bluuebunny/crossref_metadata_2025 huggingface dataset:
Effect of ECC Injection Angle on the Flow Distribution in a Downcomer Annulus During LBLOCA Reflood Phase in the Air-Water Test
Authors: Kwon T. S. , Yun B. J. , Song C. H. , Choi K. Y. , Cho H. K.
Published: March 2009

The comparison tests for the direct ECC bypass fraction were experimentally performed with a typical DVI nozzle and an ECC column nozzle having injection angle to the gravity axis. The ECC column nozzle is newly introduced to make an ECC water column in the downcomer region. The injection angle of the ECC water relative to the gravity axis is varied from 0 to ±90 degrees stepped by 45 degrees. The tests are performed in the air-water separate effect test facility (DIVA), which is 1/7.07 linearly scaled-down of the APR1400 nuclear reactor. The test results show that the direct ECC bypass fraction is affected by the ECC injection angle when the ECC water is injected using an ECC column nozzle as a single water column. The injection angle of an ECC water column relative to the circumferential air jet in the DVI system affects the direct ECC bypass fraction during the reflood phase of a LBLOCA.
Expert System for Diagnostics and Status Monitoring of NPP Water Chemistry Condition
Authors: Shvedova M. N. , Kritski V. G. , Zakharova S. V. , Nikolaev F. V. , Benediktov V. B.
Published: March 2009

Water chemistry condition (WCC) has been the subject of constant study and improvement up to the present day. It is connected with the presence of a direct relationship between the violation of water chemistry regulation on the one hand and components reliability of the circuit’s equipment and cost-effectiveness of their operation on the other. It dictates the necessity to apply different optimization methods in the field of monitoring and use of information — analytical and diagnostic systems to assess WCC quality, control and support. By now NPP experts have broad experience in revealing and removing the causes of WCC disturbances. However this knowledge is often of an intuitive, non-classified nature, scattered among various working documents, which makes their transfer difficult. Based on what has been mentioned above, special attention is currently being paid to the problem of creating expert diagnostic systems for supporting the optimum WCC. The existing developments in this field (DIWA, Smart chem Works, the water quality control system at the Onagava NPP etc. [1,3,4,5] are based on wide use of experts’ knowledge. Such expert diagnostic systems for supporting WCC refer to the new generation of intellectual control methods, which allow the incorporation of the latest achievements both in the field of water chemistry simulation and in the field of artificial intelligence and computer technologies. LI “VNIPIET” employees have, for several years, been developing an expert diagnostic system for supporting WCC and status monitoring of RBMK - reactor NPPs [2]. This system has not only conveniently organized the traditional functions of information acquisition and storage, a complete presentation of information in the form of tables, graphs of a dynamical changes of parameters and formation regular reports, diagnostic functions and issuing recommendations on WCC correction, but it also allows the assessment of confidence in the diagnosis made, relying on a wide range of numerical estimates, which were calculated by the use of expert data, and to make a credible prediction of an existing situation development. The integrated use of analytical methods and artificial intelligence methods is one of the system’s advantages. This combination allows the successful implementation of one of the main purposes of the system: the early detection of deviations from specified process conditions and the taking into account of even minor changes in parameters to provide an advanced WCC control and to prevent non-regular situations.
Ensuring Integrity and Functional Capacity of the Containment Under Beyond Design Basis Accident in the Design of New Generation NPP With WWER-1000
Authors: Berkovich V. M. , Malyshev A. B. , Taranov G. S. , Podshybyakin A. K. , Shumsky A. M.
Published: March 2009

A method for provision of containment integrity and localizing capability of the new generation NPPs with WWER reactor units under the beyond design accidents is discussed. The aim is achieved by equipment of NPPs by additional passive systems.
Seismic Behavior of Spent Fuel Dry Cask Storage Systems
Authors: Shaukat Syed K. , Luk Vincent K.
Published: March 2009

