Pressurized Water Reactor Plant Research Articles

Automated analysis of Baffle Bolts

The Baffle bolts (BB) are components of the lower internals of the vessel of Pressurized Water Reactor (PWR) nuclear power plants. They are subjected to ultrasonic testing (UT) inspection as part of the Long-Term Operation (LTO) of the PWR plants worldwide. Unlike other areas undergo inspection, usually there are more than one thousand bolts in a typical PWR reactor. The goal of the paper is to present how the Artificial Intelligence (AI) can be integrated into the inspection of the baffle bolts described in the introduction of the present document. The scope of the project was limited to the analysis of the signals acquired during the inspection to discriminate between healthy BB and those showing signals that need to be evaluated by an expert. Some different independent algorithms are used by the automated analysis system, which counts with a support decision system to achieve the most robust result possible. First inspection of its kind was carried out in winter 2021 with promising results, being scheduled the next one in Fall 2022. The expectation is to get the system qualified according to the criteria stablished by the normative. AI can be applied to UT inspections being the major limitation the amount of data, in a dual sense: train the data needed for the algorithms and obtain a confident improvement of the productivity. A key point of the project has been to create working teams of professionals that bring together knowledge in UT and AI.

Comparison of grain structure models for wave propagation analysis in centrifugally cast stainless steel

Centrifugally cast stainless steel (CASS) is widely used in primary coolant piping of pressurized water reactor plants because of its high corrosion resistance and high strength. An in-service inspection based on ultrasonic testing (UT) has to be conducted for weld joints of primary coolant piping on the basis of JSME Rules on Fitness-forService for Nuclear Power Plants. However, it is difficult to detect and size flaws in CASS components with high accuracy because of the following reasons: Ultrasonic waves are scattered and attenuated due to coarse grains, and anisotropic and heterogeneous properties in CASS lead to ultrasonic beam skewing. Numerical simulations are useful and reasonable ways for better understanding the ultrasonic wave propagation behavior in CASS. To effectively achieve this, the simulation model should include a three-dimensional (3D) grain structure. In this study, we modeled three kinds of the solidification grain structures of centrifugally CASS. One is obtained by using a cellular automaton method, another consists of many hexagonal columns with the same dimensions, and the other is transversely isotropic material. Then these structures were fed into an explicit finite element model for simulating wave propagation and the simulated results were compared with those measured by a laser Doppler vibrometer. Through the comparison, we investigated the applicability of these three kinds of solidification grain structure models to simulation for wave propagation.

A Pragmatic Approach to Update Support Studies and Human Reliability Analyses for a PSA Model

A probabilistic safety assessment (PSA) is used to estimate potential risks associated with a nuclear power plant. Event trees (ETs) and fault trees (FTs) describe accident scenarios following initiating events. They allow for identifying and quantifying the consequence frequencies. Based on success criteria, support studies are used to determine each scenario’s consequence. They are also used to define the time available for the operators to carry out the actions involved in the scenario, and consequently, the corresponding human failure probability. In this paper, we use a dynamic PSA toolbox to optimize support studies for a classical (ET/FT) PSA model. We analyze medium-break loss-of-coolant accident (LOCA) initiating events with the possible failure of the high-pressure safety injection (HPSI) system for an internal event Level 1 PSA for a pressurized water reactor plant. In order to prevent core uncovery and subsequent core damage, operators must depressurize the primary circuit to reach the low-pressure safety injection set point. This scenario represents a significant contribution to the core damage frequency. The dominant cut sets corresponding to this scenario involve HPSI failure to run over 24 h, while the thermal-hydraulic analysis supporting the human reliability assessment (HRA) analysis for this sequence assumes the unavailability of HPSI pumps at the beginning of the accident. We update the thermal-hydraulic assumptions in the support study by performing simulations for different values of the HPSI failure time. We find that even a short HPSI operation time buys significant (from the HRA point of view) available time for operators and drives significant improvement in estimating the human error probability (HEP). We postprocessed the minimal cut-set probability by integrating the obtained HEP. This result allows for more realistic quantification of the contribution of the medium-break LOCA in the core damage frequency.

