bluuebunny/crossref_metadata_2025_split huggingface dataset:
Study on Neutronics and Thermalhydraulics Characteristics of 1200-MWel Pressure-Channel SuperCritical Water-Cooled Reactor (SCWR)
Authors: Miletic Marija , Peiman Wargha , Farah Amjad , Samuel Jeffrey , Dragunov Alexey 
Published: November 2014

Nuclear power becomes more and more important in many countries worldwide as a basis for current and future electrical-energy generation. The largest group of operating Nuclear Power Plants (NPPs) equipped with water-cooled reactors (96% of all NPPs) have gross thermal efficiencies ranging from 30% and up to 36%. Such relatively low values of thermal efficiencies are due to lower pressures/temperatures at the inlet to a turbine (4.5–7.8 MPa / 257–293°C). However, modern combined-cycle power plants (Brayton gas-turbine cycle and subcritical-pressure steam Rankine cycle, fuel – natural gas) and supercritical-pressure coal-fired power plants have reached gross thermal efficiencies of 62% and 55%, respectively. Therefore, next generation or Generation IV NPPs with water-cooled reactors should have thermal efficiencies as close as possible to those of modern thermal power plants.
A significant increase in thermal efficiencies of water-cooled NPPs can be possible only due to increasing turbine inlet parameters above the critical point of water, i.e., SuperCritical Water-cooled Reactors (SCWRs) have to be designed. This path of the thermal-efficiency increasing is considered as a conventional way through which coal-fired power plants gone more than 50 years ago.
Therefore, an objective of the current paper is a study on neutronics and thermalhydraulics characteristics of a generic 1200-MWel Pressure-Channel (PCh) SCWR.
Standard neutronics codes DRAGON and DONJON have been coupled with a new thermalhydraulic code developed based on the latest empirical heat-transfer correlation, which allowed for more accurate estimation of basic characteristics of a PCh SCWR. In addition, the CFD Fluent code has been used for better understanding of specifics of heat transfer in supercritical water.
Future studies will be dedicated to materials and fuels testing in an in-pile supercritical-water loop and developing passive-safety systems.
Preliminary Analyses of Single Droplet Evaporation and Movement in the Steam Generator of HTR-PM
Authors: Xu Wei , Shi Lei , Zheng Yanhua , Liu Peng , He Liang 
Published: November 2014

Water ingress accident is one of the most severe accidents which must be analyzed in high temperature gas-cooled reactor pebble-bed modular (HTR-PM). The droplet could enter the primary circuit under the design basis accident of a double-ended guillotine break of a heat transfer tube. This paper simulates the behavior of single droplet evaporation and movement in the steam generator by numerical methods. Based on the structure characteristics of steam generator, the life time of droplet and the distance that the single droplet could move have been analyzed. The important parameters such as the droplet diameter, helium temperature, helium pressure and helium velocity which have an influence on the behavior of droplet evaporation and movement have also been discussed in detail.
The preliminary numerical simulation results indicate that the droplet diameter, helium velocity and helium temperature play an important role in the life time of droplet in the accident situation in the primary circuit. Helium pressure has a little effect on droplet evaporation in practical situation. The numerical simulation results demonstrate that only certain droplets with a diameter in certain range could arrive to the bottom of the steam generator pressure vessel (SGPV) and enter into the steam generator annular channel after collision with the bottom of the SGPV. The distance that the single droplet could move in the primary circuit is decided by a various complex factors such as the structure of the primary circuit, the droplet diameter and helium velocity. The preliminary analyses indicate that there is little probability for the single droplet to enter into the reactor core of the HTR-PM.
The CHF Enhancement Combining Nano-Fluid and Honeycomb Porous Plate in a Saturate Pool Boiling
Authors: Mori Shoji , Aznam Suazlan Mt , Okuyama Kunito 
Published: November 2014

