bluuebunny/crossref_metadata_2025_split huggingface dataset:
Moisture Separation for Nuclear Turbine Plant Efficiency and Reliability Improvement
Authors: Takahashi Ryota , Fujita Issaku , Nakamura Taichi , Kasahara Jiro , Yoshii Taichi , Inoue Naoki , Nishida Keigo 
Published: October 2016

Feasibility of Rankine cycle improvement for LWR power plant is studied. In LWR, the main steam is saturated steam, thus steam turbine is mostly operated in wet region. Therefore, the moisture separation plays a key role to achieve better turbine plant efficiency. Typical reheating configuration for present LWR plant is single reheating after HPT by MSR.
On the other hand, it is well known that the increasing of numbers of moisture separation or reheat improves the turbine plant efficiency. Double reheat cycle is common for fossil power plant and also has been proposed for LWR. However, double reheat cycle has not been adopted to actual nuclear power plant yet since the double reheat configuration makes heating and drain system complicated, and larger MSR shell and diameter piping are required due to the increase of steam volume flow.
In this paper, improved Rankine cycle with high pressure moisture separator (HP-MS) and improved low pressure loss MSR (LP-MSR) is studied in order to increase the plant efficiency. Using CFD analysis and scale model test, we confirmed pressure loss and moisture separation characteristic of LP-MSR. The performance of improved Rankine cycle with this new MSR was calculated and the results show that the electrical output of typical LWR Plant may increase 1% or more from single reheat cycle.
Qualification of Water-Filled Tank due to Tornado Missile Impact
Authors: Schmitt Timothy A. , Al-Shawaf Taha , Rahman Syed M. , Cristobal Angelo 
Published: October 2016

As part of the post-Fukushima accident scenario, the qualification of a water-filled cylindrical tank subjected to tornado missile impact was required to ensure the availability of water inventory in the tank to mitigate the post-accident effects.
Most of the classical tornado missile impact analysis and design involves using empirical formulas that have been developed based on tests. It is recognized that water backed structures provide additional resistance to perforation of the missile due to the mass and properties of the water. Therefore, a finite element analysis was used to qualify the tank for two controlling postulated missiles, namely, 2 ½” diameter schedule 40 pipe and bolted wood decking.
The location on the tank for the missile impact, angle of impact and orientation of the missile were selected to develop the most critical response. The analysis was performed using the LS-DYNA computer program. The true stress-strain material properties were used for both the tank material and the missile types. These material properties were given a bilinear elastic-plastic curve.
It was determined that, even if an impact at the thinnest section at the top of the tank occurs and the missile penetrates, the remaining inventory of water in the tank will be sufficient to mitigate the needs for a post-Fukushima scenario. Impact on the lower elevation of the tank was investigated for any potential failure or tearing of the tank wall. The maximum of equivalent (Von Mises) stress, shear stress, and plastic strain were calculated. The results show that these values are less than the limiting values with additional available margin. Consequently the analysis shows that the tank will survive a hit in the lower portions, and the water inventory of the tank is sufficient to mitigate the effect of a post-Fukushima scenario should a missile penetrate the thinner, upper section of the tank.
AP1000® Submergence Testing of Service Level I Protective Coatings
Authors: Beenken Matthew R. , Rocco Christina L. , Andreychek Timothy S. 
Published: October 2016

A portion of the Service Level (SL) I coating systems inside the AP1000®1 containment become submerged during a design basis loss-of-coolant accident (LOCA) as a result of containment flood-up. The Design Basis Accident (DBA) qualification of SL I coatings for the AP1000 did not initially include post-DBA submergence conditions.
Submergence testing was performed using a standard test (Reference 2). Test articles were carbon steel coupons coated with one of two coating SL I coating systems; an untopcoated inorganic zinc system (Carboline Carbozinc® 11 HSN2) and a system consisting of inorganic zinc and an epoxy (Carboline Carboguard® 890N3) topcoat. Half of the inorganic zinc samples tested were irradiated to a total accumulated dose of approximately 1×109 rads prior to submergence testing.
Autoclaves were used to simulate the post-LOCA environment inside of the AP1000 containment. The working fluid was a boric acid solution buffered with trisodium phosphate (TSP) to a pH of approximately 7.8. The pressure and fluid temperature inside of the autoclaves was regulated to begin at ambient conditions followed by a pressurization and heat up following pressure and temperature conditions calculated for a cold leg break LOCA for the AP1000 plant, plus an added 10% margin, and the slowly decrease over the 30 day test period to saturation temperature at about 20 psia, simulating a cool-down of the AP1000 plant. Coolant samples of the autoclave inventory were taken and evaluated for dissolved chemical species at specific time intervals during the test.
Following exposure to the submerged test environment, each coupon was visually inspected immediately following removal from the autoclaves and was inspected again several days after completion of the test cycle for signs of blistering, rusting, lifting, peeling, discoloration, softening, etching, wrinkling, cracking, swelling, dissolving, delamination, and changes in gloss relative to their observed pre-test condition.
Investigations About Released Debris and Corrosion Products and Their Impact to Head Loss Build Up and Decay Heat Removal Under Water-Chemical PWR-LOCA Conditions
Authors: Renger Stefan , Alt Sören , Kästner Wolfgang , Seeliger André 
Published: October 2016

