bluuebunny/crossref_metadata_2025_split huggingface dataset:
The SCORPIO-VVER Core Monitoring and Surveillance System for VVER Type of Reactors
Authors: Molnar Jozef , Vocka Radim 
Published: November 2014

The SCORPIO-VVER core monitoring and surveillance system has proved since the first installation at Dukovany NPP in 1999 to be a valuable tool for the reactor operators and reactor physicists. It is now installed on four units of Dukovany NPP (EDU, Czech Republic), on two units of Bohunice NPP (EBO, Slovak Republic) replacing the original Russian VK3 system and on the full scale plant training simulator at the Centre for training and education of the reactor operators and reactor physicist in Trnava (Slovak Republic). By both Czech and Slovak nuclear regulatory bodies the system was licensed as a Technical Specification Surveillance tool.
Since it’s first installation, the development of SCORPIO-VVER system continues along with the changes in VVER reactors operation. The system is being adapted according the utility needs and several notable improvements in physical modules of the system were introduced. The most significant changes were done in support of the latest optimized Gd bearing fuel assemblies, improvements in the area of core design (neutron physics, core thermal hydraulics and fuel thermal mechanics), adaptation of the system to up-rated unit conditions (uprated power up to 107%), in design and methodology of the limit and technical specifications checking and improvements in the predictive part of the system.
After the currently finished upgrades the SCORPIO-VVER is still in focus of Central European nuclear power plants with the roadmap of upgrades and modifications up to 2016.
This paper shortly describes the system’s main functions, the history of implementation at the VVER-440 type of reactors and deals with the system’s future upgrades and plans to meet the latest requirements of efficient and safety NPP operation.
A Double Porosity Model to Describe Both Permeability Change and Dissolution Processes
Authors: Niibori Yuichi , Usui Hideo , Chida Taiji , Mimura Hitoshi 
Published: November 2014

Cement is a practical material for constructing the geological disposal system of radioactive wastes. However, such materials alter groundwater up to 13 in pH around the repository, changing the permeability of natural barrier. So far, the authors have examined the relation of permeability change with dissolution process by flowing a high pH solution (NaOH, 0.1 mM) into a bed packed with amorphous silica particles. Here, the particle diameters were adjusted to a size fraction of 74 to 149 μm by sieving. Its specific surface area was estimated as 350 m2/g by the BET method using nitrogen gas. The experimental results showed that the permeability did not immediately change although the soluble silicic acid continuously flowed out of the packed bed.
This study proposes a new mathematical model considering the diffusion and dissolution processes in the inner pore of the particle. This model assumed that each packed particle (74 to 149μm in diameter) consists of the sphere-shaped aggregation of smaller particles (20 nm in diameter). OH− ions diffuse into the pore between such small particles, and simultaneously consumed by the reaction with small particles. The radius of the each packed particle (sphere-shaped aggregation of small particles) was defined by the length from the center of the aggregation to the region where the small particles still remains. Since the outer small particles more easily dissolve than inner small particles because of diffusion process of OH− ions, each packed particle gradually shrinks. The fundamental equations consist of a simple diffusion equation of spherical coordinates of OH− ions considering the reaction term, which is linked by the equation to describe the size change of small particles with time. Here, this model also considered a change (time and space) of the diffusion oefficient caused by the change of the porosity between small particles. Besides, the change of over-all permeability of the packed bed was evaluated by Kozeny-Carman equation and the calculated radii of packed particles. The dissolution rate constant already reported was used.
The calculated result was able to well describe the experimental result, though there was no fitting parameter in the comparison with the experiment results. While the flow paths of underground cannot be simply simulated by a packed bed, this approach suggested that the dynamic behavior of permeability in a natural barrier depends also on non-uniformity of dissolution processes in inner pores (secondary pores) of minerals.
Development of Supercritical Water Heat-Transfer Correlation for Vertically-Upward Internally-Ribbed Tubes
Authors: Zhang Weiqiang , Li Huixiong , Zhang Qing , Zhang Yifang , Wang Tai 
Published: November 2014

