Search

A comparative computational fluid dynamics (CFD) study was conducted on the three different types of pressurized water reactor (PWR) upper plenum, named TYPE 1 (support columns (SCs) and control rod guide tubes (CRGTs) with two large windows), TYPE 2 (SCs and CRGTs without windows), and TYPE 3 (two parallel perforated barrel shells and CRGTs). First, three types of upper plenum geometry information were collected, simplified, and adopted into the BORA facility, which is a 1/5 scale system of the four-loop PWR reactor. Then, the geometry, including the upper half core, upper plenum region, and hot legs, was built using the Salome platform. After that, an unsteady calculation to simulate the reactor balance operation at hot full power scenario was performed. Finally, the differences of flowrate distribution at the core outlet and temperature distribution and transverse velocity inside the hot legs with different upper plenum internals were compared. The results suggest that TYPE 1 upper plenum internals cause the largest flowrate difference at the core outlet while TYPE 3 leads to the most even distributed flowrate. The distribution and evolution pattern of the tangential velocity inside hot legs is highly dependent on the upper plenum internals. Two counter-rotating swirls exist inside the TYPE 1 hot leg and only one swirl revolving around the hog leg axis exist inside the TYPE 2 hot leg. For TYPE 3, two swirls like that of TYPE 1 rotating around the hot leg axis significantly increase the temperature homogenization speed. This research provides meaningful guidelines for the future optimization and design of advanced PWR upper plenum internal structures.

The small PWR has been paid more and more attention due to its diversity of application and flexibility in the site selection. However, the large core power density, the small containment space and the rapid accident progress characteristics make it difficult to control the containment pressure like the traditional PWR during the LOCA. The pressure suppression system has been used by the BWR since the early design, which is a suitable technique that can be applied to the small PWR. Since the configuration and operating conditions are different from the BWR, the pressure suppression system should be redesigned for the small PWR. Conducting the experiments on the scale down test facility is a good choice to reproduce the prototypical phenomena in the test facility, which is both economical and reasonable. A systematic scaling method referring to the H2TS method was proposed to determine the geometrical and thermohydraulic parameters of the pressure suppression containment response test facility for the small PWR conceptual design. The containment and the pressure suppression system related thermohydraulic phenomena were analyzed with top-down and bottom-up scaling methods. A set of the scaling criteria were obtained, through which the main parameters of the test facility can be determined.

For the structural integrity evaluation of pressurized water reactor (PWR) steam generator (SG) tubes subjected to transient hydraulic loading, determination of the tube-to-tube gap velocity and static pressure distributions along the tubes is prerequisite. This paper addresses both computational fluid dynamics (CFD) and analytical approaches for predicting the tube-to-tube gap velocity and static pressure distributions during blowdown following a feedwater line break (FWLB) accident at a PWR SG. First of all, a comparative study on CFD calculations of the transient velocity and pressure distributions in the SG secondary sides for two different models having 30 or no tubes is performed. The result shows that the velocities of sub-cooled water flowing between any adjacent two tubes of a tubed SG model during blowdown can be roughly estimated by applying the specified SG secondary side porosity to those of the no-tubed SG model. Secondly, simplified analytical approximate solutions for the steady two-dimensional SG secondary flow velocity and pressure distributions under a given discharge flowrate are derived using a line sink model. The simplified analytical solutions are validated by comparing them to the CFD calculations.

The BEAVRS benchmark problem describes detailed information such as the geometry and material specifications to construct a neutronic calculation model of a commercial Pressurized Water Reactor (PWR) core. The benchmark was modified several times; the main purpose of the current study is to investigate the effect of the modifications in the BEAVRS benchmark on sensitivities and uncertainties in cross section that in turn affect the multiplication factors and reactivity coefficients. The sensitivity coefficients are calculated for keff due to perturbation in cross-sections using the KSEN card of MCNP6 code for the two cores, where all control rods are in and out. Since MCNP6 cannot calculate uncertainties, the NJOY2021 was used to compute the relative covariance matrices of the cross-sections data library, thereafter, a python script was developed to calculate the uncertainties of keff as well as sensitivities and uncertainties of reactivity due to the control rods insertion. The results indicate that the uncertainties in reactivities caused by control rods insertion for the BV2 model is greater than the BV1 model, due to the effect of using higher coolant density in the nozzle and support plate structures in addition to the effect of the burnable absorber designs in the BV2 model.

