bluuebunny/crossref_metadata_2025_split huggingface dataset:
SCORPIO-VVER: Two Decades of Experience and Enhancements in Reactor Core Monitoring and Surveillance in Central Europe
Authors: Molnar Jozef , Pecka Marek , Kment Jaroslav 
Published: October 2017

During the years 1970–80’ in the satellite countries of the former Soviet Union more than 24 new reactor builds were started. In the former Czechoslovakia, the new builds were realized with a high degree of input from the local engineers and the local industry. This territory up to today has an indigenous nuclear industry, machinery and nuclear engineering background.
Starting with the year 1972, on the territory of Czechoslovakia (now on the territory of the Czech and Slovak Republic) 12 new units of the VVER-440 type of reactors were started to build. Nowadays the 2 oldest units were already shutdown in Slovakia, 4+4 units are still operating in both countries, and 2 units of VVER-440 model V213 are still under construction in Slovakia. The reactors designed lifetime in original configuration are 30 years.
During these units’ operation period, wide range of modifications and upgrades were performed to strengthen the nuclear safety and the reactors operability. In 2015/16, activities related to the unit operation licenses extension were carried out in both of countries.
In scope of strengthening the reactor’s core monitoring and surveillance, at Dukovany NPP (CZ, 1998) and at Bohunice NPP (SK, 2001) the original Russian VK3 computation system was completely replaced with an alternative advanced Core Monitoring and Surveillance System (CMS) SCORPIO-VVER. In Hungary, a locally developed “Verona”, and on the units under construction in Slovakia the Russian “Kruiz” CMS is used.
Nowadays the SCORPIO-VVER CMS presents a nuclear fuel type and fuel vendor independent, advanced computer based reactor core monitoring system with an open and flexible framework, including the latest achievements in the fields of N/F and T/H for reliable and safe reactor operation with high efficiency of fuel cycle. The system’s framework governing the know-how and knowledge of 5 European institutes with proven experiences with reactor operation, fuel pattern and fuel campaign design and with utilizing the existing unit’s project reserves to increase the reactor operation and fuel campaign efficiency.
Since the first installation the SCORPIO-VVER CMS system has a remarkable operating history and experience. More than 18 years of experiences at 6 units of VVER-440 type of reactors in two different countries helps to put the system to a very high level of usability and reliability. Even if the SCORPIO-VVER is installed only on VVER-440 reactors, it could be adapted to the needs of other VVER type of reactors and to needs of education and training centers too.
Investigation on Heat Transfer Characteristics in the Test Section With Non-Uniform Heat Flux Distribution Under Natural Circulation
Authors: Wang Qiang , Gao Puzhen , Chen Xianbing , Wang Zhongyi , Huang Ying 
Published: October 2017

Natural circulation served as an indispensable part of nuclear, attracted much more attentions in recent years. It does not need a pump to provide power. The operating principle of natural circulation caused its complexity in analysis process. It was still a difficult issue to reveal the law of natural circulation accurately. Many experiments and calculations had to be conducted to study the basic physical regulation. This paper concentrated upon the heat transfer characteristics in the test section with two different types of heat flux distribution. The two types of heating flux distribution in the test section were linear and chopped cosine along axial direction. Based on a natural circulation experimental facility, physical models and mathematic models were established. RELAP5 code was used to calculate the thermal hydraulic state of natural circulation loop. The variation of heat transfer coefficient along flow direction was different. It was tightly related to heat flux. Some relevant experiments were conducted in many different conditions and steady sate experimental data were achieved to verified theoretical calculations. Experimental data, such as water temperature, wall temperature and flow rate were recorded when the system is stable. The heat transfer coefficients were calculated according to the experimental data. The factors that affected the heat transfer characteristics of natural circulation were analyzed by comparing the heat transfer coefficient under different conditions. The heat transfer coefficient was calculated according to the empirical correlations as well.
After a series of analysis, the results indicated heat transfer coefficient had an obvious difference, which influenced ability of natural circulation. Comparing with experimental data, the evaluation of different empirical correlations was conducted in two test sections. Some empirical correlations turned out to be suitable for the estimation of heat transfer in experiment facility. The increase of heat flux could enhance heat transfer process in the two test section under low pressure. Average heat transfer coefficient increased with the decrease of inlet subcooling degree. The system pressure effected the heat transfer characteristics of natural circulation as well. The increase of mass flux would promote heat transfer while the level was different. RELAP5 had a great agreement with experimental data in single phase flow. Natural circulation ability was influenced by the position of average heat source center, which was slightly different in the research objects. The research would lend strong empirical support to the guideline of experiment and subsequence study in natural circulation.
Turbulent Mixing Characteristics in the Converging Region of Five-Parallel Jets
Authors: Zhuang Nailiang , Yang Bonan , Wang Lianfa , Yuan Hongsheng , Tan Sichao 
Published: October 2017