Several studies have proposed the use of nanofluids to enhance the in-vessel retention (IVR) capability in the severe accident management strategy implemented at certain light-water reactors. Systems using nanofluids for IVR must be applicable to large-scale systems, i.e., infinite heated surfaces. However, the effect of the size of heater with nanoparticle deposition was revealed that the CHF is decreased with the increased heater size. On the other hand, the CHF using a honeycomb porous plate was shown experimentally to be more than twice that of a plain surface with a heated surface diameter of 30 mm, which is comparatively large compared to 10 mm. This enhancement is resulted from the capillary supply of liquid onto the heated surface and the release of vapor generated through the channels.
In the present paper, in order to enhance the CHF of a large heated surface, the effects of a honeycomb porous plate and a nanofluid on the CHF were investigated experimentally. As a result, the CHF was enhanced greatly by the attachment of a honeycomb porous plate to the modified heated surface by nanoparticle deposition, even in the case of a large heated surface.
Influence of Passive Safety Systems on a Cladding Temperature Condition of Fuel Rod Simulators
Authors: Melikhov Oleg I. , Elkin Ilya V. , Melikhov Vladimir I. , Nikonov Sergey M. , Pylev Sergey S. , Zaitsev Sergey I. , Zakutaev Maxim O. , Dorofeev Dmitry I. 
Published: November 2014

In 2012 on the large-scale thermalphysic PSB-VVER facility modelling of three different type accidents is executed:
- Guillotine rupture of the pipeline on a reactor inlet;
- A small leak from “the cold” pipeline;
- Rupture of pressurizer surge line.
In experiments work of passive safety systems of new projects of NPP with RU VVER in the conditions of loss of all sources of an alternating current was modelled.
The purpose of experiments was research of influence of new passive safety systems on a temperature condition of a fuel cladding. On the basis of the received experimental data verification of the Russian system codes TRAP-KS, KORSAR/GP and SOCRAT is executed.
In paper the basic results of researches are presented.
Experimental Research on Pulsating Two-Phase Flow Pressure Drop Characteristics in Horizontal Rectangular Channel
Authors: Zhang Siqi , Gao Puzhen 
Published: November 2014

In spite of most previous studies since 1970, the theory of pulsating pipe flows supported by experimental investigations has not yet completed in comparison with the well-defined theory of steady pipe flows. Therefore, it seems that there is much to be done about experimental research in this field. In order to determine the resistance characteristics of two-phase flow under pulsatile conditions, an experimental investigation on two-phase flow with periodically fluctuating flow rates in a narrow rectangular channel is carried out. A frequency inverter is used to obtain experimental conditions with different fluctuating frequencies, amplitudes and mean values of water mass flow rate. After obtaining experimental results, comparisons between experimental frictional pressure drop values and theoretical calculations have been done. Two-phase flow on pulsating conditions is far more complicated than that on steady conditions because pulsating flow is composed of two parts: a steady component and a superimposed periodical time varying component called oscillation. In this paper, the influence of different fluctuating frequencies, amplitudes and mean values of liquid and gas mass flow rate on two-phase flow pressure drop characteristics is also discussed.
The results show that the total pressure drop and water mass flow rate change with the same fluctuating period except for a phase difference. The phase lag also changes with the fluctuating frequencies and amplitude. The accelerating pressure drop changes dramatically in a fluctuating period, especially at the end of acceleration. Also, the time when the acceleration pressure drop has its maximum value lags the time when the acceleration reaches its peak, mainly because of the inertial of the fluid.
Transient Validation of Areva’s New Arcadia™ Code System
Authors: Kuch Sebastian , Leberig Mario , Brock Richard , Reiterer Florian , Riedmann Michael , Rooks Kenneth 
Published: November 2014