Analyses of Critical Hydrogen Enrichment in PWR Containment Compartments

During a loss-of-coolant accident in a pressurized water reactor (PWR), steam of varying quality is released from the primary circuit into the equipment compartments of the containment, followed by the release of a hydrogen-steam mixture during the core degradation phase. In the case of long-lasting accidents, findings of detailed code analyses indicate an enrichment of hydrogen in lower peripheral containment compartments in the reference PWR plant under investigation. During the late accident phase with ex-vessel molten core–concrete interaction, even in the case of an operating passive autocatalytic recombiner system, this poses a threat for local hydrogen combustion later on. Such hydrogen phenomena are not expected and have not been widely studied up to now. Therefore, corresponding experiments have been performed at the THAI test facility operated by Becker Technologies. One of these tests had been precalculated with the COntainment COde SYStem (COCOSYS) as part of the Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) code system AC2 and has been used to validate the code. The 60-m3 THAI test vessel has been divided into an inner compartment that has been connected to the surrounding vessel, simulating the upper and peripheral containment part, by very small flow openings at the bottom representing the clearance between door frames and door leaves and one opening at the top representing typical openings by burst disks. The paper discusses both the experimental findings of a test series on the potential enrichment of hydrogen in lower containment compartments and the COCOSYS calculations demonstrating the applicability of the code under complex flow conditions including stratification phenomena.

Simulated false data injections attacks on emulated and hardware programmable logic controllers of the pressurizer in a representative pressurized water reactor plant

ABSTRACT Manipulating sensors data and/or Programmable Logic Controllers (PLCs) in Instrumentation and Control (I&C) systems could potentially compromise operation and safety of nuclear reactor power plants. This work utilizes the LOBO Nuclear CyberSecurity (LOBO NCS) Platform, developed recently at the University of New Mexico’s Institute for Space and Nuclear Power Studies, to investigate and contrast responses of an emulated PLC with OpenPLC and a commercial Allen-Bradley PLC. This is during nominal operation and simulated surge-in and surge-out transients of the pressurizer in a representative PWR plant. Investigations evaluate the effect of manipulating the control of linked PLCs to a physics-based Simulink model of the pressurizer on its operation during simulated transients. Simulated FDIAs introduced during the surge-in transient manipulate either input pressure or the rate of water spray into the pressurizer, to increase system pressure beyond nominal. Simulated FDIAs inconsistently overwrite holding registers of the PLCs during the duration of the attack. As a result, the immersed heaters and the water droplets spray nozzle switch off and on repeatedly, to reduce the pressure rise within the pressurizer. Manipulating a commercial Allen-Bradley PLC is more consistent than emulated OpenPLC and percentage of FDIAs’ successful overwrites increased with increasing input scan time of the PLC. Despite noted response differences iof emulated and hardware PLCs, results demonstrate using emulated PLCs in the LOBO NCS platform is suitable for current and future cybersecurity investigations.

Thermal Fatigue of Pressurized Water Reactor Coolant System Loop Drain Lines Due to Outflow Activities

Abstract Thermal cycling induced fatigue is widely recognized as one of the major contributors to the damage of nuclear plant piping systems, especially at locations where turbulent mixing of flows with different temperature occurs. Thermal fatigue caused by swirl penetration interaction with normally stagnant water layers has been identified as a mechanism that can lead to cracking in dead-ended branch lines attached to pressurized water reactor (PWR) primary coolant system. Electric Power Research Institute (EPRI) has developed screening methods, derived from extensive testing and analysis, to determine which lines are potentially affected as well as evaluation methods to perform evaluations of this thermal fatigue mechanism for the U.S. PWR plants. However, recent industry operating experience (OE) indicates that the model used to predict thermal fatigue due to swirl penetration is not fully understood. There are limitations with the EPRI generic evaluation. In addition, cumulative effects from other thermal transients, especially those resulted from outflow activities, may also contribute to the failure of reactor coolant system (RCS) branch lines. In this paper, we report direct OE from one of our PWR units where thermal fatigue cracking is observed at the RCS loop drain line close to the welded region of the elbow. A conservative analytical approach that takes into account the influence of thermal stratification, in accordance with ASME section III class 1 piping stress method, is also proposed to evaluate the severity of fatigue damage to the RCS drain line, as a result of various transients, particularly the transients from outflow activities. Finally, recommendations are made for future operation and inspection based on results of the evaluation.

