bluuebunny/crossref_metadata_2025_split huggingface dataset:
How the Use of Limited Plant Data Can Support Limited Improvements in Combined Cycle Operation
Authors: Perez James M. , Toburen Tina L. 
Published: October 2013

Process improvement and plant optimization are common initiatives among Utilities and Independent Power Producers (IPP). This paper intends to outline the value and challenges with obtaining data to support these initiatives and provide for good business decisions.
Typical information, such as performance acceptance tests and heat balance design data is essential when evaluating plant performance. Normal, on-going operating data at full load conditions is also needed when making decisions regarding plant improvements. However, obtaining all the necessary data is not always possible and with the current market conditions, conducting performance tests may not be economical. When faced with a known performance deficiency, determining the root cause and options for performance recovery can be difficult in any situation. When design data is hard to come by and operating data difficult to obtain, making an informed decision can be more of an art than a science.
An example of the decision process used to upgrade the capacity for a gas turbine combined cycle facility is explained in detail in this paper. The results of the process show that even with limited available data for the facility, improvements in performance can be achieved.
In Vessel Loss of Coolant Accident Analysis for the JET Tokamak in Support of the Safety Case Review
Authors: Vallory Joelle , Boyer Helen , Moreno Fabien , Kadri Dominique 
Published: October 2013

The Joint European Torus is currently the largest operating tokamak in the world. JET was established with a long-term objective to contribute to the creation of safe, environmentally sound prototype fusion reactors. To meet this objective, JET operates using deuterium as a fuel but also in D-T mode where the fusion reaction is fuelled by deuterium and tritium. To justify the safety of D-T experiments, safety reports have been produced at stages to obtain approval for the past experimental campaigns. The Safety Case is currently under review in preparation of the next tritium campaign in 2014/ 2015.
In this prospect, the hazard assessments have been reviewed and when possible, the calculations in support of these assessments have been updated. For the in vessel loss of coolant accident, the state of the art CEA developed thermal hydraulics code CATHARE 2 V2.5_2, has been adapted down to pressure relevant to fusion applications. The complexity of the heat transfer mechanisms involved in the vacuum vessel has been represented with different elements available in the code to improve the 0D model. Modeling parameters have been studied and any compromise in modeling was done by increasing conservatism of the peak pressure in the VV. Some physical parametric studies have also been run especially on the failure of safety equipment designed to minimize the in vessel consequences on the LOCA. One major result of the analysis is that the initial calculations were conservative in their assumptions and that some break size initially considered as beyond design could lead to pressure peak in the vessel which could be limited by the safety systems below the bursting disk pressure.
Performance Experienced From Switching Copper Condenser Tubing to a High Performance Stainless
Authors: Goebel Summer , Criminger Colin R. 
Published: October 2013

Over a period of five years, the condensers of two sister 500 MW Units, in the Midwest, were retubed using a SEA-CURE® S44660 material. S44660 is a high alloy, super ferritic stainless steel. The S44660 material replaced the original Admiralty and copper/nickel condenser tubing. Several factors contributed to the decision to use S44660 as the condenser tube material. The critical factor in the decision was a pending lower copper discharge limit in the plant’s National Pollutant Discharge Elimination System (NPDES). The plant was unable to meet the lower copper limits with copper condenser tube material.
In service since the early 1970’s, Unit A and B condensers were subjected to conditions, including silting and operational upsets, which negatively affected condenser performance. A third 500 MW sister Unit, Unit C, operational in the mid 80’s, experienced similar conditions. All three units are based load coal units and use cooling towers to cool the circulating water. The cooling tower make-up comes from the Arkansas River, which also receives the effluent from the plant.
During the summer of 2011, shortly after the Unit B condenser retube, both A & B condensers showed degradation in performance. Through a series test and trials, the causes of degradation were identified and corrected. The cause was a combination of the tube material, lower river levels, higher than normal temperatures, macro fouling, and air in leakage.
Capability of the Integral Code ASTEC V2 to Simulate a Blow Down in Plant Scale Tests
Authors: Brähler Thimo , Risken Tobias , Koch Marco K. 
Published: October 2013