The U. S. Nuclear Regulatory Commission (NRC) is conducting a research program to investigate technical issues concerning the dry cask storage systems of spent nuclear fuel by conducting confirmatory research for establishing criteria and review guidelines for the seismic behavior of these systems. The program focuses on developing 3-D finite element analysis models that address the dynamic coupling of a module/cask, a flexible concrete pad, and an underlying soil/rock foundation, in particular, the soil-structure-interaction. Parametric analyses of the coupled models are performed to include variations in module/cask geometry, site seismicity, underlying soil properties, and cask/pad interface friction. The analyses performed include: 1) a rectangular dry cask module typical of Transnuclear West design at a site in Western USA where high seismicity is expected; 2) a cylindrical dry cask typical of Holtec design at a site in Eastern USA where low seismicity is expected; and 3) a cylindrical dry cask typical of Holtec design at a site in Western USA with medium high seismicity. The paper includes assumptions made in seismic analyses performed, results, and conclusions.
Thermal Analysis of the One-Stage Melting Converter and Related Equipment for Vitrification of High-Level Radioactive Wastes
Authors: Pioro L. S. , Pioro I. L. , P’yanykh K. E.
Published: March 2009

One-stage melting-converter method for the vitrification of high-level radioactive wastes (HLRAW) was developed and investigated. This method includes the concentration (evaporation), calcination, and vitrification of the HLRAW in one-stage process inside a melting converter. Converter consists of a water-cooled conical metal melting chamber. Capacity for the saturated salty solution is 200 l/h and the capacity for the glass-HLRAW mixture is about 100 kg/h. Calculations showed that at a melt temperature of 1150°C, a cooling-water temperature of 20°C and a thickness of the melt lining of 15 mm the heat flux through the walls is about 75 kW/m2 and wall temperature is about 55°C. Thermal analysis showed that at the burner heat transfer rate of 460 kW the heat transfer rate for heating and melting of the charge is 20 kW, the heat transfer rate for evaporation of water from the solution is 100 kW, and the heat losses are 240 kW. Therefore, a thermal efficiency of the converter is about 26%. To keep a melt lining on the vault of converter and around a burner nozzle the special design two-phase closed thermosyphons are used. Design and heat transfer characteristics of these thermosyphons are discussed. Calculations have been made for heat removal of the melted glass and fissionable nuclides poured into the metal containers during first 500 hours.
Preliminary Safety Analysis of the IRIS Reactor
Authors: Ricotti M. E. , Cammi A. , Cioncolini A. , Cipollaro A. , Oriolo F. , Lombardi C. , Conway L. E. , Barroso A. C.
Published: March 2009

A deterministic analysis of the IRIS safety features has been carried out by means of the best-estimate code RELAP (ver. RELAP5 mod3.2). First, the main system components were modeled and tested separately, namely: the Reactor Pressure Vessel (RPV), the modular helical-coil Steam Generators (SG) and the Passive (natural circulation) Emergency Heat Removal System (PEHRS). Then, a preliminary set of accident transients for the whole primary and safety systems was investigated. Since the project was in a conceptual phase, the reported analyses must be considered preliminary. In fact, neither the reactor components, nor the safety systems and the reactor signal logics were completely defined at that time. Three “conventional” design basis accidents have been preliminary evaluated: a Loss Of primary Flow Accident, a Loss Of Coolant Accident and a Loss Of Feed Water accident. The results show the effectiveness of the safety systems also in LOCA conditions; the core remains covered for the required grace period. This provides the basis to move forward to the preliminary design.
An 11% Cold Leg Break LOCA Investigation in PSB-VVER Test Facility
Authors: Lipatov Igor A. , Dremin George I. , Galchanskaia Svetlana A. , Moloshnikov Alexander S. , Nickonov Serguey M. , Rovnov Alex A. , Gudkov Victor I. , Antonova Alexandra I. , Gorbunov Yuri S. , Kapustin Andrej V. , Elkin Ilya V. , Zaitsev Serguey I.
Published: March 2009

In the PSB-VVER Integral Test Facility (ITF), extensive experimental investigations on the thermal hydraulic behavior of a VVER-1000 type reactor under LOCA accidents have been carried out during 2001. The general aim of the experiments is to contribute to a better understanding of accident sequences and to provide a detailed data base for the validation of Russian and Western computer codes. This paper presents the results of an 11% cold leg break experiment (CL-11-05) assuming loss of off-site power. The presented experiment belongs to a set of tests with a spectrum of break area in the PSB cold leg carried out in the summer 2001. The results of the post-test analyses using TRAP code are presented. A short description of the TRAP code mathematical model is also given.
The REPAS Study: Reliability Evaluation of Passive Safety Systems
Authors: Ricotti M. E. , Bianchi F. , Burgazzi L. , D’Auria F. , Galassi G.
Published: March 2009

