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Plant Operations, Maintenance and Life Cycle

2006;():1-5. doi:10.1115/ICONE14-89018.

The NPP is used to planning maintenance, in-service inspection, surveillance test, fuel handling and design modification in the refueling outage; the operator response capability will be reduced plus some of the plant systems out of service or loss of power at this time. Based on 8 times refueling outage experiences of the Qinshan NPP, this article provide some good practice and lesson learned for the nuclear safety risk management focus at four safety function areas of Residual Heat Removal Capability, Inventory Control, Power availability and Reactivity control.

Commentary by Dr. Valentin Fuster
2006;():7-14. doi:10.1115/ICONE14-89025.

The determination of the thermal reactor power is traditionally done by establishing the heat balance: • for a boiling water reactor (BWR) at the interface of reactor control volume and heat cycle; • for a pressurized water reactor (PWR) at the interface of the steam generator control volume and turbine island on the secondary side. The uncertainty of these traditional methods is not easy to determine and it can be in the range of several percent. Technical and legal regulations (e.g. 10CFR50) cover an estimated instrumentation error of up to 2% by increasing the design thermal reactor power for emergency analysis to 102% of the licensed thermal reactor power. Basically, the licensee has the duty to warrant at any time operation inside the analysed region for thermal reactor power. This is normally done by keeping the indicated reactor power at the licensed 100% value. A better way is to use a method which allows a continuous warranty evaluation. The quantification of the level of fulfilment of this warranty is only achievable by a method which: • is independent of single measurements accuracies; • results in a certified quality of single process values and for the total heat cycle analysis; • leads to complete results including 2-sigma deviation especially for thermal reactor power. This method, which is called ‘process data reconciliation based on VDI 2048 guideline’, is presented here [1, 2]. The method allows to determine the true process parameters with a statistical probability of 95%, by considering closed material, mass- and energy balances following the Gaussian correction principle. The amount of redundant process information and complexity of the process improves the final results. This represents the most probable state of the process with minimized uncertainty according to VDI 2048. Hence, calibration and control of the thermal reactor power are possible with low effort but high accuracy and independent of single measurement accuracies. Furthermore, VDI 2048 describes the quality control of important process parameters. Applied to the thermal reactor power, the statistical certainty of warranting the allowable value can be quantified. This quantification allows keeping a safety margin in agreement with the authority. This paper presents the operational application of this method at an operating plant and describes the additional use of process data reconciliation for acceptance tests, power recapture and system and component diagnosis.

Commentary by Dr. Valentin Fuster
2006;():15-21. doi:10.1115/ICONE14-89086.

This paper presents the justification for the approach, details and results of the Main Generator Seal Oil System reliability enhancements on the San Onofre Nuclear Generating Station, SONGS. The SONGS, Unit 3 experienced substantial turbine damage in early 2001 after the turbine bearings lubrication oil supply failed. During a loss of off-site power incident, power was lost to the two AC powered turbine lubrication oil pumps due to a breaker failure in the switchgear and the DC powered emergency bearing lubricating oil pump failed to start due to a breaker trip. The SONGS turbine generators coasted down from full speed to a full stop without lubricating oil. This resulted in significant bearing, journal and steam path damage that required a four-month duration repair outage during a time period where electricity was in short supply in the State of California. The generator hydrogen sealing system remained operable during this event, however it was recognized during the event follow up investigation that this system had vulnerabilities to failure similar to the bearing lubrication system. In order to prevent a reoccurrence of this extremely costly event, SONGS has taken actions to modify both of these critical turbine generator systems by adding additional, continuously operating pumps with a new, independent power source and independently routed cables. The main challenge was to integrate the additional equipment into the existing lubrication and seal oil systems. The lubrication Oil System was the first system to be retrofitted and these results already have been presented. Reference 2. This paper provides the result of the reliability enhancements for the Main Generator Seal Oil System, which concludes the turbine/generator critical oil systems reliability improvements, performed by SONGS. It is worth noting that the design team discovered and corrected a number of other significant operational issues, which had been present from the early days and also learned a great deal of detailed information about this vital system during the project. The SONGS approach and findings are discussed in this paper, as well as a summary of the work performed. This technical paper will be of interest to utilities with a need to improve turbine generator reliability issues.

Commentary by Dr. Valentin Fuster
2006;():23-29. doi:10.1115/ICONE14-89093.

We analyze a system of N components with dependent failure times. The goal is to obtain the optimal block replacement interval (different for each component) over a finite horizon that minimizes the expected total maintenance cost. In addition, we allow each preventive maintenance action to change the future joint failure time distribution. We illustrate our methodology with an example from South Texas Project Nuclear Operating Company.

Commentary by Dr. Valentin Fuster
2006;():31-39. doi:10.1115/ICONE14-89096.

At WWER NPPs the horizontal steam generators (SGs), are used that differ in design concept from vertical SGs mostly used at western NPPs. Reliable operation of SG heat-exchanging tubes is the crucial worldwide problem for NPP of various types. According to the operation feedback the water chemistry is the governing factor affecting operability of SG tubing. The secondary side corrosion is considered to be the main mechanism of SG heat-exchanging tubes damage at WWER plants. To make the assessment of the tubing integrity the combination of pressure tests and eddy-current tests is used. Assessment of the tubing performance is an important part of SG life extension practice. The given paper deals with the description of the tube testing strategy and the approach to tube integrity assessment based on deterministic and probabilistic methods of fracture mechanics. Requirements for eddy-current test are given as well. Practice of condition monitoring and implementing the database on steam generators operation are presented. The approach to tubes plugging criteria is described. The research activities on corrosion mechanism studies and residual lifetime evaluation are mentioned.

Topics: Tubing , Boilers
Commentary by Dr. Valentin Fuster
2006;():41-48. doi:10.1115/ICONE14-89121.

Process benchmarking between power plants of different utilities has been a tedious task due to the difficulty to obtain significant and suitable data. Only independent consulting agencies with unrestricted access to sensitive operational data gained through projects in various comparable power plants can perform such a task. In order to conduct a holistic, integral benchmark analysis, the basis of such an analysis — the measurement data and key performance indicators — need to be determined. In a second step the verified data needs to be standardized to ensure comparability before the organizational structure and technical processes can be adjusted to the best-practice.

Commentary by Dr. Valentin Fuster
2006;():49-57. doi:10.1115/ICONE14-89122.

Process Data Reconciliation (PDR) is a certified method that calculates the most likely values considering process measurement uncertainties and closing all energy- and material balances where all interdependencies within the entire plant process are fulfilled in a covariance matrix. There are three main factors that generate the financial benefits for the user of reconciled data, depending on the type of plant and base/peak load behaviour: • Increased efficiency / maximized output; • Time advantage in retrieving “lost” megawatts; • Reduction of maintenance costs.

Commentary by Dr. Valentin Fuster
2006;():59-66. doi:10.1115/ICONE14-89138.

A new optical torque measuring method was applied to diagnosis of thermal efficiency of nuclear power plants. The sensor allows torque deformation of the rotor caused by power transmission to be measured without contact. Semiconductor laser beams and small pieces of stainless reflector that have bar-code patterns are employed. The intensity of the reflected laser beam is measured and then input into a computer through an APD and an A/D converter having high frequency sampling rates. The correlation analysis technique can translate these data into the torque deformation angle. This angle allows us to obtain the turbine output along with the torsional rigidity and the rotating speed of the rotor. The sensor was applied to a nuclear plant of Tokyo Electric Power Company, TEPCO, following its application success to the early combined cycle plants and the advanced combined cycle plants of TEPCO. As the turbine rotor of the nuclear power plant is less exposed than that of the combined cycle plants, the measurement position is confined to a narrow gap. In order to overcome the difficulty in installation, the shape of the sensor is modified to be long and thin. Sensor performance of the nuclear power plant was inspected over a year. The value of the torsional rigidity was analyzed by the finite element method at first. Accuracy was improved by correcting the torsional rigidity so that the value was consistent with the generator output. As a result, it is considered that the sensor performance has reached a practical use level.

Commentary by Dr. Valentin Fuster
2006;():67-72. doi:10.1115/ICONE14-89145.

To support the installation and use of elbow meters to measure the high pressure emergency coolant injection flow in an operating nuclear station, a test program was performed to qualify: (i) the “hot” tapping procedure for field application and (ii) the use of elbow meters for accurate flow measurements over the full range of station ECI flow conditions. This paper describes the design conditions and major components of a flow loop used for the elbow meter calibrations. Typical test results are presented and discussed.

Commentary by Dr. Valentin Fuster
2006;():73-80. doi:10.1115/ICONE14-89148.

Industrial plants have become more complex due to technological advancement. This has made the task of maintenance more difficult. The maintenance costs in terms of resources and downtime loss are so high that maintenance function has become a critical factor in a plant’s profitability. Industry should devote as much forethought to the management of maintenance function as to production. Maintenance has grown from an art to a precise, technical engineering science. Planning, organizing scheduling and control of maintenance using modern techniques pays dividends in the form of reduced costs and increased reliability. The magnitude and the dimension of maintenance have multiplied due to development in the engineering technologies. Production cost and capacities are directly affected by the breakdown time. Total operating cost including the maintenance cost plays an important role in replacement dimension. The integrated system approach would bring forth the desired results of high maintenance standards. The standards once achieved and sustained, would add to the reliability of the plan and relieve heavy stresses and strains on the engineering logistic support.

Topics: Maintenance
Commentary by Dr. Valentin Fuster
2006;():81-90. doi:10.1115/ICONE14-89152.

HANARO, an open-tank-in-pool type research reactor of 30 MWth power in Korea, has been operating normally since its initial criticality in February, 1995. For the last ten years, HANARO has carried out ten years periodic in-service inspections (ISI as below) in accordance with Article IWD in ASME SEC. XI to verify the mechanical and structural integrities of the pressure retaining components of the safety related systems and the integral attachments of the supports and restraints of the components which are NPS 4 and above, to be within a specified boundary. This paper describes the results of the ISI including a system pressure test and a VT-3 visual inspection. From the results, it was confirmed through the ISI that the pressure retaining components and parts were stable to within the specified boundaries for their mechanical and structural integrities.

Commentary by Dr. Valentin Fuster
2006;():91-95. doi:10.1115/ICONE14-89177.

The most recent core reload design verification physics testing done at Southern Nuclear Company’s (SNC) Vogtle Unit 2, performed prior to initial power operations in operating cycle 12, was successfully completed while the reactor was at least 1% ΔK/K subcritical. The testing program used was the first application of the Subcritical Physics Testing (SPT) program developed by the Westinghouse Electric Company LLC. The SPT program centers on the application of the Westinghouse Subcritical Rod Worth Measurement (SRWM) methodology that was developed in cooperation with the Vogtle Reactor Engineering staff. The SRWM methodology received U. S. Nuclear Regulatory Commission (NRC) approval in August of 2005. The first application of the SPT program occurred at Vogtle Unit 2 in October of 2005. The results of the core design verification measurements obtained during the SPT program demonstrated excellent agreement with prediction, demonstrating that the predicted core characteristics were in excellent agreement with the actual operating characteristics of the core. This paper presents an overview of the SPT Program used at Vogtle Unit 2 during operating cycle 12, and a discussion of the critical path outage time savings the SPT program is capable of providing.

Commentary by Dr. Valentin Fuster
2006;():97-102. doi:10.1115/ICONE14-89197.

The entry-time approach to dynamic reliability is based upon computational solution of the Chapman-Kolmogorov (generalized state-transition) equations underlying a certain class of marked point processes. Previous work has verified a particular finite-difference approach to computational solution of these equations. The objective of this work is to illustrate the potential application of the entry-time approach to risk-informed asset management (RIAM) decisions regarding maintenance or replacement of major systems within a plant. Results are presented in the form of plots, with replacement/maintenance period as a parameter, of expected annual revenue, along with annual variance and annual skewness as indicators of associated risks. Present results are for a hypothetical system, to illustrate the capability of the approach, but some considerations related to potential application of this approach to nuclear power plants are discussed.

Commentary by Dr. Valentin Fuster
2006;():103-107. doi:10.1115/ICONE14-89233.

Mitsubishi Heavy Industries, Ltd. (MHI) completed replacement work of the upper and lower reactor internals of the 566 MW Ikata Unit No.1 of Shikoku Electric Power Co., Inc. The event marks the world’s first all-in-one-piece extraction and replacement work of its kind in a pressurized water reactor. The author wishes to introduce an outline and features of this milestone.

Commentary by Dr. Valentin Fuster
2006;():109-121. doi:10.1115/ICONE14-89242.

MONJU is a prototype fast breeder reactor (FBR). Modification work commenced in March 2005. Since June 2004, MONJU has changed to one-loop operation of the primary heat transport system (PHTS) with all of the secondary heat transport systems (SHTS) drained of sodium. The purposes of this change are to shorten the modification period and to reduce the cost incurred for circuit trace heating electrical consumption. Before changing condition, the following issues were investigated to show that this mode of operation was possible. The heat loss from the reactor vessel and the single primary loop must exceed the decay heat by an acceptable margin but the capacity of preheaters to keep the sodium within the primary vessel at about 200°C must be maintained. With regard to the heat loss and the decay heat, the estimated heat loss in the primary system was in the range of 90–170kW in one-loop operation, and the calculated decay heat was 21.2kW. Although the heat input of the primary pump was considered, it was clear that circuit heat loss greatly exceeded the decay heat. As for preheaters, effective capacity was less than the heat loss. Therefore, the temperature of the reactor vessel room was raised to reduce the heat loss. One-loop operation of the PHTS was able to be executed by means of these measures. The cost of electrical consumption in the power plant has been reduced by one-loop operation of the PHTS and the modification period was shortened.

Topics: Heat
Commentary by Dr. Valentin Fuster
2006;():123-131. doi:10.1115/ICONE14-89278.

To provide best knowledge about safety-related water level values in boiling water reactors (BWR) is essentially for operational regime. For the water level determination hydrostatic level measurement systems are almost exclusively applied, because they stand the test over many decades in conventional and nuclear power plants (NPP). Due to the steam generation especially in BWR a specific phenomenon occurs which leads to a water-steam mixture level in the reactor annular space and reactor plenum. The mixture level is a high transient non-measurable value concerning the hydrostatic water level measuring system and it significantly differs from the measured collapsed water level. In particular, during operational and accidental transient processes like fast negative pressure transients, the monitoring of these water levels is very important. In addition to the hydrostatic water level measurement system a diverse water level measurement system for BWR should be used. A real physical diversity is given by gamma radiation distribution inside and outside the reactor pressure vessel correlating with the water level. The vertical gamma radiation distribution depends on the water level, but it is also a function of the neutron flux and the coolant recirculation pump speed. For the water level monitoring, special algorithms are required. An analytical determination of the gamma radiation distribution outside the reactor pressure vessel is impossible due to the multitude of radiation of physical processes, complicated non-stationary radiation source distribution and complex geometry of fixtures. For creating suited algorithms Soft Computing methods (Fuzzy Sets Theory, Artificial Neural Networks, etc.) will be used. Therefore, a database containing input values (gamma radiation distribution) and output values (water levels) had to be built. Here, the database was established by experiments (data from BWR and from a test setup) and simulation with the authorised thermofluid code ATHLET.

Commentary by Dr. Valentin Fuster
2006;():133-138. doi:10.1115/ICONE14-89301.

Steam Dryers for extended power up-rated conditions, based on a design with proven performance record, have been designed by Westinghouse with performance characteristics meeting all utility requirements. Extensive CFD analysis has been used to design the new dryer intended for the BWR 3000 plants. The analysis has utilized previous knowledge, acquired from experimental work, on steam flow characteristics in the BWR 3000 with its asymmetrically placed steam line nozzles. A design has been developed that is predicted to both decrease vibration levels and to solve a water-level measurement problem. An extensive experimental verification of the design is presently done.

Commentary by Dr. Valentin Fuster
2006;():139-148. doi:10.1115/ICONE14-89312.

Planned outage performance is a key measure of how well an Nuclear Power Plant (NPP) is operated. Performance during planned outages strongly affects virtually all of a plant’s performance metrics. In recognition of this fact, NPP operators worldwide have and continue to focus on improving their outage performance. The process of improving outage performance is commonly referred to as ‘Outage Optimization’ in the industry. This paper starts with a summary of the principles of Outage Optimization. It then provides an overview of a process in common use in the USA and elsewhere to manage the improvement of planned outages. The program described is comprehensive in that it involves managing improvement in both the Preparation and Execution phases of outage management.

Topics: Optimization
Commentary by Dr. Valentin Fuster
2006;():149-157. doi:10.1115/ICONE14-89321.