In this paper the Accident Source Term Evaluation Code (ASTEC) is validated against the blowdown experiments Marviken M19 and M24. These tests mainly differed by the mass flow of the released steam from the pressure vessel and the configuration of the vent pipes in the pressure suppression chamber — while in M19 all vent pipes were arranged to one pool, in M24 they were split up by 27 pipes in one and one pipe in another pool.
For the simulation of both tests, an existing model of the facility for another lumped parameter code COCOSYS was transferred to ASTEC. In this data set a simple zone and a flow connection was used to model the pressure suppression chamber. Further simulations were performed with another approach for the pressure suppression chamber, so called “DRASYS”-zones. Using the DRASYS model, the user has to specify more inputs for the geometry of the pressure suppression chamber. The results of the simulations are in good agreement to the measured pressure and temperature of both tests. By using the DRASYS model in ASTEC, the results were improved slightly for M19 compared to the simple pressure suppression zone model. In opposite, the results of the simple model are in better agreement to M24. Overall the conclusion is that ASTEC is able to simulate a Blow Down in plant scale with both models.
Plant Dynamics Evaluation of a Monju Ex-Vessel Fuel Storage System During a Station Blackout
Authors: Mori Takero , Sotsu Masutake , Honda Kei , Suzuki Satoshi , Ohira Hiroaki 
Published: October 2013

The prototype fast breeder reactor “Monju” has an ex-vessel fuel storage system (EVSS) which consists mainly of an ex-vessel fuel storage tank (EVST) and an EVST sodium cooling system. EVST uses natural circulation of sodium for decay heat removal. Natural circulation in the EVST is generated by the decay heat from the spent fuel assemblies and the cooling of the cooling coils installed in the EVST. The EVST sodium cooling system consists of three independent loops. In each loop, sodium is circulated by electromagnetic pumps and the heat is removed by an air cooler with blowers. This system has the ability to remove the maximum decay heat using two loops, and thus, it uses two of the three loops for normal operation. During a station blackout (SBO), the pumps and blowers are stopped. However, the three air coolers are installed about 13.5 m higher than the cooling coils, and therefore, the EVST sodium cooling system potentially retains some cooling ability because of natural circulation. In this study, an analysis and evaluation of the plant dynamics for the spent fuel and the EVSS structural integrity during an SBO were performed.
The ultimate heat sink for the EVST sodium cooling system is the atmosphere, and the air coolers have an exhaust stack for efficient natural circulation caused by the chimney effect. However, the EVST sodium cooling system loses pressure and the heat transfer characteristics change if the flow rate is low. It was, therefore, necessary to confirm the temperature and flow rate behavior of EVSS in this analysis.
In the present calculations, the plant dynamics analysis program “Super-COPD” was used. The factors affecting the cooling ability were investigated and analytical cases were determined. In one case, the two operated loops were switched to natural circulation after an SBO. The number of cooling loops was then changed from two to three by having an operator open the vane and dampers of the standby loop. In this case, sodium temperature in the EVST increased to approximately 320°C. When the number of cooling loops was not changed and natural circulation occurred in only two loops, the sodium temperature in the EVST increased to approximately 450°C. In both cases, however, the structural integrity of the EVSS was maintained. These analytical results, therefore, help clarify the number of necessary cooling loops for efficient decay heat removal and sodium temperature behavior in an SBO.
Simplified Approach to Design of Crushable Material due to Dynamic Load
Authors: Vongmongkol Sivadol , Faal-Amiri Asgar , Srivastava Hari M. 
Published: October 2013