AREVA has developed a new leading edge code suite to meet the challenges arising from increasing expectations in nuclear power plant availability and fuel performance while satisfying stricter safety requirements. ARCADIA™ [1] is an advanced 3D coupled neutronics/thermal-hydraulics/thermal-mechanics code system for Light Water Reactor (LWR) fuel assembly and core design calculations as well as safety analysis, using a new software architecture allowing for nodal and pin-by-pin calculation capability. ARCADIA™ was licensed by the US Nuclear Regulatory Commission (NRC) for applications for PWR UO2 cores in 2013. It is on the way to be licensed in other countries for AREVA customers.
ARCADIA™ contains the steady-state and transient core-simulator ARTEMIS™ [2] for core design and coupled transient safety analysis. ARTEMIS™ can be used in a coupled mode with S-RELAP5 and CATHARE 2 to allow fully coupled transient analysis, combining the sophisticated 3D core model of ARTEMIS™ with the proven system thermal-hydraulics of S-RELAP5 and CATHARE 2 including a detailed simulation of the Instrumentation and Control (I&C). This allows simulating complex transients affecting the core as well as the primary and secondary side including I&C signals and responses.
For the validation of ARTEMIS™ a comprehensive set of validation cases was selected, including international benchmarks and measurements covering various classes of transients. These cases include a ‘Load Rejection to station service’ event at a German 1300 MWe plant, where a wide range of system and core parameters was measured that allow the validation of the fully coupled code system. Another validation case is a nodal recalculation of the core behavior during the pump shaft break transient that occurred in the Gösgen nuclear power plant in 1985 [3].
The paper will provide representative example results for the abovementioned validation cases.
HELENA: A Heavy Liquid Metal Multi-Purpose Loop for Thermal-Hydraulics, Corrosion and Component Test
Authors: Di Piazza I. , Tarantino M. , Agostini P. , Gaggini P. 
Published: November 2014

Since the Lead-cooled Fast Reactor (LFR) has been conceptualized in the frame of GEN IV International Forum (GIF), ENEA is strongly involved on the HLM technology development.
Currently ENEA has implemented large competencies and capabilities in the field of HLM thermal-hydraulic, coolant technology, material for high temperature applications, corrosion and material protection. In this frame, the HELENA facility is well instrumented and it represents a loop working in pure lead for experiments in the field of corrosion for LFR structural materials, component test, and thermal-hydraulic investigations. The components and the process of the facility has been also depicted in some details. The scheduled working activities has been described and the main steps focused.
A prototypical mechanical pump has been designed and manufactured to properly work in pure lead at high temperatures. The long-run test on the component will be isothermal at 400°C with low oxygen content <10−8 % in weight; the oxygen content will be monitored continuously. The pump will be gradually driven to the reference mass flow rate 35 kg/s and this mass flow rate will be maintained for 1500 h, i.e. 2 months about. After this, the pump is stopped and the loop is drained. Then, the pump impeller is disassembled by the body and it is analyzed for corrosion. Then, a test on the ball valves is carried out for a few months. At the end of 2014, the facility is upgraded with the insertion of the FPS in the heating section and with the secondary side. A 19-pin wire-spaced Fuel Pin bundle Simulator (FPS) is installed to measure clad temperature and heat transfer coefficients in different conditions in the different ranks of sub-channels of the MYRRHA bundle. Therefore a test matrix on the forced convection condition in the wire-spaced bundle will be carried out.
A 7 tube/shell-and-tube Heat Exchanger couples the primary lead loop with the secondary side with water in pressure at 100 bar. The tube-in-tube technology with lead tube side, water shell side, steel powder in the gap is adopted. Bubble tubes with flowing Argon are adopted to measure pressure losses in the different branches of the loop. Several thermocouples monitor the loop in different points. An ancillary gas system ensures the cover gas.
The paper reports the description of the experiments, the proposed test matrix and description the technological solution adopted for the HELENA implementation.
Analysis of the Control Rod Drop Accident (CRDA) for Lungmen ABWR
Authors: Ho Ai-Ling , Wang Jong-Rong , Lin Hao-Tzu , Shih Chunkuan 
Published: November 2014