Traditional light water reactor simulations are usually either high fidelity, requiring hundreds of node-hours, or low fidelity, requiring only seconds to run on a common workstation. In current re...

A Large-Break Loss of Coolant Accident (LB-LOCA) in a generic, 1000 MWe and 3-loop Chinese Pressurized Water Reactor (PWR) with the In-Vessel corium Retention (IVR) strategy was simulated by using ASTEC V2.1.1. One of the criteria of assessing the validity of IVR is that the decay heat of molten pool could be removed successfully, which means that the local heat flux will not exceed the Critical Heat Flux (CHF) of outer nucleate boiling. ASTEC determines the process of core degradation, the corium relocation and the transient heat flux to indicate whether the vessel can maintain its integrity under the LB-LOCA. From the core degradation process, for this LB-LOCA scenario, the corium with fission products slumps into the lower plenum via the downcomer, which means the sideward relocation. After the initial relocation, the mass of compositions of molten pool in lower head are almost steady except the mass of steel. The augmentation of the steel in the lower head is consistent with the melting process of the lower support plate, the fluid distribution structures and the vessel inner wall. At the azimuth from 78° to 87°, the transient heat flux is much larger compared with the other locations. The transient heat flux reaches its maximum value which is about 1.1 MW/m2 at the azimuth of 84°. And the maximal ratio of the transient local heat flux to CHF is about 0.8. For the residual thickness of the lower head, the minimum is 37.35 mm at the azimuth from 78° to 87°. The results of thermal analysis conclude that the vessel can maintain its integrity during LB-LOCA scenario because the transient heat flux from the vessel wall to the external coolant do not exceed the CHF of outer nucleate boiling.

In this paper, we validate the decay heat calculation capability via a two-step method to analyze spent nuclear fuel (SNF) discharged from pressurized water reactors (PWRs). The calculation method is implemented with a lattice code STREAM and a nodal diffusion code RAST-K. One of the features of this method is the direct consideration of three-dimensional (3D) core simulation conditions with the advantage of a short simulation time. Other features include the prediction of the isotope inventory by Lagrange non-linear interpolation and the use of power history correction factors. The validation is performed with 58 decay heat measurements of 48 fuel assemblies (FAs) discharged from five PWRs operated in Sweden and the United States. These realistic benchmarks cover the discharge burnup range up to 51 GWd/MTU, 23.2 years of cooling time, and spanning an initial uranium enrichment range of 2.100–4.005 wt percent. The SNF analysis capability of STREAM is also employed in the code-to-code comparison. Compared to the measurements, the validation results of the FA calculation with RAST-K are within ± 4%, and the pin-wise results are within ± 4.3%. This paper successfully demonstrates that the developed decay heat calculation method can perform SNF back-end cycle analyses.