Crushable material has widely been used as an engineering solution for energy absorption devices among many industries. Abnormal and severe accident loads in the design of nuclear power plants are required to be addressed in order to comply with Nuclear Regulatory Commission (NRC) requirements which makes the crushable material more suitable in its highly dynamic application. One of the severe loads is from a postulated high energy piping system rupture. Its effects are required to be mitigated so that the proper operation of safety related systems, structures and components (SSC) of these facilities is assured. The postulated pipe rupture loads are among the highest loads that need to be addressed in the design process of nuclear power plants.
The impact forces produced by the postulated pipe rupture are typically being absorbed by energy absorption devices called “Pipe Whip Restraints” in which the restraints can minimize the loads affecting the SSCs to within an acceptable limit. This paper provides a simplified closed-form solution to determine the energy absorbing characteristic that will help to design these devices. This paper will also provide a comparison between results of the proposed simplified closed-form solution equations to the experimental test results and the calculated results using finite element analysis.
Assessment of Selected Two-Phase Friction Multiplier Correlations for Steam-Water Flows in Vertical Heated and Unheated Tubes
Authors: Durmayaz A. , Odabaş I. A. , Groeneveld D. C. , Gengeç N. E. 
Published: October 2013

In this study, around 10 000 experimental two-phase (2ϕ) flow pressure drop (ΔP) data for steam-water in uniformly heated and adiabatic vertical tubes have been compiled from the literature.
The compiled data have been used to assess the prediction accuracy of two-phase friction multiplier (ϕlo2) correlations selected from the literature and the homogeneous-flow model for steam-water flow in vertical tubes. The predictions of the ϕlo2 correlations have also been compared to the predictions of the homogeneous-flow model. The assessments are based on comparisons of the average and the RMS errors.
Additionally, ΔP components (due to gravitation, acceleration and friction) have been examined for a variety of flow conditions in vertical heated and unheated tubes.
The results of the assessments show that the predictions of Muller-Steinhagen and Heck (1986) correlation yield the minimum RMS error for flow in heated tubes and the predictions of the Chisholm (1973) correlation yield the minimum RMS error for flow in adiabatic tubes when these predictions are compared with both the data base and the predictions of the homogeneous-flow model.
It is concluded from the research and a review of the literature that the current state of the art is such that none of the investigated prediction methods is very good because of their large RMS errors. More work is needed to expand the ΔP data base especially for adiabatic two-phase flow. Also, a systematic experimental study should be performed on the effect of heating on the 2ϕ-flow ΔP.
Post Test Analysis of ICE Tests
Authors: Tarantino Mariano , Del Nevo Alessandro , Forgione Nicola , Bandini Giacomino 
Published: October 2013

Since 1999 ENEA is developing the heavy liquid metal (HLM) technology aiming to support the design and implementation of a Lead cooled Fast Reactor (LFR) and an Accelerator Driven System (ADS), both in the frame of the Italian and European research programs.
In these contexts several experiments have been performed, in different fields, going from coolant thermal-hydraulic, component development and structural material characterization.
Recently, in the frame of the IP-EUROTRANS (6th Framework Program EU), domain DEMETRA, ENEA assumed the commitment to perform an integral experiment aiming to reproduce the primary flow path of a pool-type nuclear reactor, cooled by Lead Bismuth Eutectics (LBE).
This experimental activity, named “Integral Circulation Experiment (ICE)”, has been implemented thanks a joint effort of several research institutes, mainly ENEA and University of Pisa, allowing to design an appropriate test section. This has been installed in the CIRCE facility, the largest worldwide experimental facility for the HLM technology investigation.
The goal of the experiments was to demonstrate the technological feasibility of a heavy liquid metal (HLM) pooltype nuclear system in a relevant scale (1 MW), investigating the related thermal–hydraulic behavior under both steady state and transient conditions.
This paper reports a description of the experiment, as well as the results carried out in the first experimental campaign run on the CIRCE pool, which consists of a full power steady state test, an un-protected loss of heat sink (ULOH) test, and an un-protected loss of flow (ULOF) test.
The post-test analyses of the experiments is presented. The whole domain has been modeled by a suitable 1-D nodalization, and the results carried out have been studied performing numerical calculations by the REALP5 system code modified to take in account the LBE thermal-physical properties when employed as nuclear coolant.
The obtained experimental results as well as the performed post-test analysis have demonstrated the thermal-hydraulic and technological feasibility of a pool-type nuclear system cooled by HLM.
Estimation of Emergency Planning Zones (EPZs) for Nuclear Research Reactor Using Plume Dispersion Code
Authors: Hussain Mazzammal , Khan Salah Ud-Din , Syed Waqar A. Adil 
Published: October 2013