Simulation of 15% and 50% Thermal Power Dispatch to an Industrial Facility Using a Flexible Generic Full-Scope Pressurized Water Reactor Plant Simulator

Nuclear power plants in the United States are increasingly challenged to compete in wholesale electricity markets due to the low electricity costs and increasingly dynamic grid conditions from competing generation sources. An alternative market for nuclear power is industrial facilities that can use the thermal and/or electrical power generated by a nuclear power plant to offset the economic losses incurred by electricity market challenges. A generic pressurized water reactor (PWR) simulator was used to show the results of a basic design for a generic thermal power extraction system and tests were run using a set of procedures to show what happens when a nuclear power plant transitions from full electrical power dispatch to 15% and 50% thermal power dispatch. This type of operation leads to losses in turbine performance efficiency due to the deviation from the design operating point, but because the thermal power is also used by the industry load without conversion losses, the combined thermal efficiency of the PWR increases. For the 15% case, the thermal efficiency increased from 32% to 41.9%, while for the 50% case, the efficiency increased up to 60.1%. In addition, for 50% thermal power dispatch, the power dissipated by the condenser decreased from approximately 2000 to approximately 1300 MW (thermal), indicating a substantially diminished impact on the environment in terms of releasing heat into the cooling water reservoir.

A Lack of Knowledge Acquisition Can Impair Nuclear Power Plant Safety

Abstract A lack of knowledge acquisition (LOKA), among engineering staff members in supply firms (e.g., vendors), owner/operator utilities (e.g., licensees), and in regulatory agencies (e.g., the NRC), can impair nuclear power plant (NPP) safety in ways that can persist throughout the operating lifetime of an NPP. A LOKA occurs when experienced technical reviewers fail to pass enough information, or technology, to less-experienced technical reviewers. The existence of a LOKA, among technical reviewers, can lead to errors and omissions that can result in misleading or incomplete licensing bases. Eight examples of errors and omissions are presented, each of which is evaluated in the context of physical phenomena, logic, licensing strategy, and effects upon regulation. These errors and omissions could be attributed to several causes, one of which could be a LOKA. Reliable attribution to a LOKA or its causes is not directly addressed, since attention is focused principally upon the safety implications of errors and omissions that may possibly, but not exclusively, be due to a LOKA. The epistemology of a LOKA, which may consider training or human relations, is generally addressed in other studies, which apply to issues that affect more than the nuclear power industry. If those who design, analyze, license, operate, and regulate NPPs do not adequately understand and apply proven engineering principles, standards, and established regulations, critical thinking, and sound logical reasoning, then it could be said that a LOKA exists. A LOKA could hamper the development of defensible conclusions in safety analyses, viable licensing strategies, and fair regulatory judgments. Eight examples of errors and omissions are presented, each of which leads to a conclusion that seems to conflict with an industry standard, a federal regulation, an engineering principle or physical phenomenon, or just plain logic. The examples are generally evaluated in accordance with the requirements of a well-known, oft-cited nuclear industry standard, which is now almost half a century old. This standard was published in 1973 by the American Nuclear Society (ANS) (1973, Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants, La Grange Park, IL, ANS-N18.2-1973). It expresses the fundamental principle of nuclear safety and licensing, which is applied by vendors, licensees, and regulators alike. This Standard defines nuclear safety criteria and plant design requirements for plant operating situations or events according to their expected frequencies of occurrence. Those events that have high frequencies of occurrence must not pose a danger to the public. Events that could pose the greatest danger to the public must be limited, by design, to extremely low expected frequencies of occurrence. This concept is implemented by grouping postulated plant situations (or events) into categories that are defined according to their expected frequencies of occurrence. Licensees are required to present analyses and evaluations of the events, in each category, to demonstrate that the events’ consequences do not exceed the category’s specified acceptance limits. Furthermore, licensees are required to demonstrate that certain events would not develop into more serious events (e.g., events that would be grouped into more serious, higher-consequence categories), without the concurrent occurrence of independent faults. That is, the Standard (American Nuclear Society, 1973, Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants, La Grange Park, IL, ANS-N18.2-1973) requires that NPPs must be designed in a way that does not allow high-frequency, low-consequence events to degrade into high-frequency, high-consequence events. The errors and omissions that are considered in the example evaluations could be due to a LOKA, among other possible causes. In each example, a LOKA is sufficient; but not necessary, to produce the noted errors and omissions. Attention is focused upon the safety implications of a LOKA, not its epistemology. The LOKA in the nuclear power industry, including its regulators, is ongoing, since it is not recognized and remedied.