The study is aimed at performing calculated substantiation of the thermal reliability of a pressurized water reactor (PWR) with spherical fuel elements, commonly known as a fixed-bed nuclear reactor, and determining the most advantageous design of the spherical packing from the view point of thermal hydraulics. Spherical fuel elements have a number of advantages over cylindrical fuel rods; in particular, they feature better retention of fission products, enhanced nuclear safety (due to a high melting temperature of ceramic materials), more intense heat transfer due to increased coolant flow turbulence, and a reduced influence of thermal cyclic loads on fuel elements. To confirm these statements, a thermohydraulic calculation of the KLT-40S type reactor with a modified intra-channel filling of fuel assemblies (FAs) consisting of spherical fuel elements was carried out. To determine the optimal spherical filling, two types of spherical packing were calculated, with three different diameters of spherical fuel elements. In the course of the calculation, solutions to the following issues were proposed: how to take into account the channels flow area variability, how to calculate the Reynolds number for channels of a given shape, what formulas should be used to determine the Nusselt number, and how to determine the hydraulic resistance in the channels. As a result of the calculation, data on the following main thermohydraulic characteristics have been obtained: surface heat flux density, heat transfer coefficient, maximum fuel temperature, and hydraulic losses. These results were compared with the results of calculations for cylindrical fuel rods. The obtained results demonstrate the advantage of spherical fuel elements over cylindrical fuel rods in a number of basic parameters, which gives prospects for further study of the use of spherical fuel elements in reactors of this type. The obtained study results can be applied in designing reactor plants of low and medium capacity, as well as in modernizing the existing reactor plants.

This paper presents a numerical prediction of the transient hydraulic loads acting on the tubes and external supports of a pressurized water reactor (PWR) steam generator (SG) during blowdown following a sudden feedwater line break (FWLB). A simplified SG model was used to easily demonstrate the prediction. The blowdown discharge flow was treated as a flashing flow to realistically simulate the transient flow fields inside the SG and the connected broken feedwater pipe. The effects of the SG initial pressure or the broken feedwater pipe length on the intensities or magnitudes of transient hydraulic loads were investigated. Then predictions of the decompression pressure wave-induced impulsive pressure differential loads on SG tubes and the transient blowdown loads on SG external supports were demonstrated and the general aspects of transient responses of such transient hydraulic loads to the FWLB were discussed.

The article focused on the development of information technology for the optimization of control over complex dynamic systems at the stage of their design that should realize possibilities of modeling of linear and nonlinear dynamic systems, the analysis and synthesis of such systems, their optimization on various quality criteria. The purpose of this article is to develop the structure and elements of information technology to optimize the control of complex dynamic systems, including automated control systems. The general structure and functional model of information technology of the analysis and synthesis of control systems includes modeling of dynamic systems in the form of systems of differential equations and transfer functions, integration of systems of differential equations, calculation of system quality criteria, methods for solving various optimization tasks on solutions of optimization problems of dynamic systems is offered.

Grain boundary (GB) oxidation of proton-irradiated 304 nuclear grade stainless steel in primary water of pressurized water reactor was investigated. The investigation was conducted by studying microstructure of the oxide and oxide precursor formed at GB on an "atom-by-atom" basis by a combination of atom-probe tomography and transmission electron microscope. The results revealed that increasing irradiation dose promoted the GB oxidation, in correspondence with a different oxide and oxide precursor formed at the GB. Correlation of the oxide and oxide precursor with the GB oxidation behavior has been discussed in detail.

The analysis of the fluid flow characteristics in reactor pressure vessel is an important part of the hydraulic design of nuclear power plant, which is related to the structure design of reactor internals, the flow distribution at core inlet and the safety of nuclear power plant. The flow distribution and mixing characteristics in the pressurized reactor vessel for the 1000MWe advanced pressurized water reactor is analyzed by using Computational Fluid Dynamics (CFD) method in this study. The geometry model of the full-scaled reactor vessel is built, which includes the cold and hot legs, downcomer, lower plenum, core, upper plenum, top plenum, and is verified with some parameters in DCD. Under normal condition, it is found that the flow skirt, core plate holes and outlet pipe cause pressure loss. The maximum and minimum flow coefficient is 1.028 and 0.961 respectively, and the standard deviation is 0.019. Compared with other reactor type, it shows relatively uniform of the flow distribution at the core inlet. The coolant mixing coefficient is investigated with adding additional variables, showing that mass transfer of coolant occurs near the interface. The coolant mainly distributes in the 90° area of the corresponding core inlet, and mixes at the interface with the coolant from the adjacent cold leg. 0.1% of corresponding coolant is still distributed at the inlet of the outer-ring components, indicating wide range of mixing coefficient distribution.