After Fukushima nuclear power plant accident, a demand for re-evaluation of emergency planning zones for nuclear facilities has emerged to ensure that in case of nuclear accident, the population, the environment and the property should lie in the safe zone. Emergency planning zones (EPZs) around a facility are the designated areas where protective measures are adopted according to predefined emergency plan for different emergency situations.
Deterministic and probabilistic approaches have been adopted to assess and design emergency planning zones for design base accidents (DBAs), however, after Fukushima nuclear power plant accident, multi-layer failure of the plant systems could not be ruled out. To estimate the dispersion of radioactive material and respective radiation doses received by the affected people, different techniques are used including on-field measurements through survey meters, on-line monitoring network and state-of the art codes for plume dispersion modeling.
In this study plume dispersion modeling and estimation of emergency planning zones has been carried out using plume dispersion code for hypothetical accident scenarios at a 10MW nuclear research reactor. Different accident scenarios were considered with different release characteristic and environmental conditions to study the affect of the parameters including release height, heat content, release time, atmospheric stability. The simulation results have been analyzed to assess the existing emergency planning zones of nuclear research reactor.
Scientific Bases of Corrosion Control for Water-Cooled Fusion Reactors Such as ITER
Authors: Saji Genn 
Published: October 2013

Scientific bases for corrosion control for water-cooled fusion reactors, such as ITER, are discussed in this paper.
In the previous overview papers [1,2], the author has identified that ‘long cell action’ corrosion plays a pivotal role in practically all unresolved corrosion issues, irrespective of fission reactor types and operation. In trying to confirm the existence of radiation-induced ‘long-cell’ action (macro) corrosion cell in the primary cooling system of LWRs, the author attempted to theoretically reproduce the electrochemical potential differences demonstrated during experiments at the INCA Loop in Sweden and the NRI-Rez Loop in the Czech Republic [3,4].
Based on these knowledge bases, characterization of water chemistry for fusion was made through radiation- and electro-chemistry analysis following the methodology explained in the companion paper also presented at this conference [5].
Approximately 70 mV of potential difference is predicted between the high dose rate and the out-of-flux regions through the illustrative model examined in this paper, which supports the existence of the LGCA mechanism. This value is about 1/5 of the potential difference for a typical PWR. For further verification of this approach, ‘international benchmark in-pile tests’ are highly awaited. Further optimization studies as well as material testing under the influence of the LGCA mechanism should be performed to establish corrosion control.
Design and Control of Bypass Condensers
Authors: Nadig Ranga , Phipps Michael 
Published: October 2013

