ASME Conference Presenter Attendance Policy and Archival Proceedings

2014;():V001T00A001. doi:10.1115/ICONE22-NS1.

This online compilation of papers from the 2014 22nd International Conference on Nuclear Engineering (ICONE22) represents the archival version of the Conference Proceedings. According to ASME’s conference presenter attendance policy, if a paper is not presented at the Conference, the paper will not be published in the official archival Proceedings, which are registered with the Library of Congress and are submitted for abstracting and indexing. The paper also will not be published in The ASME Digital Collection and may not be cited as a published paper.

Commentary by Dr. Valentin Fuster

Plant Operations, Maintenance, Engineering, Modifications, Life Cycle and Balance of Plant

2014;():V001T01A001. doi:10.1115/ICONE22-30043.

Flow accelerated corrosion (FAC) is a major degradation form of carbon steel and low alloy pipes in the secondary circuit of pressurized water reactor (PWR) plants, which has great impact on plant safety and reliability. For the purpose of effectively monitoring FAC in nuclear power plants, a statistical model for accessing FAC wall thinning rate using plant inspection data is proposed in this paper. The presented model is developed based on Gaussian stochastic process models. Wall thinning rate is considered as a function of key factors that have important influence on the FAC process (i.e., temperature, pH, mass transfer coefficient, etc.). The Kriging method, which has been widely applied in the domain of spatial analysis, is used to model the relationship between wall thinning rate and its impact factors. Model parameters are determined through maximum likelihood estimation using the inspection data. Since the likelihood function of the Kriging model is usually complicated in form, the genetic algorithm is employed to find parameter values that maximize this function. From the presented model, residual lifetime distributions of pipes affected by FAC can be derived, and conditions that may lead to high FAC rate can be found, which provides decision-making support for maintenance strategies optimization in life cycle management of the feed water system. Wall thinning data simulated from a physical-chemical mechanism model presented in literature are used to verify the presented model. Results of validation show that reasonable wall thinning rates and lifetime distributions can be obtained using this model.

Commentary by Dr. Valentin Fuster
2014;():V001T01A002. doi:10.1115/ICONE22-30118.

A PWR reactor coolant system is a highly complex physical process: heterogeneous power, flow and temperature distributions are difficult to be accurately measured, because instrumentations are limited in number, thus leading to the relevant safety and protection margins. This situation is in many ways similar to climate and weather models: a complex process that is not possible to sample and measure as finely as wanted. Meteorology and climate sciences have adapted and improved the Data Assimilation techniques in order to improve the accuracy of description and prediction in their fields.

EDF R&D is seeking to assess the potential benefits of applying Data Assimilation to a PWR’s RCS (Reactor Coolant System) measurements: is it possible to improve the estimates for parameters of a reactor’s operating set-point, i.e. improving accuracy and reducing uncertainties of measured RCS parameters?

In this paper we study the feasibility of enhanced estimation of PWR primary parameters, by using twin experiments for assessing Data Assimilation benefits. We simulated test samples with a 0D-Model, and used these samples in a Monte-Carlo approach to get background terms for Data Assimilation.

This successful preliminary study will lead to further assessments with real plant data.

Commentary by Dr. Valentin Fuster
2014;():V001T01A003. doi:10.1115/ICONE22-30241.

Nuclear power plants are no longer immune to cycling operation. While certain nuclear power plants in Europe have been performing load following operation, this type of operation has largely been avoided in the United States. Due to increasing contribution of nuclear generation in the mix, European operators were forced to make modifications to increase the maneuverability of their nuclear generation assets. However, in the United States, nuclear generation is still a relatively smaller contributor (19%). Still, with rapid increase in renewable generation, some nuclear plants are being asked to operate at reduced power and cycle to lower power levels. With most future renewable integration studies advocating for increased flexibility on the grid, nuclear generation maneuverability will allow system operators with another resource to mitigate system costs.

This paper presents the results of a detailed study of a 1,150 MW boiling water reactor nuclear plant when cycled to low loads. The authors present the relative damage of cycling to various reduced power levels 80% to 15% power levels compared to a cold startup and shutdown of a nuclear plant. An assessment was made of the systems that had fatigue damage and costs. We also discuss some of the limitations of cycling that a nuclear plant has and present and discuss recommendations to reduce damage and costs.

Commentary by Dr. Valentin Fuster
2014;():V001T01A004. doi:10.1115/ICONE22-30247.

Six nuclear power reactors in Taiwan have been operating over beyond thirty years. They are all operated by Taiwan Power Company (TPC) and expected to have 40-year lifetimes. The limited original suppliers and obsolete components are the challenge to comply with current licensing basis and maintaining a high level reliability. Therefore, the procurement of basic components from the second source is very important to the plant safety and operation.

This paper describes the dedication process applied to commercial power distribution panels in a harsh environment. The safety functions of power distribution panels provided backup power input connection for mobile diesel generators while station blackout (SBO). After Fukushima-Accident, the utility needs to setup diversity power to comply with regulatory requirements in Taiwan. The power distribution panels dedication activity include the function testing, aging, seismic qualifications (SQ), and environmental qualifications (EQ) based on EPRI NP 5652, IEEE Std. 323, and IEEE Std. 344 standards. Some subcomponents could not meet the acceptance criteria during testing and the anomalies were noticed to the customer and the utility. One of these anomalies reported to regulatory due to the subcomponent failure after accident radiation endurance test.

Commercial-Grade Item dedication is second source to obtain safety related components according to 10 CFR 21.3 definitions. In the past nineteen years, Institute of Nuclear Energy Research (INER) has actively performed the dedication service to help local nuclear power plants solve their procurement problems of nuclear grade items, due to reduced availability of qualified suppliers and/or obsolete issues of qualified components. Although the codes and standards for dedication in Taiwan refer to those in USA, the challenges may happen due to different regulators, utility, manufacture’s quality culture, and personal responsibility. This paper introduces the self-reliant experiences in dedication and economic benefit to local nuclear power plants.

Topics: Safety
Commentary by Dr. Valentin Fuster
2014;():V001T01A005. doi:10.1115/ICONE22-30304.

A remote operated quadruped robot has been developed for disaster site which can move on stairs, slopes, and uneven floor under the radiation-polluted environment, such as TEPCO Fukushima Daiichi nuclear power plants [1][2].

In particular, the control method for stable walking and the remote operation system have been developed to move on stairs in the reactor building.

We applied this robot to investigation of suspicious water leakage points in reactor building at Fukushima Daiichi nuclear power plants unit2[3]. In this investigation, a small vehicle equipped with camera and a manipulator which is connected the vehicle with cable were mounted on the robot and were carried to near the target by the quadruped robot and the investigation was carried out with the small vehicle.

Topics: Robots
Commentary by Dr. Valentin Fuster
2014;():V001T01A006. doi:10.1115/ICONE22-30431.

The SCORPIO-VVER core monitoring and surveillance system has proved since the first installation at Dukovany NPP in 1999 to be a valuable tool for the reactor operators and reactor physicists. It is now installed on four units of Dukovany NPP (EDU, Czech Republic), on two units of Bohunice NPP (EBO, Slovak Republic) replacing the original Russian VK3 system and on the full scale plant training simulator at the Centre for training and education of the reactor operators and reactor physicist in Trnava (Slovak Republic). By both Czech and Slovak nuclear regulatory bodies the system was licensed as a Technical Specification Surveillance tool.

Since it’s first installation, the development of SCORPIO-VVER system continues along with the changes in VVER reactors operation. The system is being adapted according the utility needs and several notable improvements in physical modules of the system were introduced. The most significant changes were done in support of the latest optimized Gd bearing fuel assemblies, improvements in the area of core design (neutron physics, core thermal hydraulics and fuel thermal mechanics), adaptation of the system to up-rated unit conditions (uprated power up to 107%), in design and methodology of the limit and technical specifications checking and improvements in the predictive part of the system.

After the currently finished upgrades the SCORPIO-VVER is still in focus of Central European nuclear power plants with the roadmap of upgrades and modifications up to 2016.

This paper shortly describes the system’s main functions, the history of implementation at the VVER-440 type of reactors and deals with the system’s future upgrades and plans to meet the latest requirements of efficient and safety NPP operation.

Commentary by Dr. Valentin Fuster
2014;():V001T01A007. doi:10.1115/ICONE22-30462.

Atmosphere isolation is one of the pivotal operations during the maintenance of HTR-10 helium circulator, which main objects are to keep the high helium purity within the primary loop of HTR-10 and avoid radioactive contamination and diffusion of graphite dust. Several skillful and comprehensive processes are applied to achieve the object successfully.

Topics: Maintenance , Helium
Commentary by Dr. Valentin Fuster
2014;():V001T01A008. doi:10.1115/ICONE22-30470.

The thermal stratification phenomenon widely exists in the pressurizer surge line (PSL) during the steady-state conditions and transient conditions, which could result an unexpected stress distribution. Especially, the maximum stress range is most likely to appear nearby the weld joint area on the PSL. This paper mainly focuses on the pressurizer surge line stress distribution variation caused by the weld joint. The temperature distribution and stress distribution of the surge line are calculated by using the finite element method (FEM) and finite volume method (FVM). Steady-state operating conditions are considered. The conclusion of this work is that the stress distribution is quite different from the traditional stress profile merely caused by the upper and lower temperature stratification. Moreover, the stress is much higher in the interact-area between PSL matrix and the weld joint than in any other areas.

Commentary by Dr. Valentin Fuster
2014;():V001T01A009. doi:10.1115/ICONE22-30489.

A new prediction model for the oxidation layer thickness of carbon steel is developed, that is based on the parabolic time law of corrosion and the mass transport balance theory. The relationship between the oxidation layer thickness and temperature, pH, and flow velocity is discussed. The predicted results show that the oxidation layer thickness increases exponentially with increasing temperature and decreases exponentially with increasing flow velocity. The oxidation layer thickness increases with increasing pH until pH=10.5 and then decreases. The predicted results agree with experimental results.

Commentary by Dr. Valentin Fuster
2014;():V001T01A010. doi:10.1115/ICONE22-30593.

A simple and highly reliable motor diagnostic method was developed. The method monitors low-frequency components of leakage flux from motors. Small wear of bearing housing was detected by the method which could not be detected by the conventional vibration or electric current monitoring methods.

Commentary by Dr. Valentin Fuster
2014;():V001T01A011. doi:10.1115/ICONE22-30638.

We developed a new laser processing system to repair various damages of metal parts in aging nuclear facilities. The system consisted of some components; a laser torch, a composite-type optical fiber (COF), a quasi-continuous wave (QCW) fiber laser, a coupling device. All components were installed in a mobile rack, so we can carry it to a place where we want to use. The COF that had the hybrid-delivery functions of heating laser beam and observing visible image was proved to be very useful for accuracy wire-feeding in limited tubular space. A welding expert succeeded to demonstrate manually a line laser cladding on the inner wall of 1-inch tube, observing the work surface through fiberscope images. And also we succeeded to make clad layers in 1-inch tube by automatic control of devices. The clad on the inner wall of 1-inch tube could not be made by arc welding tools. We proposed to apply the laser cladding system to maintenance of aging industrial plants and nuclear facilities.

Commentary by Dr. Valentin Fuster
2014;():V001T01A012. doi:10.1115/ICONE22-30702.

This paper addresses several key issues relevant to piping vibration: (1) vibration properties and their identification; (2) the dynamic susceptibility of a piping system; (3) standard velocity criteria for vibration damage. Practical approaches for identifying vibration properties and practical measures for preventing vibration damages are overviewed. An approach is presented for conducting vibration design reducing measures prior to installation of a piping system, and for planning post-installation vibration recording activities.

This approach is based on “theoretical predictions” of “critical” sections of a piping system, namely the sections most sensitive to vibration. The critical sections can be ranked from most vulnerable to least vulnerable. Combined with knowledge of typical vibration sources, this is a cost effective way for preventing vibration damages and for forming a base for controlling actual vibrations in operation of plant, especially for closed premises. Examples are given to exemplify vibration risk.

Commentary by Dr. Valentin Fuster
2014;():V001T01A013. doi:10.1115/ICONE22-30735.

Competition in the electricity market forces producers to achieve — in compliance with safety — efficiency of production as high as possible. This efficiency and heat rate is an important indicator of both the condition of the power plant equipment and the quality of power plant operation. To cope with these challenges, powerful methods are process data reconciliation, statistical data processing of large data sets and process simulation. These functions and methods can be used to obtain useful information about process quality and equipment and sensor health.

The paper discusses practical experience from six years of using a thermal performance monitoring and optimization system in the Dukovany nuclear power plant. The system is integrated into the overall nuclear power plant process information system and data warehouse. The system provides information in near real time.

The major benefit of the system lies in a deep view into equipment behaviour and process which ensures timely detection and identification of functionality degradation of process and equipment or sensor faults. The system also helps to find and use margins of equipment operation and the overall thermal cycle.

Selected practical examples are used to demonstrate specific benefits of the system for operation and maintenance of the Dukovany nuclear power plant. There are examples of equipment fault detection and sensor degradation detection. The optimization function is explained with an example of cooling circuit optimization aimed to increase the delivery of electrical power into the grid. A detailed description of behaviour of the main components can be used for their performance evaluation and their repair planning. The benefit of more accurate determination of parameter values is reflected in more accurate determination of reactor thermal output.

The conclusion of the paper provides an overall evaluation of system benefits for operation and maintenance of a nuclear power plant.

Commentary by Dr. Valentin Fuster
2014;():V001T01A014. doi:10.1115/ICONE22-30799.

Qinshan phase III is the only CANDU nuclear power plant (NPP) in China, which has two 728MW units, and starts commercial operation on Dec. 31st, 2002 and July 24th, 2003 respectively. According to the Periodic Safety Review of Nuclear Power Plants (NS-G-2.10) issued by IAEA in 2003 and the corresponding Chinese edition HAD103/11 issued by NNSA in 2006, the first PSR of Qinshan III should be carried out in 2012 and 2013 after 10 years of commercial operation. Comprehensive assessment of plant safety is a complex task and is facilitated by dividing it into a number of factors. The equipment qualification (EQ) factor is an important one of them. In this paper, the equipment qualification safety factor review of PSR is carried out for Qinshan III. Firstly, the project background is described. Then objectives, scopes and main review elements of EQ factor review are summarized. Also, the EQ factor review process is emphasized. Finally, the project team and outcome of EQ factor review are given.

Commentary by Dr. Valentin Fuster
2014;():V001T01A015. doi:10.1115/ICONE22-30816.

Angra-1 Nuclear Power Station (Westinghouse PWR-600 MW, 2 loops) started commercial operation in 1985, being property of Eletronuclear, subsidiary of Eletrobras in Brazil. Angra-1 has been preparing the necessary measures to renew the operating license and to apply for a lifetime extension up to 60 years.

Among the many activities to perform, there are some related to fulfilling the requirements of the Brazilian regulator, the CNEN. These include requirements related to Human Factors Engineering (HFE) that included the preparation of a Chapter 18 of HFE, to become part of the plant’s Final Safety Analysis Report (FSAR).

In the framework of the Instrument for Nuclear Safety Cooperation (INSC), created and funded by the European Union (EU) to enhance nuclear safety world-wide, cooperation activities between the EU and the Government of Brazil were set up in 2009. One of the INSC projects funded was to support the Brazilian nuclear operator of Angra-1 in the field of HFE. In 2010, the implementation of the project was awarded to a consortium lead by Tecnatom for performing a HFE Safety Evaluation to the plant and to provide support for preparing this Chapter 18.