The strategy of approach to the problem moves from the consideration that a passive system should be theoretically more reliable than an active one. In fact it does not need any external input or energy to operate and it relies only upon natural physical laws (e.g. gravity, natural circulation, internally stored energy, etc.) and/or “intelligent” use of the energy inherently available in the system (e.g. chemical reaction, decay heat, etc.). Nevertheless the passive system may fail its mission not only as a consequence of classical mechanical failure of components, but also for deviation from the expected behaviour, due to physical phenomena mainly related to thermalhydraulics or due to different boundary and initial conditions. The main sources of physical failure are identified and a probability of occurrence is assigned. The reliability analysis is performed on a passive system which operates in two-phase, natural circulation. The selected system is a loop including a heat source and a heat sink where the condensation occurs. The system behavior under different configurations has been simulated via best-estimate code (Relap5 mod3.2). The results are shown and can be treated in such a way to give qualitative and quantitative information on the system reliability. Main routes of development of the methodology are also depicted.
NDE Performance (POD) Curves for Fatigue Cracks in Piping Based on Industry Performance Demonstration Data
Authors: Gosselin S. R. , Simonen F. A. , Heasler P. G. , Doctor S. R. , Becker F. L.
Published: March 2009

This paper evaluates non-destructive examination (NDE) detection capabilities of fatigue cracks in piping. Industry performance demonstration initiative (PDI) data for fatigue crack detection were used to develop a matrix of statistically based probability of detection (POD) curves that consider various NDE performance factors. Seven primary performance factors were identified — Material, Crack Geometry/Type, NDE Examination Access, NDE Procedure, Examiner Qualification, Pipe Diameter, and Pipe Wall Thickness. A database of 16,181 NDE performance observations, with 18 fields associated with each observation, was created and used to develop statistically based POD curves for 42 stainless steel and 14 carbon steel performance cases. Subsequent comparisons of the POD fits for each of the cases showed that excellent NDE performance for fatigue cracks can be expected for ferritic materials. Very little difference was observed between the POD curves for the 14 carbon steel performance cases considered in this study and NDE performance could therefore be represented by a single POD curve. For stainless steel, very good performance can also be expected for circumferential cracks located on the same side of the weld from which the NDE examination is made. POD depended primarily on component thickness. Three POD curves for stainless steel were prepared. Best estimate and the associated 95% confidence bounds POD verses through-wall depth logistic regression digital data are provided. Probabilistic fracture mechanics (PFM) calculations were performed to compare best estimate leak probabilities obtained from both the new performance-based POD curves and previous PFM models. This work was performed under joint funding by EPRI and the U.S. Department of Energy (DOE), Office of Nuclear Energy Science and Technology’s Nuclear Energy Plant Optimization (NEPO) program.
Transient Heat Transfer From a Horizontal Cylinder in the Low-Reynolds-Number Flow of Helium Gas
Authors: Liu Qiusheng , Fukuda Katsuya , Hata Koichi
Published: March 2009

The knowledge of forced convection transient heat transfer at various periods of exponentially increasing heat input to a heater is important as a database for understanding the transient heat transfer process in a high temperature gas cooled reactor (HTGR) due to an accident in excess reactivity. In this study, the transient heat transfer coefficients for Helium gas flowing perpendicular to a horizontal cylinder were measured in the low-Reynolds-number region. The platinum heater with a diameter of 1.0 mm was heated by electric current with an exponentially increasing heat input of Q0exp(t/τ). It was clarified that the heat transfer coefficient approaches the quasi-steady-state one for the period τ over around 1 s, and it becomes higher for the period of τ shorter than about 1 s. The transient heat transfer shows less dependent on the gas flowing velocity when the period becomes very short. Based on the experimental data, the ratio of transient heat transfer to the quasi-steady-state one was correlated as a function of Reynolds number of the gas flow and the non-dimensional period of increasing heat input. For the non-dimensional period larger than about 300, the transient heat transfer approaches the steady-state one, and shows no dependence on the Reynolds number.
Application of Ozone Chemical Decontamination (T-OZON ) to the Equipment for Disposal
Authors: Enda M. , Yaita Y. , Sakai H. , Aoi H. , Inami I. , Nakagami M. , Kani K.
Published: March 2009

