thermal stratified flows. However, with the present V&V study it was possible to identify objectively the strengths and weaknesses of the model.

Thin layer activation for probing flow accelerated corrosion of carbon steel

Carbon steel feeder elbows and feeder piping of the primary coolant system of pressurized heavy water reactors undergo significant wall thinning due to flow accelerated corrosion. The rate of corrosion in certain locations exceeds the design corrosion allowance. Corrosion rate is dependent on several parameters like pH, temperature, dissolved oxygen, chromium content in material, pipe geometry, etc. Thin layer activation, utilizing the 57Fe(p,n) 57Co reaction, enabled rapid online measurement of wall thinning in a simulated straight feeder pipe. Significant velocity and pH dependence was observed for corrosion rates and the same were compared with the predictions by Sanchez-Caldera model.

Microstructures and Mechanical Properties of Dissimilar Metal Weld A508/52M/316L Used in Nuclear Power Plants

The microstructures and mechanical properties of a dissimilar metal weld A508/52M/316L used in the primary water system of pressurized water reactor (PWR) nuclear power plants were investigated. The weld exhibits complicated microstructures, with significant change around the interfaces A508/52M and 52M/316L. The variations of main elements in 52M weld metal are greater than those in the A508 and 316L, with significant changes in the zones closed to the interfaces. The bulk 52M weld metal has higher and more uneven hardness than both of the base metals A508 and 316L. The HAZ of A508 exhibits the highest hardness value in the weld. The area around the A508/52M interface is the most weak part for stress corrosion cracking (SCC) resistance of the weld in simulated PWR primary water at 290°C. SCC was only found in the specimens tested at +200mV(SHE) but not in those tested at both -780mV and Ecorr (about -500mV).

Corrosion Test of 304 Stainless Steel in Hydrazine and Methanol Solutions at 320 .DEG.C under Gamma-Irradiation and Gamma Radiolysis of Hydrazine and Methanol

It has been pointed out that dissolved hydrogen is one of the key factors of PWSCC (primary water stress corrosion cracking) in the primary systems of pressurized water reactors. The authors consider that application of alternative reductant for hydrogen may mitigate PWSCC. The corrosion tests of 304 stainless steels in 2 mmol dm−3 methanol solution and 2 mmol dm−3 hydrazine solution at 320 °C were carried out under γ-ray irradiation, and the corrosion environment was evaluated. Electrochemical corrosion potential (ECP) of the stainless steels was measured during the immersion tests. The ECP values were −605 mV and −643 mV vs. SHE in 2 mmol dm−3 hydrazine solution and 2 mmol dm−3 methanol solution at 320 °C, respectively. After the immersion tests, oxide films formed on the stainless steel specimens were analyzed with scanning probe microscope and X-ray diffraction. It is concluded that from the ECP measurement and comparison with previous results corrosion environment under following conditions is similar: (1) DH 1.5 ppm without irradiation, (2) methanol 2.9 ppm without irradiation, (3) hydrazine 2 mmol dm−3 under γ-ray irradiation and (4) methanol 2 mmol dm−3 under γ-ray irradiation. During the immersion tests, the test waters were sampled several times, and analyzed, in order to understand radiolysis of the methanol solution and the hydrazine solution at 320 °C. Hydrazine is decomposed predominantly through thermal decomposition, and an ammonia molecule is formed from a hydrazine molecule. Formaldehyde and ethylene glycol were detected in the methanol solution. But, carboxylates were not detected. The present results do not demonstrate oxidation of methanol to CO or CO2 . It is pointed out that irradiation experiments of higher absorbed dose are necessary.Copyright © 2010 by ASME

Efficiency Upgrade in PWRs

Most of the light water reactor power plants now operating or under construction use pressurized-water reactor (PWR). They are suffering of relatively low thermal efficiency which is around 33%. This would not only have a negative impact economically but also incurs environmental burden in terms of thermal pollution. In this paper, nuclear steam supply system of a typical PWR has been taken into consideration using 1000 MWe Bushehr nuclear power plant (BNPP) data. It is shown thermal efficiency could conceivably be increased by superheating live steam with natural gas up to around 40%, competing with similar fossil-fueled power plants. It is further shown that fuel cost (natural gas) as low as 0.12 Cent/MWe, extra power generated is feasible.