South Texas Project uses a large fault tree to produce scenarios (minimal cut sets) used in quantification of plant availability and event frequency predictions. On the other hand, the South Texas Project probabilistic risk assessment model uses a large event tree, small fault tree for quantifying core damage and radioactive release frequency predictions. The South Texas Project is converting its availability and event frequency model to use a large event tree, small fault in an effort to streamline application support and to provide additional detail in results. The availability and event frequency model as well as the applications it supports (maintenance and operational risk management, system engineering health assessment, preventive maintenance optimization, and RIAM) are briefly described. A methodology to perform availability modeling in a large event tree, small fault tree framework is described in detail. How the methodology can be used to support South Texas Project maintenance and operations risk management is described in detail. Differences with other fault tree methods and other recently proposed methods are discussed in detail. While the methods described are novel to the South Texas Project Risk Management program and to large event tree, small fault tree models, concepts in the area of application support and availability modeling have wider applicability to the industry.

Commentary by Dr. Valentin Fuster
2006;():159-168. doi:10.1115/ICONE14-89326.

During the 2003 outage at the Ringhals Nuclear Plant in Sweden, a leak was found in the vicinity of a Control Rod Drive Mechanism (CRDM) housing nozzle at Unit 1. Based on the ALARA principle for radioactive contamination, a unique repair process was developed. The repair system includes utilization of custom, remotely controlled GTAW-robots, a CNC cutting and finishing machine, snake-arm robots and NDE equipment. The success of the repair solution was based on performing the machining and welding operations from the inside of the SCRAM pipe through the CRDM housing since accessibility from the outside was extremely limited. Before the actual pipe replacement procedure was performed, comprehensive training programs were conducted. Training was followed by certification of equipment, staff and procedures during qualification tests in a full scale mock-up of the housing nozzle. Due to the ingenuity of the overall repair solution and training programs, the actual pipe replacement procedure was completed in less than half the anticipated time. As a result of the successful pipe replacement, the nuclear power plant was returned to normal operation.

Commentary by Dr. Valentin Fuster
2006;():169-174. doi:10.1115/ICONE14-89355.

The fatigue effects have to be revaluated in preparing the license renewal application for the continued operation of an old vintage nuclear power plant. This paper presents a complete fatigue analysis for a branch piping with the effect of thermal stratification, induced by turbulent penetration, and environmental factors on fatigue. Three-dimensional computational fluid dynamics and finite element analyses were performed for the branch line to evaluate the thermal stratification loading. Proposed is a supplementary methodology of considering the effect of environmental factor on the combined conventional peak stress intensity range, based on the NB-3600 of ASME Section III Code, with thermal stratification loading. It can be used for safety enhancement of old vintage nuclear power plants.

Commentary by Dr. Valentin Fuster
2006;():175-178. doi:10.1115/ICONE14-89375.

Improvement of residual stress is effective in a countermeasure to deal with the stress corrosion cracks in pipe welds. A irradiated laser stress improvement process (L-SIP) will be introduced as a method to improve residual stress inside steel pipes. This work method is to improve inner surface residual stress from tensile stress to compressive stress by irradiating laser beam around the welds of steel pipe and utilizing the temperature differences between inner and outer surface.

Topics: Lasers , Stress
Commentary by Dr. Valentin Fuster
2006;():179-186. doi:10.1115/ICONE14-89444.

A Control Rod Control System (CRCS) is one of the most important pieces of equipment in a nuclear power plant because it controls the nuclear reaction by moving the Control Rod Drive Mechanism (CRDM) in the reactor with speed and direction signals from the Reactor Regulating System (RRS). This paper introduces a CRCS with full-duplex configuration and a Local Operator Module (LOM) computer to enhance reliability in comparison to existing simplex systems. The duplex configuration, LOM program, and maintenance test program of the CRCS are explained in detail. This duplex system can perform a failsafe changeover in only a few milliseconds, which results in ‘bumpless’ CRDM coil current control. In particular, because all the control cards are hot-swappable and power converter modules are drawer type, the operator can replace a malfunctioning module without a system shutdown. For ease of operation, the CRCS has its own panel computer called Local Operator Module (LOM) which has a monitoring program installed. The LOM shows all the CRCS operating information such as urgent/non-urgent alarm status, controller operation status, CRDM coil voltage/current, data logs and detailed event alarms. In the CRCS power cabinet, there is an extra panel for easy and simple module maintenance. If the operator suspects that any module of a power cabinet such as a power control card or a power converter module is abnormal, he has only to place the module in the maintenance panel and run the test program to ascertain the module’s sanity. The operator also can utilize the maintenance panel for spare module testing before the spare module is installed in the system and thus avoid installing a defective module.

Commentary by Dr. Valentin Fuster
2006;():187-194. doi:10.1115/ICONE14-89475.

An attempt was made to develop an integrated simulator for maintenance optimization of LWRs (Light Water Reactors) based on PFM (Probabilistic Fracture Mechanics). The concept of the simulator is to provide a method to optimize maintenance activities for representative components and piping systems in nuclear power plants totally and quantitatively in terms of safety, availability and economic efficiency (both from cost and profit). The simulator will also provide a guideline regarding social acceptance of risk-based decision makings. This study has been conducted under “Innovative and Viable Nuclear Energy Technology (IVNET) Development Project” financially supported by Japanese METI.

Commentary by Dr. Valentin Fuster
2006;():195-203. doi:10.1115/ICONE14-89543.

The Best Estimate Power Monitor (BEPM) is a tool that was developed to maximize nuclear power plant generation, while ensuring regulatory compliance in the face of venturi fouling, industry ultra-sonic flowmeter issues and other technical challenges. The BEPM uses ASME approved “best estimate” methodology described in PTC 19.1-1985, “Measurement Uncertainty”, Section 3.8, “Weighting Method.” The BEPM method utilizes many different and independent indicators of core thermal power and independently computes the core thermal power (CTP) from each parameter. The uncertainty of each measurement is used to weight the results of the best estimate computation of CTP such that those with lower uncertainties are weighted more heavily in the computed result. The independence of these measurements is used to minimize the uncertainty of the aggregate result, and the overall uncerainty can be much lower than the uncertainties of any of the individual measured parameters. Examples of the Balance of Plant parameters used in the BEPM are turbine first stage pressure, venturi feedwater flow, condensate flow, main steam flow, high pressure turbine exhaust pressure, low pressure turbine inlet pressure, the two highest pressure feedwater heater extraction pressures, and final feedwater temperature. The BEPM typically makes use of installed plant instrumentation that provide data to the plant computer. Therefore, little or no plant modification is required. In order to compute core thermal power from the independent indicators, a set of baseline data is used for comparison. These baseline conditions are taken from a day when confidence in the value of core thermal power is high (i.e., immediately post outage when venturi fouling is not an issue or from a formal tracer test). This provides the reference point on which to base the core thermal power calculations for each of the independent parameters. The BEPM is effective only at the upper end of the power range, where the independent parameters generally vary in a highly predictable way with changes in core thermal power. This paper will present a detailed description of the BEPM methodology, examples of the BEPM output, and examples of field application. Industry applications of the BEPM include monitoring venturi fouling, verification of an ultrasonic flow meter when used as in input to the secondary calorimetric, and monitoring the performance of other plant equipment that can affect core thermal power. When used routinely as part of a thermal performance monitoring program, the BEPM can be extremely effective in generation maximization, identification of equipment degradation or failure, and identification of potential overpower conditions.

Commentary by Dr. Valentin Fuster
2006;():205-209. doi:10.1115/ICONE14-89554.

Proper and rapid identification of malfunctions is of premier importance for the safe operation of Nuclear Power Plants (NPP). Many monitoring or/and diagnosis methodologies based on artificial and computational intelligence have been proposed to aid operator to understand system problems, perform trouble-shooting action and reduce human error under serious pressure. However, because no single method is adequate to handle all requirements for diagnostic system, hybrid approaches where different methods work in conjunction to solve parts of the problem interest researchers greatly. In this study, Multilevel Flow Models (MFM) and Artificial Neural Network (ANN) are proposed and employed to develop a fault diagnosis system with the intention of improving the success rate of recognition on the one hand, and improving the understandability of diagnostic process and results on the other hand. Several simulation cases were conducted for evaluating the performance of the proposed diagnosis system. The simulation results validated the effectiveness of the proposed hybrid approach.

Commentary by Dr. Valentin Fuster
2006;():211-218. doi:10.1115/ICONE14-89597.

Within the scope of PLEX, a systematic and efficient ageing and plant life management system is becoming more and more important to ensure a safe and economical power plant operation in spite of continuous plant ageing. For the methodical implementation of PLIM & PLEX strategies, AREVA NP has developed the software tool COMSY. This knowledge-based program integrates degradation analysis tools with an inspection data management system. COMSY provides the capability to establish a program guided technical documentation by utilizing a virtual plant model which includes information regarding thermal hydraulic operation, water chemical conditions and materials applied for mechanical components. It provides the option to perform a plant-wide screening for identifying system areas, which are sensitive for degradation mechanisms typically experienced in nuclear power plants (FAC, corrosion fatigue, IGSCC, Pitting, etc.). If a system area is identified as being susceptible to degradation, a detailed analysis function enables the condition-oriented service life evaluation of vessels and piping systems in order to localize and conservatively quantify the effect of degradation. Based on these forecasts with COMSY, specific strategies can be developed to mitigate the effect of degradation and inspection activities can be focused on degradation sensitive areas. In addition, a risk-informed assessment tool serves to optimize inspection activities in respect to degradation potential and the associated damage consequence. After an in-service inspection is performed for a distinct location, the inspection data is to be evaluated according to generally accepted procedures. For this purpose an integrated inspection data management system module provides standardized, interactively operated evaluation functions. The key inspection results are transmitted as feedback in respect to the as-is condition of the component. Subsequently, all further life evaluations of the associated component are calibrated against the inspection results. The compiled condition-oriented knowledge provides the basis for a continuous optimization resulting in tailored inspection and maintenance programs geared to the specific plant. The systematic closed loop process ensures the generation of up-to-date plant documentation relating to the technical as-is status of the plant, as all the data involved in the process are compiled in a “living” documentation structure. The implementation of COMSY in various nuclear power plants has confirmed that systematic plant life management makes good economic sense, as cost reductions can be achieved while increasing the plants availability.

Commentary by Dr. Valentin Fuster
2006;():219-225. doi:10.1115/ICONE14-89624.

This study provides the insights gained from the Probabilistic Safety Assessment (PSA) model update of several Entergy Nuclear South (ENS) plants with respect to truncation convergence based on the limited guidance on the issue in the industry. The industry rule of thumb, the ASME and NRC guidance and requirements on the subject have been reviewed. The recent model updates performed at some of the ENS plants (River Bend, ANO 1 and 2) considered these criteria. Based on the current criteria used in the industry for truncation convergence, the recent PSA model update results for the River Bend Station (RBS) and ANO-1 are not converging even at a low truncation limit of 1E−11/reactor-year (yr). Many improvements were introduced in the recent model updates and convergence was expected at higher truncation values. This paper discusses the issues identified that are related to the convergence of the PSA results at low truncation limits.

Commentary by Dr. Valentin Fuster
2006;():227-232. doi:10.1115/ICONE14-89636.

This paper provides an overview of the experience in developing and using the initiating event fault trees (IEFTs) in the base and EOOS Probabilistic Safety Assessment (PSA) models at the Entergy Nuclear South (ENS) sites. Simple fault tree models can be developed and linked to the master fault tree in order to provide a better estimate of plant-specific initiating event frequencies. If developed considering certain aspects, there can be significant advantages to using IEFT methodology for some initiators, especially in the plant EOOS models. The most important advantage is to have EOOS re-calculate and use such initiating event frequencies for different alignments taking the components that are out of service into consideration while keeping the quantification time and complexity of the cutsets at an acceptable level. The important steps for developing the IEFTs, some insights gained through the experience at the ENS sites, and the advantages of using IEFTs will be discussed in this paper.

Commentary by Dr. Valentin Fuster
2006;():233-237. doi:10.1115/ICONE14-89643.

The cooling capacity of cooling towers is influenced by multiple constructive and atmospheric parameters in a very complex way. This leads to strong variations of the measured cold-water temperature and causes unacceptable unreliability of conventional acceptance tests, which are based on single point measurements. In order to overcome this lack of accuracy a new approach to acceptance test based on process data reconciliation has been developed by BTB Jansky and applied at a nuclear power plant. This approach uses process data reconciliation according to VDI 2048 [1, 2] to evaluate datasets over a long period covering different operating conditions of the cooling tower. Data reconciliation is a statistical method to determine the true process parameters with a statistical probability of 95% by considering closed material-, mass- and energy balances. Datasets which are not suitable for the evaluation due to strong transient gradients are excluded beforehand, according to well-defined criteria. The reconciled cold-water temperature is then compared, within a wet bulb temperature range of 5°C to 20°C to the manufacturer’s guaranteed temperature. Finally, if the average deviation between reconciled and guaranteed value over the evaluated period is below zero, the cooling tower guarantee is fulfilled.

Topics: Cooling towers
Commentary by Dr. Valentin Fuster
2006;():239-248. doi:10.1115/ICONE14-89644.

Significant differences have been identified in loss of offsite power (LOSP or LOOP) event description, category, duration, and applicability between the LOSP events used in NUREG/CR-6890 and ENS’ LOSP packages, which were based on EPRI LOSP reports with plant-specific applicability analysis. Thus it is appropriate to reconcile the LOSP data listed in the subject NUREG and EPRI reports. A cross comparison showed that 62 LOSP events in NUREG/CR-6890 were not included in the EPRI reports while 4 events in EPRI reports were missing in the NUREG. Among the 62 events missing in EPRI reports, the majority were applicable to shutdown conditions, which could be classified as category IV events in EPRI reports if included. Detailed reviews of LERs concluded that some events did not result in total loss of offsite power. Some LOSP events were caused by subsequent component failures after a turbine/plant trip, which have been modeled specifically in most ENS plant PRA models. Moreover, ENS has modeled (or is going to model) the partial loss of offsite power events with partial LOSP initiating events. While the direct use of NUREG/CR-6890 results in SPAR models may be appropriate, its direct use in ENS’ plant PRA models may not be appropriate because of modeling details in ENS’ plant-specific PRA models. Therefore, this paper lists all the differences between the data in NUREG/CR-6890 and EPRI reports and evaluates the applicability of the LOSP events to ENS plant-specific PRA models. The refined LOSP data will characterize the LOSP risk in a more realistic fashion.

Commentary by Dr. Valentin Fuster
2006;():249-252. doi:10.1115/ICONE14-89680.

The Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) has provided the capability for irradiation testing of nuclear fuels and materials since 1967, and is expected to operate for several more years. Within the scope of extending the life of a nuclear plant is dealing with aging and obsolescence issues. A component can be considered obsolete if the manufacturer no longer supports the component, or if the manufacturer does not even exist anymore. Though these components can be considered obsolete, the cost of obsolescence may or may not be significant; it may be more cost-effective to leave and/or repair the component rather than to replace it. The project at hand is to develop a tool that will not only identify these components, series of components, or entire systems that are obsolete, but to quantify the cost of obsolescence. This engineering tool will be based on empirical formulas created from data collected from factors that deal with obsolescence. These factors are primarily, the cost of item replacement, current cost of maintenance, cost of maintenance of the replacement, cost of failure, risk of failure, safety, increase in performance/efficiency, length of manufacturer’s support, and so forth. The objective is to be able to look at the outcome of this engineering tool and clearly see what needs to be replaced, be it a component, series of components, or an entire system. If there are several such replacements needed, which one(s) have the greatest priority for replacement. Therefore the engineering tool will identify, quantify, and prioritize the cost of obsolescence in the plant. An engineering tool of this type should find application in a number of nuclear and non-nuclear facilities. While the engineering tool is being developed, the first stage of development will be on system components. Once the foundation is set it will be used to evaluate other systems and eventually expand and develop the engineering tool for the entire plant.

Commentary by Dr. Valentin Fuster
2006;():253-259. doi:10.1115/ICONE14-89683.

The concept of Standard Maintenance Windows has been widely used in the planned outage of light water reactor in the world. However, due to the specific feature of Pressurized Heavy Water Reactor (PHWR), it has not come to a consensus for the PHWR owners to adopt Standard Maintenance Windows for planned outage aiming at the optimization of outage duration. Third Qinshan Nuclear Power Company (TQNPC), with their experience gained in the previous outages and with reference to other PHWR power plants, has identified a set of Standard Maintenance Windows for planned outage. It can be applied to similar PHWR plants and with a few windows that are specific to Qinshan Phase III NPP. The use of these Standard Maintenance Windows in planned outage has been proved to be effective in control shutdown nuclear safety, minimize the unavailability of safety system, improve the efficient utilization of outage duration, and improved the flexibility of outage schedule in the case of emergency issue, which forced the revision of outage schedule. It has also formed a solid foundation for benchmarking. The identification of Standard Maintenance Windows and its application will be discussed with relevant cases for the common improvement of outage duration.