The new decontamination process, T-OZON was applied to the decontamination of PLR pump impellers and bearings with casing covers. As the results, it was confirmed T-OZON was effective to reduce dose rate and secondary waste. The results make the further applications of T-OZON process on the decontamination for equipment disposal and system decontamination for primary coolant loops.
Fuel Rod Melt Progression Simulation Using Low-Temperature Melting Metal Alloy
Authors: Lee Seung Dong , Suh Kune Y. , Park GoonCherl , Lee Un Chul
Published: March 2009

The TMI-2 accident and various severe fuel damage experiments have shown that core damage is likely to proceed through various states before the core slumps into the lower head. Numerous experiments were conducted to address when and how the core can lose its original geometry, what geometries are formed, and in what processes the core materials are transported to the lower plenum of the reactor pressure vessel. Core degradation progresses along the line of clad ballooning, clad oxidation, material interaction, metallic blockage, molten pool formation, melt progression, and relocation to the lower head. Relocation into the lower plenum may occur from the lateral periphery or from the bottom of the core depending upon the thermal and physical states of the pool. Determining the quantities and rate of molten material transfer to the lower head is important since significant amounts of molten material relocated to the lower head can threaten the vessel integrity by steam explosion and thermal and mechanical attack of the melt. In this paper the focus is placed on the melt flow regime on a cylindrical fuel rod utilizing the LAMDA (Lumped Analysis of Melting in Degrading Assemblies) facility at the Seoul National University. The downward relocation of the molten material is a combination of the external film flow and the internal pipe flow. The heater rods are 0.8 m long and are coated by a low-temperature melting metal alloy. The electrical internal heating method is employed during the test. External heating is adopted to simulate the exothermic Zircaloy-steam reaction. Tests are conducted in several quasi-steady-state conditions. Given the variable boundary conditions including the heat flux and the water level, observation is made for the melting location, progression, and the mass of molten material. Finally, the core melt progression model is developed from the visual inspection and quantitative analysis of the experimental data. As the core material relocates downwards a blockage may be formed and grow both radially and axially. The velocity of the melt can be calculated from a force balance between the gravity and frictional losses at the melt-rod interface. When the heater rod is uncovered completely, the melt progression is initiated at the mid-point, which is the hot spot in the rod. However, the melting location is elevated as the water level rises because of the downward heat transfer. Considering the melt flow as a film, the steady-state film thickness on the cylindrical heater rod and the average velocity are computed. The steady-state film flow rate is determined in terms of the density, film thickness, and film velocity.
Comparative Study of Loss-of-Coolant Accident Using MAAP4.03 and RELAP5/MOD3.2.2
Authors: Park Chang Hwan , Lee Doo Yong , Jeong Ik , Lee Un Chul , Suh Kune Y. , Park Goon Cherl
Published: March 2009

Analysis was performed for a large-break loss-of-coolant accident (LOCA) in the APR1400 (Advanced Power Reactor 1400 MWe) with the thermal-hydraulic analysis code RELAP5/ MOD3.2.2 and the severe accident analysis code MAAP4.03. The two codes predicted different sequences for essentially the same initiating condition. As for the break flow and the emergency core cooling (ECC) flow rates, MAAP4.03 predicted considerably higher values in the initial stage than RELAP5/ MOD3.2.2. It was considered that the differing break flow and ECC flow rates would cause the LOCA sequences to deviate from one another between the two codes. Hence, the break flow model in MAAP4.03 was modified with partly implementing the two-phase homogeneous critical flow model and adopting a correction term. The ECC flow model in MAAP4.03 was also varied by changing the hardwired friction factor through the sensitivity study. The modified break flow and ECC flow models yielded more consistent calculational results between RELAP5/MOD3.2.2 and MAAP4.03. It was, however, found that the resultant effect is rather limited unless more mechanistic treatments are done for the primary system in MAAP4.03.
An Experimental Study of Critical Heat Flux in Narrow Gap With Two-Dimensional Slices
Authors: Kim Yong Hoon , Suh Kune Y. , Park Rae Joon , Kim San Baek , Kim Hee Dong
Published: March 2009