Corrosion of carbon steel feeders during dilute chemical decontamination of primary heat transport system of PHWRs

Carbon steel feeders in the primary heat transport system of pressurized heavy water reactors (PHWRs) show significant wall thinning due to flow accelerated corrosion (FAC). This is of great concern, as the wear rate in certain locations exceeds the corrosion allowance by design. This necessitates periodic measurement of wall thickness and in some cases even mid course enmasse replacement of feeders. While analyzing the data on wall thicknesses and in arriving at the wall thinning rate during operation of the reactor, sufficient care has to be taken to account for the wall thinning occurring during full system chemical decontamination campaign which is carried out occasionally to reduce dose rates during reactor shut down. Chemical decontamination of primary heat transport system is carried out using a mixture of organic acids at a total concentration of about 0.1 g/L and at 85 °C. The results of experiments carried out under simulated conditions for estimating the wall thinning occurring in carbon steel feeder elbow during dilute chemical decontamination are described in this work. The corrosion rates are quantified.

A method for ranking components importance in presence of epistemic uncertainties

Importance Measures (IMs) are used to rank the contributions of components or basic events to the system performance, e.g. its reliability or risk. Most times, IMs are calculated without due account of the uncertainties in the model of the behavior of the system. The objective of this work is to investigate how uncertainties can influence IMs and to develop a method for giving them due account in the corresponding ranking of the components or basic events. The uncertainties considered in this work affect the model parameters values and are assumed to be described by probability density functions. The method for ranking the contributors to the system performance measure is applied to the auxiliary feedwater system of a nuclear pressurized water reactor.

Validation of Analytical Capabilities of RELAP5/MOD3.3 on Boron Dilution during SBLOCA and Loss of Residual Heat Removal System

An analysis has been performed for the OECD/PKL experiments on boron dilution during small break loss-of-coolant accidents (SBLOCA) and loss of residual heat removal system (RHRS). The objective of the analysis is to evaluate and validate the system code capabilities on boron dilution/concentration transients. A focus is set on both the correct prediction of the thermal-hydraulic (T-H) parameters and of the boron dynamics in the pressurized water reactor (PWR) system. The experiments chosen for the analyses are the test E3.1 for loss of RHRS and the tests F1.1 and F1.2 for SBLOCA of the OECD/PKL project. The RELAP5/MOD3.3 (Patch02) was used for the calculations. For the SBLOCA simulating tests F1.1 and F1.2, calculations have shown that the general trends of the system behavior and T-H phenomena, such as water level in SG, coolant temperature, boron concentration, and coolant flow rate, were well simulated. For the loss of RHRS simulating test E3.1, the equilibrium pressure and core outlet temperature generally agreed with the test results. Therefore, the amount of heat removal from the core thorough the water-filled steam generators (SGs) by reflux condenser heat transfer mode was simulated by the code. However, some discrepancies between the analysis and the test were found for the local parameters, such as the water level in the pressurizer, the amount of deboration, and local temperature fluctuation in the loop seal.

Nanofluids for Enhanced Economics and Safety of Nuclear Reactors: An Evaluation of the Potential Features, Issues, and Research Gaps

Nanofluids are engineered colloidal suspensions of nanoparticles in water and exhibit a very significant enhancement (up to 200%) of the boiling critical heat flux (CHF) at modest nanoparticle concentrations ($0.1% by volume). Since CHF is the upper limit of nucleate boiling, such enhancement offers the potential for major performance improvement in many practical applications that use nucleate boiling as their prevalent heat transfer mode. The Massachusetts Institute of Technology is exploring the nuclear applications of nanofluids, specifically the following three:1. main reactor coolant for pressurized water reactors (PWRs)2. coolant for the emergency core cooling system (ECCS) of both PWRs and boiling water reactors3. coolant for in-vessel retention of the molten core during severe accidents in high-power-density light water reactors.The main features and potential issues of these applications are discussed. The first application could enable significant power uprates in current and future PWRs, thus enhancing their economic performance. Specifically, the use of nanofluids with at least 32% higher CHF could enable a 20% power density uprate in current plants without changing the fuel assembly design and without reducing the margin to CHF. The nanoparticles would not alter the neutronic performance of the system significantly. A RELAP5 analysis of the large-break loss-of-coolant accident in PWRs has shown that the use of a nanofluid in the ECCS accumulators and safety injection can increase the peak-cladding-temperature margins (in the nominal-power core) or maintain them in uprated cores if the nanofluid has a higher post-CHF heat transfer rate. The third application can increase the margin to vessel breach by 40% during severe accidents in high-power density systems such as Westinghouse AP1000 and the Korean APR1400. In summary, the use of nanofluids in nuclear systems seems promising; however, several significant gaps are evident, including, most notably, demonstration of the nanofluid thermal-hydraulic performance at prototypical reactor conditions and the compatibility of the nanofluid chemistry with the reactor materials. These gaps must be closed before any of the aforementioned applications can be implemented in a nuclear power plant.