Investigations about the release, transportation and deposit of fibrous insulation material (FIM), corrosion products as well as resulting compounds and debris mixtures become more important to reactor safety research, when considering long-term behavior of emergency core cooling systems (ECCS) during loss of coolant accidents (LOCA). Debris released by a leakage jet leads to head loss buildups at the sump strainers, the debris filters and the spacers of fuel assemblies. However, these complex processes may influence the decay heat transfer out of the reactor core.
A similar but newer scenario implies that the boric acid coolant in pressurized water reactors (PWR) can support corrosion processes at hot-dip galvanized installations, leading to a significant release of ionic zinc into the coolant during the sump recirculation phase. A long-term change of chemical properties of the coolant (e.g. pH, zinc ion concentration) has to be considered in safety analyses. Chemical analyses showed that the solubility of zinc in boric acid coolant is inversely proportional to the coolant temperature. Consequently, zinc ions can be dissolved at lower temperatures in the containment sump. Precipitations of zinc borate (ZBP) are possible at hot spots in the reactor core. The ZBP can be formed as a flocculent disperse phase in the coolant or as solidified layers at hot fuel rod surfaces. Layer spalling could lead to the release of further solid particles into coolant flow inside the reactor core.
In several joint research projects between the Zittau/Görlitz University of Applied Sciences (HSZG) and the Helmholtz-Zentrum Dresden-Rossendorf (HZDR) investigations on the impact of these processes to the head loss buildup and the heat transfer from core were done at laboratory and semi-technical scaled test facilities. Generic experiments showed the formation of ZBP in heating rod configurations. The ZBP may remain in the core structures or can be transported on debris filter cakes in upstream and downstream components of ECCS and influence the head loss. After this, research addressed the systematical clarification of physico-chemical mechanisms and their influence on thermal-hydraulic-dynamic processes occurring as a consequence of flow induced corrosion effects during LOCA.
This paper includes a description of the most important involved test facilities, applied measuring techniques, an overview of boundary conditions considered experimentally and selected results.
Utilization of Thorium in LWR Fuels Aiming at Thermal Conductivity Improvements
Authors: Kubáň Jan , Škoda Radek 
Published: October 2016

One of the main drawbacks of uranium dioxide, which is used in almost all nuclear power reactors, is its low thermal conductivity. As a consequence, temperature at the center of fuel pellet is relatively high, because heat is poorly conducted away. To reach a higher level of safety, maximal temperature in any fuel pellet is one of the main limiting parameters, which restrict the fuel thermal output. This paper deals with the use of thorium in LWR fuels with the objective of fuel pellet maximal temperature reduction.
Research investigating homogenous distribution of thorium dioxide (thoria) in uranium dioxide fuel has already been done and did not lead to considerable thermal conductivity improvements. The aim of this study is to investigate heterogeneous distribution of thorium in commonly used uranium dioxide fuel in the form of uranium and thorium pellets placed together.
Hafnium in Nuclear Fuel Cladding Used Both as Oxidation Protection and Burnable Absorber
Authors: Závorka Jiři , Škoda Radek 
Published: October 2016

The problem with higher nuclear fuel enrichment is its high initial reactivity. It has a negative effect on the peaking factor, which is one of the license conditions. The second major problem is the ability to control the reactivity of the reactor, and thereby maintaining the multiplication factor in the core equal to 1. Long-term control of the reactivity in PWR reactors is typically conducted by the concentrated boric acid (H3BO3) in the coolant; its highest possible concentration is determined by the requirement to maintain a negative reactivity coefficient. Another option are burnable absorbers.
This work deals with usage of hafnium as an advanced type of burnable absorber. Based on the model of computing code UWB1 for the study of burnable absorbers, a new cladding of nuclear fuel is designed with a thin protective layer made of hafnium. This cladding is used as a burnable absorber that helps reducing excess of fuel reactivity and prolongs the life of the fuel assemblies, which increases economic coefficient of the use of nuclear power plants. This cladding would also work as a protective layer increasing endurance and safety of nuclear power plants. Today zirconium alloys are exclusively used for this purpose. The main disadvantages of zirconium alloys include rapid high temperature oxidation of zirconium — a highly exothermic reaction between zirconium and water steam at temperatures above 800 °C. During this reaction hydrogen and inconsiderable amount of heat are released. Hydrogen excess, released heat, and damaged cover of fuel may deepen the severity and consequences of possible accidents. Another disadvantage of zirconium alloys is their gradual oxidation under standard operating conditions and ZrH formation, which leads to cladding embrittlement.
Investigation of Thermal Efficiency of Generic Pressure-Channel Reactors With Steam Reheat
Authors: Zhou Yifeng , Ponomaryov Paul , Mazza Cristina , Pioro Igor 
Published: October 2016