For this Project a specific methodology was developed for the execution of the Safety Evaluation. The methodology has been developed for evaluating — from the HFE viewpoint — a plant in operation, from the beginning of commercial operation until nowadays, including the design modifications performed to date. The obtained results have been used for developing the aforementioned Chapter 18.

The main results of the Project Execution have been:

1. The developed methodology has made it possible to perform a comprehensive HFE evaluation of Angra-1, including the analysis of Post-TMI requirements, the design included in the current FSAR, the existing Angra-1 procedures and the verification of the current Main Control Room.

2. Technical support has been provided to Angra-1 for the preparation of Chapter 18 of the FSAR, following the structure of NUREG-0711, and using the results of the HFE Safety Evaluation.

3. An Action Plan has been developed for identifying and addressing in the future all those deficiencies found during the HFE Safety Evaluation, as well as those activities that are the consequence of the new FSAR Chapter 18.

Commentary by Dr. Valentin Fuster
2014;():V001T01A016. doi:10.1115/ICONE22-30854.

Flow-accelerated corrosion (FAC) is a degradation mechanism that attacks carbon steels under conditions often found in nuclear and fossil power plants. FAC has been responsible for a number of significant accidents in nuclear power plants. The most recent noteworthy accident was at the Mihama Unit 3 (Japan) where the catastrophic failure of a pipe in the condensate system resulted in five fatalities.

FAC can affect virtually all of the carbon steel piping and components in the power cycle of nuclear reactors. The presence of wall thinning caused by FAC is determined through the use of non destructive examination (NDE) techniques. For large-bore piping components the most commonly used approach is ultrasonic technique (UT) using an inspection grid applied to the piping components.

As FAC is a reasonably well-understood degradation mechanism, a number of computer programs have been developed to help utility engineers determine the inspection locations and managing the data. CHECWORKS™ — the most commonly used computer program for this purpose — is the program discussed in this paper.

The use of CHECWORKS™ in utilities’ FAC programs will be described. Particular emphasis will be placed on: inspection planning, handling power uprates and other changes to operating conditions.

Inspection planning is one of the most common uses of the CHECWORKS™ software. As there are typically around 5,000 FAC-susceptible components in a reactor system, utility engineers must select the components with the highest FAC-rates for inspection. CHECWORKS™ uses its predictions combined with plant inspection data to provide a best estimate of FAC rates. From these FAC rate predictions and knowledge of the piping schedule and allowable wall thicknesses, inspection locations are determined.

Both the water chemistry used and the local operating conditions strongly influence the rate of FAC. CHECWORKS™ is used to study potential changes to water chemistry, system operation or power level to determine the impact of such changes on FAC rates and hence inspection locations. Of particular interest is the use of CHECWORKS™ to determine the impacts of power uprates. Because of the complicated parametric behavior of FAC rates, changing the power level will likely increase the FAC rates in some areas of the plant while other areas will likely see a decrease in FAC rates. This fact requires a pre-uprate analysis to determine how an inspection program will need to be modified.

This paper provides a description of how CHECWORKS™ is used in the above applications as well as showing typical examples of its usefulness in these analyses.

Commentary by Dr. Valentin Fuster
2014;():V001T01A017. doi:10.1115/ICONE22-30879.

The present work describes an experimental investigation of the dynamic characteristics of check valves, which means experimental examination of closing function and ability to generate pressure transients under different flow decelerations in the pipeline. Two designs of check valves are tested: a swing disc and a tilted disc check valve.

The valves are mounted on discharge pipe of a centrifugal pump. The fluid transient is generated by stopping the pump motor from actual velocity to completely stop in a prescribed time.

Each of the check valves is subjected to tests covering different pressure levels in the upper reservoir, initial flow rates in the pipeline, several decelerations of pump rotation, three settings of torque acting on the valve disc, three values of the moment of friction forces acting on the valve axis and finally free fall of the disc in stagnant water and air.

The test stand, the instrumentation and chosen valves as well as scope and conditions for performing the experiments are described. Selected measured results like angular velocities of the discs, pressures in the pipe at different conditions, and volumetric flow rates are presented and discussed.

The dynamic behaviors of the tested valves were compared with each other.

Topics: Valves , Disks
Commentary by Dr. Valentin Fuster
2014;():V001T01A018. doi:10.1115/ICONE22-30896.

The pre-rotation phenomenon found at the inlet pipe of a pump under small flow rate was observed but the mechanism was not discovered. To do this, an unsteady CFD simulation with the flow rate decreasing from the best design point to a small flow rate was carried out. The numerical results show that a critical point when pre-rotation occurs exists. The flow pattern evolution on the S1 and S2 stream surface, as well as the pressure distribution at in the inlet, was analyzed. It is shown that it is the reverse flow appeared near the shroud at the leading edge which leads to the occurrence of pre-rotation. The pre-rotation only exist in the periphery of the inlet pipe and the propagation length is limited.

Topics: Rotation
Commentary by Dr. Valentin Fuster
2014;():V001T01A019. doi:10.1115/ICONE22-30901.

Current pressurized water reactors utilize sintered UO2 that has a number of advantages and disadvantages. Uranium Dioxide’s low thermal conductivity results in a large thermal gradient within the fuel pellet corresponding to higher centerline temperatures compared to other potential fuel forms. These gradients result in non-uniform thermal expansion leading to large internal stresses resulting in cracking of the pellet and fuel-clad interaction, which can lead to loss of the integrity of the fuel pin. Higher fuel temperatures also increase the release of fission gases. Fuels with higher thermal conductivity may alleviate or reduce the severity of these adverse conditions. It is shown that higher thermal conductivity can be obtained by adding BeO to the basic UO2 matrix. This paper focuses on WWER1000 hexagonal fuel geometry. Improvements when using 10% of BeO, as proposed in this paper, reduce the centerline nuclear fuel temperature by 234°C and improve the fuel economy while reducing its cost by 7%. The study was done for NPP Temelín which has two units WWER1000/320.

Topics: Nuclear fuels
Commentary by Dr. Valentin Fuster
2014;():V001T01A020. doi:10.1115/ICONE22-30913.

Prototype Fast Breeder Reactor (PFBR), India’s first commercial fast breeder reactor employing fast fission is a challenging project from technological point of view to meet the energy security of the country. PFBR is a sodium cooled fast reactor. There are 198 fuel sub assemblies with mixed oxide fuel in the reactor. The fuel is provided with a leak tight metal cladding for containment of the fission products. There are risks of sodium circuit contamination and the fission products blocking the coolant flow to the fuel sub assemblies in case of clad rupture and release of solid fission products into the coolant. Hence PFBR is equipped with an elaborate failed fuel detection and location system. Failed Fuel Location Module is one of the subsystems used to identify the sub-assembly having fuel pins with clad failure. This paper discusses about the conceptual design, design specifications, detailed design, manufacture, assembly and some of the results of functional testing of failed fuel location module of PFBR.

Commentary by Dr. Valentin Fuster
2014;():V001T01A021. doi:10.1115/ICONE22-30916.

In a typical pressurized water reactor commercial nuclear plant, a number of components such as CRDMs, a lift rig to lift the Reactor Vessel Closure Head (RVCH), seismic restraints, missile shield, and a cooling system with large air ducts are installed on or directly over the RVCH. These components and systems are typically designed and installed individually to perform designated functions during plant operation.

During refueling outages the removal of the RVCH from the pressure vessel and its subsequent re-installation on the pressure vessel for fuel loading requires individual dismantling and reassembly of these components resulting in an expensive and time-consuming process. Prior to detensioning the RVCH from the vessel, a lengthy series of steps or detailed procedures must be followed to safely remove the head area components and to store them in their designated spaces inside containment. The procedure generally includes: removal and storage of the concrete missile shield; removal and storage of CRDM cooling ducts; removal of seismic restraints; removal of head area cables; installation of the tripod assembly over the service structure; disconnecting the vent and level indicator lines; and installation of temporary lead blankets around the RVCH. Once the refueling is complete, these procedural steps are repeated in reverse order.

Each procedure in the refueling process contributes significantly to the total cost associated with personnel time required to perform the refueling, power plant down time and consequent loss of electricity production, radiation exposure to personnel, and risks and costs associated with potential human errors. In addition, these components require a large amount of storage space inside containment raising the risk of having inadvertent contamination of work and storage areas.

To reduce the outage duration and the associated radiation exposure to the workers, the authors have designed an Integrated Head Assembly (IHA) for Callaway nuclear plant based on Mr. Baliga’s patented design as disclosed in U.S. Patents. The IHA is an assembly of all head area components integrally attached to the RVCH so that all these components can be lifted with the RVCH in one assembly (see Figure 1). The IHA also provides a forced air convection system that improves the efficiency of the CRDM cooling. The IHA reduces a significant amount of critical path time and radiation dosage during refueling outages.

Mr. Baliga’s invention has been implemented at several commercial nuclear plants in the USA (Turkey Point Units 3 and 4; Salem Units 1 and 2; DC Cook Units 1 and 2; Diablo Canyon Units 1 and 2; Davis Besse Unit 1; and Callaway Unit 1). This paper provides details of the IHA design implemented at Callaway nuclear plant in the USA.

Commentary by Dr. Valentin Fuster
2014;():V001T01A022. doi:10.1115/ICONE22-30918.

The purpose of issuing a quality manual for the radioisotope production plants is to define and describe the quality system implemented by the plant. It provides general procedures for all activities comprising the quality system, defines the authorities and responsibilities of all personnel affected by the system and provides a way to inform our customers of the specific controls that are in place at radioisotope production plants to assure continued product quality. Such a quality manual is regularly updated to depict the quality management system implemented by radioisotope production plant as accurately as possible.

Example of this type of quality manual is developed for the radioisotope production facility (RPF) located in ETRR-2 Complex site – Egyptian Atomic Energy Authority – Inshas-EGYPT.

Topics: Radioisotopes
Commentary by Dr. Valentin Fuster
2014;():V001T01A023. doi:10.1115/ICONE22-31000.

Since pipe bend has a characteristic that extrados becomes thinner and intrados thicker after fabrication process, it can be expected to be vulnerable to extrados wall thinning due to corrosion or erosion during its operation. In this paper, limit loads of pipe bend with the thinning are computed under the loading conditions of internal pressure and bending moment. Several case studies with varying geometries and wall thinning shapes are presented. The difference in the limit loads behavior between pipe bend and welded elbow is also reviewed. The calculated plastic limit loads of pipe bend are compared with other research results for the welded elbow. The results show that pipe bend can be applied to safety-related piping systems as far as the internal pressure and bending moment only are considered.

Topics: Stress , Pipe bends
Commentary by Dr. Valentin Fuster
2014;():V001T01A024. doi:10.1115/ICONE22-31057.

The AP1000® plant is an 1100-MWe pressurized water reactor (PWR) with passive safety features and extensive plant simplifications that enhance construction, operation, maintenance and safety. One of the key design approaches in the AP1000 plant is to use passive features to mitigate design basis accidents. Active defense-in-depth (DiD) features provide investment protection, reduce the demands on the passive features and support the PRA. The passive features are classified as safety-related in the US. The active defense-in-depth features are classified as non-safety (with supplemental requirements) in the US. The AP1000 design has incorporated a standardization approach, which together with the level of safety achieved by the passive safety features, results in a plant design that can be applied to different geographical regions with varying regulatory standards and utility expectations without major changes.

While the first deployments of the AP1000 plant are ongoing in China and the United States, Westinghouse has remained active in also pursuing European opportunities for the AP1000 plant. In particular, Westinghouse has cooperated for almost two decades with European utilities to ensure adaptation of the AP1000 plant to the European market. This cooperation has resulted in progress towards AP1000 plant deployment in European countries.

The AP1000 plant is recognized worldwide and has been reviewed by regulators around the world, including China, the United Kingdom (UK), Canada as well as the US. The AP1000 PWR is the only Generation III+ reactor design to obtain final design approval from the United States Nuclear Regulatory Commission (US NRC) and interim approval from UK regulatory authorities as part of the Generic Design Assessment (GDA) process. It is the only technology to be licensed for construction in the United States in more than 30 years, and the only Generation III+ technology worldwide to receive an operating license, as well as construction approval in China. The AP1000 plant has been independently assessed and confirmed to meet the requirements of the European Utilities Requirements (EUR) document and the Electric Power Research Institute (EPRI) Advanced Light Water Reactor Utility Requirements Document (URD). The AP1000 plant has also been successfully assessed against multiple European industry guidelines such as the WENRA safety objectives, the IAEA safety standards, the ENSREG stress tests and the UK Weightman Report. In support of multiple ongoing request for proposal (RFP) and pre-RFP activities in European countries, Westinghouse has focused design effort and customer interactions in several European countries to adapt the AP1000 plant to European requirements. Review of the AP1000 plant design with regulators around the world, European Standards compliance activities, and continued cooperation and interaction with European Utilities provide confidence that the AP1000 plant can be successfully licensed and deployed in Europe.

The AP1000 50Hz standard plant design (also referred to as European Passive Standard or EPS) is the resulting adaptation of the AP1000 60 Hz US standard plant design to European market needs and requirements, addressing both customer input from such programs as the European Passive Plant (EPP) program in addition to regulatory and Utility needs identified though RFP and pre-RFP activities. The AP1000 50Hz standard plant design retains the overall AP1000 plant design (safe, simple, standard), the use of proven components and its cost, safety and operability advantages, while incorporating some changes to adapt to the European environment.

This paper will discuss some of the key changes that have been incorporated into the AP1000 50Hz plant design as necessary to adapt to the European market and demonstrate that the vast majority of the standard AP1000 plant design being built in China and the US is not impacted.

Commentary by Dr. Valentin Fuster
2014;():V001T01A025. doi:10.1115/ICONE22-31083.

Weld residual stress is a troublesome problem in nuclear power plant, because it can accelerate crack growth in weld region. For low alloy steel, Post Weld Heat Treatment (PWHT) is essentially needed to relieve residual stress and to temper the hard regions in the heat affected zone (HAZ). Local PWHT is used when it is impractical to heat the whole component in a furnace. The rules and practices of related codes and standards, such as ASME and AWS, associated with local PWHT are quite different. For example, according to ASME Section III, the minimum width of heated band at each side of the weld shall be the thickness of the weld or 2 in., whichever is less. While, according to ASME B31.1, the width of heated band shall be at least three times the wall thickness at the weld of the thickest part being joined.

In this paper, the status of the related code and standard associated with local PWHT is briefly summarized, and baseline information on local PWHT is explained based on FEA (Finite Element Analysis) results and optimized local PWHT parameter is suggested to support current code of practices.

Commentary by Dr. Valentin Fuster
2014;():V001T01A026. doi:10.1115/ICONE22-31097.