The investigation on the heat transfer characteristics for supercritical pressure water (SCW) is of value for the development of the supercritical water-cooled nuclear reactor (SCWR). As an important heat transfer enhancement element, heat transfer for SCW in internally-ribbed tubes was still not solved, though lots of experimental studies have been published and a great many heat transfer correlations were proposed.
This paper presented an analysis of heat transfer in the internally-ribbed tubes, through comparing heat transfer correlations for SCW gained from different internally-ribbed tubes under the same operating condition. It was found that all existing heat transfer correlations reported could not been well applied for various internally-ribbed tubes with large deviation between prediction results and experimental values, because rib geometry had a great influence on heat transfer of internally-ribbed tubes. On the basis of experimental data collected from open literature for internally-ribbed tubes, a new general calculation correlation of heat transfer coefficient for SCW was developed for various internally-ribbed tubes by combining an optimized empirical correlation for vertically-upward smooth tubes and four dimensionless numbers of rib geometry.
The results show that the calculated values of the new present correlation is in reasonable agreement with available experimental data collected. Moreover, the new correlation was verified well by experiment data of two new-type internally-ribbed tubes performed beyond the above experimental database.
Development of Instrumentation System for Severe Accidents
Authors: Wada Shohei , Murata Akira , Arita Setsuo , Fushimi Atsushi , Suzuki Hirotsugu , Fujishima Yasutake 
Published: November 2014

The accident at Tokyo Electric Power Company’s Fukushima Dai-ichi Nuclear Power Plant (TF-1 accident) caused severe situations and resulted in a difficulty to measure important parameters to monitor plant conditions. Therefore, we have studied to select the important parameters (SA parameters) that should be monitored at the severe accident and are developing the Severe Accident Instrumentations and Monitoring Systems (SA-keisou) that could measure the parameters in severe accident conditions. (SA-keisou in Japanese means “severe accident instrumentation and monitoring systems). In this study, the boiled water reactor (BWR) is mainly explained.
This study is a part of the results of the collaborative project by domestic electric power companies and plant manufacturers that is carried out as the Safety Enhancement for LWRs program by Agency for Natural Resources and Energy.
Buckling Analysis of Tube for Steam Generator According to ASME B&PV Code
Authors: Kong Jungpyo , Ha Changhoon , Cho Minki , Yeo Seunghoon , Park Tae-Jung 
Published: November 2014

The U-tubes installed inside the steam generator experience high temperature and pressure as a role of heat transfer. Specially, during the secondary side hydrostatic test which ensures the integrity of steam generator, the U-tube is subjected to high external pressure. The purpose of this paper is to investigate the allowable external pressure of the U-tube in steam generator.
In the ASME B&PV Code Section III [1], the allowable external pressure is determined by the rules of NB-3133. Alternatively, NB-3228 analysis may be applied. In order to determining the allowable external pressure of steam generator tube, the buckling analysis is performed. The analysis consists of the collapse pressure and elastic instability pressure analyses. In this study, these pressures are determined by finite element analysis (FEA) using ANSYS computer program.
The non-linear static analysis is performed with ideally elastic-plastic material properties for the collapse analysis. On the other hand, the elastic instability pressure is calculated by eigenvalue analysis in elastic range. These allowable pressures are found to be 24.1 ksi and 10.5 ksi. Therefore the lower pressure of 24.1 ksi is the allowable external pressure of tube. In addition, the results of analysis are compared with other research [6] and Det Norske Veritas (DNV) offshore standards [11].
In conclusion, the results of buckling analysis are well matched with other research [6] and standard [11]. For steam generator tubes, the collapse pressure is dominant factor in failure. Also, the collapse pressure is largely influenced by the ovality of tube.
Analysis of Accident Progression of Fukushima Daiichi NPP by the SAMPSON Code
Authors: Naitoh Masanori , Pellegrini Marco , Suzuki Hiroaki , Mizouchi Hideo , Okada Hidetoshi 
Published: November 2014