The operational assessment of steam generator tubes based on a stochastic crack growth model

In pressurized water reactor plants, the degradation of steam generator tubes has been one of the main economic and safety problems. Steam generator tubes are subjected to the degradation mechanism of stress corrosion cracking, in which an unstable crack growth may lead to their structural failure. In general, deterministic criteria provide the least conservative assessment in order to plug or sleeve the cracked tubes. As the crack growth rate under stress corrosion cracking is of high random nature, the structural integrity assessment of steam generator tubes must be performed taking into account the failure probability at current and next inspections. In this paper, a stochastic model is applied for performing the operational assessment of steam generator tubes for any time interval. The results obtained with this model are compared with others, such as Monte-Carlo simulation and non-evolutionary statistical model. The proposed methodology is able to provide efficiently a less conservative failure probability. The plausibility of stochastic models in the structural integrity assessment of steam generator tubes is reached in favor of their operational life extension.

On the validation of BEPU methodologies through the simulation of integral experiments: Application to the PKL test facility

Validation is a key step in the field of thermal hydraulics and has played a crucial role in the implementation of system codes. These codes are then included in licensing methodologies being those either best estimate plus uncertainties (BEPU) or conservative approaches. However, only a small amount of efforts has been done on the validation of the methodologies as a whole. Such validation would imply the comparison of the bounding limits and the experimental data. Due to the lack of actual data of accidental situations, such validation should be carried out with data from integral test facility (ITF) experiments.The OECD/NEA PKL-4 test programme is investigating safety issues relevant for current pressurized water reactor (PWR) plants as well as for new PWR design concepts focusing on complex heat transfer mechanisms under two-phase flow. Within the project, a blind benchmark activity has been launched where an IBLOCA experiment has been simulated by 12 different organizations. The Universitat Politècnica de Catalunya (UPC) participated in the activity with a blind BEPU approach with the main objective of assessing the validity of the methodology as a whole.At first, the scalars that characterize the phenomenology of the particular scenario have been considered. Secondly, in order to broaden the extent of the validation, the results of the different participants have been interpreted as if they were blind experiments as well. Finally, different output parameters have been considered providing a matrix of results that is used to evaluate the validity of the methodology.