The EPR™ reactor features a fixed incore instrumentation, composed of 72 Self Powered Neutron Detectors (SPND), that provides the online reconstruction of the core maximum Linear Power Density (LPD) and minimum Departure from Nucleate Boiling Ratio (DNBR). The Instrumentation and Control (I&C) systems of the EPR™ reactor use this online reconstruction in surveillance and protection functions. The onsite thresholds of those I&C functions have to take into account all the uncertainties affecting the online reconstruction of core power distribution measured by SPNDs. One of these uncertainties is the so-called Loss Of Representativeness (LOR). This uncertainty is defined as the difference between the LPD (respectively DNBR) physical value and the LPD (respectively DNBR) computed value using SPND signals. The LOR parameter is mostly linked to the difference between the core power distribution at the time where SPNDs are calibrated and the core power distribution at the time where their signals are used. For the DNBR, LOR also takes into account the use of a simplified on-line DNBR calculation algorithm. A statistical approach is used in order to define this uncertainty. The analysis is based on the evaluation of different sets of core power distributions generated thanks to random drawings of the plant state parameters (including power level, core inlet temperature, pressure, control rod insertion and xenon distribution). The sets of core configurations representative of normal plant operation are used to define the surveillance thresholds. The sets representative of accidental transients (for which the LPD and DNBR protections are claimed) are used to define the protection thresholds. The analysis of LOR values provides an envelop probability law covering a minimum of 95% of LOR values. In order to derive the on-site threshold for LPD and DNBR, a Monte Carlo method is used to propagate the LOR probability law and the other uncertainties. Sensitivity calculations have been performed in order to cover a large spectrum of fuel loading patterns and to take into account SPND failures. In conclusion, this approach allows defining an optimized and robust set of thresholds for the on-line surveillance and protection system of EPR™ reactor.

Topics: Instrumentation
Commentary by Dr. Valentin Fuster
2014;():V001T01A027. doi:10.1115/ICONE22-31277.

The large quantities of measurement information gathered throughout a plant process make the closing of the mass and energy balance nearly impossible without the help of additional tools. For this reason, a variety of plant monitoring tools for closing plant balances was developed. A major problem with the current tools lies in the non-consideration of redundant measurements which are available throughout the entire plant process.

The online monitoring reconciliation system is based on the process data reconciliation according to VDI 2048 standard and is using all redundant measurements within the process to close mass and energy balances. As a result, the most realistic process with the lowest uncertainty can be monitored. This system is installed in more than 35 NPPs worldwide and is used

○ as a basis for correction of feed water mass flow and feed water temperature measurements (recover of lost Megawatts).

○ as a basis for correction of Taverage (Tav) (recover of steam generator outlet pressure in PWRs).

○ for maintaining the thermal core power and the feed water mass flow under continuous operation conditions.

○ for automatic detection of erroneous measurements and measurement drift.

○ for detection of inner leakages, non-condensable gases and system losses.

○ for calculating non measured values (e.g. heat transfer coefficients, ΔT, preheater loads,…).

○ as a monitoring system for the main thermodynamic process.

○ for verifying warranty tests more accurate.

○ as a application of condition-based maintenance and component monitoring.

○ for What-if scenarios (simulation, not PDR)

This paper describes the methodology according to VDI 2048 (use of Gaussian correction principle and quality criterias). The benefits gained from the use of the online monitoring system are demonstrated.

Commentary by Dr. Valentin Fuster

Nuclear Fuel and Materials

2014;():V001T02A001. doi:10.1115/ICONE22-30005.

In order to develop a minor actinide (MA) containing MOX (MA-MOX) fuel design method, the analysis models to predict irradiation behavior of MA-MOX fuel have to be developed and the accuracy of the irradiation behavior analysis code should be evaluated with the results of post-irradiation examinations (PIEs) for MA-MOX fuels. In this study, we developed the computer module TRANSIT (Thermal Property and Vapor Pressure Analysis Module for Minor Actinide Containing MOX Fuel) to compute thermal properties of MA-MOX fuel. TRANSIT can give thermal conductivity, melting temperature and vapor pressures of MA-MOX. By using this module, we improved the thermal behavior analysis code DIRAD and developed the DIRAD-TRANSIT code system to compute the irradiation behavior of MA-MOX fuel. This system was verified with the results of PIEs for the conventional MOX fuels and the MA-MOX fuels irradiated in the experimental fast reactor JOYO. As the result of the verification, we determined that the DIRAD-TRANSIT code system would precisely predict the fuel thermal behavior, i.e. fuel temperature and fuel restructuring, for oxide fuels containing several percent minor actinides.

Topics: Fuels
Commentary by Dr. Valentin Fuster
2014;():V001T02A002. doi:10.1115/ICONE22-30022.

This paper presents evaluation methods of fatigue strength of similar and dissimilar welded joints of modified 9Cr-1Mo steel which is a candidate structural material for a demonstration fast breeder reactor being developed in Japan. The discontinuity of mechanical properties across welded joint causes a non-homogeneous strain distribution, and this effect should be taken into account for evaluation of fatigue strength of weld joints. In this study, ‘2-element model’, which is consisted of base metal and weld metal, was employed. Firstly, strain ranges of each element are calculated, and secondly fatigue lives of each element are evaluated. Finally, shorter fatigue life is chosen as fatigue life of the weld joint. Failure position can be also estimated by this model. Evaluation results were compared with experimental data at elevated temperature, and it was shown that they agree well.

Commentary by Dr. Valentin Fuster
2014;():V001T02A003. doi:10.1115/ICONE22-30023.

A three-dimensional finite element model is being developed for a quarter fuel element, which is equivalent to a full fuel element using symmetry. The model uses the Multiphysics Object-Oriented Simulation Environment (MOOSE) framework developed at Idaho National Laboratory. The model facilitates an in-depth investigation into a variety of deformation phenomena for a horizontal nuclear fuel element including bowing, sagging, and stresses and strains.

This paper presents a preliminary analysis of the local stresses and strains of the sheath (clad) at the pellet-to-pellet interfaces for low, normal and high linear powers. During irradiation the fuel pellets thermally expand and take on an hourglass shape. The hourglassing behaviour leads to higher local stresses and strains in the sheath at the locations of the pellet-to-pellet interfaces. The purpose of this work is to quantify these stresses and strains for varying linear powers, and to illustrate the effect that the material model chosen for the cladding has on the results. Preliminary results are presented for two sheath types: elastic, and elastic including diffusional creep. These models are benchmarked against a validated industry code called ELESTRES. The results indicate that the predicted sheath hoop strain is about half of what is determined by ELESTRES in both the elastic and elastic-creep cases. This highlights the requirement of a pellet cracking model in three-dimensional simulations. The elastic-creep model predicts less stress within the sheath than the elastic model as expected.

Topics: Stress , Nuclear fuels
Commentary by Dr. Valentin Fuster
2014;():V001T02A004. doi:10.1115/ICONE22-30119.

The French sodium prototype ASTRID should work from 2026; it must satisfy the requirements of “Generation IV” with regard to lifetime, safety and efficiency; means are sought in order to optimize the ageing behavior of components.

Coatings that seem favorable concerning the fuel claddings and wrappers of ASTRID must be further checked.

Other surface treatments are considered for less irradiated components, in order to improve the residual stresses near the surface of metallic pieces; if the stresses get compressive, these treated components get less brittle and “stress corrosion cracking” (SCC) may be avoided.

French and Chinese establishments have developed “ultrasonic shot peening” using stainless steel balls or needles. Concurrently, US and Japanese companies have carried out “water jet peening” and nanopulse “Laser Peening” treatments; These “laser peening” treatments have already enhanced efficiently SCC and fatigue behavior of both AISI 316 pieces and 12%Cr ferrochromium blades.

Laser treatments with lower power requirements improve also the behaviour of surfaces with regard to SCC, roughness or pitting corrosion.

Passivating chemical techniques are required too in order to stabilize the protective chromine layer at the surface of metallic pieces. Peening and passivating techniques should be combined in the future.

Commentary by Dr. Valentin Fuster
2014;():V001T02A005. doi:10.1115/ICONE22-30218.

In this paper we report computational fluid dynamics and thermal stress finite element analyses of a T-joint component as used in a Swiss nuclear power plant under realistic loading conditions and simplified boundary conditions. In this report the focus is on the thermal stress analyses, and therefore only those fluid dynamics simulations used for thermal stress analyses are presented. The local stress amplitudes, simulated using elastic finite element simulations, are compared to a regulatory fatigue life curve to estimate the local distribution of low cycle fatigue damage and crack initiation probability in the T-joint. The case studies reported here have been selected with the purpose of analyzing the different ways in which the fluid structure interaction induces thermal stress in the mixing tee. In case I only the thermal stress induced by the mixing turbulence is simulated, i.e. with constant mass flow rate boundary conditions. In case II a peak in the flow rate in the main pipe, typical for transient (startup) plant conditions, is addressed and analyzed in detail. Only in the latter case is low cycle fatigue cracking to be expected in the form of non-propagating (i.e. not throughwall) surface cracks or crazing.

Commentary by Dr. Valentin Fuster
2014;():V001T02A006. doi:10.1115/ICONE22-30322.

This work summarizes results of the work carried out in CVR (Centrum výzkumu Rez) on the evaluation of the susceptibility to LME (Liquid Metal Ebrittlement) of the ferritic/martensitic steel T91 in contact with LBE (Lead-Bismuth Eutectic) at 300 and 350°C. The work, carried out in the past few years, through tensile and fracture toughness tests, revealed a marked sensitivity to embrittlement for the T91 in contact with LBE. Tests were carried out in well defined conditions and according to ASTM standards. It was observed that the LBE decreased the fracture toughness, JIC, by more than 30%, compared to the value in air. The results are discussed based on examinations of the fracture surface evidencing LME occurrence. The fracture toughness measured in LBE is proposed to be the appropriate value for fracture resistance evaluation, applied to extension of a pre-existing LME crack.

Commentary by Dr. Valentin Fuster
2014;():V001T02A007. doi:10.1115/ICONE22-30374.

The martensitic lath width (0.83 ± 0.45μm ∼ 0.48 ± 0.14 μm) and dislocation density (1.3 ± 0.3 × 1015 m−2 ∼ 6.4 ± 1.6 ×1015 m−2) change of Super-clean Reduced Activation Martensitic (SCRAM) steel caused by warm deformation on Gleeble-3500 thermo-simulation machine have been examined. The irradiation-induced helium bubbles and hardening were observed in all the specimens after helium implantation to 1e + 17/cm2 at 723 K. The helium bubbles became smaller and more numerous while the distribution was more homogeneous when the lath width decrease and dislocation density increase. The nano-indentation hardness indicated that the sample, the martensitic lath width is 0.83 ± 0.45μm and the dislocation density is 1.3 ± 0.3 × 1015 m−2, exhibited the maximum nano-indentation variation (ΔH) and the ΔH decreased with the lath width decreasing and dislocation density increasing. The hardening occurred in all helium implanted samples can mainly be ascribed to helium bubbles.

Commentary by Dr. Valentin Fuster
2014;():V001T02A008. doi:10.1115/ICONE22-30394.

Li2TiO3 has been recognized as one of the most promising tritium breeding materials for D-T fusion reactor blanket. In this study, ultra-fine Li2TiO3 powder was prepared rapidly by microwave-induced solution combustion synthesis (MSCS) using nitrates of lithium and titanate as raw materials, citric acid as fuel. The as-synthesized Li2TiO3 powder exhibits an average crystalline size as small as 20 nm with uniform distribution. Computer-assisted 3D printing technology was employed to fabricate Li2TiO3 ceramic breeding pebbles. This computer-assisted process is precise, efficient and controllable, which offers an alternative for the mass production of Li2TiO3 pebbles. The pebbles exhibit good sphericity, relatively small grain size, preferable sinterability and crushing load strength.

Commentary by Dr. Valentin Fuster
2014;():V001T02A009. doi:10.1115/ICONE22-30414.

The TerraPower Traveling Wave Reactor (TWR) is a sodium-cooled fast reactor design that utilizes a high-burnup metallic uranium fuel cycle. The fuel system depends on a cladding material with demonstrated swelling resistance to high doses as well as adequate thermal creep strength. HT9 steel is a leading cladding candidate for the first TWR, having demonstrated excellent swelling and strain performance to doses > 200 dpa. A strain model was developed as a design tool to predict fuel pin deformation as a function of irradiation dose, stress, and temperature. The sources of strain deformation will be described along with the uncertainties in utilizing existing data to build a mechanistic model. The strain model is then incorporated into a fuel performance code to provide new insight in deformation behavior of HT9 fuel pins.

Topics: Deformation , Fuels , Modeling
Commentary by Dr. Valentin Fuster
2014;():V001T02A010. doi:10.1115/ICONE22-30499.

Spent fuel storage system of pebble-bed high temperature gas-cooled reactor needs to retrieve fuel elements and spent fuel elements from storage tank during fuel reloading condition and some other special status. This function is to be achieved by the negative pressure suction system. Research in depth is needed towards the negative pressure suction system design and experiment in order to reliably supply the system with enough suction capability and meanwhile prevent the fuel element from over-speed impact damage. A comprehensive experimental facility of negative pressure suction was built to investigate and verify the designed system. The facility mainly comprises a fuel canister, a suction tube, a tube feeder, a gas isolator, a Roots blower and pipelines. The negative pressure suction force was provided by the Roots blower and drove the fuel elements out of the canister through the tube. The Roots blower was driven by a frequency converter so that the suction flow rate could be adjusted as wanted. The dynamics model of spherical element in the tube was established. The pressure drop distribution of the negative pressure suction system was also calculated. Then the pressure drops and the sphere velocity were measured at different air flow rates. Based on the experimental results and calculation analysis, parameter requirements for the Roots pump were concluded. Therefore, fuel elements could be successfully retrieved without over-speed impact damage. These results provide useful experience for engineering design of negative pressure suction system in the spent fuel storage system.

Commentary by Dr. Valentin Fuster
2014;():V001T02A011. doi:10.1115/ICONE22-30588.

This study was carried out to establish an electron beam welding process for nuclear plate-type fuel assembly fabrication. A preliminary investigation for plate fuel fabrication was conducted with a consideration of the weld performance using AA6061-T6 aluminum alloy made through the EBW (Electron Beam Welding) process. The optimum welding parameters for the plate-type fuel assembly were obtained in terms of the accelerating voltage, beam current and welding time. The welds made by the optimum parameters showed slightly lower tensile strengths than those of the un-welded specimens. The integrity of the welds by the EBW process was confirmed by the results of the tensile test, an examination of the macro-cross sections and the fracture surfaces of the welded specimens.

Commentary by Dr. Valentin Fuster
2014;():V001T02A012. doi:10.1115/ICONE22-30641.

Plutonium and oxygen diffusion with the high temperature gradient is one of the important fuel performance concerns in fast reactor (U, Pu)O2 fuel during irradiation, and will affect nuclear fuel materials properties, power distribution and overall performance of the fuel pin. This paper focuses on the plutonium, oxygen and heat diffusion within (U, Pu)O2 fast reactor fuel pellets. In this study, the correlations from the literature are used for thermal conductivity, specific heat, density, plutonium diffusion and oxygen diffusion. Three dimensional burnup dependent oxygen diffusion, plutonium diffusion and heat diffusion models are fully-coupled in steady states and transients to account for the effects on each other. The models are implemented into COMSOL Multiphysics to perform this analysis.

Commentary by Dr. Valentin Fuster
2014;():V001T02A013. doi:10.1115/ICONE22-30647.

An enhanced thermal conductivity UO2-BeO composite nuclear fuel was studied. A methodology to generate ANSYS (an engineering simulation software) FEM (Finite Element Method) thermal models of enhanced thermal conductivity oxide nuclear fuels was developed. Two fabrication methods to produce high thermal conductivity UO2-BeO oxide nuclear fuels were summarized. These two processing routes generated pellets with two different microstructures. The characteristics and microstructures of the fuel are determined for use in FEM thermal models, and the relevant thermal properties for UO2-BeO fuels by two different fabrication methods were determined. The results showed significant increase in the fuel thermal conductivities. The reactor performance analysis showed that the decrease in centerline temperature was 250–350K for the UO2-BeO composite fuel, and thus we can improve nuclear reactors’ performance and safety, and high-level radioactive waste generation.