Commentary by Dr. Valentin Fuster
2006;():261-269. doi:10.1115/ICONE14-89697.

This paper presents a nonlinear structural damage identification technique, based on an interactive data mining approach, which integrates a human cognitive model in a data mining loop. A mining control agent emulating human analysts is developed, which directly interacts with the data miner, analyzing and verifying the output of the data miner and controlling the data mining process. Additionally, an artificial neural network method, which is adopted as a core component of the proposed interactive data mining method, is evolved by adding a novelty detecting and retraining function for handling complicated nuclear power plant quake-proof data. Plant quake-proof testing data has been applied to the system to show the validation of the proposed method.

Commentary by Dr. Valentin Fuster
2006;():271-276. doi:10.1115/ICONE14-89711.

The safety upgrading of the Phenix plant undertaken between 1994 and 1997 involved a vast inspection programme of the reactor, the external storage drum and the secondary sodium circuits in order to meet the requirements of the defence-in-depth safety approach. The three lines of defence were analysed for every safety related component: demonstration of the quality of design and construction, appropriate in-service inspection and controlling the consequences of an accident. The in-service reactor block inspection programme consisted in controlling the core support structures and the high-temperature elements. Despite the fact that limited consideration had been given to inspection constraints during the design stage of the reactor in the 1960’s, as compared to more recent reactor projects such as the European Fast Reactor (EFR), all the core support line elements were able to be inspected. The three following main operations are described: Ultrasonic inspection of the upper hangers of the main vessel, using small transducers able to withstand temperatures of 130 °C. Inspection of the conical shell supporting the core diagrid. A specific ultrasonic method and a special implementation technique were used to control the under sodium structure welds, located up to several meters away from the scan surface. Remote inspection of the hot pool structures, particularly the core cover plug after partial sodium drainage of the reactor vessel. Other inspections are also summarized: control of secondary sodium circuit piping, intermediate heat exchangers, primary sodium pumps, steam generator units and external storage drum. The pool type reactor concept, developed in France since the 1960’s, presents several favourable safety and operational features. The feedback from the Phenix plant also shows real potential for in-service inspection. The design of future generation IV sodium fast reactors will benefit from the experience acquired from the Phenix plant.

Commentary by Dr. Valentin Fuster
2006;():277-283. doi:10.1115/ICONE14-89724.

In the nuclear energy field, there are so many difficult things that even people who are working in this field are not much familiar with, such as, Dose evaluation, Dose management, etc. Thus, so many efforts have been done to achieve the knowledge and data for understanding. Although some data had been achieved, the applications of these data to necessary cases were more difficult job. Moreover, the type of Dose evaluation program until now was ‘Console type’ which is not easy enough to use for the beginners. To overcome the above causes of difficulties, the window-based integrated program and database management were developed in our research lab. The program, called as INSREC, consists of four sub-programs as follow; INSREC-NOM, INSREC-ACT, INSREC-MED, and INSREC-EXI. In ICONE 11 conference, INSREC-program (ICONE-36203) which can evaluates on/off-site dose of nuclear power plant in normal operation was introduced. Upgraded INSREC-program which will be presented in ICONE 14 conference has three additional codes comparing with pre-presented INSREC-program. Those subprograms can evaluate on/off-site Dose of nuclear power plant in accident cases. And they also have the functions of ‘Dose evaluation and management’ in the hospital and provide the ‘Expert system’ based on knowledge related to nuclear energy/radiation field. The INSREC-NOM, one of subprograms, is composed of ‘Source term evaluation program’, ‘Atmospheric diffusion factor evaluation program’, ‘Off-site dose evaluation program’, and ‘On-site database program’. The INSREC-ACT is composed of ‘On/Off-site dose evaluation program’ and ‘Result analysis program’ and the INSREC-MED is composed of ‘Workers/patients dose database program’ and ‘Dose evaluation program for treatment room’. The final one, INSREC-EXI, is composed of ‘Database searching program based on artificial intelligence’, ‘Instruction program,’ and ‘FAQ/Q&A boards’. Each program was developed by using of Visual C++, Microsoft Access mainly. To verify the reliability, some suitable programs were selected such as AZAP and Strardose programs for the comparison. The AZAP program was selected for the on/off-site dose evaluation during the normal operation of nuclear reactor and Stardose program was used for the on/off-site dose evaluation in accident. The MCNP code was used for the dose evaluation and management in the hospital. Each comparison result was acceptable in errors analysis. According to the reliable verification results, it was concluded that INSREC program had an acceptable reliability for dose calculation and could give many proper dada for the sites. To serve the INSREC to people, the proper server system was constructed. We gave chances for the people (user) to utilize the INSREC through network connected to server system. The reactions were pretty much good enough to be satisfied. For the future work, many efforts will be given to improve the better user-interface and more necessary data will be provided to more people through database supplement and management.

Commentary by Dr. Valentin Fuster
2006;():285-294. doi:10.1115/ICONE14-89726.

This work presents results of robustness verification of artificial neural network correlations that improve the real time prediction of the power peak factor for reactor protection systems. The input variables considered in the correlation are those available in the reactor protection systems, namely, the axial power differences obtained from measured ex-core detectors, and the position of control rods. The correlations, based on radial basis function (RBF) and multilayer perceptron (MLP) neural networks, estimate the power peak factor, without faulty signals, with average errors between 0.13%, 0.19% and 0.15%, and maximum relative error of 2.35%. The robustness verification was performed for three different neural network correlations. The results show that they are robust against signal degradation, producing results with faulty signals with a maximum error of 6.90%. The average error associated to faulty signals for the MLP network is about half of that of the RBF network, and the maximum error is about 1% smaller. These results demonstrate that MLP neural network correlation is more robust than the RBF neural network correlation. The results also show that the input variables present redundant information. The axial power difference signals compensate the faulty signal for the position of a given control rod, and improves the results by about 10%. The results show that the errors in the power peak factor estimation by these neural network correlations, even in faulty conditions, are smaller than the current PWR schemes which may have uncertainties as high as 8%. Considering the maximum relative error of 2.35%, these neural network correlations would allow decreasing the power peak factor safety margin by about 5%. Such a reduction could be used for operating the reactor with a higher power level or with more flexibility. The neural network correlation has to meet requirements of high integrity software that performs safety grade actions. It is shown that the correlation is a very simple algorithm that can be easily codified in software. Due to its simplicity, it facilitates the necessary process of validation and verification.

Commentary by Dr. Valentin Fuster
2006;():295-301. doi:10.1115/ICONE14-89802.

Many safety requirements for research reactors are quite similar to those of power reactors. For research reactors with a higher hazard potential, the use of safety codes and guides for power reactors is more appropriate. However, there are many important differences between power reactors and research reactors that must be taken into account to ensure that adequate safety margins are available in design and operation of the research reactor. Most research reactors may have small potential for hazard to the public compared to power reactors but may pose a greater potential hazard to the plant operators. The need for greater flexibility in use of research reactors for individual experiments requires a different safety approach. Safety rules for power reactors are required to be substantially modified for application to specific research reactor. Following the intent of the available safety guides for surveillance and In-Service Inspection of Nuclear Power Plants, guidelines were formulated to develop surveillance and In-Service Inspection programme for research reactors Cirus and Dhruva. Based on the specific design of these research reactors, regulatory requirements, the degree of sophistication and experience of the technical organization involved in operating the research reactor, guidelines were evolved for developing and implementing the surveillance and In-Service Inspection programme for research reactors Cirus (40 MWt) and Dhruva (100 MWt) located at Bhabha Atomic Research Centre, Trombay, Mumbai, India. Paper describes the approach adopted for formulation of surveillance and In-service Inspection programme for Dhruva reactor in detail.

Commentary by Dr. Valentin Fuster
2006;():303-309. doi:10.1115/ICONE14-89805.

Power Economic Dispatch (ED) is vital and essential daily optimization procedure in the system operation. Present day large power generating units with multi-valves steam turbines exhibit a large variation in the input-output characteristic functions, thus non-convexity appears in the characteristic curves. Various mathematical and optimization techniques have been developed, applied to solve economic dispatch (ED) problem. Most of these are calculus-based optimization algorithms that are based on successive linearization and use the first and second order differentiations of objective function and its constraint equations as the search direction. They usually require heat input, power output characteristics of generators to be of monotonically increasing nature or of piecewise linearity. These simplifying assumptions result in an inaccurate dispatch. Genetic algorithms have used to solve the economic dispatch problem independently and in conjunction with other AI tools and mathematical programming approaches. Genetic algorithms have inherent ability to reach the global minimum region of search space in a short time, but then take longer time to converge the solution. GA based hybrid approaches get around this problem and produce encouraging results. This paper presents brief survey on hybrid approaches for economic dispatch, an architecture of extensible computational framework as common environment for conventional, genetic algorithm & hybrid approaches based solution for power economic dispatch, the implementation of three algorithms in the developed framework. The framework tested on standard test systems for its performance evaluation.

Commentary by Dr. Valentin Fuster
2006;():311-315. doi:10.1115/ICONE14-89838.

Mitsubishi Heavy Industries, Ltd. (MHI) completed replacement work of upper reactor internals (UCI) and lower reactor internals (LCI) of the pressurized water reactor in Shikoku Electric Power Company’s Ikata Unit No.1 by “the all-in-one-piece extraction method” introduced in the document of [ICONE14-89233]. In the pressurized water reactor (PWR) plant, the UCI are usually removed from the reactor vessel (RV) independently and reinstalled into the RV again every refueling outage. The LCI are independently able to be removed from the RV and reinstalled again during in-service inspection, too. In the boiling water reactor (BWR) plant, there are several cases of replacing BWR shrouds by cutting small and containing in a container. But no replacement of all reactor internals (CI) for the PWR, in one piece without splitting or cutting, has been reported. The purpose of this paper is to introduce the key points about the design and manufacture of the storage cask for old reactor internals in the replacement work by “the all-in-one-piece extraction method”.

Topics: Design , Storage
Commentary by Dr. Valentin Fuster
2006;():317-325. doi:10.1115/ICONE14-89842.

This paper addresses the necessity and feasibility of the first major overhaul on the GIS based on the analysis of the special conditions and the issues we confronted; After the comparison of various schemes, the optimized scheme is put forward; the paper also expounds the proper preparation and cautious practice which led to the hard but final accomplishment of the initial overhaul on the GIS; this article further explains the necessity of the major overhaul on the GIS through the disposal of abnormalities during the execution of this major overhaul.

Commentary by Dr. Valentin Fuster
2006;():327-332. doi:10.1115/ICONE14-89903.

The Quad Cities Units 1 and 2 have a history of steam line vibration issues. The implementation of an Extended Power Uprate resulted in significant increases in steam line vibration as well as acoustic loading of the steam dryers, which led to equipment failures and fatigue cracking of the dryers. This paper discusses the results of extensive data collection on the Quad Cities Unit 2 replacement dryer and the Main Steam Lines. This data was taken with the intent of identifying acoustic sources in the steam system. Review of the data confirmed that vortex shedding coupled column resonance in the relief and safety valve stub pipes were the principal sources of large magnitude acoustic loads in the main steam system. Modifications were developed in subscale testing to alter the acoustic properties of the valve standpipes and add acoustic damping to the system. The modifications developed and installed consisted of acoustic side branches that were attached to the Electromatic Relief Valve (ERV) and Main Steam Safety Valve (MSSV) attachment pipes. Subsequent post-modification testing was performed in plant to confirm the effectiveness of the modifications. The modifications have been demonstrated to reduce vibration loads at full Extended Power Uprate (EPU) conditions to levels below those at Original Licensed Thermal Power (OLTP).

Topics: Acoustics , Stress , Steam
Commentary by Dr. Valentin Fuster

Component Reliability and Materials Issues

2006;():333-339. doi:10.1115/ICONE14-89013.

A specific tribometer equipped for electrochemical measurements in pressurized high temperature water (PHTW) is being used to study the effect of the water chemistry on the wear rate of a stainless steel. First results indicate that a slightly acidic pH increases drastically the wear rate compared to slightly alkaline pH. The influence of hydrogen partial pressure and contact kinematics is also described.

Commentary by Dr. Valentin Fuster
2006;():341-347. doi:10.1115/ICONE14-89014.

As it was observed in the past Gas Cooled Reactors, it is expected that the VHTR’s cooling gas will be polluted by air ingresses, in-leakages or the degassing of adsorbed species out of the large amount of graphite. These impureties are reactive and may interact with the core graphite and with the metallic materials and may cause some loss or damage of their properties.

Commentary by Dr. Valentin Fuster
2006;():349-359. doi:10.1115/ICONE14-89028.

This paper describes the results of investigations of 08Cr16Ni11Mo3 (AISI 316 steel analogue) austenitic stainless steel irradiated in BN-350 breeder reactor at irradiation conditions close to that for Light Water Reactor (LWR) Internals. The pores were found in 08Cr16Ni11Mo3 steel irradiated at temperature 280°C up to rather low damage 1.3 dpa and with dose rate 3.9×10−9 dpa/s. There were obtained dose rate dependencies of yield strength, ultimate strength and ductility for 08Cr16Ni11Mo3 steel irradiated up to 7–13 dpa at 302–311°C. These dependencies show a decrease in both yield strength and ultimate strength when dose rate decreases. There was observed an apparent decrease in total elongation when dose rate decreases, which was presumably connected with the pores formation in the steel at low dose rates.

Commentary by Dr. Valentin Fuster
2006;():361-369. doi:10.1115/ICONE14-89045.

A novel type of Bolted Flanged Connection with bolts and gasket manufactured on a basis of advanced Shape Memory Alloys is examined. Presented approach combined with inverse flexion flange design of plant/piping joint reveals a significant increase of internal pressure under conditions of a variety of operating temperatures relating to critical plant/piping systems.

Commentary by Dr. Valentin Fuster
2006;():371-378. doi:10.1115/ICONE14-89085.

It is commonly requested that steam generator tubes with defects exceeding 40% of wall thickness in depth should be plugged to sustain all postulated loads with appropriate margin. The critical defect dimensions have been determined based on the concept of plastic instability. This criterion, however, is known to be too conservative for some locations and types of defects. In this context, the accurate failure estimation for steam generator tubes with a defect draws increasing attention. Although several guidelines have been developed and are used for assessing the integrity of defected tubes, most of these guidelines are related to stress corrosion cracking or wall-thinning phenomena. As some of steam generator tubes are also failed due to fretting and so on, alternative failure estimation schemes for relevant defects are required. In this paper, three-dimensional finite element (FE) analyses are carried out under internal pressure condition to simulate the failure behavior of steam generator tubes with different defect configurations; elliptical wastage type, wear scar type and rectangular wastage type defects. Maximum pressures based on material strengths are obtained from more than a hundred FE results to predict the failure of the steam generator tube. After investigating the effect of key parameters such as wastage depth, wastage length and wrap angle, simplified failure estimation equations are proposed in relation to the equivalent stress at the deepest point in wastage region. Comparison of failure pressures predicted according to the proposed estimation scheme with some corresponding burst test data shows good agreement, which provides a confidence in the use of the proposed equations to assess the integrity of steam generator tubes with wear-type defects.

Commentary by Dr. Valentin Fuster
2006;():379-388. doi:10.1115/ICONE14-89124.

To identify thermal loadings (thermal shocks and thermal stratification), in German NPPs special fatigue monitoring systems have been installed. The detailed temperature measurement uses sheathed thermocouples, which are located on the external component surface. Tightening straps are used for the widespread method of locking the thermocouples into position. The basic objective of the paper is to identify several sources of error in fatigue analyses and to evaluate their specific importance. In fatigue analyses based on the outer-surface temperature, the thermal situation at the inner-surface has to be determined. This leading analysis step is not regulated in the ASME III code. In the paper, the variance in the calculated fatigue is quantified for different inner surface temperature profiles with a common good compliance to the outer surface temperature. Another focus of the paper is the presentation of findings concerning the exactness of fatigue analyses based on FEA. The effect of geometrical truncation (mesh density) and the time-increments on the calculated fatigue level is shown. Furthermore, regarding discontinuities with thermal shock loading, the differences in the calculated fatigue between applying ASME NB-3200 on the one hand and NB-3600 on the other are quantified. Finally, the uncertainty in the calculated fatigue associated with the known variance of the physical material property values (E-Modulus, thermal conductivity, thermal expansion) is highlighted.

Commentary by Dr. Valentin Fuster
2006;():389-400. doi:10.1115/ICONE14-89176.

