bluuebunny/crossref_metadata_2025 huggingface dataset:
On-Line Workload Assessment by the PET Method
Authors: Petkov Gueorgui , Georgiev Vladislav , Stefanov Emil
Published: February 2010

The paper presents benchmarking workload results obtained by a method for operator workload assessment - NASA TLX and a method for human reliability assessment - Performance Evaluation of Teamwork - PET. Based on the archives of nuclear power plant full-scope simulator training in the accident “Main Steam Line Tube Rupture at the WWER-1000 Containment” the capacities of the two methods for indirect and direct or even on-line workload assessment are presented and discussed.
Design by Analysis of Waste Packages at Yucca Mountain for Impact Loads
Authors: James Randy J. , Jaquay Kenneth , Anderson Michael J.
Published: February 2010

The proposed geologic repository under development at Yucca Mountain, Nevada, will employ multiple shell metallic containers (waste packages) for the disposal of nuclear waste. The waste packages represent a primary engineered barrier for protection and containment of the radioactive waste, and the design of these containers must consider a variety of structural conditions to insure structural integrity. Some of the more challenging conditions for structural integrity involve severe impact loading due to hypothesized event sequences, such as drops or collisions during transport and placement. Due to interactions between the various components leading to complex structural response during an impact sequence, nonlinear explicit dynamic simulations and highly refined models are employed to qualify the design for these severe impact loads. This paper summarizes the Design by Analysis methodologies employed for qualification of waste package design under impact loading and provides several illustrative examples using these methods. Example evaluations include a collision of a waste package by the Transport and Emplacement Vehicle (TEV) and two scenarios due to seismic events, including WP impact within the TEV and impact by falling rock. The examples are intended to illustrate the stringent Design by Analysis methods employed and also highlight the scope of structural conditions included in the design basis for waste packages to be used for proposed nuclear waste storage at Yucca Mountain.
Sources of Uncertainty in a Fire Probabilistic Safety Assessment
Authors: Meyer Stephen P.
Published: February 2010

Probability Safety Assessments (PSA) by their nature are approximations of the actual risk and consequences of an accident at a nuclear power plant. Today the PSA models are becoming more complex as the computers and PSA software are faster and PSA modeling techniques are improved. Many PSAs now integrate into a single model internal initiators, internal floods, and internal fires. Each of these initiator types has some uncertainties in common and some that are primarily associated with a specific initiator type. This paper discusses some of those uncertainties found in Fire PSAs. Uncertainties in Fire PSAs arise from the broad categories of phenomenological modeling and assumptions, PSA model development and assumptions, data, and human failure events (HFEs). Phenomenological development and assumptions would include the heat release rates, fire durations, zones of influence, damage criteria, fire propagation, and the impact of smoke on equipment. Aside from the model development and assumptions in the traditional internal events PSA, the inclusion or exclusion of mitigating systems into a Fire PSA model generally has greater impact on the results than in the internal events PSA model. Most plants do not have readily available cable information for non-safety related cables. The cost to determine cable routing for non-safety related cable can be very high and, therefore, most Fire PSAs do not include non-safety mitigating systems except where needed. This increases the level of uncertainty and can skew the results. Data uncertainty arises from ignition frequency determinations, fuel available to a fire, the probability of success/failure of fire suppression systems, and the probability of hot shorts and consequential spurious operation. In addition to the HFEs included in an internal events PSA, there are other HFEs specific to fires. These include new human error probabilities (HEPs) for those HFEs that are part of the internal events PSA model due to the fact that there is a fire occurring and added stress and instrumentation failure may result. There are also HFEs associated directly with the fire such as fire detection and suppression. New HFEs will be needed to be modeled for Control Room evacuation. PSAs are being used more and more in the decision making processes of operating nuclear power plants. It is often required that initiators other than the traditional “internal events” be included in these processes. Understanding the uncertainties that are part of the Fire PSA is needed to make an informed decision. This paper addresses each of these in greater detail and provides techniques in understanding the impacts of the uncertainties.
RELAP Model for Atucha II PHWR
Authors: Schivo M. , Bedrossian G. , Gonza´lez J. , Zanocco P. , Gime´nez M. , Caputo M. , Serrano P. , Mazzantini O.
Published: February 2010

In this work, a model of Atucha II PHWR plant (which is under construction), for RELAP5 MOD3.3 code is presented. The implemented nodalization will be used for supporting probabilistic safety analysis, needed in the licensing process, commissioning and support to future operation. The main components and systems of the plant are presented, with their corresponding nodalization. The main characteristics of the plant are discussed, together with the adaptation of the nodalization in order to obtain a correct representation of the expected phenomenologies. Also an integral platform of data management is implemented. The system includes the entry of geometry and process data, their prosecution for calculations needed for the input data and finally the automatic generation of the code input. Then, a complete tracking of data is possible, from the corresponding referenced report to the input deck. This tool facilitates quality-assuring the model by an independent reviewer. The verification of information sources and documentation can be straightforwardly implemented. Finally, the obtained steady state is analyzed, and the main representative variables compared with their respective design values.
Equipment Reliability Based Ageing and Lifetime Management for Asco´ and Vandello´s Nuclear Power Plants
Authors: Jardi´ Xavier , Garci´a Miguel , Castelo Xavier , Marti´nez Luis , Mach Isabel
Published: February 2010