Commentary by Dr. Valentin Fuster
2014;():V001T02A014. doi:10.1115/ICONE22-30665.

In order to fulfil requirements for corrosion resistance for new reactor GIV, the austenitic 304L stainless steel and 18Cr-20Mn austenitic steel were improved by oxide dispersion strengthening (ODS), using two nano-oxide types: titanium and yttrium oxides.

Two new ODS steels and a reference material, A/SA-270 grade 304L SS as plate, were characterised by different techniques and its behaviour in SCWR environment was considered. Coolant compatibility studies have been performed in demineralised water at supercritical conditions: temperature of about 550°C and 25 MPa pressure. The oxide developed on the 304ODS samples is layered, thicker and more uniform than on 304L SS. Some oxides grown on 18Cr-20MnODS steel are un-adherently and they are lost in the simulated water coolant.

The weight gains of ODS samples are positive and higher than 304L SS up to approximately 1320 hours while on 18Cr-20MnODS steel is negative. The oxide films were investigated by SEM and EDS techniques.

Commentary by Dr. Valentin Fuster
2014;():V001T02A015. doi:10.1115/ICONE22-30781.

Titanium stabilized 1.4970 ‘15-15Ti’ stainless steel cladding is the primary choice for fuel cladding of several current fast spectrum research reactor projects. The choice of cladding material is based on past experiences and the availability of material databases from similar steel grades that have proven their reliability in past sodium-cooled fast reactors programs. However the last production in Europe of nuclear-grade 15-15Ti was more than 20 years ago and it remained to be seen if the know-how to produce such steel with the strict specifications for nuclear fuel cladding was still available. Results of a new production of nuclear-grade 15-15Ti cladding tubes at Sandvik for SCK•CEN is presented in this paper. It is shown that materials properties are within the strict specifications similar to the ones used during past sodium-cooled fast reactors programs. Special attention is given to microstructural analysis of the newly produced steel which contains a large number of stabilizing Ti(C-N) precipitates known for their beneficial effect on in-pile material properties and thermal creep. Results from metallography, SEM and TEM investigations are presented.

Commentary by Dr. Valentin Fuster
2014;():V001T02A016. doi:10.1115/ICONE22-30807.

The power increase rate of the reactor is often derived using the fuel performance code. Too restrictive rates are not desirable since they lead to the loss of production. On the other hand fast increase or not well controlled axial power oscillation may result in the rod failure due to pellet-cladding interaction.

Most of the currently used fuel performance codes treat the stack of the fuel pellets using a simplified “1.5D” approach where individual pellets are not distinguished and the fuel stack is taken to be symmetrical. In reality, several effects must be taken into account when more accurate description is required. These local effects include contact at pellet-pellet interface, fuel pellet cracking under thermal stress, fabrication defects pellets or azimuthal asymmetry in the heat generation or heat transfer conditions due to rod bowing or presence of control rods.

Detailed models of the local phenomena are therefore being developed at ÚJV Řež using the ABAQUS 6.12 code and used to improve the predictions of the codes routinely used for the core design assessment. For example the impact of the use of the advanced pellet materials on the peak loading that the cladding will experience during the power ramp has been quantified.

Commentary by Dr. Valentin Fuster
2014;():V001T02A017. doi:10.1115/ICONE22-30839.

The use of heavy liquid metals (HLM), such as Lead, Pb, is being considered as a coolant in Generation IV fast Reactors. However, structural materials suffer significant damage when in contact with HLMs. Both austenitic and ferritic-martensitic steels are considered and are susceptible to corrosion and/or degradation of mechanical properties. One of the approaches to mitigate erosion/corrosion problems is the use of coatings. This is a layer of a different chemical composition between the base material and the liquid metal coolant in order to prevent the bulk metal from corrosive effects. Coatings are proposed as a valid protection against high temperature damage in this environment. Their capability to grow more stable and protective oxides, by introducing the oxide forming elements (e.g. Al, Si,…) in higher amount, is proven to be an effective alternative to material engineering.

For this specific application, several coatings deposition techniques and compositions have been proposed and tested. In this work the High Velocity Oxygen Fuel, HVOF, combined with laser melting was selected for deposition of FeCrAlY coatings. The combination of the two technologies lead to a compact and adherent coating with an enriched content of Al. The method was evaluated in terms of the corrosion resistance of the coating and also its effect on the microstructure of the substrate alloy. Several attempts were carried out to modify spraying and laser parameters in order to minimise the effect on the substrate and keep the protective properties.

Results are discussed in terms of deposition parameters and protection characteristics.

Topics: Coatings
Commentary by Dr. Valentin Fuster
2014;():V001T02A018. doi:10.1115/ICONE22-30848.

Zirconium is important because of its mechanical and neutronics properties combined with its extraordinary resistance against corrosion. It’s mainly usage in nuclear engineering is as a nuclear fuel cladding. Even though Zr layers are very resistant to neutron fluence at reactor operating temperatures, high pressure and high radiation doses, exothermic Zr oxidation in accident scenario still occurs. Hydrogen, and a newly created layer of zirconium dioxide, are undesirable products of the high temperature steam oxidation. ZrO2 adopts a monoclinic crystal structure at room temperature and transitions to tetragonal and cubic at higher temperatures. The volume expansion caused by the cubic to tetragonal to monoclinic transformation induces large stresses and these stresses cause ZrO2 to crack upon cooling from high temperatures. To stabilize ZrO2 other oxides may be added. Therefore ZrO2 with its high ionic conductivity makes it one of the most promising ceramics to protect fuel cladding and the paper focuses on properties of ZrO2 used on Zr fuel cladding of common power reactors.

Commentary by Dr. Valentin Fuster
2014;():V001T02A019. doi:10.1115/ICONE22-30859.

Ultrasonic tests were conducted for 304 type austenitic stainless steels with different annealing conditions and effects of carbide precipitate formation on the velocity changes were evaluated. The velocity increased with higher annealing temperature and/or longer annealing time. SEM observations indicated that carbide precipitates were formed mainly on grain boundaries. Results show that it is not the precipitation itself but the removal of carbon from the matrix that determines the velocity change.

Topics: Stainless steel
Commentary by Dr. Valentin Fuster
2014;():V001T02A020. doi:10.1115/ICONE22-30870.

Research of nuclear reactor fuel depletion aims at development and introduction of advanced types of burnable absorbers (BA) applied within nuclear fuel. BAs compensate for the initial reactivity excess and consequently may allow for lower power peaking factors and longer fuel cycles with higher fuel enrichments.

Modern computer codes for nuclear fuel depletion calculation require a substantial amount of computational time. Therefore, any parametric calculations for BA selection need to be carried out only with a fast depletion code. The main purpose of the newly developed UWB1 code is a rapid calculation of nuclear fuel depletion, which is achieved by the approximations in the equations describing transport part of fuel depletion. Microscopic cross sections are assumed to be constant through depletion calculation steps.

The paper describes the first step of analysis using a new version of UWB1, that was accomplished with the assumption of uniformly distributed BA in fuel. BA elements, nuclides and nuclide mixtures were compared and their performance was consequently evaluated based on multiplication coefficient behavior during depletion.

Commentary by Dr. Valentin Fuster
2014;():V001T02A021. doi:10.1115/ICONE22-30873.

Polycrystalline diamond coating is a promising possibility for prevention, or reduction of high temperature oxidation of zirconium alloys. Zirconium alloys are used as cladding material in almost all types of nuclear reactors, where creates a barrier between nuclear fuel and cooling water in the primary circuit. Hydrogen and considerable amount of heat is released during steam oxidation that may occur in an eventual accident. In this paper Zircaloy-2 alloy was covered by polycrystalline diamond layer using Plasma Enhanced Linear Antennas Microwave Chemical Vapor Deposition system reactor. X-Ray Diffraction and Raman spectroscopy measurements confirmed coverage of the surface area with crystalline and amorphous carbon layer. Characterizations were done for zirconium alloy covered with diamond layer before and after corrosion and irradiation tests - ion beam irradiation tests and high temperature steam exposure.

Commentary by Dr. Valentin Fuster
2014;():V001T02A022. doi:10.1115/ICONE22-30894.

Research of fuel depletion aims at development and introduction of advanced types of burnable absorbers (BA) in nuclear fuel. BAs compensate for the initial reactivity excess and consequently allow for lower power peaking factors and longer fuel cycles with higher fuel enrichments.

The paper describes the comparison of selected BA elements (Gd, Eu, Er) for 3 different nuclear fuel types (VVER, PWR and SFR). The effect of harder VVER spectra and the effect of thermal vs fast neutron spectra is evaluated and commented. Uniformly distributed BAs in the fuel were assumed.

Comparison calculations were performed with the newly developed UWB1 fast fuel depletion code and with the SERPENT Monte Carlo code. The comparison of UWB1 and SERPENT codes is included in the analysis. Next steps of improving the UWB1 code are suggested.

Commentary by Dr. Valentin Fuster
2014;():V001T02A023. doi:10.1115/ICONE22-30952.

During its lifetime in the core of a nuclear reactor, the fuel undergoes significant changes in its physical, chemical and morphological characteristics. In outer regions of the fuel pellets the so called “high burn-up” or “rim” structure may form. In this region UO2 grains, with an original size of about 10 μm in fresh dense fuel, are reorganized into a porous structure with grain size 0.1–0.3 μm and porosity fraction up to 20 %. The mechanical, thermal and fission product retention properties of the high burn-up structure have encouraged further interest and attempts to mimic this morphology in fresh fuel. The JRC-ITU has studied various techniques for the synthesis of uranium and thorium dioxide in aqueous or nonaqueous media. Such nanoparticles can serve as starting material for production of material having similar characteristics as the high burn-up structure, as has been proved using Zr(Y)O2 nanoparticles. Recently, efforts have been focused on the compaction of the nanoparticle powders. A spark plasma sintering device (SPS, FCT Systeme GmbH) has been commissioned in the JRC-ITU and tested using various nonradioactive materials. The present study is oriented on pressing and sintering of Hf(Y)O2 and ZrO2 into nanostructured pellets. Final products have been characterized by optical and electron microscopy, X-ray powder diffraction and density measurements.

Topics: Fuels
Commentary by Dr. Valentin Fuster
2014;():V001T02A024. doi:10.1115/ICONE22-30954.

Primary Water Stress Corrosion Cracking (PWSCC) is one of important ageing issues in PWR (Pressurized Water Reactor) primary systems. It has been pointed out that high concentration dissolved hydrogen may lead to occurrence of PWSCC. The authors have proposed to substitute hydrogen by methanol as a fundamental countermeasure of PWSCC. So far corrosion tests of stainless steels and Zircaloy-4 in methanol solutions at 320 °C were conducted under γ-ray irradiation and without irradiation. The test results show that methanol is promising.

In the present paper, γ-ray irradiation experiments of methanol solution at 320 °C were done up to 100 kGy. A study on the radiolysis of methanol solution is important from two aspects. One concerns corrosion of structural materials. The radiolysis of methanol may result in formation of harmful compounds to the structural materials, such as carboxylic acids. It is necessary to know the yields of such compounds. The other concern is possible polymerization of methanol and formation of organic polymer deposit on fuel claddings. Large amount of the deposit on fuel claddings should be avoided to keep integrity of fuel claddings. Therefore, it should be clarified whether gaseous species are major products and whether polymerized species of methanol such as ethylene glycol is formed.

After the γ-ray irradiation of methanol solution, following species were analyzed: CO2 and H2, methanol, formaldehyde, formate and acetate, and ethylene glycol and glycerin.

Without γ-ray irradiation, the major process of the thermal decomposition of methanol at 320 °C is oxidation of methanol by water and generation of one CO2 molecule and three H2 molecules. Under γ-ray irradiation, the decomposition of methanol is accelerated; little methanol remains after 10 kGy irradiation. The major product is CO2, and polymerization of methanol unlikely occurs. After methanol is completely decomposed, the hydrogen yield still increases. The reducing environment is maintained. Probably, transient organic species play important roles. The addition of low concentration methanol may be sufficient to maintain reducing environment of the PWR primary systems.

Commentary by Dr. Valentin Fuster
2014;():V001T02A025. doi:10.1115/ICONE22-30990.

This paper updates scientific bases of water chemistry in applying the author’s recent theory, which integrates the elemental radiation- and electro-chemistry reactions in the “Butlar-Volmer equation,” presented in ICONE21-16525.

For the past several years the author has been trying to establish that the “long-cell” (a kin to macro-cell) corrosion mechanism is inducing practically all sorts of accelerated corrosion phenomena widely observed in water-cooled reactors, especially in aged plants.

The theoretical electrochemical potential differences have been benchmarked with the published in-pile test results for both PWR- and BWR water chemistry environments. However the author’s previous verification efforts were limited to the extent that the curves were fitted with experimental results at a single point. The author re-formulated the basic theory and found that the redox potential difference consists of an electrochemical part (e.g., Nernst equation of dissolved hydrogen or oxygen) and radiation-induced perturbation term, the latter diminishes to zero without radiation.

The author continued his studies to clarify whether our current scientific knowledge is sufficient to explain the in-core “chemistry” to reproduce the experimental results without the fitting parameter. Through his study he realized that the basic mechanism of the potential difference is still not sufficiently known.

No fitting parameter was used for the PWR water chemistry in the DH region for practical engineering applications, although it is indispensable to confirm the results with an in-pile test loop. In the BWR-NWC the theoretical redox potential out of core was still necessary to be fitted with the experimental results, due to an effect of residual hydrogen peroxide detected by the reference electrode. In addition the calculated potential shift is several times larger than the experimental observation. With the reformulation the scientific validity of the author’s theory is further confirmed. He believes that there is no doubt that the “long-cell” takes place in LWRs, although details are still debatable.

Commentary by Dr. Valentin Fuster
2014;():V001T02A026. doi:10.1115/ICONE22-30991.

Although the water radiolysis, decomposition of water by radiation, is a well-known phenomenon the exact mechanism is not well characterized especially for severe accidents. The author first reviewed the water radiolysis phenomena in LWRs during normal operation to severe accidents (e.g., TMI- and Chernobyl accidents) and performed a scoping estimation of the amount of radiological hydrogen generation, accumulation and release for the Fukushima Daiichi accident. The estimation incorporates the decay heat curve after a reactor trip combined with G-values. As much as 450 cubic meters-STP of accumulated hydrogen gas is estimated to be located inside the PCV just prior to the hydrogen explosion which occurred a day after the reactor trip in Unit 1.

When a set of radiological chain reactions are incorporated the resultant reverse reactions substantially reduce the hydrogen generation, even when removal of molecular products (i.e., oxygen and hydrogen) is assumed stripped rapidly from boiling water through bubbles. Even in the most favorable configuration a typical amount of hydrogen gas reduces to only several tens of cubic meters.

Finally, the author tested a new mechanism, “radiation-induced electrolysis,” which had been applied to his corrosion studies for last several years. His theory has been verified with the published in-pile test data, although he has never tried to apply this to his severe accident study. The predicted results indicated that the total inventory of hydrogen gas inside RPV may reach as much as 1000 cubic meters in just 3 hours during the SBO due to a high decay heat soon after the reactor trip through this process.

Commentary by Dr. Valentin Fuster
2014;():V001T02A027. doi:10.1115/ICONE22-31108.