In nuclear power plants the design extension conditions are more complex and severe than those postulated as design basis accidents, therefore, they must be taken into account in the safety analyses. In this study, many hypothetical investigated transients are applied on KONVOI pressurized water reactor during a 6 inches (182 cm2) cold leg small break loss-of-coolant-accident to revise the effects of all safety systems ways through their availability/non-availability on the thermal hydraulic behaviour of the reactor. The investigated transients are represented through three cases of small break loss-of-coolant-accident as, Case-1, without scram and all of the safety systems are a failure, Case-2, the normal scram actuation with a failure of all safety systems (non-availability), and finally Case-3, with normal actuation scram sequence and normal sequential actuation of all safety systems (availability). These three investigated transient cases are simulated by creating a model using analysis of thermal-hydraulics of leaks and transient code. In all transient cases, all types of reactivity feed- backs, boron, moderator density, moderator temperature, and fuel temperature are considered. The steady-state results are nearly in agreement with the plant parameters available in previous literature. The results show the importance of the reactivity feedback effects in loss-of-coolant-accident on the fallouts power as they are considered the key parameters for controlling the clad and fuel temperatures to maintain them below their melting point. Moreover, the calculated results in all cases show that the thermal-hydraulic parameters are in acceptable ranges and encounter the safety criterion during loss-of-coolant-accident design extension conditions accidents processes. Furthermore, the results show that the core uncover and fuel heat up do not occur in KONVOI pressurized water reactor the design extension conditions simulations, as all safety systems provide adequate core cooling by sufficient water inventory into the core to cover it.

Many experiments have been conducted on accidents and transients of pressurized water reactor (PWR) employing the rig of safety assessment/large-scale test facility (ROSA/LSTF). Recent research activities concerned with the OECD/NEA international joint projects included experimental investigation via the ROSA and ROSA-2 Projects, and counterpart testing with thermal-hydraulic integral test facilities under collaboration of the PKL-2, PKL-3, ATLAS, and ATLAS-2 Projects. Major results of the related integral effect tests (IETs) with the LSTF were reviewed to experimentally identify thermal-hydraulic phenomena involved, regarding the PWR accident sequences in accordance with the new regulatory requirements for the Japanese light-water nuclear power plants. Future separate effect test using the LSTF is planned to simulate loss of emergency core cooling system (ECCS) recirculation functions in a large-break loss-of-coolant accident (LOCA). Key results of the recent IETs utilizing the LSTF and future plans were presented relevant to multiple steam generator tube rupture accident with recovery operation, small-break LOCA with accident management measure on core exit temperature reliability, and small-break LOCA with thermal stratification under cold water injection from ECCS into cold legs. Also, main outcomes of the LSTF IETs were indicated for wide spectrum LOCA with core uncovery and anticipated transient without scram following small-break LOCA under totally failed high-pressure injection system.

In case of a postulated loss-of-coolant accident (LOCA) in a light water reactor, core degradation might occur if emergency measures fail. During this process a large amount of hydrogen might be generated by the oxidation of zirconium in the fuel rod cladding with steam. If the hydrogen leaks into the containment of a pressurized water reactor or the reactor building of a boiling water reactor and if the design concept of passive autocatalytic recombining is not adequate or fails due to adverse atmospheric conditions, it might form a combustible mixture with air. In case of ignition, pressure peaks can occur that are relevant for the integrity of the containment or the reactor building and safety systems. In the past 20 years 45 hydrogen deflagration tests (HD tests) have been performed in the THAI/THAI+ test facility with a free volume of 60–80 m3 which is operated by Becker Technologies GmbH in Germany. The lumped parameter code COCOSYS V3.0 (Containment Code System) which is part of the software package AC2 2019 developed by Gesellschaft fur Anlagen-und Reaktorsicherheit (GRS) gGmbH offers the models FRONT and COMB for the simulation of hydrogen deflagrations in the containment or reactor building of a nuclear power plant (NPP). COCOSYS simulations of select THAI+ HD tests are shown, and model development potential is deduced.