Ageing management is an important issue since the beginning of the operation of the NPPs to ensure that the design functions remain available throughout the service life of the plant. Utilities have selected different strategies to manage plant components ageing degradation, assuring that systems, structures and components important to safety remain within acceptable limits, and that procedures and personnel training remain adapted. At the same time, effective management of degradation processes is an important requirement of Safety authorities, including Nuclear Regulatory bodies. Spanish utilities developed an ageing management methodology in the earlier 90’s. In the new century, this methodology has been adapted to fit 10CFR54 requirements [1] as well as specific necessities of each plant. Although 10CFR54 is formally set for plant life extension, these requirements are a good reference for plant life management activities. In order to take advantage of synergies between different plant programmes and activities, including ageing management, maintenance rule implementation, etc... Asco´ and Vandello´s NPPs (ANAV) have performed an important effort to integrate plant activities during the last four years. The motor driving such integration is the NEI Standard Nuclear Plant Model [2] and particularly the equipment reliability process, AP-913 [3], recommendations. The paper will describe the activities performed by the plants and the support engineering companies, and the expected and achieved results during this process.
ASTEC, COCOSYS, and LIRIC Interpretation of the Iodine Behaviour in the Large-Scale THAI Test Iod-9
Authors: Weber G. , Bosland L. , Funke F. , Glowa G. , Kanzleiter T.
Published: February 2010

The large-scale iodine test Iod-9 of the German THAI programme was jointly interpreted by means of post-test analyses within the THAI Circle of the SARNET/WP16. In this test, molecular iodine (I2) was injected into the vessel dome of the 60 m3 THAI vessel to observe the evolution of its distribution between water, gas, and surfaces. The main processes addressed in Iod-9 are (a) mass transfer of I2 between the gas and the two sumps, (b) iodine transport in the main sump when it is stratified and then mixed, and (c) I2 adsorption onto, and desorption from, the vessel walls in the presence and absence of wall condensation. The codes applied by the THAI Circle partners were ASTEC-IODE (IRSN), COCOSYS-AIM (GRS) and LIRIC (AECL). IODE and AIM are semi-empirical iodine models integrated in the lumped-parameter codes ASTEC and COCOSYS respectively. With both codes multi-compartment iodine calculations can be performed. LIRIC is a mechanistic iodine model for single stand-alone calculations. The simulation results are compared with each other and with the experimental measurements. Special issues that were encountered during this work were studied in more detail: I2 diffusion in the sump water, I2 reaction with the steel of the vessel wall in gaseous and aqueous phases, and I2 mass transfer from the gas to the sump. Iodine transport and behaviour in THAI test Iod-9 are fairly well simulated by ASTEC-IODE, COCOSYS-AIM and LIRIC in post-test calculations. The measured iodine behaviour is well understood and all measured data are found to be consistent. The very slow iodine transport within the stratified main sump was simulated with COCOSYS only, in a qualitative way. Consequently, this work highlighted the need to improve modelling of (a) the wet iodine adsorption and the washdown from the walls, (b) the I2 mass transfer between gas and sump, and (c) the I2/steel reaction in the gaseous and aqueous phases. In any case, the analysis of the large-scale iodine test Iod-9 has been an important validation step for the codes applied.
Test Results From the Idaho National Laboratory 15kW High Temperature Electrolysis Test Facility
Authors: Stoots Carl M. , Condie Keith G. , O’Brien James E. , Herring J. Stephen , Hartvigsen Joseph J.
Published: February 2010

A 15 kW high temperature electrolysis test facility has been developed at the Idaho National Laboratory under the United States Department of Energy Nuclear Hydrogen Initiative. This facility is intended to study the technology readiness of using high temperature solid oxide cells for large scale nuclear powered hydrogen production. It is designed to address larger-scale issues such as thermal management (feedstock heating, high temperature gas handling, heat recuperation), multiple-stack hot zone design, multiple-stack electrical configurations, etc. Heat recuperation and hydrogen recycle are incorporated into the design. The facility was operated for 1080 hours and successfully demonstrated the largest scale high temperature solid-oxide-based production of hydrogen to date.
The 30 MeV Proton Beam Emittance Measurement via a Designed and Constructed Wire Scanner and Using WinAGILE Beam Transport Code
Authors: Ghergherehchi Mitra , Mohammadzadeh Ahmad , Afarideh Hossein
Published: February 2010

