bluuebunny/crossref_metadata_2025_split huggingface dataset:
Preliminary Analysis of In-Vessel Corium Confinement and Cooling in a Large Sodium Fast Reactor
Authors: Polidoro Franco , Parozzi Flavio 
Published: October 2013

Considering a reasonable range of core meltdown accidents that can be postulated for GenIV sodium fast reactors, good safety margins exist for corium confinement and cooling inside the reactor vessel. Coolable conditions can be reached with the adoption of an ad-hoc device in the lower plenum, i.e. core catcher, capable to intercept the downward motion of the molten material and assure its cooling. Such device has to be designed to withstand to extreme thermal-mechanical conditions that rise as consequence of the large mechanical energy release and high temperature of molten corium.
As this study has been carried out in the frame of the Collaborative Project on European Sodium Fast Reactor (CP ESFR) of the 7th Framework Programme Euratom, on the basis of the postulated accident conditions assumed for a reference 1500 MWe pool-type sodium fast reactor, the present work provides a preliminary analysis of the thermal response of a possible core catcher placed within the vessel. The dynamic thermal behaviour of the corium-structure-coolant system is analyzed with the computer code CORIUM-2D, an original simulation tool developed by RSE - Ricerca Sul Sistema Energetico, with the aim to assess the thermal interaction among corium, structures and coolant under severe accident conditions in both Light Water Reactors (LWRs) and Liquid Metal Fast Breeder Reactors (LMFBRs).
The results of the numerical simulations show that the steady-state coolable configuration of core debris and the structural integrity of main containment structures can be reached in a number of partial core meltdown situations.
EC6 Instrumentation, Control, and Electrical Design for Station Blackouts
Authors: Foster Philip , Harber John , Ford Steven , Lekakh Boris , Nie Chunlei 
Published: October 2013

The Enhanced CANDU 6 (EC6) plant design includes complementary design features that assist the goals of preventing severe accident scenarios and mitigating their consequences if they do occur. These features include design or procedural considerations (or both), and are based on a combination of phenomenological models, engineering judgments, and probabilistic methods. This paper provides an overview of the EC6 instrumentation and control (I&C) and electrical design specifically related to prevention of severe accidents during station blackouts (SBOs).
The I&C and electrical components and devices may include those used in normal operation, those of the safety systems used during design basis accidents, and those added solely for preventing or mitigating severe accidents. They are evaluated by an equipment survivability assessment to demonstrate a reasonable level of assurance for their operability, and a seismic fragility evaluation in consideration of a seismic initiating event.
The requirements for the I&C and electrical design for SBOs are developed by reviewing EC6 event sequences leading to SBOs, identifying the available systems, and identifying the actions needed and whether they are manual or automatic, passive or active. Monitoring plays a critical role in the prevention of severe accidents. The correct operation of instrumentation and associated display devices for key parameters is essential in diagnosing the plant state and safety challenges.
Accounting Nuclear Fuel Burnup for Criticality Safety Justification of Spent Nuclear Fuel Storages
Authors: Solovyov Vladislav , Galchenko Vitaliy 
Published: October 2013

In this paper we have analyzed accounting spent nuclear fuel (SNF) burnup of VVER-1000 and RBMK-1000 only with actinides.
The following characteristics were analyzed: initial fuel enrichment, burnup fraction, axial burnup profile in the fuel assembly (FA) and fuel weight.
As the results show, in the first 400 hours after stopping the reactor, there is an increase in the effective neutron multiplication factor (keff) due to beta decay of 239Np into 239Pu. Further, from 5 to 50 years, decrease in keff due to beta decay of 241Pu into 241Am. In the future there is a slight change in criticality of the system.
Accounting nuclear fuel burnup in the justification of nuclear safety of SNF storages will provide an opportunity to increase the volume of loaded fuel and thus significantly reduce technology costs of handling of SNF.
Generation of In-Structure Response Spectra for Nuclear Power Plants Subjected to High-Frequency Ground Motion
Authors: Moreschi Luis M. , Pan Qin , Wang Shen , Malushte Sanjeev R. 
Published: October 2013