The main goals of fuel development for pressurized water reactor are effectiveness and economic efficiency. Both requirements can be achieved by gradual increase of discharged fuel burn-up and prolongation of fuel cycle. The mentioned effects can be reached by optimisation of fuel assembly profiling, fuel enrichment raise, and by parasitic absorption reduction. These methods were used in VVER-440 fuel assembly optimisation, described in this paper. Fuel pin configurations with enrichment limit 5 % and also enlarged one up to 5.95 % U235 were designed. Reduction of parasitic absorption was limited by carcass frame of the assembly. Basic characteristics of the best assembly proposals are presented and effects on equilibrium fuel cycle of VVER-440 reactor are characterized.

Topics: Fuels , Manufacturing
Commentary by Dr. Valentin Fuster
2014;():V001T02A028. doi:10.1115/ICONE22-31244.

In the traditional closed fuel cycle, based on REMIX-technology (REgenerated MIXture of U and Pu oxides) the fuel composition is produced on the basis of a uranium and plutonium mixture from spent Light Water Reactor (LWR) fuel and additional natural uranium. In this case, there is some saving in the amount of natural uranium used.

The basic features of the WWER-1000 fuel loadings with a new variant REMIX-fuel during multiple recycle in the closed nuclear fuel cycle are described in this paper. Such fuel compositions are produced on a basis of a uranium and plutonium mixture allocated at processing the spent fuel after irradiation in the WWER-1000 core, depleted uranium and fission material such as: 235U as a part of high-enriched uranium from the warheads superfluous for defense.

Also here variants are considered of the perspective closed fuel cycle in which fissile feed materials for fuel manufacture is produced in the blankets of fast breeder reactors. The fissile material is 233U or Pu. The raw material is depleted uranium from the stocks of enrichment factories, or thorium. Natural uranium is not used in this case. The minimum feed material required for the REMIX technology in a closed fuel cycle was determined through calculations of different types of fissile and raw materials, with different cycle lengths and fuel-water ratios.

Commentary by Dr. Valentin Fuster
2014;():V001T02A029. doi:10.1115/ICONE22-31265.

Corrosion products deposit on Pressurized Water Reactor (PWR) fuel rod cladding surfaces and can create a number of issues including increased cladding temperature, elevated cladding corrosion, and the precipitation of boron species within the deposits. The deposits can also release and lead to increased radiation fields on system ex-core surfaces. These effects can vary widely from plant-to-plant. The amount of the deposits, commonly known as crud, is an important factor in determining the impact, but other parameters such as crud thickness, porosity, and composition are also thought to be important.

The Electric Power Research Institute (EPRI) has sponsored a number of programs to better understand the characteristics of crud and its effects. Crud has been sampled by fuel scraping and by collecting suspended crud during operation and during fuel cleaning. The chemistry and structure of the crud was then characterized. These data were then used to create simulated crud in laboratory heated rod tests. These tests explored how the crud deposits affected heat transfer at the rod surface and the interaction between the crud and the simulated coolant.

This paper discusses the nature of PWR crud and some of the practical aspects of crud simulation. Different approaches to laboratory crud creation will be reviewed, and the success in matching plant crud characteristics will be shown, with special emphasis on the production of crud for thermal conductivity measurement.

Commentary by Dr. Valentin Fuster

Plant Systems, Structures and Components

2014;():V001T03A001. doi:10.1115/ICONE22-30053.

As an essential component in the system of the HTR-10 helium turbine generator electromagnetic bearings, auxiliary bearings provide mechanical backup protection in case of the events of magnetic bearing failure happen. When contact events happen, highly localized and transient temperatures will arise from frictional heating over the dynamically varying contact area in the very short term and dissipate through the bearing in the longer term. When excessive temperature level occurs, rapid failure may be anticipated, thus it will become a serious threat to the safety of the HTR. This paper presents a detailed analysis of thermal growths due to the mechanical rub for a rotor drop on auxiliary bearings. With the aim to numerically analyze the heat generation and temperature rise, a 1D thermal model of the ball bearing composed of heat transfer network and heat sources based on heat transfer equations is established. The Matlab codes are developed to complete the numerical analysis, and an infrared method is utilized to investigate the temperature rise at the rotor/inner race contact surface. By the comparison between simulation results and the experimental data, this paper illustrates the thermal growth during a rotor drop process, which is highly non-linear. The results reveal that the axial contact force is critical to the bearing heat generation, and the ceramic balls with superior thermal properties are recommended.

Commentary by Dr. Valentin Fuster
2014;():V001T03A002. doi:10.1115/ICONE22-30058.

Magnetic bearings are widely applied in High Temperature Gas-cooled Reactor (HTGR) and auxiliary bearings are important backup and safety components in AMB systems. The dynamic analysis of the AMB rotors touchdown process is an important foundation for designing auxiliary bearings. In this paper, a data-based dynamic analysis of the touchdown process is proposed. The dynamic model of the touchdown process is firstly established and then the nonlinear extended Kalman filtering technique is applied. Based on the dynamic model and Kalman filtering technique, the proposed method can offer estimations of rotor’s displacements, velocities and accelerations from noisy observations. The proposed method is validated by the experiment data from touchdown experiments. The touchdown experiments are performed on an experimental system with a 440kg heavy rotor, the rotational speed in the experiments is 5000RPM and no brake is applied.

Commentary by Dr. Valentin Fuster
2014;():V001T03A003. doi:10.1115/ICONE22-30062.

The active magnetic bearing (AMB) system is a crucial part in the helium circulator system of the 10MW high temperature gas-cooled reactor (HTR-10). Though the AMB has been widely used in industrial fields, it is still limited in the research of the dynamic behavior of AMB’s vertical arranged rotor with axial magnetic load during its drop process. This paper establishes the dynamic model of such drop process by Matlab. Meanwhile using the Hertz contact theory establishes the contact model of different configurations. Analyze the axial friction between the rotor and thrust interface of the inner ring of Auxiliary Bearing System (ABS). Besides, the numerical model is verified by the drop experiment with the axial magnetic force. Moreover, this paper analyzes the influence of the rotor’s drop rotational frequency and the axial bracing features including stiffness and damping on the dynamic behavior during vertical arranged rotor’s drop process. Moreover, the paper provides the optimal axial stiffness and damping for the ABS satisfying the experimental conditions so as to reduce the contact force. Such results provide important references to the design of the ABS with a vertical arranged rotor and its application in HTR-10 and High Temperature Reactor-Pebblebed Modules (HTR-PM).

Topics: Rotors
Commentary by Dr. Valentin Fuster
2014;():V001T03A004. doi:10.1115/ICONE22-30077.

HTGR, short for high temperature gas cooled reactor, has gained a lot of attention in nuclear industry. Gas helium, 7MPa in pressure, is used as primary coolant of HTR-PM in where there are a lot of electrical equipment. Insulating property of helium is worse than that of air according to Paschen curves and there are very few articles or related standards about insulating property of high pressure gas helium, which makes the electrical equipment structure design lack of basis. In this study, an experimental platform for testing insulating performance is designed, based on which the experiments of testing the withstanding voltages of penetration assemblies and the breakdown voltages of parallel plane electrodes at different pressures are carried out. The results show that for both the penetration assemblies and the parallel plane, their breakdown voltages in helium are far lower than in air under the same condition of 15°C /0.1MPa. For the penetration assemblies, their insulating properties in helium at 150°C/7MPa are better than those in air at 15°C/0.1MPa.

Commentary by Dr. Valentin Fuster
2014;():V001T03A005. doi:10.1115/ICONE22-30085.

The active magnetic bearing (AMB) is a new kind of high-performance bearing which suspends the rotor with controlled electromagnetic force. It was chosen to support the rotor of the helium blower in HTR-PM instead of conventional bearings. The power losses in the active magnetic bearings compose of three components: copper loss, iron loss and windage loss. In this paper, the iron loss, which composes of the eddy current loss and the hysteresis loss, is researched. The power loss of silicon steel lamination (35H300) was measured. Experimental data was taken over a range of 50Hz to 25,000Hz (sinusoidal current) for several magnetic field intensities. According to the experimental data, the eddy current loss and hysteresis loss increase with the frequency. And the hysteresis loss in the silicon steel lamination occupies the major part when the frequency of current is low, however the growth rate of eddy current is much faster than that of the hysteresis loss. And the FEM calculation of power loss in the magnetic bearing, which rotor and stator are made from silicon steel lamination (35H300), is also presented. The result shows the core loss of magnetic bearing also follow the separation theory. We can separate the core loss of magnetic bearing into two parts: hysteresis loss and eddy current loss. It will be very useful to calculate the power loss in the magnetic bearing.

Commentary by Dr. Valentin Fuster
2014;():V001T03A006. doi:10.1115/ICONE22-30123.

The eddy effect of the magnetic thrust bearing (MTB) is researched by the finite element analysis (FEA). The active magnetic bearing (AMB) is an advanced bearing, used in HTR-PM. The alternating current in the bearing windings will decrease the electromagnetic force and cause phase lag, especially in the MTB which has no lamination structure. According to the calculation in this paper, simple sinusoidal current has large eddy effect. The force decreases obviously and the phase lag is large. However, the current containing direct part and sinusoidal part, which is closer to actual current, has less eddy effect. That is to say, because of the direct part in the current, the eddy effect of the sinusoidal part decreases.

Commentary by Dr. Valentin Fuster
2014;():V001T03A007. doi:10.1115/ICONE22-30141.

The purpose of this study is to identify hydrodynamic loads due to steam condensation acting on the Suppression Pool (S/P) in Primary Containment Vessel (PCV) of Advanced-Boiling Water Reactor (ABWR) by using general-purpose analysis code, ANSYS™. The source load methodology has been used to evaluate the hydrodynamic loads, which are classified into condensation oscillation (CO) and chugging (CH). When setting the design source from the confirmation test, the calculation method due to the eigenfunction of the cylindrical coordinate system was used. Since, there were various limitations in the previous approaches, a new analysis approach has been expected. In this study, the pool of ABWR horizontal-vent confirmatory test facility is modeled with the ANSYS™ acoustic elements. The calculation results are in good agreement with the test pressure oscillations. It is confirmed that the proposed approach can create the design source enveloping the PSD of test results.

Commentary by Dr. Valentin Fuster
2014;():V001T03A008. doi:10.1115/ICONE22-30240.

As a result of catastrophic events on the nuclear power plant “Fukushima” the European organizations on regulation of nuclear power (ENSREG) initiated wide-scale measures for complex designs revision of already operating and under construction European and Russian NPPs. Inspection was made about resistance of power units to external influences of the natural character, being accompanied by multiple failures of safety systems.

Within these works stress tests for constructed power units of LAES-2 and the Baltic NPP were executed. The structure of these checks included the settlement analysis of a condition of NPP at accident with loss of all AC power supply sources which results are presented in report materials.

Accident calculations with a full blackout were executed on the best-estimated heat-hydraulics code KORSAR/GP for justification of power unit preservations in the intact condition within 72 hours from the accident beginning by means of SG PHRS.

The system is developed for feed of the SG PHRS tanks and the fuel pool for working capacity extension the SG PHRS and power unit preservation in a stable condition more than 72 hours from the accident beginning.

Use of system for feed of tanks the SG PHRS and the fuel pool allows to increase significantly resistance of the NPP to external influences of the natural character and to increase time of preservation of the blackout power unit in a stable condition more than 5 days.

Commentary by Dr. Valentin Fuster
2014;():V001T03A009. doi:10.1115/ICONE22-30251.

Almost all industrial products are assembled from multiple parts, and this is true for all sizes of products. As an example, a nuclear facility is a large structure consisting of more than 10 million components. This paper discusses a method to analyze an assembly by gathering data on its component parts. Gathered data on component may identify ill conditioned meshes for connecting surfaces between components. These ill meshes are typified by nodal point disagreement in finite element discretization. A technique to resolve inconsistencies in data among the components is developed. By using this technique, structural analysis for an assembly can be carried out, and results can be obtained by the use of supercomputers, such as the K computer. Numerical results are discussed for components of the High Temperature Engineering Test Reactor of the Japan Atomic Energy Agency.

Commentary by Dr. Valentin Fuster
2014;():V001T03A010. doi:10.1115/ICONE22-30341.

Advanced PWR nuclear power plant has very-large steam generators of which tube-sheet is diameter 5000mm / thickness 500 mm / weight 100 ton or larger. Drilling of very thick tube sheet needs deep-hole drilling technique with strict dimensional tolerance control. In addition, the tube sheet has 10,000 or more tube holes, which means that stable drilling operation is mandatory for productivity control. Upon such background, steam generator tube sheet drilling operation needs precise dimensional control technique with highly efficient productivity.

In this study, authors did (1) feasibility-study and development of deep hole drilling machine with the Boring and Trepanning Association (BTA) method, (2) cutting parameter evaluation such as feed rate and cutting velocity, (3) tube hole measurement system development, and (4) drill exchange program development with monitoring drilling machine motor loads and setting alarm level for the loads.

Based on techniques developed by this study, IHI has achieved high-accuracy and stable deep-hole drilling technology for very-thick tube sheets applicable to the very-large steam generators.

Topics: Drilling , Boilers
Commentary by Dr. Valentin Fuster
2014;():V001T03A011. doi:10.1115/ICONE22-30356.

Reactor coolant pump (RCP) is designed for the heat transfer of heat which is generated from reactor vessel to steam generators by circulating the coolant water. RCP is the only rotating equipment in the nuclear steam supply system (NSSS). Therefore, the problem of vibration has arisen caused by the hydraulic forces of the working fluid. These forces can drastically alter the critical speeds and stability characteristics and can act as significant destabilizing forces. So, vibration evaluation of RCP has been considered as a very important issue [1]. Among them, unbalance response caused by weight of unbalancing of rotating shaft could have serious effects on the entire rotor system. Thus, precise unbalance response spectrum analyses are required. In general, in order to evaluate the unbalance response characteristics for centrifugal pump, finite element analysis was performed according to the ISO 1940-1 standard. However, finite element analysis according to the ISO 1940-1 standard does not considering fluid flow effect. So, finite element analysis result and experimental results may be some differences. Vibration characteristics of RCP has affected by fluid flow effect induced from working fluid. Therefore, in order to understand vibration characteristics for the RCP shaft assembly considered in actual operating condition, rotor dynamic analysis should be performed considering the fluid flow effect. In this research, owing to accurately evaluate the vibration characteristics for the RCP considering hydro forces due to the fluid flow, we measured the bearing force and moment take into account the fluid-induced force. And then response spectrum analysis of RCP shaft assembly was performed considering fluid induced bearing radial forces which are measured values. Lastly, evaluate the vibration characteristics considering effect of fluid flow according to the number of revolution.

Commentary by Dr. Valentin Fuster
2014;():V001T03A012. doi:10.1115/ICONE22-30359.

The flywheel of latest coolant pump provides high inertia to ensure a slow decrease in coolant flow to prevent fuel damage after the loss of power. Flywheel comprises a hub, twelve tungsten alloy blocks and a retainer ring shrink-fit assembled on the outer surface of blocks. In the structural integrity analysis, the shrinkage load due to shrink-fit and the centrifugal load due to rotation are considered, so the wall thickness of retainer ring and the magnitude of shrink-fit are key variables. In particular, these variables will change the flywheel running state. This paper considers the influence of these variables, we employ Latin hypercube design to obtain the response surface model and analyze the influence of these variables. Finally we obtain the magnitude of wall thickness of retainer ring and the range of shrink-fit.

Commentary by Dr. Valentin Fuster
2014;():V001T03A013. doi:10.1115/ICONE22-30367.