A designed and constructed wire scanner is used to measure the 30 MeV Cyclotron proton beam profiles in both x- and y-directions prior to injection to another accelerator for Proton Therapy purposes. The position control and control of movement of the two orthogonal wire scanners are done through an interface board to the PC that communicates with both the photo-couple sensors and the designed driver board, mounted and connected to the step motor set up, respectively. The Win AGILE beam transport code is used to measure the beam profiles 407cm upstream from the wire scanner. The beam sigma’s are evaluated to be σx = 2.234 cm2 and σy = 1.850 cm2. The beam emittance in both x- and y-directions obtained are εx = 1.67×10−5 m-rad and εy = 5.64×10−5 m-rad, respectively. The extract Twiss parameters are calculated as αx = −1.28, βx = 12.7, γx = 0.21, αy = −0.55, βy = 3.28, and γy = 0.10.
Stress Intensity Factors for Unusual Semi-Elliptical Surface Cracks With Ratio a/l Greater Than 0.5
Authors: Malekian Christian , Wyart Eric , Savelsberg Michael , Lacroix David , Teughels Anne , Minjauw Nicolas , Wendling Audrey
Published: February 2010

Most of the literature about fracture mechanics treats cracks with a flaw aspect ratio a/l lower or equal to 0.5 where a is the crack depth and l the total length of the crack. The limitation to 0.5 corresponds to a semi-circular shape for surface cracks and to circular cracks for subsurface cracks. This limitation does not seem to be inspired by a theoretical limitation nor by a computational limit. Moreover, limiting the aspect ratio a/l to 0.5 may generate some unnecessary conservatism in flaw analysis. The present article deals with surface cracks in plates with more unusual aspect ratios a/l>0.5 (narrow cracks). A series of Finite-Elements calculations is made to compute the stress intensity factor KI for a large range of crack depths having an aspect ratio greater than 0.5. The KI values can be used with the same formalism as the ASME XI Appendix A, such that this approach can provide an extension above the inherent limitation to 0.5. Some of the results obtained are checked by using two different Finite-Elements softwares (Systus and Ansys), each one with a different cracked mesh. In addition, a comparison is made for some cases with results obtained by a XFEM approach (eXtended Finite-Element Method), where the crack does not need to be meshed in the same way as in classical Finite-Elements. The results show a reduction of stress intensity factor, sometimes significant, when considering a flaw aspect ratio above 0.5 instead of the conventional semi-circular flaw. They also show that it is not always possible to reduce the analysis of KI to only 2 points, namely the crack surface point and the crack deepest point. The growth by fatigue or by corrosion of a crack with such unusual shape should still be investigated.
Analysis of a BWR ATWS With Boron Injection
Authors: Eriksson Marcus , Bredolt Ulf
Published: February 2010

Analyses are reported for an Anticipated Transient Without Scram (ATWS) event in a Boiling Water Reactor (BWR). The plant response is investigated for a loss-of-feedwater transient without either hydraulic or mechanical control rod insertion. Reactor control is achieved by the injection of soluble boron. Calculations are carried out with the POLCA-T computer code. The computational tool and methodology for carrying out the analyses were established at Westinghouse Electric Sweden. The simulation details all critical phases of the event leading up to stable shutdown conditions. The analysis shows the possibility for reactivity excursions due to sudden dilution of boron in the core, although the safety challenges were found to be limited for the case considered. The reactivity excursions only occur for a short time and are due to the inherently complex events involved. The paper demonstrates the utility of the POLCA-T code to analyze this type of ATWS sequences.
Ex-Vessel Coolability Analysis for a Belgian NPP
Authors: Verpoorten Jarne R. , Auglaire Miche`le , Bertels Frank
Published: February 2010

During a hypothetical Severe Accident (SA), core damage is to be expected due to insufficient core cooling. If the lack of core cooling persists, the degradation of the core can continue and could lead to the presence of corium in the lower plenum. There, the thermo-mechanical attack of the lower head by the corium could eventually lead to vessel failure and corium release to the reactor cavity pit. In this paper, it is described how the international state-of-the-art knowledge has been applied in combination with plant-specific data in order to obtain a custom Severe Accident Management (SAM) approach and hardware adaptations for existing NPPs. Also the interest of Tractebel Engineering in future SA research projects related to this topic will be addressed from the viewpoint of keeping the analysis up-to-date with the state-of-the art knowledge.
Evaluation of the Isolation System Influence in the Seismic Loading Analysis Applied to a Near Term Deployment Reactor
Authors: Lo Frano R. , Forasassi G.
Published: February 2010