The object of this paper is to understand the realistic behavior in Lungmen ABWR (Advanced Boiling Water Reactor) during a control rod drop accident (CRDA) transient. The CRDA transient would lead the reactor through an extremely fast and localized power excursion, requiring an accurate core modeling. The CRDA analysis for Lungmen ABWR was performed by coupling the 3D neutron kinetic code, PARCS, and two-phase thermal-hydraulic (T-H) code, TRACE. After TRACE/PARCS coupling calculation, the output data from TRACE/PARCS would be inputted into FRAPTRAN code as a function of time-dependent fuel rod power and coolant boundary conditions to calculate the fuel damage.
The CRDA analysis for Lungmen ABWR was performed for two conditions: a) case1: hot-full-power (HFP) at beginning of cycle (BOC); b) case2: hot-zero-power (HZP) at BOC. Under these conditions, the damage mechanisms of fuel rod are: 1) cladding ballooning and burst; 2) embrittlement and failure by high-temperature oxidation; 3) melting of cladding and/or fuel pellets. And the relevant quantities for fuel performance are the maximum fuel enthalpy and the melting temperatures of cladding and fuel pellet.
The results of CRDA analysis show that a) case1: no fuel failure occurs under HFP condition at BOC; b) case2: the fuel rod nearby the dropped control rod failed under HZP condition at BOC, and the FRAPPTRAN data exposes that the main reason of rod failure is the cladding high temperature.
Burnable Absorber Selection With UWB1 Depletion Code
Authors: Lovecký Martin , Prehradný Jan , Škoda Radek 
Published: November 2014

Research of nuclear reactor fuel depletion aims at development and introduction of advanced types of burnable absorbers (BA) applied within nuclear fuel. BAs compensate for the initial reactivity excess and consequently may allow for lower power peaking factors and longer fuel cycles with higher fuel enrichments.
Modern computer codes for nuclear fuel depletion calculation require a substantial amount of computational time. Therefore, any parametric calculations for BA selection need to be carried out only with a fast depletion code. The main purpose of the newly developed UWB1 code is a rapid calculation of nuclear fuel depletion, which is achieved by the approximations in the equations describing transport part of fuel depletion. Microscopic cross sections are assumed to be constant through depletion calculation steps.
The paper describes the first step of analysis using a new version of UWB1, that was accomplished with the assumption of uniformly distributed BA in fuel. BA elements, nuclides and nuclide mixtures were compared and their performance was consequently evaluated based on multiplication coefficient behavior during depletion.
Burnable Absorber Comparison Between VVER, PWR and SFR With UWB1 and SERPENT Codes
Authors: Prehradný Jan , Lovecký Martin , Škoda Radek 
Published: November 2014

Research of fuel depletion aims at development and introduction of advanced types of burnable absorbers (BA) in nuclear fuel. BAs compensate for the initial reactivity excess and consequently allow for lower power peaking factors and longer fuel cycles with higher fuel enrichments.
The paper describes the comparison of selected BA elements (Gd, Eu, Er) for 3 different nuclear fuel types (VVER, PWR and SFR). The effect of harder VVER spectra and the effect of thermal vs fast neutron spectra is evaluated and commented. Uniformly distributed BAs in the fuel were assumed.
Comparison calculations were performed with the newly developed UWB1 fast fuel depletion code and with the SERPENT Monte Carlo code. The comparison of UWB1 and SERPENT codes is included in the analysis. Next steps of improving the UWB1 code are suggested.
Review on Water Radiolysis in the Fukushima Daiichi Accident: Potential Cause of Hydrogen Generation and Explosion
Authors: Saji Genn 
Published: November 2014

Although the water radiolysis, decomposition of water by radiation, is a well-known phenomenon the exact mechanism is not well characterized especially for severe accidents. The author first reviewed the water radiolysis phenomena in LWRs during normal operation to severe accidents (e.g., TMI- and Chernobyl accidents) and performed a scoping estimation of the amount of radiological hydrogen generation, accumulation and release for the Fukushima Daiichi accident. The estimation incorporates the decay heat curve after a reactor trip combined with G-values. As much as 450 cubic meters-STP of accumulated hydrogen gas is estimated to be located inside the PCV just prior to the hydrogen explosion which occurred a day after the reactor trip in Unit 1.
When a set of radiological chain reactions are incorporated the resultant reverse reactions substantially reduce the hydrogen generation, even when removal of molecular products (i.e., oxygen and hydrogen) is assumed stripped rapidly from boiling water through bubbles. Even in the most favorable configuration a typical amount of hydrogen gas reduces to only several tens of cubic meters.
Finally, the author tested a new mechanism, “radiation-induced electrolysis,” which had been applied to his corrosion studies for last several years. His theory has been verified with the published in-pile test data, although he has never tried to apply this to his severe accident study. The predicted results indicated that the total inventory of hydrogen gas inside RPV may reach as much as 1000 cubic meters in just 3 hours during the SBO due to a high decay heat soon after the reactor trip through this process.
AP1000® Plant Application of Defense in Depth
Authors: Schulz Terry L. , Gorgemans Julie 
Published: November 2014