Plate-type reactor fuel is getting increasing attentions as it features excellent heat transfer ability and compact structure. Turbulence mixing accompanied with momentum and mass exchange is typical phenomena in the converging region downstream of parallel plane fuels. To deepen understanding turbulence mixing characteristics and obtain more benchmark data for CFD, an experimental study was conducted by 2D particle image velocimetry (2D-PIV). Ensemble average profile of velocity, vorticity, turbulence intensity, as well as Reynolds stress are analyzed, respectively.
Results reveal that two kinds of merging points (mp) were found, i.e., flows from two side jets merge at y/d = 3 (mp1), and flows from middle three jets merge at y/d = 7.6(mp2). The decay of vertical velocity decreases fast primarily until y/d = 6, then slightly increases until y/d = 8.8, and finally decreases gradually. Decay of velocity magnitude of present study decreases sharply until y/d = 6.4, then decreases gradually. This tendency of five-parallel jets is similar to the result of two-parallel jets. From flow visualization, the vortex scale decreases once the vortex formed. The vorticity reaches its maximum at about y/d = 1.3 and then decreases gradually. Consistently with the indicated of velocity distribution, the vorticity distribution tends to mild with the flow developing downstream. However, the location of vortex center is affected by spanwise momentum exchange. Vortices started moving closer around the mp2 indicating a strong combination of activities. Spanwise turbulence intensity shows strong fluctuations existed around mp1 implying that the main momentum transfer happened in the merging region.
Evaluation on CAP1400 Passive Containment Cooling System Capability
Authors: Di Zhang , Xin Liu , Zhe Wang , Guodong Wang , Chenxiao Ni , Shengjie Wei , Benxue Hu , Zhangli Wang 
Published: October 2017

CAP1400 is the large advanced passive demonstration plant, which is one of the National Science and Technology Major Projects. CAP1400 is an innovative design and development based on AP1000 advanced design philosophy with passive safety features. While keeping the same safety level as AP1000, CAP1400 aims to improve plant economy by means of enlargement of reactor core, increase of unit power output and optimization of plant overall design. The CAP1400 nuclear power plant uses the passive containment cooling system (PCS) to remove heat from the containment after accidents when high mass and energy fluid is released from primary or secondary systems. On account of the more mass and energy release and the internal structures design changes of CAP1400, the operating parameters may exceed the range of AP600 and AP1000 containment tests. In order to further study the performance of the CAP1400 PCS system, a series of verification tests have been carried out in the National PWR Project.
This paper describes the background of the test setup, the really focused issues of the experiment and the main test results. The experimental data from separate effect tests and large scaled test were analyzed and evaluated in depth. The performance of the PCS was studied, including water film coverage outside the containment shell, water film evaporation and heat transfer characteristics, air natural circulation and convection outside the containment vessel, convection and condensation of the air / steam mixture on the inner surface of the containment vessel, as well as the overall performance of PCS systems. The results show that the important correlations applied in CAP1400 containment pressure and temperature response analysis could predict the heat and mass transfer reasonably and the envelope factors used for safety analysis are conservative. Parameters affecting the containment pressure and temperature have been investigated, and the reasonability and conservation of containment thermal and hydraulic analysis code has been evaluated by comparing the calculated results with the measured results. The results show that containment analysis code could simulate the important heat transfer phenomena of the test containment well and predict the temperature and pressure response reasonably with some conservation.
Containment pressure and temperature response analysis after typical design basis accident analysis have been preformed, and the results show that the CAP1400 PCS is sufficient to remove the heat from the containment which ensure the safety of the plant. All of the results provide firm basis for the development of CAP1400 and sufficient support for the safety review. This work also lays the foundation for the further development of relevant analytical tools and the development of advanced and passive technologies with higher power levels.
The Numerical Simulation of the Detachment Characteristic of Bubble in Narrow Channels With Different Width
Authors: Jie Yang , Wang Jianjun , Gao Puzhen , Ying Huang 
Published: October 2017

The plate-type fuel element is widely used in ship-based nuclear reactors. The typical coolant channel in such reactor is a narrow rectangular one with a relatively large aspect ratio. The thermal-hydraulic characteristics in narrow rectangular channels may be different from that in conventional pipe. It could be expected that the geometry of narrow channels has some impact on the bubble behaviors in subcooled flow boiling.
In this paper, the bubble behaviors were simulated by using CFD software FLUENT in narrow rectangular channel with different widths under both stagnation conditions and flow conditions. The main concern was placed on the parameters of bubble departure time, departure diameter and departure velocity.
In the model setup process, the VOF model and PISO algorithm were selected. The results were compared against the experimental data and showed the models good capability of capturing the interface between gas and liquid. Furthermore, for stagnation cases, the width of the channel played an important role for bubble departure diameter. In flow cases, the width also had a great impact on departure diameter and departure velocity as well. The departure time and departure velocity showed strong relevance with each other under different contact angle conditions.
Fundamental R&D Work on Methods for State Monitoring of Transport and Storage Containers for Spent Fuel and Heat-Generating High-Level Radioactive Waste on Prolonged Intermediate Storage
Authors: Fiß Daniel , Schmidt Sebastian , Reinicke Sebastian , Kratzsch Alexander 
Published: October 2017