Structural integrity of piping systems is important to plant safety and operability. Information on degradation and failure of piping components and systems is collected and evaluated by regulatory agencies, international organizations (e.g., OECD/NEA and IAEA) and industry organizations worldwide to establish systematic feedback to reactor regulation and research and development programs associated with non-destructive examination (NDE) technology, in-service inspection (ISI) programs, leak-before-break evaluations, risk-informed ISI, and probabilistic safety assessment (PSA) applications involving passive component reliability. In 2002, the Nuclear Energy Agency (NEA) of the Organization for Economic Co-Operation and Development (OECD) has initiated an international pipe failure data collection and exchange project. The OECD Pipe Failure Data Exchange (OPDE) Project has been established to encourage multilateral co-operation in the collection and analysis of data relating to pipe failure events in commercial nuclear power plants. At present, the database contains 3644 records to which twelve participating countries contributed. This paper presents a brief description of the ODPE project objectives and work scope, as well as the Canadian contribution on data validation with respect to development and application of the pipe failure data collection on which OPDE is based. It gives a number of tables and figures that can be obtained from these records, with selected data ranging from a very broad (i.e. level of participation in the database from each member country), to very specific (i.e. plant operational state at time of pipe failure discovery for CANDU reactors).

Topics: Pipes , Failure data
Commentary by Dr. Valentin Fuster
2006;():401-407. doi:10.1115/ICONE14-89187.

Equilibrium evaporation behavior was experimentally investigated for polonium (210 Po) in liquid lead-bismuth eutectic (LBE) and for rare-earth elements gadolinium (Gd) and europium (Eu) in LBE to understand and clarify the transfer behavior of toxic impurities from LBE coolant to a gas phase. The experiments utilized the ‘transpiration method’ in which saturated vapor in an isothermal evaporation pot was transported by inert carrier gas and collected outside of the pot. While the previous paper ICONE12-49111 has already reported the evaporation behavior of LBE and of tellurium in LBE, this paper summarizes the outlines and the results of experiments for important impurity materials 210 Po and rare-earth elements which are accumulated in liquid LBE as activation products and spallation products. In the experiments for rare-earth elements, non-radioactive isotope was used. The LBE pool is about 330–670 g in weight and has a surface area of 4cm × 14cm. 210 Po experiments were carried out with a smaller test apparatus and radioactive 210 Po produced through neutron irradiation of LBE in the Japan Materials Testing Reactor (JMTR). We obtained fundamental and instructive evaporation data such as vapor concentration, partial vapor pressure of 210 Po in the gas phase, and gas-liquid equilibrium partition coefficients of the impurities in LBE under the temperature condition between 450 and 750°C. The 210 Po test revealed that Po had characteristics to be retained in LBE but was still more volatile than LBE solvent. A part of Eu tests implied high volatility of rare-earth elements comparable to that of Po. This tendency is possibly related to the local enrichment of the solute near the pool surface and needs to be investigated more. These results are useful and indispensable for the evaluation of radioactive materials transfer to the gas phase in LBE-cooled nuclear systems.

Topics: Evaporation
Commentary by Dr. Valentin Fuster
2006;():409-414. doi:10.1115/ICONE14-89195.

Search of new energy sources draws the increasing attention to use for this purpose of reactors. In the Europe some years the program EUROATOM uniting scientific of the many countries for the decision of constructive problems at designing of fusion reactors operates. One of the main things in this program is the problem of liquid metals breeder blanket behaviour. Structural material of blanket should meet high requirements because of extreme operating conditions. Therefore the knowledge of the effect of metals flow velocity, temperatures and also a neutron irradiation and a magnetic field on the corrosion processes are necessary. At the moment the eutectic lead -lithium (Pb-17Li) is considered as the most suitable tritium breeder material [1–3]. In turn as a structural material have been tested both many austenitic and ferritic-martensitic steels [2–4]. As the optimum variant is considered steel EUROFER 97, which corrosion rate in liquid Pb-17Li eutectic is the least [3,4]. However, these results have been received without taking into account influence of a strong magnetic field. At the same time, this influence should be essential, as because of change of hydrodynamics of a liquid metal flow, and because of interaction of a magnetic field with a ferromagnetic steel. It has been shown in [5,6] that the magnetic field leads to increase of corrosion rate for austenitic (316L) and martensitic (1,4914) steels. Experimental data for EUROFER 97, and also a theoretical substantiation of the phenomenon are absent, that creates essential difficulties for forecasting working capacity of blanket construction. The aim of presented work were the theoretical and experimental investigations of magnetic field influence on the corrosion of EUROFER 97 steel exposed to flowing Pb-17 Li in specific designed loop.

Commentary by Dr. Valentin Fuster
2006;():415-418. doi:10.1115/ICONE14-89211.

A fretting/wear degradation at the tube support in the U-bend region of a steam generator (SG) of a pressurized water reactor (PWR) has been reported. Simulated fretted flaws were machined on SG tubes of 195 mm in length. A pressure test was carried out with the tubes at room temperature by using a high pressure test facility which consisted of a water pressurizing pump, a test specimen section and a control unit. Water leak rates just after a ligament rupture or a burst were measured. Tubes degraded by up to 70% of the tube wall (TW) showed a high safety margin in terms of the burst pressure during normal operating conditions. Tubes degraded by up to 50% of the TW did not show a burst. Burst pressure depended on the defect depths rather than on the wrap angles. The tube with a wrap angle of 0° showed a fish mouth fracture, whereas the tube with a 45° wrap angle showed a three way fracture.

Commentary by Dr. Valentin Fuster
2006;():419-424. doi:10.1115/ICONE14-89228.

Laser peening is a process to improve residual stress from tensile to compressive in surface layer of materials by irradiating high-power laser pulses on the material in water. Toshiba has developed a laser peening system composed of Q-switched Nd:YAG laser oscillators, laser delivery equipment and underwater remote handling equipment. We have applied the system for Japanese operating BWR power plants as a preventive maintenance measure for stress corrosion cracking (SCC) on reactor internals like core shrouds or control rod drive (CRD) penetrations since 1999. As for PWRs, alloy 600 or 182 can be susceptible to primary water stress corrosion cracking (PWSCC), and some cracks or leakages caused by the PWSCC have been discovered on penetrations of reactor vessel heads (RVHs), reactor bottom-mounted instrumentation (BMI) nozzles, and others. Taking measures to meet the unconformity of the RVH penetrations, RVHs themselves have been replaced in many PWRs. On the other hand, it’s too time-consuming and expensive to replace BMI nozzles, therefore, any other convenient and less expensive measures are required instead of the replacement. In Toshiba, we carried out various tests for laser-peened nickel base alloys and confirmed the effectiveness of laser peening as a preventive maintenance measure for PWSCC. We have developed a laser peening system for PWRs as well after the one for BWRs, and applied it for BMI nozzles, core deluge line nozzles and primary water inlet nozzles of Ikata Unit 1 and 2 of Shikoku Electric Power Company since 2004, which are Japanese operating PWR power plants. In this system, laser oscillators and control devices were packed into two containers placed on the operating floor inside the reactor containment vessel. Laser pulses were delivered through twin optical fibers and irradiated on two portions in parallel to reduce operation time. For BMI nozzles, we developed a tiny irradiation head for small tubes and we peened the inner surface around J-groove welds after laser ultrasonic testing (LUT) as the remote inspection, and we peened the outer surface and the weld for Ikata Unit 2 supplementary. For core deluge line nozzles and primary water inlet nozzles, we peened the inner surface of the dissimilar metal welding, which is of nickel base alloy, joining a safe end and a low alloy metal nozzle. In this paper, the development and the actual application of the laser peening system for PWR power plants will be described.

Commentary by Dr. Valentin Fuster
2006;():425-432. doi:10.1115/ICONE14-89230.

In-service inspection (ISI) is carried out to confirm the integrity of the main components of the Fast Breeder Reactor (FBR) “MONJU”. The weld-joints are examined by using an inspection device which has a glass fiber scope for visual examination and a horizontally polarized shear (SH) wave electromagnetic acoustic transducer (EMAT) for volumetric testing. The ambient temperature during the inspection is 200°C and the irradiation field is 10 Sv/hr. A new inspection device has been developed in order to improve the visual test performance, volumetric test performance and controllability of the inspection device reflecting the experience of the original test. In this paper, detail of the new inspection device and the test results of sensors such as the CCD camera, EMAT and bead sensor are reported. The paper also reports on the CCD camera cooling system and other components.

Commentary by Dr. Valentin Fuster
2006;():433-441. doi:10.1115/ICONE14-89250.

Four types of surveillance programs were (are) realized in Slovak NPP’s: • “Standard Surveillance Specimen Program” (SSSP) was finished in Jaslovské Bohunice V-2 Nuclear Power Plant (NPP) Units 3 and 4; • “Extended Surveillance Specimen Program” (ESSP), was prepared for Jaslovské Bohunice NPP V-2 with aim to validate the SSSP results; • For the Mochovce NPP Unit 1 and 2 was prepared completely new surveillance program “Modern Surveillance Specimen Program” (MSSP), based on the philosophy that the results of MSSP must be available during all NPP service life; • For the Bohunice V-1 NPP was finished “New Surveillance Specimen Program” (NSSP) coordinated by IAEA, which gave arguments for prolongation of service life these units for minimum 20 years; • New Advanced Surveillance Specimen Program (ASSP) for Bohunice V-2 NPP (units 3 and 4) and Mochovce NPP (units 1, 2) is approved now. ASSP is dealing with the irradiation embrittlement of heat affected zone (HAZ) and RPV’s austenitic cladding, which were not evaluated till this time in surveillance programs. SSSP started in 1979 and was finished in 1990. ESSP program started in 1995 and will be finished in 2007, was prepared with aim of: • increasing of neutron fluence measurement accuracy; • substantial improvement the irradiation temperature measurement; • fixed orientation of samples to the centre of the reactor core; • minimum differences of neutron dose for all the Charpy-V notch and COD specimens; • the dose rate effect evaluation. In the year 1996 was started the new surveillance specimen program for the Mochovce RPV’s unit-1 and 2, based on the fundamental postulate — to provide the irradiation embrittlement monitoring till the end of units operation. The “New Surveillance Specimen Program” (NSSP) prepared in the year 1999 for the Bohunice V-1 NPP was finished in the year 2004. Main goal of this program was to evaluate the weld material properties degradation due to the irradiation and recovery efficiency by annealing too. The results showed us that the NPP V-1 RPV’s can be operated without any technical limitation for minimum next 20 years . Advanced Surveillance Specimen Program (ASSP) for Bohunice V-2 NPP (units 3 and 4) and Mochovce NPP (units 1 and 2) was prepared as the part of research project dealing with the WWER-440 units ageing management.

Commentary by Dr. Valentin Fuster
2006;():443-450. doi:10.1115/ICONE14-89271.

Corrosion integrity of canister in the concrete cask for spent fuel storage is very important because the canister serves to maintain the sealability over the storage period of 40 to 60 years. Natural exposure and accelerated corrosion tests of conventional stainless steels for canister, that are Type 304, 304L, and 316(LN), for concrete cask’s canister have been conducted by using many three Point Bending (3PB) test specimens and compared. The SCC propagation rates in Type 304 and 304L at the natural condition were about 1.2E−12 to 1.8E−11 m/s at the K (Stress Intensity Factor) range of 0.6 to 9.0 MPa√m, and that of the accelerate test (60 degrees C, 95%RHS., filled with NaCl mist) were about 1.0E−10 to 3.5E−9 m/s at the K range of 0.3 to 32 MPa√m. The SCC propagation rates under both natural and accelerated conditions were independent with K. Both da/dt values of the direct exposure test and of the under glass exposure test were in the same scattering band.

Commentary by Dr. Valentin Fuster
2006;():451-456. doi:10.1115/ICONE14-89286.

PSA studies, that were developed for some NPPs, permit the using of the created models to perform many research tests, in order to optimize the structures, systems and components (SSCs) operation or to identify the NPP or systems weaknesses, due to specific or special factors. SSCs that influence decisively the NPP reliability are considered as critical. Also, for the accident conditions, the SSC, which have a major influence to the system availability or operability, are considered as critical. Many worldwide NPPs reached the life time or are very close to do that. Several SSCs have shorter life times than NPP’s life time. Ageing is one of the factors that decrease the SSC life time. Due to ageing, if are not replaced, some SSCs, or groups of redundant SSCs, become critical looking to safety. Some questions for what to do in the situation when a SSC must be replaced and the SSC specific manufacturer doesn’t exist, could also be put. The paper tried to solve the problem of SSC modeling by introducing of an ageing factor in SSC model. Fault tree (F/T) modeling approach is assumed. There are two possibilities for modeling: failure rates that are changed or specific MCS (minimal cut set) term modified by ageing. Risk analysis and PSA techniques are used as a basis for analysis. The paper includes: the steps to establish the systems or components that suffer ageing; methods to identify CSSC taking into account ageing; the events associated to ageing/degradation and presentation of method to determine the ageing related events, selection of the SSCs that are important for analyses; selection of the most significant ageing events; ranking of ageing events; association of events to these components in order to decide for the CSSC detailed analyses; ranking / ordering of the ageing related events; optimization of NPP systems design and operation considering ageing; impact of ageing to NPP operation/safety/safety margins and to manufacturer technical specifications. The paper presents a brief description of the most important aspects of the methods, used to analyze the ageing effects on appearing of CSSCs, taking into account the previous developed NPP PSA models and PSA modeling tools.

Commentary by Dr. Valentin Fuster
2006;():457-465. doi:10.1115/ICONE14-89294.

In order to evaluate the water chemistry in the irradiation field during IASCC irradiation test, a water radiolysis code for IASCC irradiation loop system was developed. In the water radiolysis code, a multiple node model was introduced since the irradiation loop system has a wide rage temperature distribution as well as the dose distribution. To investigate the applicability of developed water radiolysis code, water chemistry at the water sampling point of the irradiation loop system was measured and compared with analytical results under several water chemistry conditions. Further, water chemistry distribution in the in-pile region as well as in the out-pile region was calculated by the developed water radiolysis code.

Topics: Water
Commentary by Dr. Valentin Fuster
2006;():467-476. doi:10.1115/ICONE14-89297.

As a counter measurement of intergranular stress corrosion cracking (IGSCC) in boiling water reactors, the induction heating stress improvement (IHSI) has been developed as a method to improve the stress factor, especially residual stresses in affected areas of pipe joint welds. In this method, a pipe is heated from the outside by an induction coil and cooled from the inside with water simultaneously. By thermal stresses to produce a temperature differential between the inner and outer pipe surfaces, the residual stress inside the pipe is improved compression. IHSI had been applied to weld joints of austenitic stainless steel pipes (P-8+P-8). However IHSI had not been applied to weld joints of nickel-chromium-iron alloy (P-43) and austenitic stainless steel (P-8). This weld joint (P-43+P-8) is used for instrumentation nozzles in nuclear power plants’ reactor pressure vessels. Therefore for the purpose of applying IHSI to this one, we studied the following. i) Investigation of IHSI conditions (Essential Variables); ii) Residual stresses after IHSI; iii) Mechanical properties after IHSI. This paper explains that IHSI is sufficiently effective in improvement of the residual stresses for this weld joint (P-43+P-8), and that IHSI does not cause negative effects by results of mechanical properties, and IHSI is verified concerning applying it to this kind of weld joint.

Commentary by Dr. Valentin Fuster
2006;():477-482. doi:10.1115/ICONE14-89303.

Corrosion tests of Al and SS-304-sputtering-surface treated STBA26 (9Cr.1Mo.0.1Si) and SiC/SiC composites with BN (boron nitide) coating has been conducted in high temperature LBE of 700°C at low oxygen concentration of 6.8 × 10−7 wt% and the behavior was analyzed. The sputtering technique was used to protect the steel from corrosion. The thickness of sputtering-treated layer was 21.45 μm. All specimens were immersed in LBE in a pot for 1000 hours. The STBA26 (9Cr.1Mo.0.1Si) without surface treated were also tested for comparison with sputtering-treated steels. The results showed that sputtering-treated layer still remained on the base of STBA26. No penetration of LBE was observed in this layer. The layer could protect the steel from penetration of LBE. The result also showed that thin layer which contains aluminum oxide and chromium oxide was formed on the surface-treated layer, and it protected the base area. On the contrary, the penetration in base area was observed in the as received STBA26. In SiC/SiC composites, there appeared cracks in a thin surface area and LBE penetrated deeply into the material. The corrosion did not occur in this SiC/SiC composite in the high temperature LBE.

Commentary by Dr. Valentin Fuster
2006;():483-489. doi:10.1115/ICONE14-89338.