Reliable prediction of nuclear fuel rod behavior is of prime concern for assessing proper design parameters as well as operating conditions and sustainability aspects in a nuclear reactor. In this study, a numerical analysis has been conducted to investigate the thermomechanical behavior of solid and annular fuel rods individually, considering a pressurized water reactor (PWR). The transient heat conduction equation has been employed for the heat-generating uranium dioxide pellet along with the surrounding zircaloy cladding and solved numerically using the finite element method. Loss of coolant conditions has been simulated by reducing the coolant heat transfer coefficient and compared with the normal operating conditions. In all the cases considered, the annular fuel has shown superior properties in terms of minimized fuel surface heat flux, lower but relatively uniform radial temperature, and thermal strain distribution.

In nuclear power plants, steam generators are important parts. This paper introduces a U-tube steam generator (UTSG) model implemented completely using MATLAB environment. The UTSG consists of four regions: upward and downward primary regions, upward and downward metal tube regions, and secondary regions, which contain heat transfer region, steam separation region, and subcooled water region. Governing equations are derived by applying energy and mass conservation equations in all regions. Accurate functions that describe the relationships between thermodynamic properties of the saturated steam are introduced instead of interpolation method that is widely used. Steady state and one transient case are presented as well.

Qinshan nuclear power plant is the first large Chinese-designed nuclear power station based on pressurized water reactor, and the second generation main stream active safety injection system is adopted for Qinshan nuclear power plant. In this paper, a novel passive safety injection system (PSIS) has been proposed for ocean-based Qinshan Phase One nuclear power plant to replace the original active one. The PSIS contains high-pressure, medium-pressure, and lower-pressure safety injection systems, and a two-stage automatic depressurization system. To evaluate the system performance, small-break LOCA has been investigated using RELAP5. Various break sizes and locations including 2-inch, 10-inch cold leg break, and double-ended direct vessel injection line break were analyzed. Key safety parameters such as safe injection mass flow rates, coolant level of the core, system pressure, and fuel cladding temperature were monitored during the accident process. The results illustrate that the performance of the safety injection system can guarantee the effective core cooling and submerged under different LOCA even with only half of the safety injection system, which can fulfill the single failure criteria. The thermal-hydraulic analysis for the Qinshan passive safety injection system is significant to master the related technologies for Chinese engineer and develop the Chinese-designed third-generation nuclear power plants, and the PSIS can guarantee the reactor submerged under LOCA even plus the station block out accident.

Reactors are widely used in the nuclear power plant due to the rapid demand for electricity by reducing the greenhouse effect. However, the effectiveness of the nuclear reactor depends on an adequate safety system. Hence, temperature and heat transfer are two critical parameters for any reactor in operation for which intelligent temperature control with an integrated safety system is essential. Therefore, the present study has emphasized the development of a simulated small-scale water-based reactor with intelligent control and safety system and examined through the analysis of thermal-hydraulic parameters. Radial heat transfer of an electric rod used as fuel in the primary circuit has been analyzed by taking sensor reading in various positions of the core. The developed system is self-controlled with all possible active and passive safety systems. Consecutively, the prototype has also been designed including manual adjustment to ensure a fail-safe environment. The system is capable to operate at temperatures between 80°C and 120°C, although the design can withstand up to 200°C. The data of the experiment are taken under the pressure of 200 kPa at 120°C temperature. Results show that heat output of 2116.09 kJ has been obtained from the system against heat input of 2514.80 kJ, which gives an efficiency around 16% of the developed system.