The stator and rotor cans in canned motor reactor coolant pump are assumed to be elastic coaxial cylindrical shells due to their particular geometric structures in present study. Thin shell structures such as cans are prone to buckling instabilities. Furthermore, a lot of accidents were caused by losing stability. The dynamic behavior of coaxial circular cylindrical shells subjected to axial fluid flow in the annular gap between two shells is investigated in this paper. The outer shell is stiffened by ring-ribs because of its instability easily. The shell is modeled based on Donnell’s shallow theory. The “smeared stiffeners” approach is used for ring-stiffeners. The fluid is assumed to be an incompressible ideal fluid and the potential flow theory is employed to describe shell-fluid interaction. Numerical analyses are conducted by means of energy variation to obtain the critical flow velocity of losing stability with aid of Hamilton principle. This study shows effects of geometrical parameters on stability of shells. The size and number of ring-stiffeners on dynamic stability are examined. It is found that stiffeners can vary modes instability and enhance the stability of shells. The flow velocities of losing stability with different boundary conductions can be calculated and compared. The results show clamped shells are more stable than simply supported shells. The results presented are in reasonable agreement with those available in the literature.

Commentary by Dr. Valentin Fuster
2014;():V001T03A014. doi:10.1115/ICONE22-30382.

In this work the structural reliability of the circumferentially cracked core support mount of Monju Fast Breeder Reactor (FBR) is analyzed using Finite Element Analysis (FEA). The 3D shell model employed was derived after detailed evaluation of the core support mount behavior with a specific 3D solid model. First, elastoplastic static analysis results show that, under nominal operating conditions, the overall structure would be able to survive a total loss of the core support mount. Second, using the double elastic slope method it was inferred that earthquake loading integrity could be warranted up to a crack representing more than 50% of the total circumference. Both results highlight the ample primary loading margins taken in the design of Monju’s reactor core support structures. Furthermore, the developed 3D shell FEA model will be applied to study other extreme cases such as those under severe accident conditions.

Commentary by Dr. Valentin Fuster
2014;():V001T03A015. doi:10.1115/ICONE22-30385.

In order to gain the best use of filtered containment venting systems (FCVSs), the decomtamination factor of FCVSs is to be investigated as a function of system parameter including steam flow rate, pressure, temperature, water level, and operating time. A full-height test facilities were designed and constructed in Central Research Institute of Electric Power Industry (CRIEPI), Japan to evaluate the decontamination factor (DF) in FCVSs. The target types are the orifice and the venturi FCVSs. The height and the internal diameter of the cylindrical test vessel is 8 m and 0.5 m. Bubbly flows were visualized through the view window up to 0.8 MPa and 170 °C. Steam bubbles in 0.2 wt% sodium thiosulfate and 0.5 wt% sodium hydroxide were found to be much smaller than those in water. The DF were evaluated for the aerosol, elemental iodine and organic iodine. The installed aerosol optical spectrometer measures the number density and the diameter of aerosols. The concentrations of elemental iodine were quantified with an inductively-coupled plasma with mass spectrometry (ICP-MS). The concentration of organic iodine was quantified with a gas chromatography with mass spectrometry (GC-MS). In order to investigate two-phase flow dynamics in the vessel, separate effect tests were conducted with air-water test facility. The height of cylindrical test vessel is 8 m. Visual observation was conducted for two internal diameter levels: 0.05 and 0.5 m. High speed video frames were recorded through the transparent (acrylic) vessel wall. Wire-Mesh Sensors (WMS) were installed to acquire a cross-sectional void fraction to compare with DF in the facility. On the basis of the obtained database, we develop the FCVSs performance evaluation technique and propose an optimal FCVSs operation method for a further safety improvements of the nuclear power plant.

Commentary by Dr. Valentin Fuster
2014;():V001T03A016. doi:10.1115/ICONE22-30446.

Metallic insulation is commonly used in reactor vessel because of its resistance to radiation and corrosion. Since the main mode of heat loss of reactor vessel is thermal radiation, the ability to prevent radiation heat transfer is important for metallic insulation. But the thermal conductivity of metallic insulation is difficult to calculate owing to their complex geometry. This article uses FLUENT 14.0 to obtain the important parameter “view factor”, and then develops a computational model of effective conductivity of metallic insulation. Heat transfer test of metallic insulation was done, and the numerical simulation of metallic insulation was also performed. Based on results of test and simulation, the computational model is modified. The modified model can fit the test result better. Based on the modified model, the effective conductivity of metallic insulation increases with the increase of temperature of hot side and cold side, among which the temperature of hot side influences more. And when the temperature is high, the effective conductivity increases much faster.

Commentary by Dr. Valentin Fuster
2014;():V001T03A017. doi:10.1115/ICONE22-30502.

One water-lubricated tilting-pad radial bearing is studied in this paper to learn the effect of temperature-viscosity effect caused by the water friction loss on the surface of the pads on the carrying capacity and dynamic characteristics of the bearing. The Reynolds equation considering the turbulent stress term and the energy equation are established based on a simplified model of the tilting-pad radial bearing. The influence of the viscosity change to the dynamic characteristics of the bearing is analyzed while considering the temperature rise in case of water friction loss.

Result shows that the temperature and water friction loss will increase in a large degree when shaft neck eccentricity is large. The carrying capacity of the bearing will decrease when the temperature increases. However, the rise of the temperature will not change the stiffness and damping coefficients. The maximal rise of the temperature reaches 293.15K (20 °C) when shaft neck eccentricity angle is 180°and eccentricity ratio is 0.9 leading to the reduction of bearing capacity by 8%. Thus, to maintain the stability and avoid the obvious temperature rise by water friction, the load force upon the shaft should be less than 280kN when the surrounding temperature is 353.15K (80°C) and shaft neck eccentricity angle is 180° and eccentricity ratio is less than 0.8.

Commentary by Dr. Valentin Fuster
2014;():V001T03A018. doi:10.1115/ICONE22-30511.

The operating performance and safety characteristics of canned motor Reactor Coolant Pump (RCP) are vital to the nuclear reactor. The annular flow, which is between the rotor and stator, makes substantial effects on the rotordynamic characteristics of RCP. In this work, the annular flow is simulated by means of 3-dimensional CFD approaches. The annular flow field and the auxiliary impeller are modeled. The results show that the pre-swirl strength at the inlet of the annular flow is evident. Another model without the auxiliary impeller is analyzed under constant pre-swirl ratio with various axial flow rates and whirl speeds. The rotordynamic coefficients of the annular flow are obtained. The results are compared with theoretical predictions. Numerical results show that the fluid in the annular region makes remarkable effects on the rotordynamic characteristics of the canned motor RCP.

Commentary by Dr. Valentin Fuster
2014;():V001T03A019. doi:10.1115/ICONE22-30516.

In this paper, the theoretical study and experimental investigation on the rod drop performance of high-temperature gas-cooled reactor (HTGR) pebble-bed module have been presented. The control rod drive mechanisms (CRDMs), serving as the first shutdown system of the reactor, are positioned above the reactor pressure vessel. When the reactor is operated at the power regulation mode, the control rods are pulled up-and-down in their channels around the reactor core. The CRDM provides a fail-safe operational mode for the control rod system. If the reactor emergency shutdown is required the control rods could drop into their channels by gravity. Thus the key factor, emergency insertion time of the whole control rod stroke, which represents the inherent safety of the CRDM, is crucially important and should be measured precisely. In the final objective of ensuring reliability of the CRDM, a full size drive line had been built and tested to obtain the overall performance function of the CRDM. Every component of the CRDM test line was simulated at the scale 1:1, including a 15 meters high test bench that was used as the substitution of the pressure vessel. At current stage, the rod drop performance had been experimental investigated at ambient temperature and pressure. The emergency insertion time of an 8 meters stroke was measured to be less than 50 seconds. A mathematical model of CRDM also had been developed. The rod motion characteristic equations show that the rod dropping speed approaches to a constant during the emergency insertion. The theoretical results are in agreement with the test results.

Commentary by Dr. Valentin Fuster
2014;():V001T03A020. doi:10.1115/ICONE22-30598.

The U-tubes installed inside the steam generator experience high temperature and pressure as a role of heat transfer. Specially, during the secondary side hydrostatic test which ensures the integrity of steam generator, the U-tube is subjected to high external pressure. The purpose of this paper is to investigate the allowable external pressure of the U-tube in steam generator.

In the ASME B&PV Code Section III [1], the allowable external pressure is determined by the rules of NB-3133. Alternatively, NB-3228 analysis may be applied. In order to determining the allowable external pressure of steam generator tube, the buckling analysis is performed. The analysis consists of the collapse pressure and elastic instability pressure analyses. In this study, these pressures are determined by finite element analysis (FEA) using ANSYS computer program.

The non-linear static analysis is performed with ideally elastic-plastic material properties for the collapse analysis. On the other hand, the elastic instability pressure is calculated by eigenvalue analysis in elastic range. These allowable pressures are found to be 24.1 ksi and 10.5 ksi. Therefore the lower pressure of 24.1 ksi is the allowable external pressure of tube. In addition, the results of analysis are compared with other research [6] and Det Norske Veritas (DNV) offshore standards [11].

In conclusion, the results of buckling analysis are well matched with other research [6] and standard [11]. For steam generator tubes, the collapse pressure is dominant factor in failure. Also, the collapse pressure is largely influenced by the ovality of tube.

Topics: Boilers , Buckling
Commentary by Dr. Valentin Fuster
2014;():V001T03A021. doi:10.1115/ICONE22-30600.

The HTR-PM (High Temperature Reactor Pebble bed Module) reactor core consists of assemblies that include graphite bricks, carbon bricks and keys/dowels. The double-layer structure, which is composed of two layers of bricks connected by keys and dowels, is the basic load-bearing unit of the graphite internals. A series of seismic experiments have been carried out to study the dynamic characteristic of this multi-body structure. As a part of them, the present study aimed to evaluate the nonlinear dynamic response of a full-size graphite/carbon unit and investigate the integrity of this assembly.

Dynamic excitations were applied at the bottom of the double-layer structure. The experimental random-motion excitation ranged from 0.2g to 0.75g. The sine-sweep excitation frequency band ranged from 5–100Hz, sweep rate was 1oct/min, with acceleration magnitudes of 0.3g and 0.5g. The sine beat wave excitation frequencies were 5Hz, 10Hz, 20Hz and 40Hz with acceleration of 0.5g. Response data of acceleration and displacement at certain points on bricks were measured. PSD curves and displacement time-histories were acquired. Results indicated that the fundamental frequency of the assembly is 42∼49Hz and the damping rate is 4.9%. The dynamic response of the assembly exhibited as an integrated component only when the excitation is lower than 0.35g.

Topics: Graphite
Commentary by Dr. Valentin Fuster
2014;():V001T03A022. doi:10.1115/ICONE22-30643.

The probabilistic fracture mechanics analysis code PASCAL-SP is improved by introducing crack-growth evaluation methods based on J-integrals, including calculation functions of J-integral values for semi-elliptical surfaces and through-wall cracks in pipes. Using the improved PASCAL-SP, sensitivity analyses that varied parameters such as earthquake magnitude were carried out on the basis of probabilistic evaluation. Results obtained from sensitivity analyses are also presented, e.g., the effect of earthquake magnitude on failure probability. The improved PASCAL-SP makes evaluation of the failure probability of piping under large seismic loading possible.

Commentary by Dr. Valentin Fuster
2014;():V001T03A023. doi:10.1115/ICONE22-30700.

Jet pumps are key components to feed cooling water into reactor core in boiling water reactor. Inside condition of jet pumps is high flow condition. Therefore, jet pumps have risk of damages by flow-induced vibration, especially, the leakage-flow-induced vibration at the slip joint between the inlet mixer and the diffuser in extended power uprating condition with increasing core flow rate or particular operating condition such as single loop operation that increases differential pressure of the slip joint. To mitigate the risk of the leakage-flow-induced vibration, slip joint extension which can be installed on the top of diffuser was developed (See Figure 1). Self-excited vibration is treated as negative damping i.e. unstable state. It is well-known that the leakage flow through divergent gap flow passage causes the negative damping. However, the configuration of the gap flow passage of the slip joint with slip joint extension is complicated flow passage which consists of convergent, divergent and parallel flow passage region. To addition to this, the leakage flow direction in normal or power uprating condition is opposite to in abnormal operating condition such as single loop operating. Therefore, it is necessary to identify the optimum configuration of gap flow passage of the slip joint extension to suppress leakage-flow-induced vibration for various operating conditions. To achieve this goal, the gap flow passage of the slip joint extension was determined using transfer matrix method based on the leakage-flow-induced vibration theory. The effect and characteristic of vibration suppression for the slip joint extension was confirmed by fundamental tests that simulated the slip joint configuration.

Commentary by Dr. Valentin Fuster
2014;():V001T03A024. doi:10.1115/ICONE22-30829.

Buckling failure load of stainless steel columns under compressive stress was experimentally measured in severe accident conditions, which addresses the accidents in Fukushima Daiichi nuclear power plants. Firstly, buckling failure load defined as load which causes failure of the column (plastic collapse) was measured in a wide range of temperatures from 25 °C up to 1200 °C. The load values measured in this study were compared to numerical estimations by eigenvalue simulations (for an ideal column) and by nonlinear simulations (for a column with initial bending). Two different methods for measurement of the buckling failure load were employed to examine the effect of thermal history on buckling failure. Different load values were obtained from two methods in high temperature conditions over 800 °C. The difference in the buckling failure load between two methods increased with temperature, which was explained by the effect of creep at high temperatures. Moreover, the influence of asymmetric temperature profiles along a plate column was also explored with regard to the failure mode and the buckling failure load. In present study, all of the buckling processes were visualized by a high speed camera.

Commentary by Dr. Valentin Fuster
2014;():V001T03A025. doi:10.1115/ICONE22-30893.

There is a possibility that water in a spent fuel storage pool may overflow due to sloshing during long-period earthquakes. Therefore, this paper presents two sloshing suppression methods for a rectangular pool to reduce the volume of overflow water. Vibration tests were carried out to evaluate the volume of overflow water. The 1/20-scale model pool is used. First method is applying immersed blocks on the bottom of the rectangular pool. The volume of the water over the sidewall should be the maximum when the 1st sloshing mode is excited, and this behavior has significant influence on the volume of overflow water. The immersed blocks suppress the 1st sloshing mode, thereby reducing the over flow water. Vibration test were conducted by changing the following conditions: height of blocks, open area ratio, block position in the excitation direction, and number of opening sections. Changing of the natural frequencies and the amplification ratios are confirmed by the sinusoidal sweep test. In random wave excitation tests, the volume of overflow water from the pool with the best configuration blocks is lower than about 60% of that from the regular pool. Second method is applying horizontal baffle plates on the sidewall of rectangular pool. Horizontal baffle plate is a well-known sloshing suppression method. However, there is a little information about the relationship between installation condition of baffle plate and volume of overflow water Vibration test were conducted by changing the following conditions: installation height level, overhang length of baffle plate. In random wave excitation tests, the volume of overflow water from the pool with the best installation condition baffle plate is lower than 40% of that from the regular pool.

Commentary by Dr. Valentin Fuster
2014;():V001T03A026. doi:10.1115/ICONE22-30904.