Flaw Evaluation Procedure for Cast Austenitic Stainless-Steel Materials Using a Newly Developed Statistical Thermal Aging Model

Abstract Thermal embrittlement of some cast austenitic stainless steels (CASS) occurs at reactor operating temperatures leading to very low fracture toughness. Because of their low aged toughness with high variability, flaw evaluations of CASS material need to be established with an understanding of the materials aged condition, especially since most U.S. pressurized water reactor (PWR) nuclear plants have been given plant life extensions for 60-year operation. A flaw evaluation procedure for CASS materials is presented here using a new statistical model developed to predict the toughness of fully aged CASS using the materials' chemical compositions. In this procedure, the dimensionless-plastic-zone-parameter (DPZP) analysis is used to determine when limit-load is applicable and also approximate the elastic-plastic correction factor (Z-factor) to predict the failure stress for CASS pipe/fittings with a circumferential surface crack. The procedure was validated against several CF8m pipe test results which include various pipe diameters, crack sizes, ferrite contents, failure modes. The as-developed flaw evaluation procedure was also used to determine the Z-factors for four different pipe diameters for a database of 274 pipe/elbows in U.S. PWR plants –solving 1096 sample problems to understand what range of Z-factors in U.S. PWR plants (for CF8m CASS materials). Finally, the applicability of the CF8m-based statistical model for use with CF3 and CF8 CASS materials was also verified with available test results. The outcome of this work has been implemented in an ASME boiler and pressure vessel (BPV) code case and has been approved in 2020 as Code Case N-906.

Estimation of fission product source terms for the SGTR accident of a reference PWR plant using MELCOR and MAAP5

A steam generator tube rupture (SGTR) is a type of containment bypass accidents that if core damage is not prevented during this accident progression, has the potential to lead to a significant release of fission products (FPs) to the environment through the faulted steam generator (SG) or other possible paths in the secondary system. This paper presents a best practice modeling approach to more realistically quantify the FP source terms expected during an SGTR accident of a reference pressurized water reactor (PWR) plant, and also investigates the impact of dedicated accident mitigation actions on the results of interest through the established modeling approach. To consider the integral responses of the reactor, SGs, and containment to a severe accident initiated by SGTR, detailed modeling was performed for a broad spectrum of plant safety/mitigation systems and relevant phenomena influencing the evolution of the accident. As the modeling and simulation tools for the SGTR accident analysis, MELCOR 2.2 and MAAP5 were employed to compare the results of interest. The results of the analysis demonstrate that the release of FPs into the environment could be estimated more realistically as a result of the best practice modeling for the SGTR accident and also it could be greatly decreased through dedicated mitigation actions. Relevant insights are summarized in terms of particular points of interest and remaining uncertainty issues for further study.

Estimation of fission product source terms for the ISLOCA of a reference PWR plant using MELCOR and MAAP5

An interfacing systems loss of coolant accident (ISLOCA) is a type of containment bypass accident that may lead to a significant release of fission products (FPs) to the environment. This paper presents a best practice modeling approach to more realistically quantify the FP source terms expected during the ISLOCA of a reference pressurized water reactor (PWR) plant and investigates the influence of a dedicated accident mitigation action on the results of interest through the established modeling approach. In addition to this, the key plant features and phenomenological mechanisms that might have a significant influence on the magnitude and timing of FP releases to the environment during an ISLOCA were examined through a couple of sensitivity analyses. To consider the integral responses of the reactor, containment, and auxiliary building to a severe accident initiated by an ISLOCA, detailed modeling was performed for a broad spectrum of relevant phenomena, plant systems, and structures that promote the deposition and retention of FPs along the flow path to the environment. As the modeling and simulation tools for the ISLOCA analysis, MELCOR 2.2 and MAAP5.04 were employed to compare the results of interest. The results of the analysis shows that the present best practice modeling approach proposed here has the potential to significantly decrease the release of FPs into the environment in the event of an ISLOCA. Relevant insights are summarized in terms of particular points of interest and remaining uncertainty issues for further study.