This paper describes analysis results of the early phase accident progression of the Fukushima Daiichi Nuclear Power Plant (NPP) Unit 1 by the severe accident analysis code SAMPSON.
The isolation condensers were the only devices for decay heat removal at Unit 1, but they stopped after the loss of AC and DC powers. Since there were no decay heat removal for about 14 hours after their termination until the start of alternative water injection into the core by the fire engine, the core melt and the reactor pressure vessel (RPV) bottom failure occurred resulting in large amount of fission products release into the environment. The original SAMPSON was improved by adding new modellings for the phenomena which have been deemed specific to the Fukushima Daiichi NPP: (1) deterioration of SRV gaskets and (2) buckling of in-core-monitor housings which caused the early steam leakage from the core into the drywell, and (3) melt of the in-core-monitor housings in the lower plenum of the RPV.
The analysis results showed that (1) 55.3% of UO2 of the initial loading and 66.1% of the core material including UO2, zircaloy, steel and control materials had melted down into the pedestal of the drywell, (2) the amount of Hydrogen generated by Zr-H2O reaction was 686 kg, (3) amount of Cs element released from fuels was 61 kg which was 72% of the total Cs element which was included in fuels at the initiation of the accident, and (4) 18.3% of the corium which fell into the pedestal was one large lump and the 81.7% was particulate corium.
Modelling of the Impact of Local Effects on Fuel-Cladding Interaction During Power Ramp
Authors: Klouzal Jan , Dostál Martin 
Published: November 2014

The power increase rate of the reactor is often derived using the fuel performance code. Too restrictive rates are not desirable since they lead to the loss of production. On the other hand fast increase or not well controlled axial power oscillation may result in the rod failure due to pellet-cladding interaction.
Most of the currently used fuel performance codes treat the stack of the fuel pellets using a simplified “1.5D” approach where individual pellets are not distinguished and the fuel stack is taken to be symmetrical. In reality, several effects must be taken into account when more accurate description is required. These local effects include contact at pellet-pellet interface, fuel pellet cracking under thermal stress, fabrication defects pellets or azimuthal asymmetry in the heat generation or heat transfer conditions due to rod bowing or presence of control rods.
Detailed models of the local phenomena are therefore being developed at ÚJV Řež using the ABAQUS 6.12 code and used to improve the predictions of the codes routinely used for the core design assessment. For example the impact of the use of the advanced pellet materials on the peak loading that the cladding will experience during the power ramp has been quantified.
Cold Water Injection and Rod Ejection Analysis of Annular Fueled PWRs by a Hybrid Lumped Parameter Model
Authors: Duarte Juliana P. , Rivero José de Jesus , Alvim Antonio Carlos M. , Piqueira José Roberto C. , Frutuoso e Melo Paulo F. F. 
Published: November 2014

Annular fuels are being studied to increase the power of advanced third-generation reactors by 50%. This paper aimed to analyze transient scenarios through a hybrid lumped parameter-finite difference model in a pressurized water reactor with annular fuel. The model used in this work is more detailed than the double lumped parameter one, but still simple enough to model some transients in PWR fuels, as rod ejection accident and cold water insertion accident. The heat transfer equations are solved by the numerical semi-implicit Crank-Nicolson method together with point kinetics equations with six groups of delayed neutrons and a lumped parameter model for the reactor coolant.
The model takes into account in an approximate way the hot spot by using a composed peaking factor equal to 2.5. The reactivity feedback is taken into account by considering the Doppler effect of fuel temperature, and also moderator temperature variation. The results were compared with solid fuel performance and showed that the annular fuel reached considerable lower fuel temperature profiles even for 150% power, as compared to 100% power for solid fuel, thus showing that this kind of fuel has a better safety performance for the transients analyzed. The rod ejection accident showed that feedback effects can lead the reactor to a new safe steady state condition.
The Implementing Geological Disposal Technology Platform: Key Challenges in Research and Development in Radioactive Waste Management
Authors: Delay Jacques , Slovak Jiri , Kowe Raymond 
Published: November 2014