Test interval optimization of safety systems of nuclear power plant using fuzzy-genetic approach

Probabilistic safety assessment (PSA) is the most effective and efficient tool for safety and risk management in nuclear power plants (NPP). PSA studies not only evaluate risk/safety of systems but also their results are very useful in safe, economical and effective design and operation of NPPs. The latter application is popularly known as “Risk-Informed Decision Making”. Evaluation of technical specifications is one such important application of Risk-Informed decision making. Deciding test interval (TI), one of the important technical specifications, with the given resources and risk effectiveness is an optimization problem. Uncertainty is inherently present in the availability parameters such as failure rate and repair time due to the limitation in assessing these parameters precisely. This paper presents a solution to test interval optimization problem with uncertain parameters in the model with fuzzy-genetic approach along with a case of application from a safety system of Indian pressurized heavy water reactor (PHWR).

One-Year Outcomes With Abciximab vs. Placebo During Percutaneous Coronary Intervention After Pre-treatment With Clopidogrel

Light water reactors (LWRs) have been generating electricity for over five decades for the electrical grid of more than 20 countries. Most of these reactors are built using a series of alloys and materials that have changed very little over the many decades. The most common structural materials are based on the three elements, iron (Fe), chromium (Cr) and nickel (Ni) such as stainless steels and nickel-based alloys. Most of the Fe-Cr-Ni alloys contain enough Cr to render them passive with low general corrosion in typical LWR environments. Since the chemistry of the water in the reactor is highly controlled, without aggressive impurities such as chloride or sulfate ions, localized corrosion is not a problem for the common structural alloys. The most common failure mode of the Fe-Cr-Ni alloys is environmentally assisted cracking or stress corrosion cracking (SCC), which is currently well understood and successfully mitigated, mainly by dissolving hydrogen gas into the coolant.The other fundamental material in the reactor core is the zirconium-based alloy used for cladding of the fuel rods. The traditional fuel rod alloys were Zircaloy-2 and Zircaloy-4, which were used regularly in boiling water reactor and pressurized water reactor systems, respectively. Very early in the development of the alloys (early 1950s), it was understood that a small amount (<2%) of alloying elements greatly reduced the general corrosion rate of the zirconium alloys. The traditional corrosion problems of zirconium alloys, such as nodular corrosion, shadow corrosion from the waterside, and SCC from the fuel side, are now well understood and under control. Debris fretting from the coolant side remains the main failure mode of zirconium alloys fuel cladding. Over time the degradation processes of zirconium alloys were successfully identified and managed to allow for their use in nuclear power generation for more than six decades.

The Effect of Surgical Case Volume on Outcome After the Norwood Operation

Light water reactors (LWRs) have been generating electricity for over five decades for the electrical grid of more than 20 countries. Most of these reactors are built using a series of alloys and materials that have changed very little over the many decades. The most common structural materials are based on the three elements, iron (Fe), chromium (Cr) and nickel (Ni) such as stainless steels and nickel-based alloys. Most of the Fe-Cr-Ni alloys contain enough Cr to render them passive with low general corrosion in typical LWR environments. Since the chemistry of the water in the reactor is highly controlled, without aggressive impurities such as chloride or sulfate ions, localized corrosion is not a problem for the common structural alloys. The most common failure mode of the Fe-Cr-Ni alloys is environmentally assisted cracking or stress corrosion cracking (SCC), which is currently well understood and successfully mitigated, mainly by dissolving hydrogen gas into the coolant.The other fundamental material in the reactor core is the zirconium-based alloy used for cladding of the fuel rods. The traditional fuel rod alloys were Zircaloy-2 and Zircaloy-4, which were used regularly in boiling water reactor and pressurized water reactor systems, respectively. Very early in the development of the alloys (early 1950s), it was understood that a small amount (<2%) of alloying elements greatly reduced the general corrosion rate of the zirconium alloys. The traditional corrosion problems of zirconium alloys, such as nodular corrosion, shadow corrosion from the waterside, and SCC from the fuel side, are now well understood and under control. Debris fretting from the coolant side remains the main failure mode of zirconium alloys fuel cladding. Over time the degradation processes of zirconium alloys were successfully identified and managed to allow for their use in nuclear power generation for more than six decades.