An assessment of radiological source terms for loss of cooling and coolant accidents of a PWR spent fuel pool

This paper presents the MELCOR code simulation study regarding how a dedicated severe accident mitigation (SAM) measure the release of radiological FPs to the environment in the spent fuel pool (SFP) of a reference pressurized water reactor (PWR) plant. To draw up relevant insights, a series of accident scenarios have been investigated, taking into account a combination of two initiating events (loss-of-cooling and loss-of-coolant accidents), two SFP operating modes (normal and refueling), and three external water makeup strategies through an external cooling water injection as one of dedicated SAM strategies for a long-term cooling of the SFP. For the analysis, relevant code modeling has been made for a broad spectrum of severe accident-related phenomena being expected during the SFP accidents. As a result, the evolution of relevant physical phenomena has been investigated and discussed in light of the release of FPs, and several points of uncertain issues, which need to address more carefully (a) either when implementing a cross-comparison for different code predictions (such as MELCOR and MAAP) for SFP accident progression or (b) when assessing radiological risk and/or developing dedicated SAM strategies for the SFP of PWR plant, have been drawn up.

Simulation of Porous Magnetite Deposits on Steam Generator Tubes in Circulating Water at 270 °C

In the secondary side of pressurized water reactors (PWRs), the main corrosion product accumulated on the steam generator (SG) tubes is magnetite, which has a porous structure. The purpose of this work is to simulate the porous magnetite deposited to the SG tubes using a loop system. We newly developed a circulating loop system for a porous magnetite deposition test. A test section was designed as a single hydraulic flow channel, and a cartridge heater was fabricated and mounted into a commercial SG tube to provide an equal heating source for the primary water. After the deposition test, the simulated magnetite deposits were characterized for comparison to real SG tube deposits collected from an operating PWR plant. The magnetite deposits produced using the loop system were appropriate for simulating the real SG tube deposits because the particle characteristics, phase, and porous morphology are closely similar to those of real deposit samples. Using the loop system, the chemical impurities such as Na and Cl can be easily concentrated within the pores of the simulated magnetite deposits. These simulated magnetite samples are expected to be widely utilized in various research fields such as the heat transfer degradation and magnetite accelerated corrosion of SG tubes.

Extension of a Level 2 PSA Event Tree Based on Results of a Probabilistic Dynamic Safety Analysis of Induced Steam Generator Tube Rupture

This paper presents the approach of extending a classical generic event tree (ET) of a Level 2 Probabilistic Safety Analysis to the results of a probabilistic dynamic safety analysis. The example of creep-induced steam generator tube rupture has been chosen. The results of an Analysis of Thermal Hydraulics of Leaks and Transients with Core Degradation (ATHLET-CD)/Monte Carlo Dynamic Event Tree (MCDET) simulation analyzing the failure of reactor coolant system components by creeping in a scenario of a high-pressure core melt accident in a generic pressurized water reactor (PWR) have been implemented in the ET. From the results of these simulations, a set of parameters has been extracted and integrated into the ET along with their probability distributions. The effect of these parameters on both the progression of the severe accident sequence under consideration and the release categories of a generic German PWR plant, respectively, is discussed in this paper.

Risk Reduction Assessment for Installing RCP Shutdown Seal

As one of the lessons learned from the Fukushima Daiichi accident, long-term station blackout (SBO) and subsequent loss of ultimate heat sinks have prompted discussion on this topic in the nuclear industry. The SBO sequences for a Westinghouse three-loop pressurized water reactor (PWR) have been under investigation for a long period. To cope with the long-term SBO issue, many nuclear power plants have decided to replace the reactor coolant pump (RCP) seal by the new passive thermal shutdown seal (PSDS). The PSDS is a fail-safe protection device that will significantly reduce leakage from the RCP seal in case of loss of cooling. This makes a seal loss-of-coolant accident no longer a risk-significant event, and the relevant probabilistic risk assessment (PRA) models need to be modified to reflect the associated plant change. The PRA model of a Westinghouse PWR plant has been reviewed to reflect this more strongly; i.e. loss of component cooling water (CCW), loss of 4-kV vital alternating-current power, and loss of off-site power are revised for their sequences. According to the Westinghouse analysis, the PSDS temperature must be maintained below 104°C, and the operators have to control the RCP speed. In this paper, those factors are incorporated into the event tree structure revision of the loss of CCW event (TC) and loss of power either off-site (TP) or vital power A train (TAPB). In another case, LOOP initiating events need to consider the time span that the blackout conditions would affect the RCP seal integrity. Because the RCP will be tripped automatically, the limitation associated with RCP speed will not be applicable. Compared with the TC, TP, and TAPB event tree base cases, RCP speed slowdown or available time span is introduced into the PSDS model. The relevant part in the PRA model is subject to review and modification. The risk reduction associated with the PSDS is found to be significant.