Irradiation assisted stress corrosion cracking (IASCC) is one of the critical concerns when stainless steel components have been in service in light water reactors (LWRs) for a long period. In general, IASCC can be reproduced on the materials irradiated over a certain threshold fluence level of fast neutron by the post-irradiation examinations (PIEs). It is, however, considered that the reproduced IASCC by PIEs must be carefully compared with the actual IASCC in nuclear power plants, because the actual IASCC occurs in the core under simultaneous effects of radiation, stress and high temperature water environment. In the research field of IASCC, mainly PIEs for irradiated materials have been carried out, because there are many difficulties on SCC tests under neutron irradiation. Hence as a part of the key techniques for in-pile SCC tests, we have embarked on a development of the test technique to obtain information concerning effects of applied stress level, water chemistry, irradiation conditions, etc. A high temperature water loop facility was installed at the Japan Materials Testing Reactor (JMTR) to carry out the in-pile IASCC testing under a framework of cooperative research program between JAERI and the JAPC. In-pile IASCC growth tests have been successfully carried out using the compact tension (CT) type specimens of type 304 stainless steel that had been pre-irradiated up to a neutron fluence level around 1×1025 n/m2 before the in-pile testing since 2004. The tests were carried out in pure water simulated boiling water reactor (BWR) coolant condition. In the paper, results of the in-pile SCC growth tests will be discussed comparing with the result obtained by PIEs from a viewpoint of the synergistic effects on IASCC.

Commentary by Dr. Valentin Fuster
2006;():491-494. doi:10.1115/ICONE14-89346.

Stress corrosion cracking (SCC) is one of the major reasons to reduce the reliability of aged reactor components. Toshiba has been developing underwater laser welding onto surface of the aged components as maintenance and repair techniques. Because most of the reactor internal components to apply this underwater laser welding technique have 3-dimensional shape, effect of welding positions and welded shapes are examined and presented in this report.

Commentary by Dr. Valentin Fuster
2006;():495-511. doi:10.1115/ICONE14-89350.

In spite of industries’ effort over the last 40 years, corrosion-related issues continue to be one of the largest unresolved problems for nuclear power plants worldwide. There are several types of strange corrosion phenomena from the point of view of our current understanding of corrosion science established in other fields. Some of these are IGSCC, PWSCC, AOA, and FAC (Erosion-Corrosion). Through studying and coping with diverse corrosion phenomena, the author believes that they share a common basis with respect to the assumed corrosion mechanism (e.g., ‘local cell action’ hypothesis). In general, local cell action is rarely severe since it produces a fairly uniform corrosion. The ‘long cell action’ that transports electrons through structures far beyond the region of local cell corrosion activities has been identified as a basic mechanism in soil corrosion. If this mechanism is assumed in nuclear power plants, the structure becomes anodic in the area where the potential is less positive and cathodic where this potential is more positive. Metallic ions generated at anodic corrosion sites are transported to remote cathodic sites through the circulation of water and deposits as corrosion products. The SCC, FAC (E-C) and PWSCC occur in the anodic sites as the structure itself acts as a short-circuiting conductor between the two sites, the action is similar to a galvanic cell but in a very large scale. This situation is the same as a battery that has been short-circuited at the terminals. No apparent external potential difference exists between the two electrodes, but an electrochemical reaction is still taking place inside the battery cell with a large internal short current. In this example what is important is the potential difference between the local coolant and the surface of the structural material. Long cell action corrosion is likely enhancing the local cell action’s anodic corrosion activities, such as SCC, FAC/E-C, and PWSCC. It tends to be more hazardous because of its localized nature compared with the local cell action corrosion. There exist various mechanisms (electrochemical cell configurations) that induce such potential differences, including: ionic concentration, aeration, temperature, flow velocity, radiation and corrosion potentials. In this paper, the author will discuss these potential differences and their relevance to the un-resolved corrosion issues in nuclear power plants. Due to the importance of this potential mechanism the author is calling for further verification experiments as a joint international project.

Commentary by Dr. Valentin Fuster
2006;():513-522. doi:10.1115/ICONE14-89382.

3-D virtual analysis system to analyze the motion of Control Rod Drive Mechanism (CRDM) was developed. The analysis system consists of 3-D model established per the actual dimensions and interfaces of CRDM parts and a routine to calculate forces acting on the mechanism, and was verified by mock-up test using equipment same to actual product. The analysis system is useful for functional evaluation in maintenance or to factor out root causes in case of malfunction of CRDM.

Commentary by Dr. Valentin Fuster
2006;():523-530. doi:10.1115/ICONE14-89384.

An analytical study on micro-indentation method to integrity evaluation for graphite components was carried out. The indentation method is used as simplicity test to measure mechanical properties of materials. This method is thought to be applicable to evaluate the residual stress from the relationship between indentation load and indentation depth. In this study, in order to confirm the applicability of the micro-indentation method for lifetime evaluation of the graphite component, indentation load-depth behavior under stress/strain condition was evaluated taking account of the specified minimum ultimate strength of IG-110 graphite. Moreover, analytical investigations of indentation load-depth behavior for oxidized graphite and oxidized graphite with residual strain were also carried out. As a result, it can be said that the indentation method is potentially applicable to evaluate the integrity of graphite components.

Commentary by Dr. Valentin Fuster
2006;():531-537. doi:10.1115/ICONE14-89398.

The experimental study has been carried out to investigate reaction, transport and settling behavior of lead-bismuth eutectic (LBE) in flowing liquid sodium. In the test, 168g of LBE were poured into flowing sodium from the top of a vertical-type sodium loop which contained 23.2 kg of sodium. The initial temperature of LBE and sodium was 673K. The flow rate and the maximum velocity of sodium in the loop were controlled and measured at 20 dm3 /min and 1 m/sec, respectively, using an electro-magnetic pump and an electro-magnetic flow meter. The sodium loop has a settling chamber at the lower part to investigate the concentration decrease behavior of solid particle reaction products in the sodium due to the settling effect. The concentration was measured by sodium sampling from the 11 positions of the loop during the experiment and its post-test chemical analysis. The temperature changes at the various parts of the loop were also measured during the experiment by thermo-couples attached on the outer surface of the loop. Ultrasonic detectors were attached on the outer surface of the loop below the position of a LBE pour nozzle to demonstrate the utility as a leak detector.

Topics: Sodium
Commentary by Dr. Valentin Fuster
2006;():539-545. doi:10.1115/ICONE14-89434.

The present paper details the results of electroslag cladding and tube-to-tubesheet welding qualification tests conducted by Ansaldo-Camozzi ESC with Alloy 690 (Alloy 52 filler metal) on steel for nuclear power stations’ steam generators shell, tubesheet and head; the possibility of submerged arc cladding on first layer was also considered. Test results, in terms of chemical analysis, mechanical properties and microstructure are reproducible and confidently applicable to production cladding and show that electroslag process can be used for Alloy 52 cladding with exceptionally stable and regular operation and high productivity. The application of submerged arc cladding process to the first layer leads to a higher base metal dilution, which should be avoided. Moreover, though the heat affected zone is deeper with electroslag cladding, in both cases no coarsened grain zone is found due to recrystallisation effect of second cladding layer. Finally, the application of electroslag process to cladding of Alloy 52 with modified chemical composition, was proved to be highly beneficial as it strongly reduces hot cracking sensitivity, which is typical of submerged arc cladded Alloy 52, both during tube-to-tubesheet welding and first re-welding.

Commentary by Dr. Valentin Fuster
2006;():547-552. doi:10.1115/ICONE14-89464.

Low cycle fatigue tests were conducted to investigate the cyclic behavior and the fatigue life of type 316LN stainless steel (SS) at various strain rates in 310°C low oxygen-containing water. The strain rates were 0.008, 0.04, and 0.4%/s, and the applied strain amplitude was varied from 0.4 to 1.0%. The dissolved oxygen concentration of the test water was maintained below 1 ppb. The test material in 310°C low oxygen-containing water experienced a primary hardening, followed by a softening. From our data, we confirm the occurrence of the dynamic strain aging (DSA), and finally it can be considered that the primary hardening was brought about by the DSA. The secondary hardening was observed distinctly for 0.4%/s and 0.4%. The improvement of fatigue resistance and the secondary hardening occurred under the same loading condition. Therefore, the improvement of fatigue resistance may be related to the occurrence of the secondary hardening. When the secondary hardening occurs, intense slip bands are replaced by the corduroy structure. The corduroy structure can induce retardation of crack initiation, and ultimately the fatigue resistance is improved. Comparative study between the fatigue life generated in the current study and some prediction models was performed to evaluate the reliability of our data.

Commentary by Dr. Valentin Fuster
2006;():553-561. doi:10.1115/ICONE14-89481.

Assessing the structural integrity of a nuclear Reactor Pressure Vessel (RPV) subjected to pressurized-thermal-shock (PTS) transients is extremely important to safety. In addition to conventional deterministic calculations to confirm RPV integrity, Electricité de France (EDF) carries out probabilistic analyses. Probabilistic analyses are interesting because some key variables, albeit conventionally taken at conservative values, can be modeled more accurately through statistical variability. One variable which significantly affects RPV structural integrity assessment is cleavage fracture initiation toughness. The reference fracture toughness method currently in use at EDF is the RCCM and ASME Code lower-bound KIC based on the indexing parameter RTNDT . However, in order to quantify the toughness scatter for probabilistic analyses, the master curve method is being analyzed at present. Furthermore, the master curve method is a direct means of evaluating fracture toughness based on KJC data. In the framework of the master curve investigation undertaken by EDF, this article deals with the following two statistical items: building a master curve from an extract of a fracture toughness dataset (from the European project “Unified Reference Fracture Toughness Design curves for RPV Steels”) and controlling statistical uncertainty for both mono-temperature and multi-temperature tests. Concerning the first point, master curve temperature dependence is empirical in nature. To determine the “original” master curve, Wallin postulated that a unified description of fracture toughness temperature dependence for ferritic steels is possible, and used a large number of data corresponding to nuclear-grade pressure vessel steels and welds. Our working hypothesis is that some ferritic steels may behave in slightly different ways. Therefore we focused exclusively on the basic french reactor vessel metal of types A508 Class 3 and A 533 grade B Class 1, taking the sampling level and direction into account as well as the test specimen type. As for the second point, the emphasis is placed on the uncertainties in applying the master curve approach. For a toughness dataset based on different specimens of a single product, application of the master curve methodology requires the statistical estimation of one parameter: the reference temperature T0 . Because of the limited number of specimens, estimation of this temperature is uncertain. The ASTM standard provides a rough evaluation of this statistical uncertainty through an approximate confidence interval. In this paper, a thorough study is carried out to build more meaningful confidence intervals (for both mono-temperature and multi-temperature tests). These results ensure better control over uncertainty, and allow rigorous analysis of the impact of its influencing factors: the number of specimens and the temperatures at which they have been tested.

Commentary by Dr. Valentin Fuster
2006;():563-567. doi:10.1115/ICONE14-89486.

A key problem in the development of heavy liquid metal cooled reactors is a corrosion of the structural and fuel cladding materials by the coolants. Thus, the problem has been considered as an important design-factor that limits the operational temperature and flow velocity of the next generation nuclear reactors using lead-alloys. Corrosion data has been obtained on as-received and active coating materials of HT9 and 316L in a stagnant lead-alloy containing a reduced atmosphere of oxygen with an exposure time of 1500 hours at 600°C. After each test, the specimens were analyzed metallurgically by using a scanning electron microscopy (SEM) with a energy dispersive X-ray analysis (EDX) for the cross sections of the specimens. In addition, X-ray diffraction (XRD) was performed to evaluate the phase composition of the steels.

Commentary by Dr. Valentin Fuster
2006;():569-578. doi:10.1115/ICONE14-89499.

This study couples structural analyses of four typical U.S. nuclear power plant containments with existing probabilistic risk analysis (PRA) models to assess the effects of degradation. This method is used to determine the increase in the early release frequencies (risk) due to postulated cases of corrosion in the steel liners and shells, as well as other forms of degradation.

Commentary by Dr. Valentin Fuster
2006;():579-585. doi:10.1115/ICONE14-89520.

Environmental qualification testing was performed on a modified Limitorque torque switch for the torque switch safety functions in the Limitorque type SMB actuators located inside and outside containment in a nuclear power plant. The torque switch specimen was installed in a Limitorque SMB-1 electric actuator mounted on an 8” Velan gate valve and operated with a customized programmable logic controller to allow normal torque switch behaviour to be observed. The present paper describes the qualification testing performed. The modified torque switch was aged to a 30-year service life at the normal service conditions for both inside and outside containment. Aging included radiation, thermal and cycle aging. A seismic test and then a combined Loss of Coolant Accident (LOCA) and Main Steam Line Break (MSLB) steam accident simulation were followed. After each stage of aging, functional tests were done to confirm normal insulation resistance, normal contact resistance and normal operation.

Topics: Torque , Actuators , Switches
Commentary by Dr. Valentin Fuster
2006;():587-592. doi:10.1115/ICONE14-89578.

Pressure vessel integrity assessment after long-term service irradiation is commonly based on surveillance program results. Nevertheless, only the investigation of RPV material from decommissioned NPPs enables the evaluation of the real toughness response. Such a chance is given now through the investigation of material from the former Greifswald NPP (VVER-440/230) to evaluate the material state of a standard RPV design and to assess the quality of prediction rules and assessment tools. The operation of the four Greifswald units was finished in 1991 after 12–15 years of operation. In autumn 2005 the first trepans (diameter 120 mm) were gained from the unit 1 of this NPP. Some details of the trepanning procedure will be given. The paper mainly deals with the retrospective dosimetry based on Niobium, which is a trace element of the RPV material. The reaction 93 Nb(n,n′)93m Nb with an energy dependence highly correlated to radiation damage and a half life of the reaction product of 16.13 years is well suited for retrospective fast neutron dosimetry. Fluence calculations using the code TRAMO were based on pin-wise time dependent neutron sources and an updated nuclear data base (ENDF/B-VI release 8). The neutron spectra were determined at the trepan positions. The different loading schemes of unit 1 (standard and with 4 or 6 dummy assemblies) were taken into account. The calculated specific 93m Nb activities for February, 2006 at the sample positions were determined to 16.3 Bq/μg Nb for sample 1, (0.1cm distance from inner wall), and 4.0 Bq/μg Nb for sample 2 (11.5 cm distance from inner wall). Unfortunately, a second neutron reaction besides 93 Nb(n,n′) leading to 93m Nb-activity is the reaction 92 Mo(n,γ)93 Mo. 93 Mo decays by electron capture to 93m Nb with a half life of 4000 years and a branching ratio br = 0.88. As (n,γ)-reactions are produced mainly by low energy neutrons, being less important for material damage, the 93m Nb-activity generated through the Mo-path should be determined separately and subtracted from the measured activity. For the sample 1 in the maximum flux azimuthal position of weld SN4 with a Nb-content of 8 ppm and an Mo-content of 4000 ppm for February 3, 2006 was obtained a Mo-induced 93m Nb-activity of 80 Bq/g steel, amounting to 37.7% of the total 93m Nb-activity. It turns out that the 93m Nb generation on the second path is nearly of the same order as the fast neutron induced generation from Niobium. For the experimental determination of the 93m Nb-activity the Nb-content was determined by ICP-MS (inductive coupled plasma mass spectrometry) after dissolution of the material sample. The radiochemical isolation of Nb was done by anion exchange separation. The radiochemical separation was accompanied by determination of the chemical yield of Nb using again the ICP-MS method. The measurement of the 93m Nb activity was realized by Liquid Scintillation Spectrometry (LSC). The first comparison between the calculated and the measured 93m Nb activities resulted in deviations between 15 and 50%. Possible reasons for the observed differences are discussed.

Commentary by Dr. Valentin Fuster
2006;():593-599. doi:10.1115/ICONE14-89583.

New plants are on the horizon! Plans for no less than thirteen (13) Combined Operating Licenses (COL’s) have been announced in 2005 with corresponding submittals to begin by 2007. Several utilities have announced intentions of breaking ground in 2010, meaning that heavy component orders will have to be placed as early as 2007! Industry analysts’ project that 100 new plants will be required in the next 25 years. With an average lead time of 10 years, the bulk of the new plant orders will have to be placed by not later than 2015. This equates to almost 8–9 new plant orders a year!!

Commentary by Dr. Valentin Fuster
2006;():601-606. doi:10.1115/ICONE14-89600.