Nuclear power plant (NPP) design is strictly dependent on the seismic hazards and safety aspects related to the external events of the site. Passive vibration isolators are the most simple and reliable means to protect sensitive equipment from environmental shocks and vibrations. This paper concerns the methodological approach to treat isolation applied to a near term deployment reactor and its internals structures in order to attain a suitable decrease of response spectra at each floor along the height of the structure. The aim of this evaluation is to determine the seismic resistance capability of as-built structures systems and components in the event of the considered Safe Shutdown earthquake (SSE). The use of anti-seismic techniques, such as seismic isolation (SI) and passive energy dissipation, seems able to ensure the full integrity and operability of important structures and systems even in very severe seismic conditions. Therefore the seismic dynamic loadings, propagated up to the main reactor system and components, may be reduced by using the developed base-isolation system (high flexibility for horizontal motions) that might combine suitable dampers with the isolating components to support reactor structures and building. To investigate and analyze the effects of the mentioned earthquake on the considered reactor internals, a deterministic methodological approach, based on the evaluation of the propagation of seismic waves along the structure, was used. To the purpose of this study a numerical assessment of dynamic structural response behaviour of the structures was accomplished by means of the finite element approach and setting up, as accurately as possible, a representative three-dimensional model of mentioned NPP structures. The obtained results in terms of response spectra (carried out from both cases of isolated and not isolated seismic analyses) were compared in order to highlight the isolation technique effectiveness.
A Hybrid Qualitative/Quantitative Uncertainty Importance Assessment Approach: Applications to Thermal-Hydraulics System Codes Calculations
Authors: Pourgol-Mohammad Mohammad , Sepanloo Kamran
Published: February 2010

Uncertainty importance is used to rank sources of uncertainty in the input variables for their degree of contribution to the uncertainty of the output variable(s). Such ranking is used to plan in reducing epistemic uncertainty in the output variable(s). In application to Thermal-Hydraulic calculations using RELAP5, TRAC and TRACE codes involving uncertainty, uncertainty ranking can be used to confirm the results of Phenomena Identification and Ranking Table (PIRT) utilized in available uncertainty quantification methodologies. Several methodologies have been developed to address uncertainty importance assessment. Existing methodologies for uncertainty importance are not practical for some applications such as some Thermal-Hydraulic calculations due to required computational time and resources. A new efficient uncertainty importance ranking method is proposed as part of a broader research conducted by the authors for comprehensive TH code uncertainty assessment. Given the computational complexities of the TH codes, the proposed uncertainty importance measure is defined in multiples of standard deviation (xσ) changes in a given input parameter or variable over the resulting changes in the standard deviation of output variable (such as a Figure of Merit). The total uncertainty range resulted from propagation of uncertainties is obtained from several available methodologies e.g., CSAU, GRS, UMAE and the recently proposed integrated methodology IMTHUA, proposed by the author. There are some difficulties in assessment of non-linearity of some input changes vs. variations in the output variables, which require special treatment. Different levels of input change (multiples of standard variation) are devised for accurate ranking of uncertainty contributors. Comparing the output change as a fraction of the overall uncertainty range will result in a ranking index to show the contribution of each uncertainty source. In this paper a brief overview of the importance analysis as well as the difference between uncertainty-importance vs. importance uncertainty will be first given. Current methodologies for uncertainty importance will be discussed and their applicability to TH analysis will be discussed. A description of the proposed methodology, along with an example of its application to LOFT LBLOCA uncertain parameters will be discussed.
Validation of Severe Accident Management Guidance in Belgium
Authors: Oury Laurence , Umidova Zeynab , Auglaire Miche`le , Cipollaro Antonio
Published: February 2010

The Belgian Safety Authorities requested to address the severe accident issue within the periodic safety review of the Nuclear Power Plants (NPPs). The key elements of the approach applied in Belgium include the implementation and the validation of the Severe Accident Management Guidance (SAMG). The present paper focuses on the validation of the SAMG that has taken place during the last four years. The validation process aims at verifying the ability of the Technical Support Center (TSC) to manage severe accidents thanks to the implemented Severe Accident Management Guidance. The validation was practically realized by table-top drills based on Westinghouse Owner’s Group SAMG scenario templates. Eight scenarios were exercised: 7 at full power and 1 at shutdown state. One of the scenarios consists in a fast evolving scenario. This paper details the methodology used for the validation process as well as the feedback of the validation exercises realized in the NPPs. In a first phase the objectives of the exercises were defined in agreement with the Safety Authorities; successively, the exercises were realized in the TSC in presence of the Safety Authorities with the active participation of the members of the crisis team that would be present in the TSC. For each exercise, a scenarist presented the chronology of events until the opening of the SAMG. Then, during the severe accident phase, the scenarist regularly provided the evolution of the parameters and was playing also the role of the control room, if needed. That organization is described and analyzed. Moreover, the paper reports on the general feedback from the exercises and the major findings of the validation. Possible adaptations and improvements of the guidelines have been identified, such as for example additional or alternative entry criteria to the guidelines, priorities definition due to the existence of the safety injection-containment spray connection, evaluation of the application of the WOG SAMG to shutdown states and supplementary information for alternative electrical supply in case of station black-out.
System Analyses of High and Low-Temperature Interface Designs for a Nuclear-Driven High-Temperature Electrolysis Hydrogen Production Plant
Authors: Harvego E. A. , O’Brien J. E. , McKellar M. G.
Published: February 2010