A cooling mechanism due to boiling in a gap between the debris crust and the reactor pressure vessel (RPV) wall was proposed for the TMI-2 reactor accident analysis. If there is enough heat transfer through the gap to cool the outer surface of the debris and the inner surface of the wall, the RPV wall may preserve its integrity during a severe core melt accident. If the heat removal through gap cooling relative to the counter-current flow limitation (CCFL) is pronounced, the safety margin of the reactor can be far greater than what had been previously known in the severe accident management arena. Should a severe accident take place, the RPV integrity will be maintained because of the inherent nature of degraded core coolability inside the lower head due to boiling in a narrow gap between the debris crust and the RPV wall. As a defense-in-depth measure, the heat removal capability by gap cooling coupled with external cooling can be examined for the Korean Standard Nuclear Power Plant (KSNPP) and the Advanced Power Reactor 1400MWe (APR1400) in light of the TMI-2 vessel survival. A number of studies were carried out to investigate the complex heat transfer mechanisms for the debris cooling in the lower plenum. However, these heat transfer mechanisms have not been clearly understood yet. The CHFG (Critical Heat Flux in Gap) experiments at KAERI were carried out to develop the critical heat flux (CHF) correlation in a hemispherical gap, which is the upper limit of the heat transfer. According to the CHFG experiments performed with a pool boiling condition, the CHF in a parallel gap was reduced by 1/30 compared with the value measured in the open pool boiling condition. The correlation developed from the CHFG experiment is based on the fact that the CHF in a hemispherical gap is governed by the CCFL and a Kutateladze type CCFL parameter correlates CCFL data well in hemispherical gap geometry. However, the results of the CHFG experiments appear to be limited in their value because the power of the heaters was restricted by the three-dimensional (3D) geometry. The two-dimensional (2D) geometry relative to the 3D geometry enables the heaters to produce higher power. Experiments were conducted to develop the CHF correlation for gap cooling with the 2D slices. The experimental facility consisted of a heater, a pressure vessel, a heat exchanger and the pressure and temperature measurement system. Tests were carried out in the pressure range of 0.1 to 1 MPa for the gap sizes of 1mm and 2mm using demineralized water.
Validating and Verifying a New Thermal-Hydraulic Analysis Tool
Authors: Schultz Richard R. , Weaver Walter L. , Ougouag Abderrafi M. , Wieselquist William A.
Published: March 2009

The Idaho National Engineering & Environmental Laboratory (INEEL) has developed a new analysis tool by coupling the Fluent computational fluid dynamics (CFD) code to the RELAP5-3D©/ATHENA advanced thermal-hydraulic analysis code. This tool enables researchers to perform detailed, three-dimensional analyses using Fluent’s CFD capability while the boundary conditions required by the Fluent calculation are provided by the balance-of-system model created using RELAP5-3D©/ATHENA. Both steady-state and transient calculations can be performed, using many working fluids and point to three-dimensional neutronics. A general description of the techniques used to couple the codes is given. The validation and verification (V&V) matrix is outlined. V&V is presently ongoing.
Study of Controlled Condensation Implosion Events
Authors: Seporaitis Marijus , Pabarcius Raimondas , Almenas Kazys
Published: March 2009

At LEI (Lithuanian Energy Institute) an experimental program has been initiated to investigate the ‘condensation implosion’ phenomena that can occur for horizontally stratified liquid-vapour flow conditions. The goal is understand the critical boundary conditions sufficiently so that the phenomenon can be controlled and initiated at will. After a reliable ‘pulser’ is developed, the follow up goal is to implement this unique component in a thermal-hydraulic system designed to perform certain tasks, e.g. to pump water or to transport energy passively in a downward direction. Experimental data obtained to data has shown that pulsers can be designed in which the vapour-liquid interface perturbation required for the initiation of condensation implosions is generated internally and depends solely on the rate at which liquid is supplied to the pulser. Data is presented which documents the conditions required for transition from a smooth to a wavy interface, and subsequently to an exponentially increasing surface distortion that culminates in a ‘condensation implosion’. The importance of the shear-stress generated by the condensation rate is illustrated.
The Effect of the Treatment of Localised Friction in Two-Phase Mixtures on the Stability of Parallel Channels
Authors: Abdou H. N. , Garea V. B. , Larreteguy A. E.
Published: March 2009