Currently, i.e., in 2016, 4361 nuclear-power reactors operate in the world. 96.6% of these reactors are water-cooled (373 reactors (280 PWRs, 78 BWRs and 15 LGRs are cooled with light water and 48 reactors — PHWRs are cooled with heavy water. 15% of all water-cooled reactors are pressure-channel or pressure-tube design, the rest — pressure-vessel design. All current NPPs with water-cooled reactors have relatively low thermal efficiencies within 30–36% compared to that of current NPPs with AGRs (42%) and SFR (40%) and compared to that of modern advanced thermal power plants: combined-cycle plants (up to 62%) and supercritical-pressure coal-fired plants (up to 55%).
Therefore, it is very important to propose ways of improvement of thermal efficiency for this largest group of nuclear-power reactors. It should be noted that among six Generation-IV nuclear-reactor concepts one concept is a SCWR, which might reach thermal efficiencies within the range of 45–50% and even beyond. However, this concept has been never tested, and the most difficult problem on the way of implementation of this type of reactor is the reliability of materials at supercritical pressures and temperatures, very aggressive reactor coolant – supercritical water, and high neutron flux. Up till now, no experiments on behavior of various core materials at these conditions have been reported so far in the open literature.
As an interim way of thermal-efficiency improvement for water-cooled NPPs nuclear steam reheat can be considered. However, this way is more appropriate only for pressure-channel reactors, for example, CANDU-type or PHWRs. Moreover, in the 60’s and 70’s, Russia, the USA and some other countries have developed and implemented the nuclear steam reheat in subcritical-pressure experimental boiling reactors.
Therefore, an objective of the current paper is to summarize this experience and to estimate effect of a number of parameters on thermal efficiencies of a generic pressure-channel reactors with nuclear steam reheat. For this purpose the DE-TOP program has been used.
Strategies to Prevent Benign Transients From Becoming Serious Accidents
Authors: Miranda Samuel 
Published: October 2016

Nuclear safety criteria are based upon the concept that plant situations that are expected to have a high frequency of occurrence must not pose a danger to the public, and that plant situations that pose the greatest danger to the public must be limited to situations that have a very low expected frequency of occurrence. This concept is implemented by grouping plant situations (or events) into categories that are defined according to their expected frequencies of occurrence. It is important not to allow the boundaries of these categories to be crossed. Plant designs cannot allow low-consequence events of high expected frequency to develop into events of more serious, high-consequence categories.
The development of this system of frequency-based categorization is discussed, followed by an evaluation of various methods that have been proposed and applied to show how minor events could be prevented from becoming major events.
Scaling Analysis and Facility Design for Stability Investigation During Accidents in a PWR-Type SMR
Authors: Yan Yikuan , Shi Shanbin , Ishii Mamoru 
Published: October 2016

Small modular reactor (SMR) concept has been developed as one of the key solutions for the growing demand of safe and clean energy. SMR designs can be applied extensively in areas such as sea water desalination and small-scale power generation etc. Unlike conventional light water reactors, most SMRs greatly simplify the structure of reactor pressure vessel, usually eliminate pumps and use natural circulation to cool down the core and transfer energy. However, flow instability may easily occur and affect the entire two phase natural circulation, which is of great importance for the start-up and normal operation process of BWR-type SMRs. For PWR-type SMRs, two-phase natural circulation could exist during accidents such as small break loss of coolant accident (SBLOCA) and loss of heat sink. Current research aims to experimentally investigate potential flow instabilities related to natural circulation for a PWR-type SMR during the accidents. For current research, the NuScale reactor design is selected as the research prototype.
In this paper, the design and scaling analysis of a scaled PWR-type experimental facility are provided. In order to experimentally study the natural circulation behavior of PWR-type SMR during accidental scenarios, detailed scaling analyses are necessary to ensure that the scaled phenomena could be obtained in a laboratory test facility. A three-level scaling method is used to get the scaling ratios derived from various non-dimensional numbers. An ideally scaled facility is first accomplished based on derived scaling ratios. RELAP5 simulations of both steady state and transient cases for the ideally scaled facility are performed and compared to the prototype to ensure the accuracy of the scaling analysis. Then the ideally scaled facility is modified under engineering considerations and an engineering scaled facility is designed. Similar RELAP5 analyses are performed on the engineering scaled facility and the results match well with those in the prototype and ideally scaled facility.
Study on Thermal Efficiency of SuperCritical Water NPPs
Authors: Mazza Cristina , Ponomaryov Paul , Zhou Yifeng , Pioro Igor 
Published: October 2016