As part of the Next Generation Nuclear Plant (NGNP) project, an evaluation of a low-temperature heat-pump interface design for a nuclear-driven high-temperature electrolysis (HTE) hydrogen production plant was performed using the UniSim process analysis software. The low-temperature interface design is intended to reduce the interface temperature between the reactor power conversion system and the hydrogen production plant by extracting process heat from the low temperature portion of the power cycle rather than from the high-temperature portion of the cycle as is done with the current Idaho National Laboratory (INL) reference design. The intent of this design change is to mitigate the potential for tritium migration from the reactor core to the hydrogen plant, and reduce the potential for high temperature creep in the interface structures. The UniSim model assumed a 600 MWt Very-High Temperature Reactor (VHTR) operating at a primary system pressure of 7.0 MPa and a reactor outlet temperature of 900°C. The low-temperature heat-pump loop is a water/steam loop that operates between 2.6 MPa and 5.0 MPa. The HTE hydrogen production loop operated at 5 MPa, with plant conditions optimized to maximize plant performance (i.e., 800°C electrolysis operating temperature, area specific resistance (ASR) = 0.4 ohm-cm2, and a current density of 0.25 amps/cm2). An air sweep gas system was used to remove oxygen from the anode side of the electrolyzer. Heat was also recovered from the hydrogen and oxygen product streams to maximize hydrogen production efficiencies. The results of the UniSim analysis showed that the low-temperature interface design was an effective heat-pump concept, transferring 31.5 MWt from the low-temperature leg of the gas turbine power cycle to the HTE process boiler, while consuming 16.0 MWe of compressor power. However, when this concept was compared with the current INL reference direct Brayton cycle design and with a modification of the reference design to simulate an indirect Brayton cycle (both with heat extracted from the high-temperature portion of the power cycle), the latter two concepts had higher overall hydrogen production rates and efficiencies compared to the low-temperature heat-pump concept, but at the expense of higher interface temperatures. Therefore, the ultimate decision on the viability of the low-temperature heat-pump concept involves a tradeoff between the benefits of a lower-temperature interface between the power conversion system and the hydrogen production plant, and the reduced hydrogen production efficiency of the low-temperature heat-pump concept compared to concepts using high-temperature process heat.
Global Balance of Plant Using Process Data Reconciliation According to VDI2048
Authors: Langenstein Magnus , Laipple Bernd
Published: February 2010

The large quantities of measurement information gathered throughout a plant process make the closing of the plant balance nearly impossible without the help of additional tools. For this reason, a variety of plant monitoring tools for closing plant balances have been developed. A major problem with the current tools lies in the non-consideration of redundant process information which is available throughout the plant. The monitoring system ProcessPlus™ is based on the process data reconciliation program VALI 4, which is certified according to the VDI 2048 standard and is using all redundant pieces of information within the process. Plausibility checks and structured quality control serve as the foundation for the system. Among other components, a procedural process image, significant diagnosis and monitoring tools have been developed and now offer a fast and economically ideal support for process optimization. This paper describes the methodology according to VDI 2048 and the benefits gained from the use of an online plant monitoring system by means of examples from day-to-day operations.
A Comparative Study of Methodologies for Determining Plant Availabilities and Initiator Frequencies
Authors: Kee Ernie , Yilmaz Fatma , Nelson Paul
Published: February 2010

Several methodologies for computing plant (un)avail-abilities and system initiator frequencies are compared to results from a Markov model of a simplified cooling-water system. All methodologies are found to be acceptable for small ratios of failure rates to repair rates, but some degenerate more rapidly than others as these ratios increase. Some, but not all, of this degeneration seems to stem from inappropriate use of various forms of a rare-event approximation. The larger errors tend to be conservative in nature, perhaps excessively so.
Closed Model for Calculation of Differential Pressure After a Loss of Coolant Accident in Boiling Water Reactors
Authors: Kratzsch Alexander , Ka¨stner Wolfgang , Hampel Rainer
Published: February 2010

The paper deals with the calculation of differential pressure on sieves after a loss of coolant accident (LOCA) in boiling water reactors. One of the main features in reactor safety research is the safe heat dissipation from the reactor core and the reactor containment of light-water reactors. In the case of loss of coolant accident the possibility of the entry of insulation material into the reactor containment and the building sump of the reactor containment and into the associated systems to the residual heat exhaust is a serious problem. This can lead to a handicap of the system functions. To ensure the residual heat exhaust it is necessary the emergency cooling systems to put in operation which transport the water from the sump to the condensation chamber and directly to the reactor pressure vessel. A high allocation of the sieves with fractionated insulation material, in the sump can lead to a blockage of the sieves, inadmissibly increase of differential pressure, build-up at the sieves and to malfunctioning pumps. Hence, the scaling and retention of fractionated insulation material in the building sump of the reactor containment must be estimated. This allows the potential plant status in case of incidents to be assessed. The differential pressure is the essential parameter for the assessment of allocation of the sieves.
An Approach for Assessing Fuel Assembly’s Structural Integrity Under In-Core Operating Conditions Using Finite Element Method
Authors: Zhao Wei , Aleshin Yuriy
Published: February 2010