A small leak occurring on the surface of a fuel rod due to damage exposes UO2 to a steam atmosphere. During this time, fission gas trapped inside the fuel rod leaks out, and the gas leakage can be increased due to UO2 oxidation. Numerous studies have focused on the steam oxidation and its thermodynamic calculation in UO2. However, the thermodynamic calculation of the UO2 oxidation in a pressurized water reactor (PWR) environment has not been studied extensively. Moreover, the kinetics of the oxidation of UO2 pellet also has not been investigated. Therefore, in this study, the thermodynamics of UO2 oxidation under steam injection due to a damaged fuel rod in a PWR environment is studied. In addition, the diminishing radius of the UO2 pellet with time in the PWR environment was calculated through an experiment simulating the initial time of steam injection at the puncture.

Recurring step straining was developed to investigate the environmentally assisted cracking (EAC) initiation of a Ni-30Cr-10Fe alloy (Alloy 690) in simulated pressurized water reactor (PWR) primary water. Thanks to this technique, the roles of corrosion and straining in the EAC initiation of Alloy 690 was decoupled and typical microstructures in stress corrosion cracking, i.e. diffusion induced grain boundary migration and oxidation into the migration zone, were reproduced. Digital image correlation using indigenous surface oxide particles as markers was successfully incorporated into this technique for high-resolution in-plane strain measurement. The preliminary results show that intergranular crack initiation is correlated with the high local strain in both adjacent grains, indicating that slip transfer across the grain boundary promotes crack initiation. Recurring step straining is desirable for the mechanistic study of EAC as it makes the separate roles of chemical and mechanical processes adjustable and traceable.

Due to excess reactivity, fresh nuclear fuel often contains burnable poisons. This research looks at six different burnable poisons and their impacts on reactivity, material attractiveness, and waste management. An MCNP simulation of a PWR fuel pin was performed with a fuel burnup of 60 GWd/MTHM to determine when each burnable poison fuel type would decrease below a k∞ of 1. For determining the plutonium material attractiveness in each burnable poison fuel type, the plutonium isotopic content of the used fuel was evaluated using Bathke's Figure of Merit formula. For the waste management analysis, the thermal output of each burnable poison fuel type was determined through ORIGEN decay simulations at 100 and 300 years after being discharged from the core. The performance of all six burnable poisons varied over the three criteria considered and no single burnable poison performed best in all three considerations.

The phenomenon identification and ranking table (PIRT) is an important basis in the nuclear power plant (NPP) thermal-hydraulic analysis. This study focuses on the importance ranking of the input parameters when lacking the PIRT, and the target scenario is the small break loss of coolant accident (SBLOCA) in a pressurized water reactor (PWR) CPR1000. A total of 54 input parameters which might have influence on the figure of merit (FOM) were identified, and the sensitivity measure of each input on the FOM was calculated through an optimized moment-independent global sensitivity analysis method. The importance ranking orders of the parameters were transformed into the Savage scores, and the parameters were categorized based on the Savage scores. A parameter importance ranking table for the SBLOCA scenario of the CPR1000 reactor was obtained, and the influences of some important parameters at different break sizes and different accident stages were analyzed.

Based on the current high-level radioactive waste management basic plan and the analysis results of spent nuclear fuel characteristics, such as dimensions and decay heat, an improved geological disposal concept for spent nuclear fuel from domestic nuclear power plants was proposed in this study. To this end, disposal container concepts for spent nuclear fuel from two types of reactors, pressurized water reactor (PWR) and Canada deuterium uranium (CANDU), considering the dimensions and interim storage method, were derived. In addition, considering the cooling time of the spent nuclear fuel at the time of disposal, according to the current basic plan-based scenarios, the amount of decay heat capacity for a disposal container was determined. Furthermore, improved disposal concepts for each disposal container were proposed, and analyses were conducted to determine whether the design requirements for the temperature limit were satisfied. Then, the disposal efficiencies of these disposal concepts were compared with those of the existing disposal concepts. The results indicated that the disposal area was reduced by approximately 20%, and the disposal density was increased by more than 20%.