In waste to energy plants and certain genre of cogeneration plants, it is mandatory to condense the steam from the boiler or HRSG in a separate bypass condenser when the steam turbine is out of service. The steam from the boiler or HRSG is attemperated in a pressure reducing desuperheating valve and then condensed in a bypass condenser. To avoid flashing of condensate in downstream piping it is customary to subcool the condensate in the bypass condenser. Circulating water from the steam surface condenser is used to condense the steam in the bypass condenser.
Some of the challenges involved in the design of the bypass condenser are:
• High shellside design pressure and temperature
• Condensate subcooling
• Large circulating water (tubeside) flow rate
• Relatively low circulating water (tubeside) inlet temperature
• Large Log Mean Temperature Difference (LMTD)
• Large shell diameters
• Small tube lengths
The diverse requirements complicate the mechanical and thermal design of the bypass condenser. This paper highlights the complexities in the design and performance of the bypass condenser. Similarities with the design and operation of steam surface condensers and feedwater heater are reviewed. The uniqueness of the bypass condenser’s design and operation are discussed and appropriate solutions to ensure proper performance are suggested.
Flow Characterization and Heat Transfer of Core-Catcher in Passive Safety System
Authors: Kurita Tomohisa , Tobimatsu Toshimi , Tahara Mika , Yamada Masato , Kojima Yoshihiro 
Published: October 2013

A mitigation system which can keep core melt stable after a severe accident is necessary to a next generation BWR design. Toshiba has been developing a compact core catcher to be placed at the lower drywell in the containment vessel. The cooling water for the core catcher is supplied from the passive flooder and PCCS drain line. After the core catcher is flooded, the molten core would be cooled by both overflooding water and inclined cooling channels, in which water is boiling and natural circulation is established. So the core catcher can operate in passive manner and has no active component inside the containment.
This paper summarizes flow dynamics and heat removal capability in an inclined cooling channel of core catcher when cooling water flows by the natural circulation.
Experimental and Analysis Study of Free Standing Rack Under Seismic Excitation
Authors: Taniguchi Katsuhiko , Okuno Daisaku , Iwasaki Akihisa , Nekomoto Yoshitsugu , Matsuoka Toshihiro 
Published: October 2013

For high earthquake resistance and ease of installation, free standing racks which are not anchored to the pool floor or walls has been adopted in many countries.
Under the earthquake, the response of the free standing rack is highly nonlinear and involves a complex combination of motions (sliding, rocking, twisting, and turning) and impacts between the fuel assemblies and the fuel cell walls, rack-to-rack, and the pit floor and rack pedestals. To obtain an accurate simulation of the free standing rack, the seismic analysis requires careful considerations of these complex phenomena (sliding, rocking, twisting, and turning), fluid coupling effects and frictional effects.
We carried out seismic experiments on the full-scale rack model in water and dry conditions to obtain the fundamental data about free standing rack (sliding, rocking and turning motions). We have developed the nonlinear dynamic analysis method to predict seismic response for the free standing rack utilizing the full-scale test result and verified the analysis evaluation method of the rack by comparison of test result.
Is ExacTrac x‐ray system an alternative to CBCT for positioning patients with head and neck cancers?
Authors: Clemente Stefania , Chiumento Costanza , Fiorentino Alba , Simeon Vittorio , Cozzolino Mariella , Oliviero Caterina , Califano Giorgia , Caivano Rocchina , Fusco Vincenzo 
Published: October 2013

Purpose:To evaluate the usefulness of a six‐degrees‐of freedom (6D) correction using ExacTrac robotics system in patients with head‐and‐neck (HN) cancer receiving radiation therapy.Methods:Local setup accuracy was analyzed for 12 patients undergoing intensity‐modulated radiation therapy (IMRT). Patient position was imaged daily upon two different protocols, cone‐beam computed tomography (CBCT), and ExacTrac (ET) images correction. Setup data from either approach were compared in terms of both residual errors after correction and punctual displacement of selected regions of interest (Mandible, C2, and C6 vertebral bodies).Results:On average, both protocols achieved reasonably low residual errors after initial correction. The observed differences in shift vectors between the two protocols showed that CBCT tends to weight more C2 and C6 at the expense of the mandible, while ET tends to average more differences among the different ROIs.Conclusions:CBCT, even without 6D correction capabilities, seems preferable to ET for better consistent alignment and the capability to see soft tissues. Therefore, in our experience, CBCT represents a benchmark for positioning head and neck cancer patients.