As the demand for emission-free energy increases, the continued improvement of Nuclear Power Plants (NPPs) and their thermal efficiencies is crucial to fulfilling that demand. Current NPPs, especially, with water-cooled reactors, have significantly lower thermal efficiencies (32–36%) compared to those of modern advanced thermal power plants (55–62%). Even Generation-III+ water-cooled NPPs will have thermal efficiencies not higher than 37–38%. Therefore, to be competitive on the energy market, new nuclear reactors and NPPs, so-called, Generation-IV concepts, should be designed and commissioned.
The paper discusses the vital role that thermal efficiency plays with respect to how far nuclear reactors can be more cost efficient and competitive. An evaluation of thermal efficiencies has been carried out for SuperCritical Water (SCW) NPPs with Rankine “steam”-turbine power cycle using the IAEA DEsalination Thermodynamic Optimization Program (DE-TOP). Various options for improving thermal efficiencies of SCW NPPs have been studied. This study was performed in support of possible designs of the first experimental SCW reactors.
GRAPE: Graphical RELAP/SCDAPSIM Analysis Platform for Education and Engineering
Authors: Tatsumi Masahiro , Tsujita Kosuke , Sato Kotaro , Tamari Yohei 
Published: October 2016

This paper describes development of GRAPE, a plant simulator for introductory education of nuclear engineering and nuclear safety. GRAPE is not a full-featured real-time simulator but a kind of scenario-based analysis and visualization platform using a best-estimate code such as RELAP/SCDAPSIM. The PWR and BWR plant models which can be adopted to the scenarios under severe accident conditions are also developed and maintained.
GRAPE is designed and implemented by taking advantage of recent technologies of the Web such as HTML5, CSS and JavaScript in combination with the modern software architecture. Intuitive graphical user interface was established to meet user needs while maintaining flexibility, extensibility and maintainability. It is also effective for study and research projects as visualization and/or sensitivity platforms since GRAPE can import existing RELAP-based calculation results using a converting utility program to visualize the calculation results. A macro language is available to automate the workflow in sensitivity analysis and make the process more efficient.
Thermal Mixing of Liquid Metal Coolant in an Annular Channel
Authors: Kashinsky O. N. , Lobanov P. D. , Pribaturin N. A. , Kurdyumov A. S. , Vorobyev M. A. 
Published: October 2016

Results of experimental studies of mixing of liquids with different temperatures are presented. Experiments were performed in a T-junction connected to an annular channel. Rose’s metal was used as the test liquid. The test section was made of thin wall stainless tubes. The distribution of wall temperatures was measured using a high speed infrared camera. Experiments were performed in transient regimes.
Results obtained provide experimental data necessary for validation of thermal hydraulic codes for heavy liquid metals.
Implementation of Small Punch Testing and Automated Ball Indentation in the Process of Irradiated NPP Materials Degradation Evaluation
Authors: Kopriva Radim , Eliasova Ivana , Kytka Milos 
Published: October 2016

In the terms of nuclear power plant operational life management, current trend of components lifetime extension requires precise and credible information of structural material degradation. Present-day standard conventional methods of mechanical testing are usually based on the use of large specimens and higher consumption of testing material, whose availability and volume is often limited.
For determination of material properties, sampling of the necessary volume of material is in most cases connected with affecting the integrity or even destruction of the assessed component. Moreover, several components are not usually covered surveillance programs, e.g. reactor pressure vessel internals.
Innovative testing methods of Small Punch Testing (SPT) and Automated Ball Indentation Test (ABIT) are based on the determination of material properties from miniaturized testing specimens and their semi-destructive approach is very promising for the possibility of present data base of irradiated materials testing results enlargement and enable the option of component in-situ testing (ABI testing).
Paper describes the process of implementation of Small Punch Testing and Automated Ball Indentation Testing techniques in the evaluation of mechanical properties of irradiated NPP structural materials degradation. Presentation depicts all necessary steps carried out at the Hot Cell facility of UJV Rez, a. s. (Integrity and Technical Engineering Division) for the employment of these innovative testing techniques to the portfolio of accredited mechanical tests performed within the frame of surveillance programs of WWER type reactors — e.g. preliminary design, modification of the testing equipment, fractography analyses etc.
Detailed comparison between the results of standard conventional testing techniques and SPT and ABIT for both initial unirradiated and irradiated state for the material A533B (JRQ) is included for determination of suitability of used methods for testing of materials within the surveillance programs of WWER type reactors.