KAERI (Korea Atomic Energy Research Institute) is developing an accelerator-driven transmutation system called HYPER (HYbrid Power Extraction Reactor). HYPER is the 1000MWth system designed to transmute the long-lived TRU (Transuranic Elements) and FP (Fission Product) included in the PWR spent fuel. LBE (Lead-Bismuth Eutectic) is used as the spallation target and coolant material in HYPER. KAERI has also investigated the conceptual design of a lead-cooled fast reactor. Lead (Pb) is used as the coolant material in that reactor. The most significant problem is a corrosion when using the lead-alloy liquid metal in those reactors. Therefore, it is necessary to study the corrosion characteristics and develop the technology to protect the steel structure materials against a corrosion. KAERI has been developing the facilities needed to study the corrosion of lead-alloy. KAERI fabricated a static corrosion test facility in 2003. The static corrosion tests of HT-9, 316L and T91 have been performed at 600 °C and 650 °C since 2003. The Pb-Bi loop was constructed in 2006. The Pb-Bi loop is an isothermal loop which can be operated at temperatures up to 550 °C. The Pb loop is designed to be operated with ΔT = 150 °C (Tmin = 450 °C and Tmax = 600 °C). The first stage of the Pb loop construction was finished and operations began in 2006. We will complete the second stage of the Pb loop construction after testing the first stage Pb loop.

Commentary by Dr. Valentin Fuster
2006;():607-614. doi:10.1115/ICONE14-89623.

The complete understanding of the incubation and growth of microstructurally short cracks is still somewhat beyond the present state-of-the-art explanations. A good example is the intergranular stress corrosion cracking of Inconel 600 in high-temperature water. An effort was therefore made by the authors to construct a computational model of the crack growth kinetics at the grain-size scale. The main idea is to divide continuum (e.g., polycrystalline aggregate) into a set of sub-continua (grains). Random grain structure is modelled using Voronoi-Dirichlet tessellation. Each grain is assumed to be a monocrystal with random orientation of the crystal lattice. Elastic behaviour of grains is assumed to be anisotropic. Crystal plasticity is used to describe (small to moderate) plastic deformation of monocrystal grains. Explicit geometrical modelling of grain boundaries and triple points allows for the development of the incompatible strains along the grain boundaries and at triple points. Finite element method (ABAQUS) is used to obtain numerical solutions of strain and stress fields. The analysis is currently limited to two-dimensional models. Numerical examples illustrate analysis of about one grain boundary long transgranular cracks. In particular, the dependence of crack tip displacements on the random orientation of neighbouring grains is studied. The limited number of calculations performed indicates that the incompatibility strains, which develop along the boundaries of randomly oriented grains, significantly influence the local stress fields and therefore also the crack tip displacements. First attempts are also made to quantify the preferential growth directions of cracks crossing the discontinuities (e.g., grain boundary).

Commentary by Dr. Valentin Fuster
2006;():615-624. doi:10.1115/ICONE14-89631.

Concrete’s properties are more complex than those of most materials because not only is concrete a composite material whose constituents have different properties, but its properties depend upon moisture and porosity. Exposure of concrete to elevated temperature affects its mechanical and physical properties. Elements could distort and displace, and, under certain conditions, the concrete surfaces could spall due to the buildup of steam pressure. Because thermally-induced dimensional changes, loss of structural integrity, and release of moisture and gases resulting from the migration of free water could adversely affect plant operations and safety, a complete understanding of the behavior of concrete under long-term elevated-temperature exposure as well as both during and after a thermal excursion resulting from a postulated design-basis accident condition is essential for reliable design evaluations and assessments of nuclear power plant structures. As the properties of concrete change with respect to time and the environment to which it is exposed, an assessment of the effects of concrete aging is also important in performing safety evaluations. The effects of elevated temperature on Portland cement concretes and constituent materials are summarized, design codes and standards identified, and considerations for elevated temperature service noted.

Commentary by Dr. Valentin Fuster
2006;():625-630. doi:10.1115/ICONE14-89634.

All over the world numerous fatalities and large property losses have taken place because of Industrial Explosions. In order to mitigate the losses of life and property due to overpressure rupture discs are used. Successful implementation would depend on many factors and one of the important factors is the selection and design of rupture discs. There are several factors that affect the disc performance. The present paper discusses the factors that effect the disc performance so that the right type of disc can be selected.

Topics: Design , Disks , Rupture
Commentary by Dr. Valentin Fuster
2006;():631-640. doi:10.1115/ICONE14-89658.

The author has investigated the characteristics of boron co-deposition with crud experienced in AOA and iron ferrite deposition in CDA. Corrosion product deposits found in cores with appreciable AOA have been reported in mostly nickel-based (as NiO or elemental nickel) as opposed to nickel ferrite deposits common to non-boiling cores. Significant quantities of meta-ZrO2 and nickel iron oxyborates (bonaccordite), notably Ni2 FeBO5 have also been found in deposits on cores with AOA. On the basis of this general characterization information, the author has constructed a potential-pH diagram of Ni2 FeB(OH)10 , which is a hydrated state of FeNi2 (BO3 )O2 as summarized in this paper. Although preliminary, the estimated E-pH diagram suggests some interesting observation, including: growth of bonaccordite “needles” on the clad is associated with a local anodic electrochemical reaction necessary to remove excess electrons from the system to a cathode. During the AOA cycle, the concentration of nickel and iron ions must have been unusually high as they should be for a significant amount of crud deposits. The author thinks such an acceleration of the anodic dissolution of metal cations is due to the effect of the long cell action corrosion mechanism. As early as 1949, an Italian scientist Petracchi demonstrated that electrochemical effects significantly influence the erosion rate. He constructed a flow nozzle with specimens kept under external electrical potential. Upon inducing as low as 0.1 mA/cm2 of the positive current, the erosion rates were reported drastically increased. No erosion was observed by reversing the polarity of the potential. As discussed in a companion paper also presented at this conference [1], the author discusses various mechanisms (electrochemical cell configurations) that induce potential differences, including those differences in ionic concentration, aeration, temperature, flow velocity, radiation and corrosion potentials. In this paper, the author discusses how these potential differences are related to the AOA/CDA issues in PWR/VVER plants. The author is calling for further verification experiments regarding this corrosion mechanism as a joint international project.

Commentary by Dr. Valentin Fuster
2006;():641-644. doi:10.1115/ICONE14-89659.

The Advanced Test Reactor (ATR) located at the Department of Energy’s Idaho National Laboratory, is the most powerful test reactor operating in the United States rated at a design power of 250 MW(t). Operating cycles are nominally seven per year with outages that last 7 to 14 days, allowing time for routine plant maintenance and experiment insertions and manipulations. While the ATR pressurized water loops can operate at the same temperature and pressure requirements of a pressurized water reactor, the loops also have the ability to operate at higher conditions. Hence, it is critical to ensure that when component replacements are called for, they can meet or exceed design requirements of a typical power reactor, while continuing to satisfy the design requirements of the ATR experiment loops.

Commentary by Dr. Valentin Fuster
2006;():645-652. doi:10.1115/ICONE14-89685.

Irradiation of materials by energetic particles causes significant degradation of the mechanical properties, most notably an increased yield stress and decrease ductility, thus limiting lifetime of materials used in nuclear reactors. The microstructure of irradiated materials evolves over a wide range of length and time scales, making radiation damage and inherently multi-scale phenomenon. At atomic length scale, the principal sources of radiation damage are the primary knock-on atoms that recoil under collision from energetic particles such as neutrons or ions. These knock-on atoms in turn produce vacancies and self-interstitial atoms, and stacking fault tetrahedra. At higher length scale, these defect clusters form loops around existing dislocations, leading to their decoration and immobilization, which ultimately leads to radiation hardening in most of the materials. All these defects finally effect the macroscopic mechanical and other properties. An attempt is made to understand these phenomena using molecular dynamics studies and discrete dislocation dynamics modelling.

Topics: Modeling
Commentary by Dr. Valentin Fuster
2006;():653-661. doi:10.1115/ICONE14-89690.

The combination of one or several processes of ruins can involve the materials failure of a nuclear power plant. These processes arise from the external agents action such as the pressure, the mechanical efforts, the heat flows and the radiations constitute the whole of the “actions” of the surrounding medium. The prolongation and the repetition of these effects can involve a deterioration of the machine. In accordance with the decree of February 26, 1974, the PWR operator must be firstly, sure that the system is controlled according to the situations considered in the file of dimensioning and secondly, be able to know anytime the life of the equipment. The physical phenomena which cause the structures ruin are less complex in the PWR than in the SFR. In the SFR, the high temperatures imposed on components for long periods can involve a significant creep. In the course of time, this deformations accelerate the release of fatigue cracks. To consider the creep, the reactor lifespan is correlated at the numbers of thermals transients envisaged initially. To realize the management of aging in Phenix power plant, it is necessary to carry out an individualized monitoring of the structures and not only on the vessel. We must ensure the good state and/or the correct operation of the significant stations for safety which are the control of the reactivity, the movement of control rods, the primary sodium containment and the decay heat removal. For that, we monitor the main vessel, the conical skirt, the IHX and the Core Cover Plug. A profound knowledge of the thermal transients of the past is necessary to carry out an effective assessment. In order to guarantee that any harmful situation is well taken into the management of aging, we monitor permanently certain measurements (primary and secondary pump speed, hot and cold pool temperatures, IHX-main vessel and reactor roof temperatures). We present in the article the scientific method used in the Physics Section. A logical diagram specific to the type of situation and the structure allows to associate the harmful transient at a identical situation which has been happened in the past. During the last two cycles, the nuclear power plant has sustained 34 startup (20 during the 51st cycle and 14 during the 52nd cycle). After two cycles of operation, there is approximately 70 to 80% of occurrences authorized for the whole of the structures. For the last 4 cycles, the number of transients to come will remain quite lower than the number dimensioned initially.

Commentary by Dr. Valentin Fuster
2006;():663-672. doi:10.1115/ICONE14-89720.

SA312 type 304LN stainless steel material, having closer control over impurities and inclusion content, is the intended piping material in the Advanced Heavy Water Reactors. Deformation, fatigue and fracture behaviour of this material and its weldments have been characterized at ambient temperature and at 558K. The details of the fractographic investigations and stretch zone width measurements are also discussed. The base metals shows high initiation toughness (>500 kJ/m2 ) and large tearing modulus at ambient and operating temperatures. Shielded Metal Arc Welding (SMAW) weld metal shows much much reduced initiation toughness and tearing resistance in comparison to base metal and Gas Tungsten Arc Welding (GTAW) welds. This is attributed to larger density of second phase inclusions in the SMAW weld metal. SZW measurements give a good alternate estimate of the toughness of the materials. Fatigue crack growth rate in SMAW weld metal was found to be comparable to base metal at higher load ratios.

Commentary by Dr. Valentin Fuster
2006;():673-678. doi:10.1115/ICONE14-89760.

For the development of the System Based Code, which was proposed by Asada and intends to optimize structural design of nuclear components by enabling margin exchange between various technical options, a tool for life cycle structural reliability evaluation method is necessary. For this purpose, the authors are developing a material strength and structural reliability evaluation system MSS-REAL. The system is primarily for fast breeder reactors but its methodologies can also be applied to the other types of reactors. This paper summarizes the features of the MSS-REAL system with examples and also describes a future development plan.

Topics: Reliability
Commentary by Dr. Valentin Fuster
2006;():679-689. doi:10.1115/ICONE14-89790.

A matrix containing inert gas bubbles dilates in direct proportion to the growth experienced by the gas bubbles. This phenomenon is termed as swelling. A model for the swelling induced by the growth of the helium gas bubbles in irradiated copper-boron alloys is presented. The bubbles grow by acquiring vacancies from the external surface, which acts as a source of vacancies. The vacancies reach the surface of the bubbles mainly via lattice diffusion and to a limited extent via diffusion through short-circuiting paths such as grain boundaries and dislocation pipes. The model predicts that overall swelling of the matrix varies as 1.5th power of time. Another consequence of the present model is that the growth rate of a gas bubble varies inversely as the cube of its distance from the external surface. The model has been applied to the data on irradiated copper-boron alloys and found to be in accord with the experimental results. The model is general and can be applied to the growth of all kinds of stationary inert gas bubbles trapped within a crystalline matrix.

Topics: Copper , Alloys , Bubbles
Commentary by Dr. Valentin Fuster
2006;():691-699. doi:10.1115/ICONE14-89827.

This paper provides information on swing check valve selection criteria suitable for nuclear power plant applications. In this project, four swing check valves were analyzed to demonstrate the implementation and application of this information. In this example, swing check valves were selected according to “ASME Boiler and Pressure Vessel Code, Section III” [1] and “ASME B16.34, Valves Flanged, Threaded, and Welding End” [2]. This paper also discusses the utilization of Computational Fluid Dynamics Software (CFD) as a means to analyze valve design. The use of CFD is a relatively new approach for validation of valve design that is becoming invaluable due to the high cost of physical bench testing. The Instrument Society of America (ISA) Analysis Division and the American Society of Mechanical Engineers (ASME) Computational Fluid Dynamics Technical Committee have taken a proactive approach in setting standards and practices for the use of CFD in design and validation.

Topics: Design , Valves
Commentary by Dr. Valentin Fuster
2006;():701-708. doi:10.1115/ICONE14-89859.

The first CANDU (CAN adian D euterium U ranium) pressurized heavy water reactor (PHWR) went into operation in July 1971. Today, there are several units in operation at the Pickering, Bruce, and Darlington sites in Ontario, Canada. The steam generator tubing materials were manufactured from Monel 400, Inconel 600, and Incoloy 800 for the Pickering, Bruce, and Darlington respectively and are subjected to different operating conditions. This paper presents a review of some of the various types of degradation mechanisms that have been observed on these tubing materials over the operating period of the respective plants. The results presented are based on the metallurgical examination of removed tubes. The mechanisms that have been observed include pitting, stress corrosion cracking, intergranular attack, fretting, and erosion corrosion. The nature of the flaws and causative factors (if known) are discussed.

Topics: Tubing , Boilers , Mechanisms
Commentary by Dr. Valentin Fuster

Codes, Standards, Licensing and Regulatory Issues

2006;():709-719. doi:10.1115/ICONE14-89254.

The ASME Board on Nuclear Codes & Standards (BNCS) has formed a Task Group on Regulatory Endorsement (TG-RE) that is currently in discussions with the United States Nuclear Regulatory Commission (NRC) to look at suggestions and recommendations that can be used to help with the endorsement of new and revised ASME Nuclear Codes & Standards (NC&S). With the coming of new reactors in the USA in the very near future we need to look at both the regulations and all the ASME NC&S to determine where we need to make changes to support these new plants. At the same time it is important that we maintain our operating plants while addressing ageing management needs of our existing reactors. This is going to take new thinking, time, resources, and money. For all this to take place the regulations and requirements that we use must be clear concise and necessary for safety and to that end both the NRC and ASME are working together to make this happen. Because of the influence that the USA has in the world in dealing with these issues, this paper is written to inform the international nuclear engineering community about the issues and what actions are being addressed under this effort.

Commentary by Dr. Valentin Fuster
2006;():721-725. doi:10.1115/ICONE14-89393.

The late Professor Emeritus Yasuhide Asada proposed the System Based Code concept, which intends the optimization of design of nuclear plants through margin exchange among a variety of technical options which are not allowed by current codes and standards. The key technology of the System Based Code is margin exchange evaluation methodology. This paper describes recent progress with regards to margin exchange methodologies in Japan.

Commentary by Dr. Valentin Fuster
2006;():727-734. doi:10.1115/ICONE14-89397.

For a leap of progress in structural deign of nuclear plant components, The late Professor Emeritus Yasuhide Asada proposed the System Based Code. The key concepts of the System Based Code are; (1) life-cycle margin optimization, (2) expansion of technical options as well as combinations of technical options beyond the current codes and standards, and (3) designing to clearly defined target reliabilities. Those concepts are very new to most of the nuclear power plant designers who are naturally obliged to design to current codes and standards; the application of the concepts of the System Based Code to design will lead to entire change of practices that designers have long been accustomed to. On the other hand, experienced designers are supposed to have expertise that can support and accelerate the development of the System Based Code. Therefore, interfacing with experienced designers is of crucial importance for the development of the System Based Code. The authors conducted a survey on the acceptability of the System Based Code concept. The results were analyzed from the possibility of improving structural design both in terms of reliability and cost effectiveness by the introduction of the System Based Code concept. It was concluded that the System Based Code is beneficial for those purposes. Also described is the expertise elicited from the results of the survey that can be reflected to the development of the System Based Code.

Topics: Cycles
Commentary by Dr. Valentin Fuster
2006;():735-744. doi:10.1115/ICONE14-89470.