A one-dimensional analytical model has been developed to be used for the linear analysis of density-wave oscillations in a parallel heated channel. The heated channel is divided into a single-phase and a two-phase region. The two-phase region is represented by the homogeneous model. The localised friction at the channel exit is treated considering the two-phase mixture. The exact equation for the total channel pressure drop is perturbed around the steady state. The stability characteristics of the heated channel are investigated using the Nyquist criterion. The marginal stability boundary (MSB) is determined in the two-dimensional thermodynamic equilibrium space parameters, the subcooled boiling number and the phase change number. The predictions of the model are compared with experimental results published in open literature. The results indicate a more stable system with (1) low system pressure, (2) high inlet restriction, (3) low outlet restriction, and (4) high inlet velocity. The results show that the model agrees well with the available experimental data. In particular, the results show the significance of correcting the localised friction due to the presence of the two-phase mixture in the two-phase region: explicit inclusion of the two-phase localised friction improves the agreement with experimental results. This effect is more important for high heating power and high inlet subcooling.
A Foundation for Stressor-Based Prognostics for Next Generation Systems
Authors: Jarrell Don , Sisk Daniel , Bond Leonard
Published: March 2009

Pacific Northwest National Laboratory (PNNL) scientists are performing research under the Department of Energy Nuclear Energy Research Initiative (NERI) program, to develop a methodology for accurate identification and prediction of equipment faults in critical machinery. The 3-year project, on-line intelligent self-diagnostic monitoring system (SDMS) for next generation nuclear power plants is scheduled for completion at the end of FY 2002. The research involves running machinery to failure in the Laboratory by the introduction of intentional faults. During testing, advanced diagnostic/prognostic sensors and analysis systems monitor the equipment stressor levels, correlate them with expected degradation rates, and predict the resulting machinery performance levels and residual lifetime. Application of a first principles physics-based approach is expected to produce prognostic methodologies of significantly higher accuracies than are currently available. This paper reviews the evolution and current state of the maintenance art. It presents a key measurement philosophy that results from the use of condition based maintenance (CBM) as a fundamental investigative precept, and explains how this approach impacts degradation and failure measurement and prediction accuracy. It then examines how this measurement approach is applied in sensing and correlating pump stressors with regard to degradation rate and time to equipment failure. The specifics are examined on how this approach is being applied at PNNL to cavitation and vibration phenomena in a centrifugal pump. Preliminary vibration analysis results show an excellent correspondence between the (laser) motor position indication, the vibration response, and the dynamic force loading on the bearings. Orbital harmonic vibratory motion of the pump and motor appear to be readily correlated through the FFTs of all three sensing systems.
Studies on Applicability of Smoke Sensors as a Sodium Leak Monitoring System
Authors: Morizono Koji , Maruta Yuji , Ikeda Makinori
Published: March 2009

The prototype sodium-cooled FBR MONJU was shut down due to a sodium leak incident on Dec. 8, 1995, and has remained shut-down since. Currently, a safety assessment of the proposed new countermeasures against future leaks is being carried out. One of these countermeasures, a new sodium leak monitoring system to detect small scale leakage with high reliability, will be installed throughout the entire secondary cooling circuit area. An initial assessment suggested that smoke sensors would best meet the requirements for this monitoring system from the standpoints of sensitivity, reliability, and installation cost. To confirm the suitability of smoke sensors for sodium leak monitoring, their sensitivity to sodium aerosol has been measured experimentally, and their ability to detect small scale sodium leakage has been estimated by analytical methods.
Evaluation of NDE Round-Robin Exercises Using the NRC Steam Generator Mockup at Argonne National Laboratory
Authors: Muscara Joseph , Kupperman David S. , Bakhtiari Sasan , Park Jang-Yul , Shack William J.
Published: March 2009