The Korea Atomic Energy Research Institute (KAERI) has developed geological repository systems for the disposal of high-level wastes and spent nuclear fuels (SNFs) in South Korea. The purpose of the most recently developed system, the improved KAERI Reference Disposal System Plus (KRS+), is to dispose of all SNFs in Korea with improved disposal area efficiency. In this paper, a system-level safety assessment model for the KRS+ is presented with long-term assessment results. A system-level model is used to evaluate the overall performance of the disposal system rather than simulating a single component. Because a repository site in Korea has yet to be selected, a conceptual model is used to describe the proposed disposal system. Some uncertain parameters are incorporated into the model for the future site selection process. These parameters include options for a fractured pathway in a geosphere, parameters for radionuclide migration, and repository design dimensions. Two types of SNF, PULS7 from a pressurized water reactor and Canada Deuterium Uranium from a heavy water reactor, were selected as a reference inventory considering the future cumulative stock of SNFs in Korea. The highest peak radiological dose to a representative public was estimated to be 8.19×10-4 mSv‧yr-1, primarily from 129I. The proposed KRS+ design is expected to have a high safety margin that is on the order of two times lower than the dose limit criterion of 0.1 mSv‧yr-1.

This paper proposes the application of a pattern recognition–based technique to enhance the process of control rod position identification in pressurized water reactors (PWRs). The proposed techniq...

The crevice stress corrosion cracking (SCC) susceptibility of austenitic stainless steels was evaluated in simulated pressurized water reactor (PWR) environments. To simulate the abnormal condition in temporary clamping devices on leaking small bore pipes, crevice bent beam (CBB) tests were performed in the oxygenated as well as hydrogenated conditions. No SCC cracks were found for SS316 in both conditions. SS304 also showed good resistance in the hydrogenated condition. However, all SS304 specimens showed SCC cracks in the oxygenated condition, indicating poor crevice SCC resistance. It was found that residual ferrites were selectively dissolved because of the galvanic corrosion coupled with the neighbouring austenite phase, resulting in SCC initiation in SS304. Crack morphologies were mostly transgranular assisted by the damaged δ-ferrite and deformation-induced slip bands.

We have developed a practical-scale dry disassembling process to dismantle PWR (Pressurized Water Reactor) spent nuclear fuel assembly in the order of several tens of kilograms of heavy metal/batch to supply rod-cuts (cladding tube and UO2 pellets) for mechanical decladding process. Dry head-end disassembling process has advantages over the wet head-end process because of the lower risk of proliferation and treatment of spent fuel with relatively high heat and radioactivity. This study describes the main design considerations for the disassembling process of the spent nuclear fuel assembly during the dry head-end process. The down-ender, dismantling, extraction, and cutting technologies are analyzed and models have been designed for testing. The purpose of dry head-end disassembly process is to test the main device performance and to obtain scale-up data for practical-scale disassembling. With this in mind, design considerations were analyzed based on remoteness, and basic verification tests were performed. However, the authors used simulated fuel, instead of the actual spent fuel, owing to a lack of joint determination. In addition, in the present study, we did not consider the heat generated from minor actinides or the radioactivity of the fission product; these aspects will be considered in a future study. During the basic test performed in this study, a simulated assembly was completely disassembled using new methods, such as dismantling, extraction, and cutting processes. The practical-scale dry disassembling technology can be tested using scale-up data for reuse of the spent fuel.

The water level control system in the pressurizer is carried out for safety purposes when the PWR reactor is operating. Control is done by maintaining the water level in the pressurizer at a certain level. The height of the water level in the pressurizer will change according to the conditions of the nuclear reactor. This research aimed to obtain the best safety control system using the fractional order PID (FOPID) control system and prove that using the control system will provide better results than other existing system applications. In this research, the optimum gain control of FOPID is obtained with K p equals 416.4, K i equals 416.4, K d equals 104.1, constitutes integral equals 0.15, and derivative order equals 1.9, which results in 324 times faster settling time and maximum overshoot 1.5 times lower when compared to conventional PID controllers combined with CMAC. DOI: http://dx.doi.org/10.17977/um024v5i22020p071

This paper is focused on overcoming reactor shutdown malfunction and diagnostics due to poor cooling and water drop level within a pressurized water reactor. So the temperature coupling analysis be...