Monitoring and fault diagnosis of the steam generator system of a nuclear power plant using data-driven modeling and residual space analysis

An advanced fault detection and isolation (FDI) technique was developed for the steam generator system of a typical pressurized water reactor (PWR) plant. The system characterization module used the group method of data handling method for modeling the interrelationship among the various process variables associated with a U-tube steam generator (UTSG). A principal component analysis algorithm was utilized for the generation of representative fault signatures. The accuracy of fault identification was quantified using normalized vector projections on to fault spaces. Other classical pattern classification methods were executed in parallel to increase the robustness of FDI results. Six typical static faults and one transient fault were successfully detected and isolated using a full-scope simulation of a four-loop PWR. The results demonstrate the implementation of the FDI algorithm for both instrument and actuator monitoring.

2D simulation of the initiation and propagation of crack array under thermal fatigue

Various components of nuclear reactors are submitted to various thermo-mechanical loadings. Thermal fatigue cracking has been clearly detected in reactor heat removal system (RHRS) of pressurized water reactors (PWRs). The present study focuses on AISI 304 L stainless steel used in PWRs. The thermal fatigue behavior of this steel has been investigated using a specific thermal fatigue facility called “SPLASH”. This test equipment allows the reproduction of multiple crack networks similar to those detected during component inspections. The present study deals with the modeling of crack networks initiation and propagation. It is structured in two parts: (i) experimental details and main characteristics of the cracks networks, and (ii) numerical simulation of multiple cracks initiation and growth problem, using an elastic–plastic thermal–mechanical computation and a generalized Paris’ law. The model presented in this study gives predictions in a good agreement with observations, as far as the evolution of the mean and deepest cracks during cycling is concerned.

Analysis of PWR RCS Injection Strategy during Severe Accident

Reactor coolant system (RCS) injection is an important strategy for severe accident management of a pressurized water reactor (PWR) system. Maanshan is a typical Westinghouse PWR nuclear power plant (NPP) with large, dry containment. The severe accident management guideline (SAMG) of Maanshan NPP is developed based on the Westinghouse Owners Group (WOG) SAMG.The purpose of this work is to analyze the RCS injection strategy of PWR system in an overheated core condition. Power is assumed recovered as the vessel water level drops to the bottom of active fuel. The Modular Accident Analysis Program version 4.0.4 (MAAP4) code is chosen as a tool for analysis. A postulated station blackout sequence for Maanshan NPP is cited as a reference case for this analysis. The hot leg creep rupture occurs during the mitigation action with immediate injection after power recovery according to WOG SAMG, which is not desired. This phenomenon is not considered while developing the WOG SAMG. Two other RCS injection methods are analyzed by using MAAP4. The RCS injection strategy is modified in the Maanshan SAMG. These results can be applied for typical PWR NPPs.

Integral Thermal-Hydraulics Tests for the Safety Evaluation of VVER-440/213 Nuclear Reactors and Safety Code Validation

The Paks nuclear power plant is equipped with pressurized water reactors of the VVER-440/213 type. These plants have a number of special features, namely, six-loop primary circuit, horizontal steam generators, loop seal in both hot and cold legs, safety injection tank setpoint pressure higher than secondary pressure, etc. As a consequence of the special design solutions, the transient behavior of such a reactor system is different from the usual pressurized water reactor system behavior. To study the transient behavior of these plants, the PMK-2 integral-type facility, a thermal-hydraulic model of the Paks nuclear power plant, was designed and constructed.A short description of the specific design solutions of the VVER-440/213-type plants is given with the modeling aspects and similarity criteria applied to the design of the PMK-2 facility. Since the startup of the facility in 1985, 48 experiments have been performed primarily in an international framework with the participation of several experts from European and overseas countries to study one- and two-phase natural circulation, loss-of-coolant accidents, special plant transients, and experiments in support of the accident management measures. The results of several experiments illustrate the system effects of special design solutions and the effectiveness of bleed-and-feed accident management measures. A brief commentary on the thermal-hydraulic system code validation is provided, and conclusions are offered.

Effects analysis fuzzy inference system in nuclear problems using approximate reasoning

In this paper a fuzzy inference system modeling technique applied on failure mode and effects analysis (FMEA) is introduced in reactor nuclear problems. This method uses the concept of a pure fuzzy logic system to treat the traditional FMEA parameters: probabilities of occurrence, severity and detection. The auxiliary feed-water system of a typical two-loop pressurized water reactor (PWR) was used as practical example in this analysis. The kernel result is the conceptual confrontation among the traditional risk priority number (RPN) and the fuzzy risk priority number (FRPN) obtained from experts opinion. The set of results demonstrated the great potential of the inference system and advantage of the gray approach in this class of problems.