This paper describes the development and application of a single fuel assembly (FA) finite element analysis (FEA) model that can be used to assess and predict the structural behavior of FA during its entire life at early design stage. This model includes all structural components of a FA and has been validated under laboratory test conditions, including spacer grid test under axial load, skeleton test under axial and lateral load, FA test under axial and lateral load, as well as the fuel rod drag test. The model is then applied to an example FA design by predicting the FA’s in-core behavior, including its growth characteristics due to irradiation, and irradiation-induce creep and hydrogen absorption; effect of spring force relaxation of structural spacer grids; effect of holddown spring force increase with growth; effect of different amounts of initial FA bow and bowed shapes; FA distortion evolution vs. burnup history; and stress and deformation of various structural components. It demonstrates that this approach can be used to assess fuel assembly’s structural behavior under various operating conditions at early design stage.
Design of CT Reconstruction Software Based on the Air Container CT Inspection System
Authors: Liu Jinhui , Xiang Xincheng , Wu Zhifang
Published: February 2010

The air container CT inspection system has been developed as a high space-resolution, high density-resolution CT scanner. The design of CT reconstruction software based on the system uses the back project algorithm to construct the CT images, and the filter functions are also included to process the CT data. Because the system can be used to inspect the large objects that cannot be penetrated easily by gamma-ray, the reconstructed CT images may have some artifacts that may damage CT image. The digital subtraction method is used to realize the CT reconstruction for the edge parts of large object that has big gamma-ray attenuation in the center part of the object. The nonlinear effect while measuring is corrected using serials of thin ion plates. The CT images for large objects are improved obviously by using the plate correction. The CT images based on the back project algorithm are also obtained by using the non-perfect projection data. The obtained CT image can have a certain precision to meet the demand. And the principal of the CT reconstruction using the non-perfect project data is also analyzed. The noise influence to the CT imaging quality is also analyzed. Different kinds of noises were added to the signals of all channels or selected channel. The experiment showed that the CT construction realized by the paper has high density resolution and high spatial resolution. The designing for CT construction software is successful.
Improvements in the German Fire PSA Methodology Demonstrated at a German BWR Built to Earlier Standards
Authors: Ro¨wekamp M. , Riese O. , Berg H. P.
Published: February 2010

Fire PSA has to be performed as part of the Probabilistic Safety Analysis (PSA) for full power and low power and shutdown states within the mandatory comprehensive Safety Reviews for German nuclear power plants (NPP) at time intervals of ten years. The German guidance document on PSA methods requires a Fire PSA of several steps starting with alternatively a qualitative and quantitative or an advanced combined screening approach providing estimated values for damage frequencies. The existing combined approach has been enhanced by automating the screening process including a systematic calculation of fire propagation probabilities and standardized simplified fire simulations to roughly predict critical parameters. Further enhancements focus on fire induced cable failures and circuit faults and improvements in the uncertainty and sensitivity analysis. Thus the uncertainties in the results of fire PSA can be reduced as far as feasible and the predicted CDF values are more reliable. The results of a Fire PSA for a German NPP with boiling water reactor designed to earlier standards (BWR-69 type) with an overall fire induced core damage frequency of 1.9 E−06/a are outlined. This is in good agreement with results publicly available from Fire PSA for NPP in other countries. However, the CDF values are higher than those for some other German plants resulting from some pessimistic assumptions made.
A Modern Approach to Condition-Based Maintenance of Reinforced Rubber Cooling System Expansion Joints Utilizing a Microwave Non-Destructive Inspection Method
Authors: Little Jack , Stakenborghs Robert
Published: February 2010

Reinforced rubber has found broad application as a construction material for expansion joints in heat exchanger inlet and outlet piping. Traditionally, the products have been replaced based on age, regardless of the service time or conditions. The fundamental reason for this has been the absence of a suitable nondestructive examination (NDE) method which characterizes the true health of the component. Until recently, the primary tools for condition assessment of such rubber products have been: • Visual examination of surface condition; • Durometer inspection in accordance with ASTM D-2240 (also a determination of surface condition). Microwave Interferometric inspection represents a method for characterizing the condition of such an industrial rubber product volumetrically. This paper will provide technical information on this new NDE technology as well as present inspection results of cooling water system expansion joints installed in US commercial nuclear power plants.
Friction Welding of Mild Steel Components
Authors: Akbar Sajjad , Khalil M. Shahid , Janjua A. Bashir , Amin M.
Published: February 2010