The continuing search for a long-term storage for highly-active nuclear waste in Germany requires a prolonged intermediate storage period of spent fuel in dry storage casks at the power plant sites. Currently, it is not sufficiently clear if there might be a loss of integrity of the fuel rods within such long periods, e.g. due to rising pressure from gaseous products of nuclear decay. Regarding a final evaluation, extrapolative modelling of the radiochemical and thermomechanical material behavior is challenging and not suitable for predictions on the condition of storage container inventory after the intermediate storage period. Therefore, it is of public interest to find measurement principles or methods which can provide information about the condition of the storage container inventory.
In line with a cooperative project (project partners: Technical University Dresden, Zittau/Görlitz University of Applied Sciences) different measurement principles and methods (radiation emission, muon transmission, thermography, acoustical spectrometry) for non-invasive condition monitoring of the storage container inventory in case of prolonged intermediate storage are going to be investigated and evaluated. The results shall help to determine suitable methods for the identification of both changes of the spent fuel and inner container structure over long periods without opening the container and would be a significant contribution for the long-term safety of intermediately stored highly radioactive waste. Furthermore, suitable methods would provide information about the transport and conditioning ability of the waste before transfer to the repository.
This paper deals with the content of the subproject of Zittau/Görlitz University of Applied Sciences as well as with the approach for project realization. A further main part of this paper is the development of experimental infrastructure to support the investigations.
Study on the Application of CDAG Methodology in EPR Nuclear Power Plant
Authors: Yuhao Yin , Yichao Huang , Feng Zhao 
Published: October 2017

The Westinghouse Owners Group Core Damage Assessment Guidance (CDAG), which has been authorized by the NRC staffs, is now used by licensee emergency response organization staff for estimating the extent of core damage that may have occurred during an accident at a Westinghouse nuclear power plant. On the other hand, EPR is a 3rd generation nuclear power plant, which applies the advanced European nuclear power technology.
This paper introduced Core Damage Assessment Guidance methodology in detail. The CDAG methodology is then attempted to apply to the EPR nuclear power plant. Detailed calculations have been performed for the setpoints of containment radiation monitors (CRM) and core exit thermocouples (CETs) with EPR design characteristics, which are the two main methods for estimation core damage amount. This paper also focuses the discussion on the reasons of difference of core damage estimating results between CRM method and CETs method; based on the discussion, several advices are provided when the two methods show a reasonable discrepancy in conclusions.
Several conclusions can be made from the discussions in this article that 1)the Westinghouse Owners Group CDAG methodology proved to be reasonable when applied to EPR power plant for core damage assessment under severe accident; 2) the CDAG methodology which reflect the latest understanding of fission product behavior, is very simple and timely for core damage assessment based on NPP (nuclear power plant) real-time parameters; 3) conservative calculation results of setpoints on CRM and CETs based on EPR design show a reasonable trend and range; 4) it is concluded that several factors such as the releasing way, RCS fission product retention, fuel burnups might have great impact on the estimating results, when the results from two main indications (CRM and CETs) show an unexpected response.
An Improved Algorithm for Potential Accident Consequence Assessment at NPPs Based on PAVAN
Authors: Qian Chen , Liang Zhai , Yuhao Yin , Yunliang Tao 
Published: October 2017