Development of critical heat flux correlation for In-vessel retention

ABSTRACT In-vessel retention (IVR) is a strategy for severe accident management in which the lower head of the reactor vessel is submerged in a water-flooded reactor cavity. Critical heat flux (CHF) data for IVR are important for estimating cooling capacity of the reactor vessel. The existing CHF data for IVR which were obtained for the specific geometries and thermal-hydraulic conditions of actual plants are difficult to be applied to plants with other specifications. Hence, the purpose of this study is to develop CHF correlations applicable to various pressurized water reactor plants in a wide range of thermal outputs based on newly obtained CHF data. A rectangular test section with a cross-section of 150 mm × 150 mm and length of 600 mm was used for simulating a cooling channel. The thermal-hydraulic conditions expected in actual plants were studied, and the results were used in the experiment. The effects of parameters such as pressure, mass flux, thermodynamic quality, and angle on CHF were investigated . Based on these results, we developed a CHF correlation formula that can be applied to a wider range than previously, up to a maximum heat flux of 3000 kW/m2, and that predicts CHF with an error of ± 10%.

A Study on the Crystalline Boron Analysis in CRUD in Spent Fuel Cladding Using EPMA X-ray Images

Chalk River Unidentified Deposits (CRUDs) were collected from the Korean pressurized water reactor (PWR) plant (A, B, and C) where the axial offset anomaly (AOA) occurred. AOA, also known as a CRUD-induced power shift, is one of the key issues in maintaining stable PWR plant operations. CRUDs were sampled from spent nuclear fuel rods and analyzed using an electron probe micro-analyzer (EPMA). This paper describes the characteristics of boron-deposits from the CRUDs sampled from twice-burnt assemblies from the Korean PWR. The primary coolant of a PWR contains boron and lithium. It is known that boron deposition occurs in a thick CRUD layer under substantial sub-cooled nucleate boiling (SNB). The results of this study are summarized as follows. Boron was not found at the locations where the existence was confirmed in simulated CRUDs, in other words, the cladding and CRUD boundaries. Nevertheless, we clearly observed the presence of boron and confirmed that boron existed as a lump in crystalline form. In addition, the study confirmed that CRUD existed in a crystal form with a unique size of about 10 μm.

EcSOD in the Health Benefits of Exercise

The coupling of the Qeshm island desalination plant with different energy sources was economically evaluated by using a Desalination Economic Evaluation Program (DEEP) package. The capacity of this plant was 18,000 m3/day and its energy source was 250 MW (th) gas power plant. Six energy source plants, namely, coal, oil, gas, combined cycle (CC), pressurized water reactor (PWR), and pebble-bed modular reactor (PBMR), were coupled with this desalination plant. The estimated annual electrical costs indicated that the optimum fossil power plant was CC with 97.72 $/kWh, which was 30% higher than 67.42 and 64.63 $/kWh in PWR and PBMR power plants, respectively. The total annual water costs and the total annual costs of coupling with the desalination plant of CC, PWR, and PMMR plants were approximately 1.78, 1.49, and 0.77 $/m3, and 94.74, 49.23, and 49.1 m$, respectively. The construction costs of these power plants are approximately 131, 510 and 553 M$, respectively, On the basis of the contraction times, costs, and lifetimes of these power plants 25, 40 and 60 years for CC, PBMR and PWR, we concluded that CC, PBMR and PWR are short-term, medium-term and long-term strategy to generate electricity and couple with desalination plants, respectively.