The purpose of seismic qualification of Structures, Systems and Components (SSCs) in nuclear power plants is to ensure that their intended safety function will not be compromised during and after a postulated earthquake event. The seismic performance of the equipment is generally evaluated using In-Structure Response Spectra (ISRS) at equipment-support locations as an input motion. Traditionally, these ISRS are generated based on design ground spectra prescribed by either U.S. Nuclear Regulatory Commission Regulatory Guide 1.60 or other design spectral shapes, which normally consider the frequencies content up to 33Hz. However, it has been recently recognized that probabilistic hazard-based site specific ground motion response spectra (GMRS) for Central and Eastern United States (CEUS) hard rock sites contains significant energy in the high frequency range, far beyond 33Hz. Since the motion at equipment support locations is highly affected by the dynamic characteristics of the soil or rock surrounding the building foundations and those of the structure itself, the adequacy of dynamic modeling and analysis techniques for determining the ISRS is critical to seismic qualification of safety-related equipment.
This paper provides examples on dynamic modeling and analysis techniques required to accurately capture the structural responses for purposes of calculating ISRS throughout the frequency range of interest, including the high frequency responses typically expected at the CEUS sites. The discussion includes the selection of finite element mesh size, and sensitivity analysis performed to demonstrate that the propagation of these high frequencies through the different levels of the structure is properly captured. Other analytical considerations, such as the selection of time step size, for conducting time-history analysis, are also presented.
TRACE and RELAP5 Codes for Beyond Design Accident Condition Simulation in the SPES3 Facility
Authors: Ferri Roberta , Mascari Fulvio , Meloni Paride , Vella Giuseppe 
Published: October 2013

Code validation on qualified experimental data is a fundamental issue in the design and safety analyses of nuclear power plants.
The SPES3 facility is being built at the SIET laboratories for an integral type SMR simulation, in the frame of an R&D program on nuclear fission, funded by the Italian Ministry of Economic Development and led by ENEA.
The facility, based on the IRIS reactor design, reproduces the primary, secondary and containment systems with 1:100 volume scale, full elevation and prototypical fluid and thermal-hydraulic conditions. It is suitable to test the plant response to design and beyond design accidents in order to verify the effectiveness of the primary and containment system dynamic coupling to cope with loss of coolant accidents.
Full and complete nodalizations of SPES3 were developed for TRACE and RELAP5 codes in order to investigate the code response to the simulation of the same accidental transient. The DVI line DEG break was simulated in beyond design conditions, assuming the failure of all emergency heat removal systems and relying on PCC intervention for containment depressurization and decay heat removal.
The comparison of the code simulation results, other than providing information on the system behavior, allowed to investigate specific phenomena evidenced by the codes, according to the related modeling approach of components with one and three-dimensional volumes.
The TRACE and RELAP5 codes will be applied for further transient analyses and will be validated on SPES3 experimental data, once the facility will be available.
Two-Phase Flow in an Annular Channel With an Obstacle
Authors: Kashinsky O. N. , Lobanov P. D. , Kurdyumov A. S. , Pribaturin N. A. 
Published: October 2013

Experimental study of gas-liquid two-phase flow in an annular channel is performed. The channel consisted of two coaxial tubes with the diameters of 42 and 20 mm. An obstacle covering a quarter of the channel section was placed in the channel to produce a strong three-dimensional disturbance of the flow. Gas-liquid flow was produced by injecting air bubbles at the channel entrance through a special mixer. Measurements of local wall shear stress are performed using an electrochemical technique. Measurements of time-averaged and fluctuational wall shear stress are performed at various points relative to the obstacle, this allowed to study the field of the hydrodynamic parameters of the flow. Local void fraction is measured using a conductivity probe which traversed across the channel. The distribution of local void fraction in the region downstream the obstacle is obtained. Increased values of local void fraction in the region close to the obstacle are detected.
The experimental data obtained can be used for validation of existing and developing computer codes accounting for a 3-D structure of two-phase flows.
Investigation of a Heavy Liquid Metal Free Surface Target Experiment
Authors: Litfin Karsten , Fellmoser Frank , Batta Abdalla , Class Andreas G. , Wetzel Thomas 
Published: October 2013