PAVAN is an atmospheric dispersion program for evaluating design basis accidental releases of radioactive materials from nuclear power plant. It was developed by Pacific Northwest Laboratory on the basic of the atmospheric dispersion models described in RG 1.145 by NRC (U.S. Nuclear Regulatory Commission). Using the joint frequency of wind direction, wind speed and atmospheric stability, the atmospheric relative concentration values for the exclusion area boundary and outer LPZ boundary of nuclear power plant are calculated and given by the program. Once the program was introduced, it has been widely used in the radioactive accident consequence assessment, especially in the FSAR (Final Safety Analysis Report) and Report of EIA (Environmental Impact Assessment) of NPPs in China.
The theory basis and general method of PAVAN is introduced in this paper. And specialty of the X/Q points based on joint frequency data is discussed. The envelope algorithm of PAVAN is also introduced and discussed.
The paper presents an improved algorithm based on PAVAN which uses the hourly meteorological data as input instead of joint frequency data. In this algorithm, the size of X/Q points is related to the quantity of the hourly meteorological data. When the quantity is large enough, e.g. 17520 sets of hourly meteorological data in two years, the envelope curve for X/Q points fit more exactly than PAVAN.
Using the observed meteorological data, the improved algorithm is compared with PAVAN. The result proves that the former is more accurate. In general, the improved algorithm is relatively conservative. In some situation, the conservativeness is not certain. The factors which result in the uncertainty are deeply discussed.
Further optimized are performed by the algorithm. The number of points to seek in envelope curve fitting is set to be dynamic and be a quarter of total number of X/Q points to be fitted. The result shows that increasing the number of points to seek in the iteration process of envelope curve fitting will lead to more conservative X/Q values.
Additionally, the optimized algorithm provides X/Q value of 50% probability level for overall site. The value is not relatively conservative. From the standpoint of statistical probability, it is more realistic and is acceptable for potential accident consequence assessment. Especially, when X/Q value of 95% probability level for overall site is too conservative to accept, the value of 50% probability level can be used to replace the conservative value.
The Need for Integrated Cybersecurity and Safety Training
Authors: Gupta Deeksha , Bajramovic Edita , Hoppe Holger , Ciriello Antonio 
Published: October 2017

Companies involved in the nuclear energy domain, like component and platform manufacturers, system integrators and utilities, have well established yearly trainings on Nuclear Safety Culture. These trainings are typically covered as part of the annual quality assurance-related refresher trainings, introductory courses for new employees, or indoctrinations of temporary staff. Gradually, security awareness trainings are also addressed on a regular basis, typically with a focus on IT, the daily office work, test bay or construction site work environment, and some data protection and privacy-related topics. Due to emerging national nuclear regulation, steadily but surely, specialized cybersecurity trainings are foreseen for integrators and utilities.
Beyond these safety, physical security and cybersecurity specific trainings, there is a need to address the joint part of these disciplines, starting from the planning phase of a new Nuclear Power Plant (NPP). The engineers working on safety, physical protection and cybersecurity, must be aware of these interrelations to jointly elaborate a robust I&C architecture (defense-in-depth, design basis events, functional categorization and systems classification) and a resilient security architecture (security by design, security grading, zone model or infrastructure domain, security conduits, forensic readiness, Security Information and Event Management).
This paper provides more in-depth justification of when and where additional training is needed, due to the ubiquitous deployment of digital technology in new NPPs. Additionally, for existing NPPs, the benefits of conveying knowledge by training on specific interfaces between the involved disciplines, will be discussed.
Furthermore, the paper will address the need of focused training of management stakeholders, as eventually, they must agree on the residual risk. The decision-makers are in charge of facilitating the inter-disciplinary cooperation in parallel to the allocation of resources, e.g. on security certifications of products, extended modeling-based safety and security analyses and security testing coverage.
Implementation of Fast Searching Method for the Cross-Section Usage in HTGR Simulators
Authors: Auve Benjamin , Wei Chunlin , Sui Zhe , Sun Jun 
Published: October 2017

The modular High-Temperature Gas-cooled Reactor (HTGR) is one of the six generation IV advanced nuclear reactors. With the final purpose of operator training and licensing, the engineering simulation system (ESS) has been studied to model the pebble-bed type reactor core and has been successfully implemented into the full scope simulator of HTR-PM.
As stated in corresponding industrial standards, one important feature of the nuclear power plant simulator is real-time calculation, and the other one is simulation results with high fidelity (compared to design parameters or operational data in different stages). In ESS, each macro cross-section was in the form of polynomial by function of several variables (like burn-up, buckling, temperatures), the expression of which was finalized by multivariate regression analysis from large scattered database generated by the VSOP. Since the polynomial is explicit and prepared in advance, the macro cross-sections are quickly calculated in running ESS. However, some variables (such as temperature) in HTGR are in larger scope so that the polynomial is not easy to meet full range accuracy. One normal idea is to optimize the expression of polynomial, while another means was proposed and tested in present paper.
Other than focusing on the polynomials, a new method, called the fast searching, was described to significantly improve the accuracy of macro cross-section calculation while it was also fast to maintain the real-time feature. Instead of setting up a regression polynomial from the large cross-section database, the fast searching method treated the database as scatted points in the multi-dimension space, and aimed to locate the target position of unknown macro cross-section by fast searching and interpolating. Searching was to find the neighbouring database points around the target point in the multi-dimension space, which naturally improved the accuracy. While interpolating was to predict the macro cross-section of target point based on those neighbouring database points. To keep the searching and interpolating fast, the original database of macro cross-sections was analysed. A series of searching and interpolating methods have been described, programmed, tested and compared to find appropriate methods to calculate all the macro cross-sections in limited time cost. Finally, the fast searching method and its program was implemented into ESS to show better performances.