The main purpose of this research is to investigate the effect of friction in the thermal stress of Reactor Coolant System (RCS) of VVER-1000. RCS is a large system connecting reactor vessel, steam generators and RC Pumps. During the heat-up of reactor, the RCS expand and during cool-down of reactor, it contracts. Because of the heavy weight of reactor and steam generator, the friction at the support of RCS affects the thermal stress of RCS. In this paper how much support friction contributes to the development of thermal stress is assessed in order to investigate the thermal stress and effect of support friction. A quarter-symmetry model of VVER-1000 RCS is developed in ANSYS and meshed with hexahedral elements to ensure better solution accuracies. The model includes reactor vessel, steam generator and reactor coolant pump. Internals of reactor vessel, steam generators and RCPs are represented by point mass to simplify the model. Temperature of inside surface of hot-leg side of reactor vessel to inlet side of steam generator is assumed same uniform hot-leg temperature, and the temperature of inside surface of outlet side of steam generator to reactor vessel is uniform cold-leg temperature. All outside surface are assumed insulated. The analysis includes neither transient thermal loading nor dynamic loadings. The analysis results show that friction at support brings little effect on the peak thermal stress. The peak thermal stress occurs at hot-leg nozzle of reactor pressure vessel and it approached near yield stress. If load combination is included the localized total stress at hot-leg nozzle could go over the yield stress. This peak stress could affect fatigue life in a long run. A recommendation is made that a detailed fatigue analysis of VVER-1000 RCS is necessary.

Commentary by Dr. Valentin Fuster
2014;():V001T03A027. doi:10.1115/ICONE22-31103.

Fretting wear is a particular type of wear that is expected to occur in fuel assemblies of pressurized water cooled nuclear reactors. Fretting damage of fuel rods may cause Nuclear Power Plant (NPP) operations problems and are very expensive to repair. As utilities and fuel vendors adopt higher utilization of uranium and improved thermal margins plants, burned fuel rods will be loaded at core the periphery as part of the margin mechanisms. Pressurized Water Reactors (PWRs) have experienced fuel rods fretting wear failures due to flow induced vibrations. This study describes basic results of theoretical analysis and describes experiments to predict thinning of the Zr cladding wall thickness performed.

Commentary by Dr. Valentin Fuster
2014;():V001T03A028. doi:10.1115/ICONE22-31157.

During a loss-of-coolant accident (LOCA) event the AP1000® passive safety features actuate to provide emergency core cooling (ECC) to the reactor core using passive features that do not rely on electrical power being available. The core makeup tanks (CMTs), and accumulators (ACCs) actuate to quench the fuel rods and refill the reactor vessel. After the CMTs and ACCs empty, the in-containment refueling water storage tank (IRWST) utilizing gravity injects a large volume of sub-cooled fluid into the reactor vessel. This floods the vessel and the lower region of containment (containment sump) initiating gravity induced long-term recirculation cooling. The discharge of high energy fluid during the blowdown, re-flood, and re-fill phases is assumed to condense on the colder structures inside the containment including the containment vessel shell. Heat is transferred through the shell to the film of water from the Passive Containment Cooling System (PCS) applied to the outside of the containment vessel shell. This results in evaporative heat transfer on the outside of the containment vessel. Due to the large heat transfer coefficients on the inside and outside of the shell the heat conduction through the shell is very important to the heat rejection capability of the PCS, and plays a large part in ensuring the containment vessel pressure is not exceeded during design basis events. The AP1000® containment vessel is forged from a high strength carbon steel alloy that is coated with an inorganic zinc coating which protects the containment vessel from corrosion during its design life. The coating acts as a sacrificial anodic layer which corrodes in lieu of corrosion of the substrate beneath it. The corrosion of the coating can potentially lead to degradation in thermal conductivity of the coating due to metallic oxides typically having a lower thermal conductivity than that of the non-oxidized state. A reduction in thermal conductivity of the protective coating will impact the overall heat transfer through the containment vessel during PCS operation. The purpose of this work is to develop a mechanistic model demonstrated against empirical validation for assessing the effects of oxidation on the thermal conductivity of the protective inorganic zinc coating (IOZ) on the AP1000® containment vessel.

Commentary by Dr. Valentin Fuster
2014;():V001T03A029. doi:10.1115/ICONE22-31167.

Neutron noise monitoring and analysis has been a valuable tool for in-service monitoring of the structural integrity of reactor internals. Several nuclear power plants have experienced structural degradation in their reactor internals. Significant signatures were noticed in the neutron noise data when the degradations occurred. This article briefly summarizes the findings based on these experiences. A significant amount of neutron noise time-history data has been obtained based on the continuous neutron noise data monitoring and analysis for pressurized water reactors (PWRs). Over the past decade, power plants have made many major component repairs and replacements. This article correlates the plant operating history with its corresponding signatures on the neutron noise data. This information is an extremely valuable reference for diagnosing the condition of the reactor internals through neutron noise data monitoring and analyses.

Commentary by Dr. Valentin Fuster
2014;():V001T03A030. doi:10.1115/ICONE22-31263.

Under the combined accident thermal and seismic loadings, the structural response of the AP1000 Auxiliary and Shield Building (ASB) is numerically investigated. A nonlinear Finite Element Model (FEM) of the AP1000 ASB is developed, in which the rebar in the reinforced concrete is explicitly described and the nonlinear behavior of the concrete is considered. The numerical modeling method and material models used by the reinforced concrete are validated by the testing results published in the literature. The propagation of the thermal loading-induced concrete cracks along the wall thickness is studied. Furthermore, the effects of thermal cracks on the wall stiffness, the development of the thermal stress and the axial forces acting on the reinforcement are fully discussed. The impact of thermal concrete cracks on the design demand of the rebar is also investigated. It is worthy of being further studied how to incorporate the appropriate reduction factor caused by concrete cracks into the linear structural design.

Topics: Accidents
Commentary by Dr. Valentin Fuster

Codes, Standards, Licensing and Regulatory Issues

2014;():V001T07A001. doi:10.1115/ICONE22-30166.

Environmentally-assisted fatigue evaluations are to be conducted for ASME Code Class 1 piping components in a pressurized water reactor. Environmental fatigue correction factor method for incorporating the effects of light water reactor coolant environments into ASME Section III fatigue evaluations was investigated in this paper. Both ASME Code NB-3200 and NB-3600 methods were used to determine the usage factors of the piping components. Considered in these calculations were the loads which are generally applied to the piping design for the nuclear power plants such as seismic, thermal expansions, thermal transients, thermal stratifications and building-filtered dynamic loadings. For the practical applications of NB-3600 method, regarded as the simple and conservative approach, to the piping components, it was presumed that the stress intensity and/or strain time histories for all or some of the external loadings were not known; therefore the time consistency might not be considered in calculating the usage factors as well as environmental correction factors (Fen). In NB-3200 method in contrast to NB-3600, the stress variations with time for all loads except for the dynamic loads were obtained for the fatigue evaluations in LWR environments, and therefore the time consistency was considered. The results showed that the environmental correction factors as well as in-air cumulative usage factors calculated from NB-3200 methods were significantly less than those from NB-3600 rules. In addition, comparing the results of conventional ASME fatigue evaluation applied until 2006 to the ones in accordance with USNRC RG 1.207 issued on 2007, one may identify that the cumulative usage factors in LWR environments were larger than the conventional one due to the change of design fatigue curves as well as Fen factors accounting for the environmental effects on fatigue. Although this work was focused on the detailed calculations of the usage factors and Fen values, one might identify or suggest a number of areas requiring further clarification or research through the efforts of this study, which were not yet addressed. A few items needed to be clarified, especially for NB-3600-based fatigue evaluations, are also discussed in this paper.

Commentary by Dr. Valentin Fuster
2014;():V001T07A002. doi:10.1115/ICONE22-30525.

In the System Based Code (SBC), a reliability target is defined according to the importance of components (risk and/or failure probability), and the grade of material, design, manufacture and maintenance are chosen to satisfy the reliability target. Therefore, a reliability assessment of the components plays an important role in the concept of SBC. In the reliability assessment guidelines of SBC under development in JSME, the Load and Resistance Factor Design (LRFD) method and the Monte Carlo method will be applied to reliability assessment. Along with a development plan of the SBC, examination of the target reliability according to the importance of components and preparation of the LRFD methods and assessment of partial safety factor for some failure modes are underway. Until now, the LRFD methods have been developed for burst due to internal pressure, plastic collapse due to membrane and bending stress, fatigue, plastic collapse of a pipe having a circumferential crack, and buckling of a thin wall cylinder. This paper describes the action plans for the development of the reliability assessment methods and examination results till date.

Commentary by Dr. Valentin Fuster
2014;():V001T07A003. doi:10.1115/ICONE22-30530.

Since introduce different technical routes, during decades of nuclear power development in our country, the French RCC series standards, American ASME standards and Russian standards are adopted, which led to the current various standards exist in their own way. To promote the building of nuclear power standards system in China, in the year of 2012, important research subject “the research on the standard system of advanced nuclear power in China” has been carried out and subject “nuclear power construction and commissioning” is one of it.. By digestion and absorption of four oversea AP1000 units of Sanmen nuclear power plant in Zhejiang province and Haiyang nuclear power plant in Shandong province, the building of standard system during nuclear power construction suitable to our national condition is studied, including the system frame and composition standards, building standard system method during construction, namely through research and example to present what kind of standard system is suitable for China standard system during construction, and what kind of method or design is used to obtain and maintain such system. The thesis is to promote the subject research methods based on examples to build China’s nuclear power standard system.

Commentary by Dr. Valentin Fuster
2014;():V001T07A004. doi:10.1115/ICONE22-30567.

This paper is the first one of the series of four papers that describe ongoing activities in the Japan Society of Mechanical Engineers (JSME) and the American Society of Mechanical Engineers (ASME) on the elaboration of the System Based Code (SBC) concept aiming at its application to nuclear structural codes and standards. This paper includes a brief introduction to the SBC concept and the technical features of structural evaluation methodologies that are being developed for use in the SBC concept; Load and Resistance Factor Design based reliability assessment methods and the JSME guidelines for reliability evaluation of static components of sodium-cooled fast reactors. Also described is the ongoing collaboration of JSME and ASME at the Joint Task Group for System Based Code established in the ASME Boiler and Pressure Vessel Code Committee which is developing alternative rules for ASME B&PV Code Section XI Division 3, inservice inspection requirements for liquid metal reactor components.

Commentary by Dr. Valentin Fuster
2014;():V001T07A005. doi:10.1115/ICONE22-30570.

This paper describes the outline of the guidelines on structural reliability evaluation for the passive components of the fast breeder reactor (FBR). The guidelines are now being prepared by the task force for the system based code in the Japan society of mechanical engineers in order to contribute to reducing differences in evaluated structural reliability by evaluators. They consist of five chapters, which are “General rules”, “Reliability evaluation”, “Failure scenario setting”, “Modeling”, and “Failure probability calculation”, respectively. In the chapter of “Reliability evaluation”, the general procedures of reliability evaluation are explained. Detailed procedures at each step are explained in the following chapters in the guidelines. Evaluation procedures for accumulation damage and crack propagation due to creep-fatigue interaction which is a typical degradation phenomenon in FBR are also provided in the appendix of the guidelines. In addition, some examples of random variables are prepared as standard input data for structural reliability evaluation.

Commentary by Dr. Valentin Fuster
2014;():V001T07A006. doi:10.1115/ICONE22-30572.

This paper describes the ongoing activities at the Joint Task Group for System Based Code established in 2012 by the Japan Society of Mechanical Engineers (JSME) and the American Society of Mechanical Engineers (ASME) in the ASME Boiler and Pressure Vessel Code Committee. The Joint Task Group aims at developing alternative rules for ASME Boiler and Pressure Vessel Code Section XI Division 3, inservice inspection requirements for liquid metal reactors. The alternative rules will be developed based on the System Based Code concept which was originally proposed in Japan and is being elaborated both in JSME and ASME. The alternative rules are for sodium-cooled fast reactors where some of the components could have difficulties in conforming to the current requirements primarily due to accessibility. The alternative requirements would consist of a set of relieved requirements and a logic flow through which the applicability of them is judged. The logic flow considers both component structural integrity and the plant safety goals. The issuance of a Code Case is envisioned around 2016. Further efforts to integrate the process into a new framework being developed in Section XI would cover various types of reactors.

Commentary by Dr. Valentin Fuster
2014;():V001T07A007. doi:10.1115/ICONE22-30640.

Design codes, such as RCC-MRx and ASME III NH, for generation IV nuclear reactors use interaction diagram based method for creep-fatigue assessment. In the interaction diagram the fatigue damage is expressed as the ratio of design cycles over the allowable amount of cycles in service and the creep damage as the ratio of time in service over the design life. With this approach it is assumed that these quantities can be added linearly to represent the combined creep-fatigue damage accumulation. Failure is assumed to occur when the sum of the damage reaches a specified value, usually unity or less. The fatigue damage fraction should naturally be unity when no creep damage is present and creep damage should be unity when no fatigue damage is present. However, strict fatigue limits and safety factors used for creep rupture strengths as well as different approaches to relaxation calculation can cause a situation where creep-fatigue test data plotted according to the design rules are three orders of magnitude away from the interaction diagram unity line. Thus, utilizing the interaction diagram methods for predicting the number of creep-fatigue cycles may be inaccurate and from design point of view these methods may be overly conservative. In this paper the results of creep-fatigue tests carried out for austenitic stainless steel 316 and heat resistant ferritic-martensitic steel P91, which are included in the design codes, such as RCC-MRx, are assessed using the interaction diagram method with different levels of criteria for the creep and fatigue fractions. The test results are also compared against the predictions of a recently developed simplified creep-fatigue model which predicts the creep-fatigue damage as a function of strain range, temperature and hold period duration with little amount of fitting parameters. The Φ-model utilizes the creep rupture strength and ultimate tensile strength (UTS) of the material in question as base for the creep-fatigue prediction. Furthermore, challenge of acquiring representative creep damage fractions from the dynamic material response, i.e. cyclic softening with P91 steel, for the interaction diagram based assessment is discussed.

Topics: Creep , Fatigue , Modeling
Commentary by Dr. Valentin Fuster
2014;():V001T07A008. doi:10.1115/ICONE22-30694.

There is a huge and rapid growing demand for energy in China. The developing of new energy, including nuclear energy, is an important way to solve China’s energy and environment problems. Currently, Nuclear power is playing a more and more important role in China, especially in the coastal areas where the economy is developing rapidly. The main nuclear reactor equipment is the key guarantee to nuclear safety, such as the reactor pressure vessel. With the continuous development of science and technology, the reactor pressure vessel uses more and more of advanced design methods, high performance materials and industrial-proven technologies. As an airtight container who is directly exposed to the great operating pressure of reactor, the reactor pressure vessel plays a crucial role to the safety of reactor. In the paper, by retrieving Chinese Intellectual Property Office patent libraries, 12 Chinese innovative patents for the reactor pressure vessel are introduced, containing four kinds of main technologies: closure, body, nozzle safe end and other technologies. The patent results could lead the corresponding advice to the design of the equipment effectively, to the system arrangement reasonably, and further to the reference for designing a safe nuclear power plants. The paper is to exchange the latest technological progress of reactor pressure vessel, in order to promote technological innovation in the nuclear power field.

Commentary by Dr. Valentin Fuster
2014;():V001T07A009. doi:10.1115/ICONE22-30696.

The first Japanese spent fuel interim storage facility away from a reactor site is about to be commissioned in Mutsu City, Aomori Prefecture. In designing, licensing and construction of the Dual Purpose Casks (DPCs, for transport and storage) for this facility, codes and standards established by the Atomic Energy Society of Japan (AESJ) and by the Japan Society of Mechanical Engineers (JSME) have been applied.