Several distinct reactor strategies are proposed within the context of the IP Eurotrans framework programme for the transmutation of nuclear waste. A pool type reactor filled with liquid heavy metal and containing a subcritical core is one of the promising designs. Additional neutrons required for the nuclear reaction are generated by a spallation reaction inside the core. A high power proton beam is guided through a vacuum tube from an accelerator into the liquid heavy metal pool i.e. into the reactor core. At the point where the beam hits the metal surface special construction effort is indispensable to handle the high heat production.
A specific target design is used to ensure a high fluid velocity and a stable surface at the beam entry. This design employs a concentric vertical feeder establishing a free conical surface with velocity up to 2.5 m/s ensuring stable surface flow and appropriate heat removal. The proposed target geometry has been constructed using underlying rules developed by the MYRRHA design group for the free surface target in the MYRRHA research reactor.
A full scale model of this design using lead bismuth eutectic (LBE) has been set up and experimentally investigated at the of the Karlsruhe Institute of Technology (KIT). Measurements taken by high speed digital imaging visualize both conical inner and outer jet free surface. They show a stable surface in a wide range of operating conditions starting from 35% of the nominal flow rate and agree well with numerical investigations using commercial CFD code Star-CD and Star-CCM+. Previous concerns related to splashing or cavitation during the start-up or shut-down procedure proofed unjustified.
Assessing Sensitivity of Observations in Source Term Estimation for Nuclear Accidents
Authors: Ma Yuanwei , Wang Dezhong , Tan Wenji , Ji Zhilong , Zhang Kuo 
Published: October 2013

In the Fukushima nuclear accident, due to the lack of field observations and the complexity of source terms, researchers failed to estimate the source term accurately immediately. Data assimilation methods to estimate source terms have many good features: they works well with highly nonlinear dynamic models, no linearization in the evolution of error statistics, etc. This study built a data assimilation system using the ensemble Kalman Filter for real-time estimates of source parameters. The assimilation system uses a Gaussian puff model as the atmospheric dispersion model, assimilating forward with the observation data.
Considering measurement error, numerical experiments were carried on to verify the stability and accuracy of the scheme. Then the sensitivity of observation configration is tested by the twin experiments.
First, the single parameter release rate of the source term is estimated by different sensor grid configurations. In a sparse sensors grid, the error of estimation is about 10%, and in a 11*11 grid configuration, the error is less than 1%.
Under the analysis of the Fukushima nuclear accident, ahead for the actual situation, four parameters are estimated at the same time, by 2*2 to 11*11 grid configurations. The studies showed that the radionuclides plume should cover as many sensors as possible, which will lead a to successful estimation.
Sensitivity Analysis of Fuel Centerline Temperature in SCWRs
Authors: Abdalla Ayman , Peiman Wargha , Pioro Igor , Gabriel Kamiel 
Published: October 2013

The Generation IV International Forum (GIF) is intended to encourage the world’s leading nuclear countries to develop nuclear energy systems that can supply future energy demands. There are six nuclear reactor concepts under research and development as part of the GIF. The SuperCritical Water-cooled Reactor (SCWR) is one of these six nuclear-reactor concepts. The proposed SCWRs operate at high temperatures and pressures at around 625°C and 25 MPa, respectively. These high operating parameters are essential in order to achieve a thermal efficiency of around 45–50%, which is significantly higher than those of the current conventional nuclear power plant (NPPs) which operate at a thermal efficiency in the range of 30–35%.
The SCWRs high operating temperatures and pressures impose many challenges. One of these challenges is the heating of the fuel to temperatures that can cause fuel melting. The main objective of this paper is to conduct a sensitivity analysis in order to determine the factors mostly affecting the fuel centerline temperature. In this process, different thermal conductivity fuels such as Mixed Oxide Fuel (MOX), Uranium Oxide + Beryllium Oxide (UO2+BeO), and Uranium Carbide (UC) will be examined enclosed in a 54-element fuel bundle. Other factors such as the sheath material and the Heat Transfer Coefficient (HTC) might also affect the fuel centerline temperature. The HTC will be increased by a multiple of two and the fuel centerline temperature will be calculated. Therefore, in this paper the HTC, bulk-fluid, sheath and fuel centerline temperature will be calculated along the heated length of a generic SCWR fuel channel at an average channel thermal power of 8.5 MWth.
Comparison of Existing Supercritical Carbon Dioxide Heat Transfer Correlations for Horizontal and Vertical Bare Tubes
Authors: Surendran Prabu , Gupta Sahil , Preda Tiberiu , Pioro Igor 
Published: October 2013