The Pressurized Water Reactor Owners Group (formerly the Westinghouse Owners Group (WOG)) methodology for extending the inservice inspection interval for welds in pressurized water reactor (PWR) reactor pressure vessel (RPV) was introduced as ICONE12-49429. The paper presented a risk informed basis for extending the interval between inspections from the current interval of 10 years to 20 years. In the paper presented at ICONE-12, results of pilot studies on typical Westinghouse and Combustion Engineering Nuclear Steam Supply System (NSSS) designs of PWR vessels showed that the change in risk associated with the proposed inspection interval extension was within the guidelines specified in the United States Nuclear Regulatory Commission (NRC) Regulatory Guide 1.174 for an acceptably small change in risk. Since the methodology was originally presented, the evaluation has been updated to incorporate the latest changes in the NRC Pressurized Thermal Shock (PTS) Risk Reevaluation Program and expanded to include the Babcock and Wilcox NSSS RPV design. The results of these evaluations demonstrate that the proposed RPV inspection interval extension remains a viable option for the industry. The updates to the methodology and input, pilot plant evaluations, results, process for demonstrating applicability of the pilot plant analysis to non-pilot lead plants and lessons learned from the evaluations performed are summarized in this paper.

Commentary by Dr. Valentin Fuster
2006;():745-749. doi:10.1115/ICONE14-89616.

On December 2005, the French regulator issued a new regulation for French nuclear power plants, in particular for pressure equipment (PE). This regulation need first to agree with non-nuclear PE regulation and add to that some specific requirements, in particular radiation protection requirements. Different advantages are in these proposal, it’s more qualitative risk oriented and it’s an important link with non-nuclear industry. Only few components are nuclear specific. But, the general philosophy of the existing Codes (RCC-M [15], KTA [16] or ASME [17]) have to be improved. For foreign Codes, it’s plan to define the differences in the user specifications. In parallel to that, a new safety classification has been developed by French utility. The consequences is the need to cross all these specifications to define a minimum quality level for each components or systems. In the same time a new concept has been developed to replace the well known “Leak Before Break methodology”: the “Break Exclusion” methodology. This paper will summarize the key aspects of these different topics.

Commentary by Dr. Valentin Fuster
2006;():751-756. doi:10.1115/ICONE14-89648.

For the last four years as reported in ICONE 13 Paper 13-50638, the ASME Board of Nuclear Codes and Standards (BNCS) has been leading an effort to identify code changes necessary to support the future nuclear plants of the world. In that paper the authors identified the results of meetings with NSSS suppliers, government regulators, engineers/constructors, and owner operators to ascertain the status of their future designs and what modifications are necessary so the right rules and materials are in ASME Nuclear Codes and Standards.

Commentary by Dr. Valentin Fuster
2006;():757-763. doi:10.1115/ICONE14-89668.

Since its foundation in 1997, the Main Committee on Power Generation Facility Codes, MC-PGFC, of the Japan Society of Mechanical Engineers, JSME, has issued a number of nuclear codes including the rules on design and construction and the rules on fitness-for-service for nuclear power plants. Some of these JSME nuclear codes have been endorsed by the regulatory body, and are now utilized in the regulatory processes of the actual plants. Among these nuclear codes recently published is the “rules on design and construction for fast reactors”. It includes as its main body design rules on class 1 components for elevated temperature services. This paper overview the main features of the code.

Commentary by Dr. Valentin Fuster
2006;():765-773. doi:10.1115/ICONE14-89707.

Comprehensive reformation of the regulatory system has been introduced in Japan in order to apply recent technical progress in a timely manner. “The Technical Standards for Nuclear Power Generation Equipments”, known as the Ordinance No.622) of the Ministry of International Trade and Industry, which is used for detailed design, construction and operating stage of Nuclear Power Plants, was being modified to performance specifications with the consensus codes and standards being used as prescriptive specifications, in order to facilitate prompt review of the Ordinance with response to technological innovation. The activities on modification were performed by the Nuclear and Industrial Safety Agency (NISA), the regulatory body in Japan, with support of the Japan Nuclear Energy Safety Organization (JNES), a technical support organization. The revised Ordinance No.62 was issued on July 1, 2005 and is enforced from January 1 2006. During the period from the issuance to the enforcement, JNES carried out to prepare enforceable regulatory guide which complies with each provisions of the Ordinance No.62, and also made technical assessment to endorse the applicability of consensus codes and standards, in response to NISA’s request. Some consensus codes and standards were re-assessed since they were already used in regulatory review of the construction plan submitted by licensee. Other consensus codes and standards were newly assessed for endorsement. In case that proper consensus code or standards were not prepared, details of regulatory requirements were described in the regulatory guide as immediate measures. At the same time, appropriate standards developing bodies were requested to prepare those consensus code or standards. Supplementary note which provides background information on the modification, applicable examples etc. was prepared for convenience to the users of the Ordinance No. 62. This paper shows the activities on modification and the results, following the NISA’s presentation at ICONE-13 that introduced the framework of the performance specifications and the modification process of the Ordinance NO. 62.

Commentary by Dr. Valentin Fuster
2006;():775-781. doi:10.1115/ICONE14-89722.

Although the government admit the benefit of construction of a nuclear facility for national electric source, related policy could be developed and carried out only if the public, especially who have some stake on it, recognize the benefit and accept the policy. For public participation, Korea has a system of public-hearing in accordance with the law. Because of the absence of the detailed way for public opinion aggregation and for the reflection of the aggregated opinion, Korean public-hearing system is only a conceptual model. Therefore, some specific system for Korean Public-Hearing should be developed and applied. In this study, to share the right of decision making, which is an ultimate concept for public participation, decision making components and the characteristics of each phase are analyzed. The criteria weight for assessment and comparison with alternatives are founded as a valuation factor of the decision making components, which should be based on the social consensus. On these foundations, a system for aggregation and reflection of the public opinion was proposed. The system named “CPDM” (Consensus based Participatory Decision Making) has three authority groups for decision making. At first, “advisory experts group” play a role for the technical assessment and the serve utility value on the criteria for each alternatives. Next, “participatory deliberation group” play a role for consensus building on the relative-importance (weight) between the criteria by feedback to promote degree of consensus. Lastly including gentlemen of the long robe, “expert group for decision making” paly a role to reflect the utility and weight and make a decision with agreement for performance of it. Also, in this study, a mathematical model for the quantification of the degree of consensus was conceptualized using Ordered Weighted Averaging (OWA) aggregation operator and fuzzy similarity theory, which is a comparison concept. Since this model enables influence of each criteria and each participant on collective consensus to be analyzed, a direction to promote consensus building can be derived. That is to say, this model can support consensus building and promote public acceptance for the nuclear industry and related policy.

Commentary by Dr. Valentin Fuster
2006;():783-790. doi:10.1115/ICONE14-89740.

Risk-informed inservice inspection (ISI) programs have been in use for over seven years as an alternative to current regulatory requirements in the development and implementation of ISI programs for nuclear plant piping systems. Programs using the Westinghouse Owners Group (WOG) (now known as the Pressurized Water Reactor Owners Group - PWROG) risk-informed ISI methodology have been developed and implemented within the U.S. and several other countries. Additionally, many plants have conducted or are in the process of conducting updates to their risk-informed ISI programs. In the development and implementation of these risk-informed ISI programs and the associated updates to those programs, the following important lessons learned have been identified and are addressed. • Concepts such as “loss of inventory,” which are typically not modeled in a plant’s probabilistic risk assessment (PRA) model for all systems. • The importance of considering operator actions in the identification of consequences associated with a piping failure and the categorization of segments as high safety significant (HSS) or low safety significant (LSS). • The impact that the above considerations have had on the large early release frequency (LERF) and categorization of segments as HSS or LSS. • The importance of automation. • Making the update process more efficient to reduce costs associated with maintaining the risk-informed ISI program. The insights gained are associated with many of the steps in the risk-informed ISI process including: development of the consequences associated with piping failures, categorization of segments, structural element selection and program updates. Many of these lessons learned have impacted the results of the risk-informed ISI programs and have impacted the updates to those programs. This paper summarizes the lessons learned and insights gained from the application of the WOG risk-informed ISI methodology in the U.S., Europe and Asia.

Commentary by Dr. Valentin Fuster
2006;():791-794. doi:10.1115/ICONE14-89778.

With the globalization of the nuclear industry, it is clear that the reactor suppliers are based in many countries around the world (such as United States, France, Japan, Canada, South Korea, South Africa) and they will be marketing their reactors to many countries around the world (such as US, China, South Korea, France, Canada, Finland, Taiwan). They will also be fabricating their components in many different countries around the world. With this situation, it is clear that the requirements of ASME Nuclear Codes and Standards need to be adjusted to accommodate the regulations, fabricating processes, and technology of various countries around the world. It is also very important for the American Society of Mechanical Engineers (ASME) to be able to assure that products meeting the applicable ASME Code requirements will provide the same level of safety and quality assurance as those products currently fabricated under the ASME accreditation process. To do this, many countries are in the process of establishing or changing their regulations, and it is important for ASME to interface with the appropriate organizations in those countries, in order to ensure there is effective use of ASME Codes and standards around the world.

Commentary by Dr. Valentin Fuster

Fuel Cycle and High Level Waste Management

2006;():795-805. doi:10.1115/ICONE14-89099.

Silicon carbide (SiC) based uranium ceramic material can be fabricated as hosts for ultra high temperature applications, such as gas-cooled fast reactor fuels and in-core materials. A pyrolysis-based material processing technique allows for the fabrication of SiC based uranium ceramic materials at a lower temperature compared to sintering route. Modeling of the process is considered important for optimizing the fabrication and producing material with high uniformity. This study presents a process model describing polymer pyrolysis and uranium ceramic material processing, including heat transfer, polymer pyrolysis, SiC crystallization, chemical reactions, and species transport of a porous uranium oxide mixed polymer. Three key reactions for polymer pyrolysis and one key reaction for uranium oxide polymer interaction are established for the processing. Included in the model formulation are the effects of transport processes such as heat-up, polymer decomposition, and volatiles escape. The model is capable of accurately predicting the polymer pyrolysis and chemical reactions of the source material. Processing of a sample with certain geometry is simulated. The effects of heating rate, particle size and volume ratio of uranium oxide and polymer on porosity evolution, species uniformity, reaction rate are investigated.

Commentary by Dr. Valentin Fuster
2006;():807-814. doi:10.1115/ICONE14-89104.

The inert matrix fuels are a promising option to reduce-eliminate worldwide plutonium stockpiles by burning it in LWRs. These fuels, where plutonium is hosted in a U-free inert matrix phase, may reach high burning efficiency while preventing new plutonium build-up under irradiation. A specific investigation on CSZ and thoria inert matrices has been developed by ENEA since several years. In-pile testing on the ENEA-conceived innovative fuels is ongoing in the OECD Halden HBWR since June 2000 (IFA-652 experiment). The registered burnup at the end of 2005 is about 38 MWd·kgUeq −1 vs. 45 MWd·kgUeq −1 (40 MWd·kgUOXeq −1 ) target. Fuel pins are equipped with fuel temperature thermocouples, internal pressure transducers and fuel stack elongation sensors, with the task of studying thermal conductivity and its degradation with burnup, densification-swelling behaviour and the FGR. In this paper, the response at low burnup (< 7 MWd·kgUeq −1 ) of CSZ-based fuels loaded in IFA-652, is analysed by means of the TRANSURANUS code. To this purpose, a comprehensive modelling of the above mentioned un-irradiated fuels, mainly relying on the thermophysical characterisation performed at the JRC/ITU-Karlsruhe, has been implemented in a custom TRANSURANUS version (TU-IMF). A comparison of the code predictions vs. the experimental data, aimed at evaluating the early-stage under irradiation phenomena, particularly densification and relocation, has been performed.

Commentary by Dr. Valentin Fuster
2006;():815-821. doi:10.1115/ICONE14-89136.

There have been a lot of tests and analyses reported for evaluation of drop tests of metal casks. However, no quantitative measurement has ever been made for any instantaneous leakage through metal gaskets during the drop tests due to loosening of the bolts in the containments and lateral sliding of the lids. In order to determine a source term for radiation exposure dose assessment, it is necessary to obtain fundamental data of instantaneous leakage. In this study, leak tests were performed by using scale models of the lid structure and a full scale cask without impact limiters simulating drop accidents in a storage facility, with aim of measuring and evaluating any instantaneous leakage at drop impact. Prior to drop tests of a full scale metal cask, a series of leakage tests using scale models were carried out to establish the measurement method and to examine a relationship between the amount of the lateral sliding of the lid and the leak rate. It was determined that the leak rate did not depend on the lateral sliding speeds. Drop tests of a full scale metal cask without impact limiters were carried out by simulating drop accidents during handling in a storage facility. The target was designed to simulate a reinforced concrete floor in the facility. The first test was a horizontal drop from a height of 1 m. The second test simulated a rotational impact around an axis of a lower trunnion of the cask from the horizontal status at a height of 1 m. In the horizontal drop test, the amount of helium gas leakage was calculated by integrating the leak rate with time. The total amount of helium gas leakage from the primary and secondary lids was 1.99×10−6 Pa · m3 . This value is 9.61×10−9 % of the initially installed helium gas. The amount of leakage was insignificant. In the rotational drop test, the total amount of leakage from the primary and secondary lids was 1.74×10−5 Pa·m3 . This value is 8.45×10−8 % of the initially installed helium gas. This value was larger than that of the horizontal drop test. Nevertheless, the amount of leakage was also insignificant. The relationship between the maximum sliding displacement of the lid and the leak rate coincided between the tests of a scale model and a full scale metal cask.

Topics: Metals , Drops , Leakage
Commentary by Dr. Valentin Fuster
2006;():823-828. doi:10.1115/ICONE14-89196.

A new fuel reprocessing system by U-Pu co-crystallization process is proposed and examined by flowsheet analysis. This reprocessing system is based on the fact that hexavalent plutonium in nitric acid solution is co-crystallized with uranyl nitrate, whereas it is not crystallized when uranyl nitrate does not exist in the solution. The system consists of five steps: dissolution of spent fuel, plutonium oxidation, U-Pu co-crystallization as a co-decontamination, re-dissolution of the crystals, and U re-crystallization as a U-Pu separation. The system requires a recycling of the mother liquor from the U-Pu co-crystallization step and the appropriate recycle ratio is determined by flowsheet analysis such that the satisfactory decontamination is achieved. Further flowsheet study using four different compositions of LWR spent fuels demonstrates that the constant ratio of plutonium to uranium in mother liquor from the re-crystallization step is achieved for every composition by controlling the temperature. It is also demonstrated by comparing to the Purex process that the size of the plant based on the proposed system is significantly reduced.

Topics: Crystallization
Commentary by Dr. Valentin Fuster
2006;():829-838. doi:10.1115/ICONE14-89212.

The target system, whose function is to supply an external neutron source to the ADS sub-critical core to sustain the neutron chain reaction, is the most critical part of an ADS being subject to severe thermo-mechanical loading and material damage due to accelerator protons and fission neutrons. A windowless option was chosen as reference configuration for the target system of the LBE-cooled ADS within the European PDS-XADS project in order to reduce the material damage and to increase its life. This document deals with the thermo-hydraulic results of the calculations performed with STAR-CD and RELAP5 codes for studying the behaviour of the windowless target system during off-normal operating conditions. It also reports a description of modifications properly implemented in the codes needed for this analysis. The windowless target system shows a satisfactory thermo-hydraulic behaviour for the analysed accidents, except for the loss of both pumps without proton beam shut-off and the beam trips lasting more than one second.

Commentary by Dr. Valentin Fuster
2006;():839-847. doi:10.1115/ICONE14-89281.

The performances of different concepts of Fast Breeder Reactor (Na-cooled, He-cooled and Pb-cooled FBR) for the current French fleet renewal are analyzed in the framework of a transition scenario to a 100% FBR fleet at the end of the XXIst century. Firstly, the modeling of these three FBR types by means of a semi-analytical approach in TIRELIRE - STRATEGIE, the EDF fuel cycle simulation code, is presented, together with some validation elements against ERANOS, the French reference code system for neutronic FBR analysis (CEA). Afterwards, performances comparisons are made in terms of maximum deployable power, natural uranium consumption and waste production. The results show that the FBR maximum deployable capacity, independently from the FBR technology, is highly sensitive to the fuel cycle options, like the spent nuclear fuel cooling time or the Minor Actinides management strategy. Thus, some of the key parameters defining the dynamic of FBR deployment are highlighted, to inform the orientation of R&D in the development and optimization of these systems.

Commentary by Dr. Valentin Fuster
2006;():849-855. doi:10.1115/ICONE14-89292.