The AESJ established the first standard for DPCs as “Standard for Safety Design and Inspection of Metal Casks for Spent Fuel Interim Storage Facilities” in 2002 (later revised in 2010). The standard provides the design requirements to maintain the basic safety functions of DPCs, namely containment, heat removal, shielding, criticality prevention and the structural integrity of the cask itself and of the spent fuel cladding during transport and storage. Inspection methods and criteria to ensure maintenance of the basic safety functions and structural integrity over every stage of operations involving DPCs including pre-shipment after storage are prescribed as well. The structural integrity criteria for major DPC components refer to the rules provided by the JSME.

JSME completed the structural design and construction code (the Code) for DPCs as “Rules on Transport/Storage Packagings for Spent Nuclear Fuel” in 2001 (later revised in 2007). Currently, the scope of the rules cover the Containment Vessel, Basket, Trunnions and Intermediate Shell as major components of DPCs. Rules for these components are based on those for components of nuclear power plants (NPP) with similar safety functions, but special considerations based on their shapes, loading types and required functions are added.

The Code has differences from that for NPP components with considerations to DPC characteristics;

- The primary stress and the secondary stress generated in Containment Vessels shall be evaluated under Service Conditions A to D (from ASME Sec III, Div.1).

- Stress generated in the seal region lid bolts of Containment Vessels shall not exceed yield strength under Service Conditions A to D in order to maintain the containment function.

- Fatigue analysis on Baskets is not required, and Trunnions can be designed only for Service Conditions A and B with special stress limits consistent with conventional assessment methods for transport packages.

- Stress limits for earthquakes during storage are specified.

- Ductile cast iron with special fracture toughness requirements can be used as a material for Containment Vessels.

DPC specific considerations in standards and rules will be focused on in this paper. Additionally, comparison with the ASME Code will be discussed.

Commentary by Dr. Valentin Fuster
2014;():V001T07A010. doi:10.1115/ICONE22-30713.

Two major Codes are used for Fitness for Service of Nuclear Power Plants: one is the ASME B&PV Code Section XI and the other one is the French RSE-M Code. Both of them are largely used in many countries, partially or totally.

The last 2013 RSE-M covers “Mechanical Components of Pressurized Water Reactors (PWRs):

- Pre-service and In-service inspection

- Surveillance in operation or during shutdown

- Flaw evaluation

- Repairs-Replacements parts for plant in operation

- Pressure tests

The last 2013 ASME Section XI covers “Mechanical components and containment of Light Water Reactors (LWRs)” and has a larger scope with similar topics: more types of plants (PWR and Boiling Water Reactor-BWR), other components like metallic and concrete containments…

The paper is a first comparison covering the scope, the jurisdiction, the general organization of each section, the major principles to develop In Service Inspection, Repair-Replacement activities, the flaw evaluation rules, the pressure test requirements, the surveillance procedures (monitoring…) and the connections with Design Codes…

These Codes are extremely important for In-service inspection programs in particular and essential tools to justify long term operation of Nuclear Power Plants.

Commentary by Dr. Valentin Fuster
2014;():V001T07A011. doi:10.1115/ICONE22-30715.

During the past 30 years the main rules for fatigue analysis of pressure vessels were based on elastic approaches in order to evaluate a cyclic strain amplitude and compare with an S-N fatigue curve for the corresponding material. After review of some rules in different Nuclear and Non Nuclear Codes, like ASME Boiler & Pressure Vessel Code Section III, French RCC-M and RCC-MRx, European Standards EN 13445, the major conservatisms and uncertainties of different rules are discussed.

All these Codes propose simple rules to evaluate the strain amplitude based on elastic approaches and simplified correction factors (Ke and Kv), transient combination rules and damage cumulating procedure.

In the other hand, the material properties are based on standard fatigue tests done on the material associated to reduction factors to consider some particular effects like scatter, scale, surface roughness, mean stress or environmental effects to transfer them from small specimen to real structures.

Concerned components are mainly piping systems.

No existing Code covers all the aspects of fatigue with similar conservatisms that can affect the in-service inspection programs and the remaining life assessment of the corresponding components.

After the review of different rules, key factors that affect the results and predictions will be identified.

Some proposals will be issued to progress in the near future. Finally, a first set of recommendation on fatigue analysis will be presented to improve existing codes on an harmonized way, associated to material properties needed, as fatigue curves associated to reduction factors.

Commentary by Dr. Valentin Fuster
2014;():V001T07A012. doi:10.1115/ICONE22-30969.

Tools for design and operation of power plants underwent considerable improvements over the past. This is on the one hand a result of the development of powerful computers and related finite-element codes, but on the other hand also a result of significant improvements in materials mechanics and understanding of materials properties. Making such improvements available for implementation into design codes needs often extended R&D projects which happen more and more on an international level. The Generation IV International Forum (GIF) for advanced nuclear plants can be considered as a very good example in this respect. Although this has been recognized by different stakeholders involved in design and operation of nuclear plants, substantial international interaction and collaboration is still not in place. Implementation of international research projects could help to establish a sound scientific basis for materials properties and materials degradation necessary to meet the challenges of current and future nuclear plants. It is the aim of this paper to highlight topics which could be considered for international collaboration of the standards development organizations for the benefit of all participants. The paper discusses possible materials research themes related to future codes which appear to be appropriate for international collaboration.

Commentary by Dr. Valentin Fuster
2014;():V001T07A013. doi:10.1115/ICONE22-31054.

This paper discusses a framework for designing artificial test problems, evaluation criteria, and two of the benchmark tests developed under a research project initiated by the Canadian Nuclear Safety Commission to investigate the approaches for qualification of tolerance limit methods and algorithms proposed for application in optimization of CANDU reactor protection trip setpoints for aged conditions. A significant component of this investigation has been the development of a series of benchmark problems of gradually increased complexity, from simple “theoretical” problems up to complex problems closer to the real application.

The first benchmark problem discussed in this paper is a simplified scalar problem which does not involve extremal, maximum or minimum, operations, typically encountered in the real applications. The second benchmark is a high dimensional, but still simple, problem for statistical inference of maximum channel power during normal operation.

Bayesian algorithms have been developed for each benchmark problem to provide an independent way of constructing tolerance limits from the same data and allow assessing how well different methods make use of those data and, depending on the type of application, evaluating what the level of “conservatism” is. The Bayesian method is not, however, used as a reference method, or “gold” standard, but simply as an independent review method.

The approach and the tests developed can be used as a starting point for developing a generic suite (generic in the sense of potentially applying whatever the proposed statistical method) of empirical studies, with clear criteria for passing those tests. Some lessons learned, in particular concerning the need to assure the completeness of the description of the application and the role of completeness of input information, are also discussed.

It is concluded that a formal process, which should include extended and detailed benchmark tests, but targeted to the context of the particular application and aimed at identifying the domain of validity of the proposed tolerance limit method and algorithm, is needed and might provide the necessary confidence in the proposed statistical procedure.

Topics: Optimization , Testing
Commentary by Dr. Valentin Fuster
2014;():V001T07A014. doi:10.1115/ICONE22-31058.

The AP1000 plant is an 1100-MWe pressurized water reactor (PWR) with passive safety features and extensive plant simplifications that enhance construction, operation, maintenance and safety. One of the key design approaches in the AP1000 plant design is to use passive features to mitigate design basis accidents. Active defense-in-depth (DiD) features provide investment protection, reduce the demands on the passive features and support the Probabilistic Risk Assessment (PRA). The passive features are classified as safety-related in the United States. The active DiD features are classified as nonsafety-related (with supplemental requirements) in the United States. The AP1000 plant design has also incorporated a standardization approach, which together with the level of safety achieved by the passive safety features, results in a plant design that can be applied to different geographical regions with varying regulatory standards and utility expectations without major changes.

This paper will discuss the approach taken to defining DiD in the AP1000 plant and the effectiveness of that approach. It will also address the capability of the AP1000 plant to meet deterministic DiD guidelines such as the ones in application in the UK or described in the Western European Nuclear Regulators’ Association (WENRA) safety objectives for new plants.

Commentary by Dr. Valentin Fuster
2014;():V001T07A015. doi:10.1115/ICONE22-31059.

The Los Alamos National Laboratory (LANL) is a one of the largest and diverse science and technological institutions in the world. The size and sophistication of LANL’s facilities and workforce present a unique challenge to develop and implement a Quality Assurance (QA) program that meets LANL’s needs. LANL has updated its QA Program to a targeted, requirements-based approach, and broadened its Quality Assurance technical expertise into essential technical areas. The expanded areas of expertise include engineering, project management, nuclear facility operations, and weapons design and fabrication. This approach is achieving success as evidenced on an institutional level by LANL’s receipt of various national, international and local awards for its products and services. Success is also realized on the QA Program level with sufficient recognition of the importance of the QA Program by the LANL workforce. However, QA program challenges remain in areas of expanding the importance of QA; streamlining the grading process and ensuring the program is commensurate with risk and customer expectations; maintaining sufficient authority and freedom from line management for deployed QA personnel while continuing to increase the technical breadth of QA personnel. These are the focus areas to continuously improve the LANL QA Program.

Topics: Quality control
Commentary by Dr. Valentin Fuster
2014;():V001T07A016. doi:10.1115/ICONE22-31074.

In this paper, fatigue assessment of Class 1 nuclear power piping according to ASME Boiler and Pressure Vessel Code, Section III, NB-3600, is discussed. Key parameters involved in the fatigue assessment, i.e. the alternating stress intensity Salt, the penalty factor Ke and the cumulative damage factor U, are addressed. In particular, a so-called simplified elastic-plastic discontinuity analysis for alternative verification when basic fatigue requirements found unsatisfactory, and the procedures for evaluating the alternating stress intensity Salt, is reviewed. Factors that can significantly affect the reliability and accuracy of the fatigue assessment are examined. It is illustrated that there is a great need of other alternatives to this simplified elastic-plastic analysis procedure. An alternative based on non-linear finite element analysis is suggested. This paper is a continuation of our work presented in ICONE16-21, which attempted to categorize design rules in the code into linear and non-linear rules and to clarify corresponding requirements that can be used in combination with non-linear finite element analysis.

Commentary by Dr. Valentin Fuster
2014;():V001T07A017. doi:10.1115/ICONE22-31143.

When the flaws are detected in Japanese nuclear power components by in-service inspection, structural integrity assessment are performed in the technical judgment on continuous service. If cyclic loading is assumed, fatigue crack growth analysis should be conducted based on the Rules on Fitness-for-Service for Nuclear Power Plants of the Japan Society of Mechanical Engineers Code (JSME FFS Code). However, fatigue crack growth analysis for BWR components consisting of Ni-base alloy is currently impossible, since the reference curve of fatigue crack growth rate for Ni-base alloy in BWR water environment is not yet prescribed in the JSME FFS Code.

In this study, fatigue crack growth behavior of Ni-base alloy used for Japanese BWR plants in BWR water environment was investigated. Based on the experimental data, the fatigue crack growth rate curve was evaluated. Four test parameters of material, corrosion potential, stress ratio and load rising time were considered. As a result of fatigue crack growth tests, the effects of all test parameters on the fatigue crack growth behavior were found. A Mean curve of fatigue crack growth rate in Paris law format, which was a function of stress ratio and rising time, was formulated based on crack growth data in normal water chemistry (corrosion potential was over 150 mVSHE) for weld metal and heat affected zone (HAZ), respectively. A reference curve of fatigue crack growth rate was also formulated by the statistical treatment considering the scatter of crack growth rate.

Further, in order to determine the threshold stress intensity factor range ΔKth of reference curve of fatigue crack growth, ΔK decreasing tests were conducted under the test condition of 1 second of rising time. As a result, the threshold value of ΔK was evaluated based on the ASTM E 647, and the ΔKth of the reference curve was conservatively determined considering the margin.

Commentary by Dr. Valentin Fuster
2014;():V001T07A018. doi:10.1115/ICONE22-31169.

Most of the current operating nuclear power plants in the United States were designed using the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section III, for fatigue design curves. These design curves were developed in the late 1960s and early 1970s. They were often referred to as “air curves” because they were based on tests conducted in laboratory air environments at ambient temperatures. In recent years, laboratory fatigue test data showed that the light-water reactor environment could have significant impact on the fatigue life of carbon and low-alloy steels, austenitic stainless steel, and nickel-chromium-iron (Ni-Cr-Fe) alloys. United States Nuclear Regulatory Commission, Regulatory Guide 1.207 provides a guideline for evaluating fatigue analyses incorporating the life reduction of metal components due to the effects of the light-water reactor environment for new reactors. It recommend following the method developed in NUREG/CR-6909 [3] when designing reactor coolant pressure boundary components. The industry has invested a lot of effort in developing methods and rules for applying environmental fatigue evaluations for ASME Class 1 components and piping. However, the industry experience in applying the environmental fatigue evaluation for reactor core support structures and internal structures has been very limited. During the recent aging management programs, reactor internal component environmental fatigue evaluations for several pressurized water reactors were evaluated. The analyses calculated the cumulative fatigue usage using the recorded plant-specific transient cycles and the projected cycles for 60 years of plant life. The study concludes that the actual fatigue usages of the components are substantially lower than the specified original design conditions. Even assuming the most severe light-water reactor coolant environmental effects, fatigue will not be a concern for 60 years of plant life. The experiences with environmental fatigue evaluation for reactor internals are still very limited. This study shall provide the industry with beneficial information to develop the approaches and rules addressing the environmental effect on the fatigue life of reactor internals.

Commentary by Dr. Valentin Fuster
2014;():V001T07A019. doi:10.1115/ICONE22-31204.

The AP1000® plant is an 1100-MWe pressurized water reactor (PWR) with passive safety features and extensive plant simplifications that enhance construction, operation, maintenance, safety, and costs.

Four AP1000 units are currently under construction on coastal sites of Sanmen and Haiyang, China. Additionally, the United States (US) Nuclear Regulatory Commission (NRC) issued combined licenses (COLs) to allow Southern Nuclear Operating Company (SNC) & South Carolina Electric & Gas Company (SCE&G) to construct and operate AP1000 plants at the existing Vogtle & VC Summer sites in Georgia and South Carolina, respectively. Although construction at both US sites is underway, the first four China AP1000 plants will become operational ahead of the U.S. Domestic AP1000 plants. Westinghouse is also actively engaged in deploying the AP1000 plant design in other regions throughout the world such as Europe. For example, the AP1000 plant design was evaluated by the UK Office for Nuclear Regulation as part of the UK Generic Design Assessment and received a statement of interim Design Acceptance in late 2011.

This paper reviews the past and on-going AP1000 plant licensing activities and discusses how the significant lessons learned gathered through the AP1000 plant worldwide deployment increase licensing certainty for any new AP1000 project.

Topics: Licensing
Commentary by Dr. Valentin Fuster
2014;():V001T07A020. doi:10.1115/ICONE22-31252.

This paper explores the relationship between the Design Specification required for each Section III component and the terms Quality, Quality Assurance, Quality Product, and Management System (Quality Management System). The author’s experience indicates that the relationship between Quality and Quality Assurance is poorly understood and that the resistance experienced to the implementation of a Quality Assurance Program is often due to the person’s lack of understanding that a Quality Assurance Program really does help people do their work “right the first time”.

This lack of understanding often results in poorly defined requirements in the purchasing documents, inadequate definition of scope and a product that does not meet requirements. Even more importantly it can lead to the development of poor relationships between the purchaser and the supplier and particularly in the nuclear field, poor relationships with Regulators. The economic impact is also significant, delays in schedule, delays in product acceptance and payment results in costs that are eventually passed on to the consumer.

Once the above relationships are properly understood and appreciated resistance can turn into positive acceptance. The Section III concept of providing a Design Specification as a complete basis for construction also takes on new meaning.

Commentary by Dr. Valentin Fuster

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