The AP1000 plant is an 1100-MWe pressurized water reactor (PWR) with passive safety features and extensive plant simplifications that enhance construction, operation, maintenance and safety. One of the key design approaches in the AP1000 plant design is to use passive features to mitigate design basis accidents. Active defense-in-depth (DiD) features provide investment protection, reduce the demands on the passive features and support the Probabilistic Risk Assessment (PRA). The passive features are classified as safety-related in the United States. The active DiD features are classified as nonsafety-related (with supplemental requirements) in the United States. The AP1000 plant design has also incorporated a standardization approach, which together with the level of safety achieved by the passive safety features, results in a plant design that can be applied to different geographical regions with varying regulatory standards and utility expectations without major changes.
This paper will discuss the approach taken to defining DiD in the AP1000 plant and the effectiveness of that approach. It will also address the capability of the AP1000 plant to meet deterministic DiD guidelines such as the ones in application in the UK or described in the Western European Nuclear Regulators’ Association (WENRA) safety objectives for new plants.
Loss-of-Flow-Accidents (LOFA) Study for 100 MW IPWR
Authors: Yin Shasha , Gao Liang , Tian Wenxi , Zhang Yapei , Qiu Suizheng , Su Guanghui 
Published: November 2014

The inherent system safety of the 100 MW integral pressurized water reactor (IPWR) can be improved by placing the core, the efficient once-through steam generators and the main coolant pumps in the reactor pressure vessel, and omitting some large pipes and valves in the primary coolant system which can prevent the occurrence of large break loss of coolant accident and reduce the possibility of core melt accident. The application of the passive safety systems simplifies the structures of IPWR and improves the economy of the reactor. In case of accidents, the primary coolant system establishes natural circulation to take the core decay heat away by passive safety systems using gravity and other natural driving forces, thereby enhancing the safety and reliability of the system IPWR. It’s of great significance to establish reasonable and correctable models, including the primary coolant system model, the second loop model and passive core cooling system model, to study thermal-hydraulic phenomena under steady state, transient state and accident conditions.
Based on transient safety analysis program RELAP5/MOD3.4, 100 MW IPWR system was simulated. A series of models of reactor coolant system and passive safety systems were established. The main system models are composed of primary coolant system model, part of second loop model, passive safety injection system model and passive residual heat removal system model. The primary coolant system model includes core, lower plenum, downcomer, region of steam generators, upper plenum, riser, pressurizer, and surge line; the second loop model includes the main feed water line, the steam line, and steam generator tubes; passive safety injection system model includes core makeup tank, accumulator, automatic depressurization system, direct vessel injection line; and passive residual heat removal system model includes passive residual heat removal heat exchanger in containment refueling water storage tank. Based on the established models, the steady state was debugged with the RELAP5 input card.
Steady state calculation was performed and the results agree well with designed values, which verifies the validity of the model and the input card. Using the steady state results as initial conditions, transient calculation was performed. Typical accidents (loss of main water accident) were calculated, which were relieved by auxiliary feedwater system (AFWS) and passive residual heat removal system (PRHR SYSTEM). The results, obtained from AFWS and PRHR SYSTEM, were contrasted and process of accident and thermal-hydraulic phenomena were analyzed according to transient calculation results. The transient calculation results showed that the integral PWR system and the passive safety system model can provide a reference for IPWR transient safety analysis.