This paper presents a thorough analysis of ability of various heat transfer correlations to predict wall temperatures and Heat Transfer Coefficients (HTCs) against experiments on internal forced-convective heat transfer to supercritical carbon dioxide conducted by Koppel [1], He [2], Kim [3] and Bae [4]. It should be noted the Koppel dataset was taken from a paper which used the Koppel data but was not written by Koppel. All experiments were completed in bare tubes with diameters from 0.948 mm to 9 mm for horizontal and vertical configurations. The datasets contain a total of 1573 wall temperature points with pressures ranging from 7.58 to 9.59 MPa, mass fluxes of 400 to 1641 kg/m2s and heat fluxes from 20 to 225 kW/m2.
The main objective of the study was to compare several correlations and select the best of them in predicting HTC and wall temperature values for supercritical carbon dioxide. This study will be beneficial for analyzing heat exchangers involving supercritical carbon dioxide, and for verifying scaling parameters between CO2 and other fluids. In addition, supercritical carbon dioxide’s use as a modeling fluid is necessary as the costs of experiments are lower than supercritical water.
The datasets were compiled and calculations were performed to find HTCs and wall and bulk-fluid temperatures using existing correlations. Calculated results were compared with the experimental ones. The correlations used were Mokry et al. [5], Swenson et al. [6] and a set of new correlations presented in Gutpa et al. [7]. Statistical error calculations were performed are presented in the paper.
Methodology for the Analysis of Fuel Behavior During LOCA and RIA for Licensing Purposes
Authors: Adorni Martina , Del Nevo Alessandro , D’Auria Francesco 
Published: October 2013

Licensing requirements vary by country in terms of their scope, range of applicability and numerical values and may imply the use of system thermal hydraulic computer codes. Depending on the specific event scenario and on the purpose of the analysis, it might be required the availability of calculation methods that are not implemented in the standard system thermal hydraulic codes, as for burst temperature, burst strain and flow blockage calculations. This may imply the use of a dedicated fuel rod thermo-mechanical computer code, which can be coupled with thermal-hydraulic system and neutron kinetic codes to be used for the safety analysis.
This paper describes the development and the application of a methodology for the analysis of the Large Break Loss of Coolant Accident (LB-LOCA) scenario in Atucha-2 Nuclear Power Plant (NPP), focusing on the procedure adopted for the use of the fuel rod thermo-mechanical code and its application for the safety analysis (Chapter 15 Final Safety Analysis Report, FSAR). The methodology implies the application of best estimate thermal-hydraulic, neutron physics and fuel pin performance computer codes, with the objective to verify the compliance with the specific acceptance criteria. The fuel pin performance code is applied with the main objective to evaluate the extent of cladding failures during the transient.
A strong effort has been performed in order to enhance the fuel behaviour code capabilities and to improve the reliability of the code results.
Effect of Pressure on Underdeveloped Nucleate Boiling on the Single Horizontal Tube Heated by Condensing Steam
Authors: Morozov A. V. , Remizov O. V. , Kalyakin D. S. 
Published: October 2013