In this work, the depletion capability implemented in Monte Carlo code MCS is investigated to predict the isotopic compositions of spent nuclear fuel (SNF). By comparison of MCS calculation results to post irradiation examination (PIE) data obtained from one pressurized water reactor (PWR), the validation of this capability is conducted. The depletion analysis is performed with the ENDF/B-VII.1 library and a fuel assembly model. The transmutation equation is solved by the Chebyshev Rational Approximation Method (CRAM) with a depletion chain of 3820 isotopes. 18 actinides and 19 fission products are analyzed in 14 SNF samples. The effect of statistical uncertainties on the calculated number densities is discussed. On average, most of the actinides and fission products analyzed are predicted within ± 6 % of the experiment. MCS depletion results are also compared to other depletion codes based on publicly reported information in literature. The code-to-code analysis shows comparable accuracy. Overall, it is demonstrated that the depletion capability in MCS can be reliably applied in the prediction of SNF isotopic inventory.

Corrosion behavior and mechanism of irradiated 304 nuclear grade stainless steel were studied in simulated pressurized water reactor primary water. The microstructure of the oxide formed on the steel irradiated to different doses over an exposure period range of 25–1500 h was analyzed and compared. It was found that the general and intergranular corrosion rates of the steel were increased with irradiation dose, in correspondence with an evolution of the general oxide and the oxide formed at the grain boundary. Correlation of the oxide evolution with the corrosion kinetics and mechanism has been discussed in detail.

The structural integrity of welded joints in the reactor pressure vessel (RPV) is directly related to the safety of nuclear power plants. The RPV is made from SA508-III steel in a pressurized water reactor. In this study, we investigated the effects of temperature on the tensile and fracture toughness properties of Chinese SA508-III welded joint in different sampling areas in order to provide reference data for structural integrity assessments of RPVs. The specimens used in tensile and fracture toughness tests were fabricated from the base metal (BM), weld metal (WM), and the heat-affected zone (HAZ) in the welded joint. The representative testing temperatures included the ambient temperature (20 °C), upper shelf temperature (100 °C), and service temperature (320 °C). The results showed that temperature greatly affected the fracture toughness (JIC) values for the SA508-III welded joint. The JIC values for BM and HAZ both decreased remarkably from 20 °C to 320 °C. The fracture morphologies showed that the BM and HAZ in the welded joint exhibited fully ductile fracture at 20 °C, whereas partial cleavage fracture was mixed in ductile fracture mode at 100 °C and 320 °C. The WM exhibited the ductile and cleavage fracture mixed mode at various temperatures, and the JIC values showed slight changes.

Nuclear power plant, nuclear industry and aerospace industry require a high level of safety so it is necessary to make risk assessment for these systems, subsystems and components. So, it is needed to take into consideration the components failure rate and its probability, and have qualitative and quantitative evaluation using fault tree analysis. The conducted study is aiming to make qualitative and quantitative evaluation for Auxiliary Feed Water system (AFWS) which is designed to supply feed water to the steam generators for Reactor Coolant System (RCS) heat removal in a case of loss of main feed water or loss of all AC (alternating current) power to the station auxiliary, (AFWS) consisted of two trains to single steam generator which achieve 100% capacity for each train, Complete redundancy and diversity achieved. Results show the qualitative evaluation of auxiliary feed water system, this qualitative assessment is supported by failure rate data of components to have quantitative evaluation, this study indicated the unavailability of the individual components and the total unavailability of the (AFWS) for 8760 hour. To be sure that the system is available during operation periods, test and repair periods must be obtain at its exact scheduled time.

The pressurizer of a pressurized water reactor (PWR), as a spray-heating degasser, has been widely used to remove dissolved gas in the primary coolant of PWRs. In the real degassing process, the bo...