Since the advent of the friction welding process in 1956 and its early application, it has been followed by comprehensive investigations and considerable research effort has been made in U.S.S.R. Japan. U.K. U.S.A. and Europe to find out the most appropriate friction welding process conditions for welding certain pairs of different sizes and configurations. Despite this considerable work, the understanding of the basic mechanism involved in friction welding did not keep pace with the rapid increase in industrial application of the process. This was Probably due to partly relative insensitivity of the weld quality to variation in process control parameters; speed, pressure and burn off and partly to the fact that complex mechanisms are involved in the friction generation of heat and the consolidation of the weld. In this research, it has been deliberated to identify basic or minimum requirements in connection with the production of sound friction welds of mild steel work Pieces. Minimum energy input per unit area of the weld has been calculated and process parameters are recommended for production of sound welds in the mild steel work pieces of round and tabular cross-sections. Mechanical tests have been carried out on test samples welded by friction welding. Results have been discussed.
A Bayesian Framework for Model Uncertainty Considerations in Fire Simulation Codes
Authors: Azarkhail Mohammadreza , Ontiveros Victor , Modarres Mohammad
Published: February 2010

Over the past twenty years, computer-based simulation codes have emerged as the leading tools to assess risks of severe events such as fire. The results of such simulation codes are usually used to estimate the magnitude, frequency and consequence of hazards. A typical simulation program/model utilizes many different sub-models, each characterizing a physical or chemical process contributing to exposure of the hazard or occurrence of certain adverse failure events. The final prediction made by such simulation codes can be temporal, spatial or just a single estimate for the measure of interest. The predictions made by the simulation codes are subject to different contributing uncertainties, including the uncertainty about the inputs to the code, uncertainty of sub-models used in the codes and uncertainty in the parameters of the probabilistic models (if applicable) used in the codes to characterize (e.g., validate) code outputs. A primary way to measure the model uncertainties is to perform independent experiments and assess conformance of the models to observations from such experiments. It is very important to note that the experimental results themselves may also involve uncertainties, for example due to measurement errors and lack of precision in instrumentation. In this research experimental data collected as part of the Fire Model Verification and Validation [1] are used to characterize the share of model uncertainty in the total code output uncertainty, when experimental data are compared to the code predictions. In this particular case, one should assume the uncertainty of experiments (e.g., due to sensor or material variability) is available from independent sources. The outcome of this study is the probabilistic estimation of uncertainty associated with the model and the corresponding uncertainty in the predictions made by the simulation code. A Bayesian framework was developed in this research to assess fire model prediction uncertainties in light of uncertain experimental observations. In this research the complexity of the Bayesian inference equations was overcome by adopting a Markov Chain Monte Carlo (MCMC) simulation technique. This paper will discuss the Bayesian framework, examples of using this framework in assessing fire model uncertainties, and a discussion of how the results can be used in risk-informed analyses.
Break-Up of Gas Stratification in LWR Containment Induced by Negatively Buoyant Jets and Plumes
Authors: Kapulla Ralf , Paladino Domenico , Mignot Guillaume , Zboray Robert , Gupta Sanjeev
Published: February 2010

For the creation of an experimental database related to physical phenomena relevant for LWR containment safety, tests are performed in MISTRA (CEA, France) and PANDA (PSI, Switzerland) facilities in the frame of the OECD/SETH-2 project. The specific purpose of these tests is to obtain data suitable to improve and validate advanced Lumped Parameter (LP) codes as well as codes with 3D capabilities with respect to the prediction of post-accident containment thermal-hydraulic conditions. The experimental data is related to hydrogen transport within containment compartments. In particular, the effect of mass sources (the release of steam and hydrogen), heat sources (hydrogen-oxygen recombiner), and heat sinks (condensation of steam caused by containment coolers and sprays or “cold” wall) on the break-up/erosion of an initially gas stratified configuration characterized by a layer with a high hydrogen content. Helium is used to simulate hydrogen in the PANDA facility. This paper presents the result of a series of SETH-2 PANDA tests attributed to “vertical fluid release” (plumes or jets). Two large containment compartments (∼180 m3) connected by a bended pipe of ∼1 m diameter are used for these tests. For all the tests, a helium-steam mixture having a thickness of 2 m is created in the upper volume of one compartment while the remaining volume is filled with steam. During the tests, steam jets or plumes are created by injecting steam from a vertical pipe located at the center of the vessel 2 m below the helium-steam mixture. The jet or plume is initially positively buoyant and becomes negatively buoyant once it reaches the helium-steam layer. These transient tests show the degradation of the helium-steam layer for different jet Reynolds numbers. The initial Froude number at the injection pipe varied in the range of ∼3 to ∼9, while the estimated Froude number at the helium-steam mixture/steam interface varied from ∼0.70 to ∼2.
Safety Analysis of the Control Rod Drop Event Applying 3-D Neutronics/Thermal-Hydraulics Coupling Code SPARKLE
Authors: Maruyama Yuta , Imura Satoshi , Ogawa Junto , Miyake Shuhei
Published: February 2010