The Implementing Geological Disposal of Radioactive Waste Technology Platform (IGD-TP) was launched in November 2009 to tackle the remaining research, development and demonstration (RD&D) challenges with a view to fostering the implementation of geological disposal programmes for high-level and long-lived waste in Europe.
The IGD-TP’s Vision is that “by 2025, the first geological disposal facilities for spent fuel, high-level waste and other long-lived radioactive waste will be operating safely in Europe”. Aside from most of European waste management organisations, the IGD-TP now has 110 members covering most of the RD&D actors in the field of implementing geological disposal of radioactive waste in Europe.
The IGD-TP Strategic Research Agenda (SRA), that defines shared RD&D priorities with an important cooperative added value, is used as a basis for the Euratom programme. It provides a vehicle to emphasise RD&D and networking activities that are important for establishing safety cases and fostering disposal implementation. As the IGD-TP brings together the national organisations which have a mandate to implement geological disposal and act as science providers, its SRA also ensures a balance between fundamental science, implementation-driven RD&D and technological demonstration.
The SRA is in turn supported by a Deployment Plan (DP) for the Joint Activities to be carried out by the Technology Platform with its members and participants. The Joint Activities were derived from the individual SRA Topics and prioritized and assigned a timeline for their implementation. The deployment scheme of the activities is updated on a yearly basis.
Characterization of In-Core Water Chemistry for Corrosion Control of LWRs
Authors: Saji Genn 
Published: November 2014

This paper updates scientific bases of water chemistry in applying the author’s recent theory, which integrates the elemental radiation- and electro-chemistry reactions in the “Butlar-Volmer equation,” presented in ICONE21-16525.
For the past several years the author has been trying to establish that the “long-cell” (a kin to macro-cell) corrosion mechanism is inducing practically all sorts of accelerated corrosion phenomena widely observed in water-cooled reactors, especially in aged plants.
The theoretical electrochemical potential differences have been benchmarked with the published in-pile test results for both PWR- and BWR water chemistry environments. However the author’s previous verification efforts were limited to the extent that the curves were fitted with experimental results at a single point. The author re-formulated the basic theory and found that the redox potential difference consists of an electrochemical part (e.g., Nernst equation of dissolved hydrogen or oxygen) and radiation-induced perturbation term, the latter diminishes to zero without radiation.
The author continued his studies to clarify whether our current scientific knowledge is sufficient to explain the in-core “chemistry” to reproduce the experimental results without the fitting parameter. Through his study he realized that the basic mechanism of the potential difference is still not sufficiently known.
No fitting parameter was used for the PWR water chemistry in the DH region for practical engineering applications, although it is indispensable to confirm the results with an in-pile test loop. In the BWR-NWC the theoretical redox potential out of core was still necessary to be fitted with the experimental results, due to an effect of residual hydrogen peroxide detected by the reference electrode. In addition the calculated potential shift is several times larger than the experimental observation. With the reformulation the scientific validity of the author’s theory is further confirmed. He believes that there is no doubt that the “long-cell” takes place in LWRs, although details are still debatable.
Molten Salt Fast Reactor Blanket Design and Proliferation Resistance Assessment
Authors: Bykov Valentyn , Křepel Jiři , Pautz Andreas 
Published: November 2014