Crystallization is one of the remarkable technologies for future fuel reprocessing process that has safety and economical advantages. Japan Atomic Energy Agency (JAEA) (former Japan Nuclear Cycle Development Institute), Mitsubishi Material Corporation and Saitama University have been developing the crystallization process. In previous study, we carried out experimental studies with uranium, MOX and spent fuel conditions, and flowsheet analysis was considered. [1, 2, 3] In association with these studies, an innovative continuous crystallizer and its system was developed to ensure high process performance. From the design study, an annular type continuous crystallizer was selected as the most promising design, and performance was confirmed by small-scale test and engineering scale demonstration at uranium crystallization conditions. In this paper, the design study and the demonstration test results are described.

Commentary by Dr. Valentin Fuster
2006;():857-863. doi:10.1115/ICONE14-89293.

Spent nuclear fuel coming from a Japanese nuclear power plant is stored in the interim storage facility before reprocessing. There are two types of the storage methods which are wet and dry type. In Japan, it is anticipated that the dry storage facility will increase compared with the wet type facility. The dry interim storage facility using the metal cask has been operated in Japan. In another dry storage technology, there is a concrete overpack. Especially in USA, a lot of concrete overpacks are used for the dry interim storage. In Japan, for the concrete cask, the codes of the Japan Society of Mechanical Engineers and the governmental technical guidelines are prepared for the realization of the interim storage as well as the code for the metal cask. But the interim storage using the concrete overpack has not been in progress because the evaluation on the stress corrosion cracking (SCC) of the canister is not sufficient. Japanese interim storage facilities would be constructed near the seashore. The metal casks and concrete overpacks are stored in the storage building in Japan. On the other hand, in USA they are stored outside. It is necessary to remove the decay heat of the spent nuclear fuel in the cask from the storage building. Generally, the heat is removed by natural cooling in the dry storage facility. Air including the sea salt particles goes into the dry storage facility (Figure 1). Concerning the concrete overpack, air goes into the cask body and cools the canister. Air goes along the canister surface and is in contact with the surface directly. In this case, the sea salt in the air attaches to the surface and then there is the concern about the occurrence of the SCC. For the concrete overpack, the canister including the spent fuel is sealed by the welding. The loss of sealability caused by the SCC has to be avoided. To evaluate the SCC for the canister, it is necessary to make clear the amount of the sea salt particles coming into the storage building and the concentration on the canister. In present, the evaluation on that point is not sufficient. In this study, the concentration of the sea salt particles in the air and on the surface of the storage facility are measured inside and outside of the building. For the measurement, two sites of the dry storage facility using the metal cask are chosen. This data is applicable for the evaluation on the SCC of the canister to realize the interim storage using the concrete overpack.

Commentary by Dr. Valentin Fuster
2006;():865-870. doi:10.1115/ICONE14-89314.

“COSI”, a software developed by the Nuclear Energy Direction at CEA, the French Atomic Energy Commission, is a code simulating a pool of nuclear electricity generating plants with its associated fuel cycle facilities. This code has been designed to study short, medium and long term options for the introduction of various types of nuclear reactors and for the usage of associated nuclear materials. It permits to study transition scenarios and gives due consideration to isotopic composition essentially of uranium, plutonium, minor actinides and some fission products.

Commentary by Dr. Valentin Fuster
2006;():871-877. doi:10.1115/ICONE14-89315.

In the frame of the French law for the researches about waste management, different dynamic scenarios have been studied [1]. These scenarios are considering the French case and start from the present situation, which consists in a single stage of Plutonium recycling in PWRs. The scenarios described in this paper take into account two main options: Continuation of nuclear energy or phase out option.

Commentary by Dr. Valentin Fuster
2006;():879-899. doi:10.1115/ICONE14-89319.

This report summarizes some of the challenges encountered and solutions implemented to ensure safe storage and handling of damaged spent nuclear fuels (SNF). It includes a brief summary of some SNF storage environments and resulting SNF degradation, experience with handling and repackaging significantly degraded SNFs, and the associated lessons learned. This work provides useful insight and resolutions to many engineering challenges facing SNF handling and storage facilities. The context of this report is taken from a report produced at Idaho National Laboratory and further detailed information, such as equipment design and usage, can be found in the appendices to that report.

Commentary by Dr. Valentin Fuster
2006;():901-905. doi:10.1115/ICONE14-89334.

The nuclear spent fuel transport and storage cask is used for transport of the spent fuel from a nuclear power station to an intermediate storage facility. Leak tightness and subcriticality on transportation required from IAEA TS-R1 [1] have to be assured by a 9m drop test and its numerical simulation. This paper describes the drop test using a full-scale prototype test cask. The test was conducted by German Federal Institute for Materials Research and Testing (BAM) at their test facility in Horstwalde, Germany and comparison of the test result with the “MH1 (Mitsubishi Heavy Industries, Ltd.)” numerical simulation using LS-DYNA code. The drop orientations of the tests were slap down and vertical. From the drop test the following is demonstrated: • The leak rate of He gas after the drop tests satisfied the IAEA’s criteria. • The numerical simulation which modeled the cask body enabled dynamic response such as acceleration and strain of the cask body. This means the simulation method qualified the relation of dynamic response of the cask body and leakage behavior.

Topics: Simulation , Drops
Commentary by Dr. Valentin Fuster
2006;():907-912. doi:10.1115/ICONE14-89374.

Japan Atomic Energy Agency (JAEA) has been leading feasibility study on commercialized fast reactor cycle systems in Japan. In this study, we have proposed a new disassembly technology by mechanical disassembly system that consists of a mechanical cutting step and a wrapper tube pulling step. In the mechanical disassembly system, high durability mechanical tool grinds the wrapper tube (Slit-cut (S/C) operation in circle direction), and then the wrapper tube is pulled out and removed from the fuel assembly. Then the fuel pins are cut (Crop-cut (C/C) operation at entrance nozzle side) and the entrance nozzle is removed. The fuel pins are transported to the shearing device in next process. The Fundamental tests were carried out with simulated FBR fuel pins and wrapper tube, and cutting performance and wrapper tube pulling performance has been confirmed by engineering scale. As results, we established an efficient disassembly procedure and the fundamental design of mechanical disassembly system.

Topics: Fuels , Design
Commentary by Dr. Valentin Fuster
2006;():913-921. doi:10.1115/ICONE14-89430.

The Tokai Vitrification Facility (TVF) is the only operating vitrification plant in Japan, constructed and operated by JAEA, to vitrify concentrated high radioactive liquid waste (HALW) in the Tokai Reprocessing Plant (TRP). JAEA started TVF hot operation in 1995 and produced 218 canisters as of March, 2006. An existing melter is the second melter, which was installed from 2002 to 2004 in place of the first melter stopped its operation by damage of a main electrode. JAEA has estimated that the damage was caused by accumulation of noble metal. Therefore, melter bottom structure was improved to get better drain ability of glass containing noble metal. Completing the melter replacement, vitrification operation was restarted in October 2004 and produced 88 canisters successfully until the end of March 2006. Through these experiences, JAEA made basic strategy to achieve stable TVF operation: keeping stable operation of the existing melter preventing adverse effect by noble metal accumulation and developing a new advanced melter with long lifetime preparing for future exchange as the third melter. Based on the basic strategy, JAEA made a decade development plan of necessary key technologies and has started the development since 2005.

Commentary by Dr. Valentin Fuster
2006;():923-928. doi:10.1115/ICONE14-89506.

Burnup limitations are normally set to limit stresses in the fuel assembly components. The defined limits provide guidance to the fuel designer to minimize fuel failure during steady sate operation, and also prevent against some thermal and mechanical phenomena that could occur during overpower transients. In particular, a LHGR limit value is set to take into account physical phenomena that could lead to pellet-cladding interaction. This limit value directly relates to a PCI limit, which may be set based on experimental ramp tests. Thus, to avoid violating the PCI limit, fuel conditioning procedures are still required for both barrier and non-barrier fuel. Simulation of the power ramp procedures to be performed by the reactor operator during startup or power increase maneuvers is advisable as a preventive measure of possible overpower consequences on the fuel thermomechanical behavior. In this paper, the thermomechanical behavior of two different kinds of BWR fuel rods is analyzed for fuel preconditioning procedures. Five different preconditioning computations were performed, each with three different ascending linear power rate ramps. The starting point of the ramps was taken from data of the Cycle 8 of the Unit 1 of the Laguna Verde Nuclear Power Plant, located in MEXICO. The top limit of the ramps was the threshold linear power at which failure by PCI could occur, as a function of burnup. The analysis was performed with the FEMAXI-V code.

Commentary by Dr. Valentin Fuster
2006;():929-934. doi:10.1115/ICONE14-89563.

Reprocessing of spent LWR fuel is an intrinsic part of the closed fuel cycle. While current technologies treat recovered minor actinides as high level wastes, the primary objective of one of the U.S. DOE Nuclear Energy Research Initiative (NERI) projects is to assess the possibility, advantages and limitations of achieving ultra-long life VHTR (Very High Temperature Reactor) configurations by utilizing minor actinides as a fuel component. The postulated principal mechanism is an enhanced involvement of self-generated fissile compositions based on spent LWR fuel. Since pebble bed and prismatic core designs permit flexibility in component configuration, fuel utilization and management, it is possible to improve fissile properties of minor actinides by neutron spectrum shifting through configuration adjustments. Depending on neutron spectra, neptunium, americium and curium may contribute to small reactivity swings (self-stabilization) over prolonged irradiation periods. The presented analysis is focused on achievability of spectral variations and their potential impact. In principle, promising core features and performance characteristics have been demonstrated.

Commentary by Dr. Valentin Fuster
2006;():935-944. doi:10.1115/ICONE14-89581.

Robust increases in energy demand, improvements in the performance of existing nuclear power plants, renewed interest in assuring domestic energy supply and concern about climate change have recently provided powerful arguments for renewing and further expanding the use of nuclear energy in the United States.

Commentary by Dr. Valentin Fuster
2006;():945-952. doi:10.1115/ICONE14-89773.

There has been a resurgence of interest in the possibility of processing the US spent nuclear fuel, instead of burying it in a geologic repository. Accordingly, key topical findings from three relevant EPRI evaluations made in the 1990–1995 timeframe are recapped and updated to accommodate a few developments over the subsequent ten years. Views recently expressed by other US entities are discussed. Processing aspects thereby addressed include effects on waste disposal and on geologic repository capacity, impacts on the economics of the nuclear fuel cycle and of the overall nuclear power scenario, alternative dispositions of the plutonium separated by the processing, impacts on the structure of the perceived weapons proliferation risk, and challenges for the immediate future and for the current half-century. Currently, there is a statutory limit of 70,000 metric tons on the amount of nuclear waste materials that can be accepted at Yucca Mountain. The Environmental Impact Statement (EIS) for the project analyzed emplacement of up to 120,000 metric tons of nuclear waste products in the repository. Additional scientific analyses suggest significantly higher capacity could be achieved with changes in the repository configuration that use only geology that has already been characterized and do not deviate from existing design parameters. Conservatively assuming the repository capacity postulated in the EIS, the need date for a second repository is essentially deferrable until that determined by a potential new nuclear plant deployment program. A further increase in technical capacity of the first repository (and further and extensive delay to the need date for a second repository) is potentially achievable by processing the spent fuel to remove the plutonium (and at least the americium too), provided the plutonium and the americium are then comprehensively burnt. The burning of some of the isotopes involved would need fast reactors (discounting for now a small possibility that one of several recently postulated alternatives will prove superior overall). However, adoption of processing would carry a substantial cost burden and reliability of the few demonstration fast reactors built to-date has been poor. Trends and developments could remove these obstacles to the processing scenario, possibly before major decisions on a second repository become necessary, which need not be until mid-century at the earliest. Pending the outcomes of these long-term trends and developments, economics and reliability encourage us to stay with non-processing for the near term at least. Besides completing the Yucca Mountain program, the two biggest and inter-related fuel-cycle needs today are for a nationwide consensus on which processing technology offers the optimum mix of economic competitiveness and proliferation resistance and for a sustained effort to negotiate greater international cooperation and safeguards. Equally likely to control the readiness schedule is development/demonstration of an acceptable, reliable and affordable fast reactor.

Commentary by Dr. Valentin Fuster
2006;():953-959. doi:10.1115/ICONE14-89792.

Verification of 235 U enrichment in uranium hexaflouride (UF6 ) cylinders is often achieved by destructive and non-destructive assay techniques. These techniques are time consuming, need suitable and similar standard, in addition to loss of the nuclear material in the case of destructive analysis. This paper introduce an innovative approach for verifying of 235 U enrichment in UF6 cylinder. The approach is based on measuring dose rate (μSv/h) resulted from the emmitted gamma rays of 235 U at the surface of the cylinder and then calculating the activity of uranium and enrichment percentage inside the cylinder by a three dimensional model. Attenuation of the main 235 U gamma transitions due to the cylinder wall (5A Type of Ni alloy) was also caculated and corrected for. The method was applied on UF6 cylinders enriched with 19.75% of 235 U. The calculated enrichment was found to be 18% with 9% uncertainty. By the suggested method, the calculated total uranium activity inside one of the investigated UF6 cylinder was found close to the target (certified) value (5.6 GBq) with 9% uncertainty. The method is being developed by taking into consideration other parameters.

Commentary by Dr. Valentin Fuster
2006;():961-969. doi:10.1115/ICONE14-89794.

The waste produced from processing spent fuel from the EBR II reactor must be processed into a waste form suitable for long term storage in Yucca Mountain. The method chosen produces zeolite granules mixed with glass frit, which must then be converted into a solid. This is accomplished by loading it into a can and heating to 900 C in a furnace regulated at 915 C. During heatup to 900 C, the zeolite and glass frit react and consolidate to produce a sodalite monolith. The resultant ceramic waste form (CWF) is then cooled. The waste form is 52 cm in diameter and initially 300 cm long but consolidates to 150 cm in length during the heating process. After cooling it is then inserted in a 5-DHLW/DOE SNF Long Canister. Without intervention, the waste takes 82 hours to heat up to 900 C in a furnace designed to geometrically fit the cylindrical waste form. This paper investigates the reduction in heating times possible with four different methods of additional heating through a center hole. The hole size is kept small to maximize the amount of CWF that is processed in a single run. A hole radius of 1.82 cm was selected which removes only 1% of the CWF. A reference computation was done with a specified inner hole surface temperature of 915 C to provide a benchmark for the amount of improvement which can be made. It showed that the heatup time can potentially be reduced to 43 hours with center hole heating. The first method, simply pouring high temperature liquid aluminum into the hole, did not produce any noticeable effect on reducing heat up times. The second method, flowing liquid aluminum through the hole, works well as long as the velocity is high enough (2.5 cm/sec) to prevent solidification of the aluminum during the initial front movement of the aluminum into the center hole. The velocity can be reduced to 1 cm/sec after the initial front has traversed the ceramic. This procedure reduces the formation time to near that of the reference case. The third method, flowing a gas through the center hole, also works well as long as the product of heat capacity and velocity of the gas is equivalent to that of the flowing aluminum, and the velocity is high enough to produce an intermediate size heat transfer coefficient. The fourth method, using an electric heater, works well and heater sizes between 500 to 1000 Watts are adequate. These later three methods all can reduce the heatup time to 44 hours allowing production to be doubled and a more uniform heating.

Commentary by Dr. Valentin Fuster
2006;():971-979. doi:10.1115/ICONE14-89813.

Vitrified Waste Storage Facilities, which store high-level radioactive vitrified waste packages with natural air-cooling system, have been constructed at JNFL’s Rokkasho Reprocessing Site. [1] Construction work for the first vitrified waste storage facility which stores the waste packages vitrified in Rokkasho site was completed in December in 2005. After the final inspection, a forced draft test was conducted to verify actual fluid resistance throughout the flow passage. The test results show that the measured pressure loss is less than 70Pa at design flow rate and that it has approximately 10% margin to the design value. This confirms that the facility has the adequate cooling performance.

Commentary by Dr. Valentin Fuster
2006;():981-990. doi:10.1115/ICONE14-89857.

The purpose of this paper is to present an initial analysis of the maximum amount of commercial spent nuclear fuel (CSNF) that could be emplaced into a geological repository at Yucca Mountain. This analysis identifies and uses programmatic, material, and geological constraints and factors that affect this estimation of maximum amount of CSNF for disposal. The conclusion of this initial analysis is that the current legislative limit on Yucca Mountain disposal capacity, 63,000 MTHM of CSNF, is a small fraction of the available physical capacity of the Yucca Mountain system assuming the current high-temperature operating mode (HTOM) design. EPRI is confident that at least four times the legislative limit for CSNF (∼260,000 MTHM) can be emplaced in the Yucca Mountain system. It is possible that with additional site characterization, upwards of nine times the legislative limit (∼570,000 MTHM) could be emplaced.

Commentary by Dr. Valentin Fuster

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