Mitsubishi Heavy Industries (MHI) has developed the SPARKLE code, which is a PWR plant system transient analysis code that includes a three-dimensional (3D) neutronics module coupled with a thermal-hydraulics module. MHI has performed a study of the applicability of the SPARKLE code to the events which are associated with dynamic changes in power distribution, such as the rod ejection event or the steam line break event. In this paper, MHI has applied the SPARKLE code to the control rod drop event (drop of multiple rods), which features such a power distribution change. In addition, the neutron flux detection is dependent on the location of the dropped rods in this event, which can be dynamically calculated in the SPARKLE code. By applying the SPARKLE code to the control rod drop event, it was confirmed that the safety margin for this event is sufficiently larger than the margin calculated using the current safety analysis method, even if the appropriate conservative assumptions are made.
Simplified Approach to Establishing Bounding Main Steam Piping Support Load Increase Due to Reactor Power Uprate
Authors: Stakenborghs Robert , Dziuba Lindsey
Published: February 2010

Many nuclear power reactors are increasing their electrical output by increasing their licensed thermal power output. This process requires submitting license change request information to the appropriate regulatory body. One of the fundamental changes to a plant operating condition that must occur to provide additional electrical power generation is an increase in the main steam mass flow. This increase in mass flow results in an increase in the fluid velocity at the main turbine stop and control valves. The increased velocity can be correlated to an increase in the piping support loads as a result of a main turbine trip with a corresponding stop valve closure. The increased loads are a concern for the regulatory body as well as the plant owners since these loads occur at some real frequency in a plant. Typically, an overly conservative scaling factor has been applied to these loads, which may result in unnecessary plant modifications. This paper discusses the fundamental transient fluid dynamic process that results in the main steam piping load. A simplified method is provided that more realistically estimates bounding magnitudes of piping support loads versus percentage power uprate. This method will allow plant design engineers to more realistically determine the increase in the main steam piping support loads that result from power uprate conditions.
Wave Propagation Analysis of Piping Structures
Authors: Nishida Akemi
Published: February 2010

It is becoming important to carry out detailed modeling procedures and analyses to better understand the actual phenomena. Because some accidents caused by high-frequency vibrations of piping have been recently reported, the clarification of the dynamic behavior of the piping structure during operation is imperative in order to avoid such accidents. The aim of our research is to develop detailed analysis tools and to determine the dynamic behavior of piping systems in nuclear power plants, which are complicated assemblages of different parts. In this study, a three-dimensional dynamic frame analysis tool for wave propagation analysis is developed by using the spectral element method (SEM) based on the Timoshenko beam theory. Further, a multi-connected structure is analyzed and compared with the experimental results. Consequently, the applicability of the SEM is shown.
Advanced Studies on Plutonium for Pu Isotope Transmutation and Pu Separation
Authors: Mohamed Ashraf Elsayed Mohamed
Published: February 2010

In this paper, I discuss and present the advanced experiments methods for thermochemical properties of plutonium in liquid gallium at 1073 k, application to plutonium cerium separation and the theoretical studies on Pu isotope transmutation. Electrochemical investigations on plutonium diluted in liquid gallium using molten chlorides showed that gallium can be used as solvent metal for pyrochemical processes involving plutonium at 1073 k. In particular, a selective extraction of plutonium from gallium by anodic oxidation is thermodynamically possible. The deposition of plutonium on a liquid gallium cathode is also possible with CaCl 2 as molten salt with a careful control of cathode potential.
Lessons Learned and Insights From Implementation of EPRI Guidelines on Internal Flooding at Fort Calhoun Station
Authors: Sadlon Daniel L. , Weston Rupert A. , Schneider Raymond E. , Fluehr John J. , Hackerott Harold A.
Published: February 2010

An Internal Flooding Probabilistic Risk Assessment (IF-PRA) study is required to meet the requirements of the Internal Flooding portions of the ASME Internal Events PRA standard [1, 2, 3] and RG 1.200 R1 [4] for risk-informed applications at Nuclear Power Plants (NPPs). An internal flooding study was performed recently for a Nuclear Power Plant (NPP), Fort Calhoun Station (FCS), with a nuclear steam supply system designed by Combustion Engineering. The study was performed using guidance that is currently being developed for industry use by EPRI [5]. The intent of the EPRI draft guidance is to describe how to perform the various tasks to meet the requirements and expectations of the ASME PRA Standard in order to comply with RG 1.200. Performing this analysis provided an opportunity to gain numerous insights on how to best perform the analysis from a technical standpoint. This paper provides lessons learned and insights gained in the implementation of the EPRI draft guidance on internal flooding for PRA. Insights and lessons learned on the assignment of initiating event frequencies, treatment of dynamic flood volumes and time-based operator response timings, and developing associated peer review materials are provided. Discussions dealing with various sources of pipe failure frequencies highlight key differences between commonly used data. Characterizations of spray scenarios were evaluated to determine their impact on plant risk caused by internal flooding events. Maintenance-induced flooding scenarios that are often neglected were evaluated to gain insights on their impact on plant risk. Insights on performing a timely completion of the analysis are provided. Optimization techniques for plant walkdowns to capture relevant information and identify high priority flood areas are discussed. The impact of the flooding scenarios on PRA-related equipment was integrated into the internal events PRA model to quantify the associated risk. The rules used to perform the integration process and the lessons learned from the integration are also discussed.