Molten Salt Reactor (MSR) designs are frequently accompanied by a blanket salt. This way the irradiation of the outer reactor wall will be strongly reduced. On the other hand, the barrier between the core and blanket will undergo higher irradiation and it will be necessary to replace it several times during the reactor lifetime. Furthermore, this blanket salt will also have a positive impact on neutron economy by improving the breeding performance.
In this paper a blanket of a generic two fluid molten salt reactor utilizing fast thorium-uranium cycle was investigated. This was done by tracking the evolution of uranium, neptunium and plutonium isotopes with burnup, which was then influenced by removal of uranium from the blanket. A significant reduction in the production of minor actinides was observed.
The uranium vector removed from the core was then investigated for proliferation resistance, using NUREC proliferation resistance metric and comparison with other weapon designs. The evaluation concluded that while the presence of U-232 increases radiological hazard associated with this uranium, thereby erecting a radiological barrier, it cannot be treated as “self-protecting” based on IAEA and NRC standards, requiring 1 Sv/h at 1m dose rate. Moreover ideas on how an interested party could reduce this radiological hazard were discussed.
A Study on Weld Residual Stress Relaxation by Local Post Weld Heat Treatment for Circumferential Weld
Authors: Lee Seung-gun , Son Youngho 
Published: November 2014

Weld residual stress is a troublesome problem in nuclear power plant, because it can accelerate crack growth in weld region. For low alloy steel, Post Weld Heat Treatment (PWHT) is essentially needed to relieve residual stress and to temper the hard regions in the heat affected zone (HAZ). Local PWHT is used when it is impractical to heat the whole component in a furnace. The rules and practices of related codes and standards, such as ASME and AWS, associated with local PWHT are quite different. For example, according to ASME Section III, the minimum width of heated band at each side of the weld shall be the thickness of the weld or 2 in., whichever is less. While, according to ASME B31.1, the width of heated band shall be at least three times the wall thickness at the weld of the thickest part being joined.
In this paper, the status of the related code and standard associated with local PWHT is briefly summarized, and baseline information on local PWHT is explained based on FEA (Finite Element Analysis) results and optimized local PWHT parameter is suggested to support current code of practices.
MTC 3D: A Fully Coupled Methodology for MSLB Safety Analysis
Authors: Royère Christian , Gonnet Michel , Manchid Brahim 
Published: November 2014

The main steam line break (MSLB) is an overcooling accident that may lead to an over criticality, and so to a power increase, after the reactor trip. The most penalizing single failure is a RCCA bank stuck out of the core when the reactor trip occurs. This configuration leads to a strong asymmetry of the radial power shape combined with a strong asymmetry of the core inlet temperature that results in a strongly distorted 3D power distribution.
In the original design, the MSLB accident was studied with a simplified and conservative 0D method. The point kinetics approach requires the use of extremely conservative assumptions in order to account for the asymmetry in the core region that takes place during the transient. The use of the coupling between a three-dimensional neutronic code (SMART), a 3D core thermal-hydraulic code (FLICA cf. ref [4]) and a reactor coolant system code (MANTA cf. ref [3]) allows representing the 3D heterogeneity of the power shape and also of the resulting cross flows. In addition, this coupling allows determining moderator and Doppler feedback effects in a much more realistic way thus limiting accident consequences estimated.
A methodology, called MTC3D (for Méthode Totalement Couplée 3D in French), has been developed using the coupling between the three codes to perform the MSLB analysis. The physical dominant parameters of the transient are identified through a comprehensive sensitivity analysis. Then, a deterministic approach is used in the entire transient simulation considering dominant parameters in a penalizing way.
In a first step, neutronic data are determined with SMART calculations. In a second step, MANTA/SMART/FLICA transients are performed with penalized neutronic and thermal-hydraulic data. In a third step, as the steam line break transient is a relatively slow transient, the core power distribution is evaluated with a steady state SMART/FLICA calculation without penalization. In a last step, safety criteria, such as minimum DNBR (Departure from Nucleate Boiling Ratio) are calculated with FLICA calculations based on core power distribution calculated at the third step and boundary conditions calculated at the second step.
The use of 3D neutronic and detailed thermal-hydraulic codes to model the reactor core allows considering a more physical representation of the core configuration for transient analysis. The coupling between 3D neutronic and core thermal-hydraulic codes allows exhibiting intrinsic margins without over penalizations related to a simplified 0D method.