In new Russian NPP with VVER reactor in the event of LOCA, provision is made for the use of passive heat removal system for necessary core cooling. In the case of leakage in the primary circuit this system assures the transition of steam generators to operation in the mode of condensation of the primary circuit steam coming to steam generator piping from the reactor. As a result, the condensate from the steam generators arrives to the core providing its additional cooling.
The steam generator off-design condensation mode has the following features: undeveloped nucleate boiling on the horizontal tubes heated by condensing steam; natural circulation processes in the both steam generator circuits; low heat fluxes and temperature differences. The experimental study of undeveloped nucleate boiling on the single horizontal tube heated by condensing steam has been carried out in the Institute for Physics and Power Engineering.
The experiments have been carried out on the GROT test facility. The heart of the test facility is the single VVER steam generator tube (length l = 10.2 m, outer diameter D = 16 mm, wall thickness d = 1.5 mm). The tube is fabricated of the original stainless steel 08Cr18Ni10Ti. The length and geometry of the test tube corresponded to that of real steam generator. The test facility was equipped with thermocouples enabling the temperatures of primary and secondary facility circuits to be controlled.
The experiments were carried out at three heating steam pressures Ps1: 0.21, 0.35, 0.55 MPa. The main task of the research was to study the pressure effect on the process of undeveloped nucleate boiling on the single horizontal tube.
On the base of the results of these experiments the empirical correlations for prediction of heat transfer coefficient and heat flux were obtained. The generalizing empirical correlations obtained can be used for the substantiation of work heat-exchanging equipment of NPP with VVER reactor in the condensation mode, and also can be applied to the verification of computer codes.
Uncertainty Evaluation of the Rod Withdrawal at Power Accident Analysis Including 3D Neutron Kinetics
Authors: Bajs Tomislav , Petruzzi Alessandro , Grgić Davor 
Published: October 2013

A continuous uncontrolled Rod Cluster Control Assembly (RCCA) bank withdrawal at power belongs to group of Reactivity Initiated Accidents (RIA). It will cause an increase in core heat flux and a reactor coolant temperature rise. Unless terminated by manual or automatic action, the power mismatch and resultant coolant temperature rise could eventually result in departure from nucleate boiling (DNB) and/or fuel centreline melt. The accident can be DNBR or overpower limiting accident depending on initial power level and rate and amount of reactivity addition.
The Rod Withdrawal At Power (RWAP) accident was analyzed for NPP Krško to evaluate possible Resistance Temperature Detectors (RTD) bypass removal and introduction of thermowell for the average temperature measurement. The influence of different coolant temperature measurement delays to related protection system response and limiting system variables was studied first using point kinetics model as implemented in RELAP5 code. The selected scenario (maximum insertion rate with rods in manual mode) has been re-calculated using RELAP5/PARCS coupled code. Core wide departure from nucleate boiling ratio (DNBR) calculation has been performed at the end of the coupled code calculation using COBRA based model to determine minimum DNBR for hot channel. In order to assess available safety margins following such accident CIAU methodology has been applied to evaluate the uncertainty of RELAP5 analysis and modified CIAU/TN methodology to evaluate uncertainty of the three-dimensional neutronics/thermal-hydraulics calculations. Differences between system and coupled code results and uncertainties is discussed.
The Safety Analysis of the 600 MWe Sodium-Cooled Fast Reactor With MARS-LMR
Authors: Lee Kwi Lim , Ha Kwi Seok , Jeong Hae Yong , Chang Won Pyo 
Published: October 2013

Korea Atomic Energy Research Institute (KAERI) has been developing a conceptual design of the demonstration fast reactor (DFR), which is the pool type sodium cooled fast reactor with the thermal power of 1548.2 MW and the core loaded with metal fuel.
The DFR is composed of a Primary Heat Transport System (PHTS), an Intermediate Heat Transport System (IHTS), a Steam Generating System (SGs) and a decay heat removal system (DHRS). The DHRS is composed of 2 units of Passive Decay-heat Removal Circuits (PDRC) and 2 units of Active Decay-heat Removal Circuits (ADRC). The PDRC consists of two independent loops with sodium-to-sodium Decay Heat eXchanger (DHX) and natural-draft sodium-to-Air Heat eXchanger (AHX). The ADRC consists of two independent loops with sodium-to-sodium DHX and Forced-Draft sodium-to-air Heat eXchanger (FDHX) located in the upper region of the reactor building. The PDRC is very different from that of KALIMER-600 on the points of the submerged location and the heat transfer mechanism.
For the identification of safety characteristics, 5 DBE’s (Design Bases Events) are analyzed using the MARS-LMR code. The representative DBE’s are TOP (Transient of Over Power), LOF (Loss Of Flow), LOHS (Loss Of Heat Sink), Reactor Vessel Leak and Pipe Break. As a result, it is identified that the DFR were appropriately performed as designed and the temperatures of the fuel and the structure were evaluated to satisfy the criteria.