This paper discusses round-robin exercises using the NRC steam generator (SG) mock-up at Argonne National Laboratory to assess inspection reliability. The purpose of the round robins was to assess the current reliability of SG tubing inservice inspection, determine the probability of detection (POD) as function of flaw size or severity, and assess the capability for sizing of flaws. For the round robin and subsequent evaluation completed in 2001, eleven teams participated. Bobbin and rotating coil mock-up data collected by qualified industry personnel were evaluated. The mock-up contains hundreds of cracks and simulations of artifacts such as corrosion deposits and tube support plates that make detection and characterization of cracks more difficult in operating steam generators than in most laboratory situations. An expert Task Group from industry, Argonne National Laboratory, and the NRC have reviewed the signals from the laboratory-grown cracks used in the mock-up to ensure that they provide reasonable simulations of those obtained in the field. The mock-up contains 400 tube openings. Each tube contains nine 22.2-mm (7/8-in.) diameter, 30.5-cm (1-ft) long, Alloy 600 test sections. The flaws are located in the tube sheet near the roll transition zone (RTZ), in the tube support plate (TSP), and in the freespan. The flaws are primarily intergranular stress corrosion cracks (axial and circumferential, ID and OD) though intergranular attack (IGA) wear and fatigue cracks are also present, as well as cracks in dents. In addition to the simulated tube sheet and TSP the mock-up has simulated sludge and magnetite deposits. A multiparameter eddy current algorithm, validated for mock-up flaws, provided a detailed isometric plot for every flaw and was used to establish the reference state of defects in the mock-up. The detection results for the 11 teams were used to develop POD curves as a function of maximum depth, voltage and the parameter mp , for the various types of flaws. The POD curves were represented as linear logistic curves, and the curve parameters were determined by the method of Maximum Likelihood. The effect of both statistical uncertainties inherent in sampling from distributions and the uncertainties due to errors in the estimates of maximum depth and mp was investigated. The 95% one-sided confidence limits (OSL), which include errors in maximum depth estimates, are presented along with the POD curves. For the second round robin a reconfigured mock-up is being used to evaluate the effectiveness of eddy current array probes. The primary emphasis is on the X-Probe. Progress with the X-Probe round robin is discussed in this paper.
CFD Predictions of Severe Accident Steam Generator Flows in a 1/7th Scale Pressurized Water Reactor
Authors: Boyd Christopher , Hardesty Kelly
Published: March 2009

Computational Fluid Dynamics (CFD) is applied to steam generator inlet plenum mixing as part of a larger plan covering steam generator tube integrity. The technique is verified by comparing predicted results with severe accident natural circulation data [1] from a 1/7th scale Westinghouse facility. This exercise demonstrates that the technique can predict the natural circulation and mixing phenomena relevant to steam generator tube integrity issues. The model includes primary side flow paths for a single hot leg and steam generator. Qualitatively, the experimentally observed flow phenomena are predicted. The paths of the natural circulation flows and the relative flow proportions are correctly predicted. Quantitatively, comparisons are made with temperatures, mass flows, and other parameters. All predictions are generally within 10% of the experimental values. Overall, there is a high degree of confidence in the CFD technique for prediction of the relevant flow phenomena associated with this type of severe accident sequence.
Containment Fire Simulation by a CFD Code
Authors: Heitsch Matthias
Published: March 2009

In the frame of an international collaborative project to evaluate fire models a code benchmark was initiated to better quantify the strengths and weaknesses of the codes involved. CFX has been applied to simulate selected cases of both parts of the benchmark. These simulations are presented and discussed in this paper. In the first part of the benchmark a pool fire just represented by a heat release table is considered. Consequently, the physical fire model within CFX is simple. Radiative heat exchange together with turbulent mixing are involved. Two cases with and without venting of the fire room are compared. The second part of the benchmark requires a more detailed fire model in order to inspect the availability of oxygen locally and to control the fire intensity. Under unvented conditions oxygen starvation is encountered and the fire oscillates. Mechanical ventilation changes this behavior and provides enough oxygen all over the simulation time. The predefined damage criteria to characterize, if a target cable in the fire room would be damaged, are not met. However, surface temperatures predicted are well above the assumed threshold temperatures. A continuation of the work presented is foreseen and will address a more complex physical modeling of relevant fire scanarios.
Evaluation of Pipe Whip Impacts on Neighboring Piping and Walls of the Ignalina Nuclear Power Plant
Authors: Dundulis Gintautas , Kulak Ronald F. , Marchertas Algirdas , Narvydas Evaldas , Petry Mark C. , Uspuras Eugenijus
Published: March 2009

Presented in this paper is the transient analysis of a Group Distribution Header (GDH) following a guillotine break at the end of the header. The GDH is the most important component of reactor safety in case of accidents. Emergency Core Cooling System (ECCS) piping is connected to the GDH piping such that, during an accident, coolant passes from the GDH into the ECCS. The GDH that is propelled into motion after a guillotine break can impact neighboring GDH pipes or the nearest wall of the compartment. Therefore, two cases are investigated: • GDH impact on an adjacent GDH and its attached piping; • GDH impact on an adjacent reinforced concrete wall. A whipping RBMK-1500 GDH along with neighboring concrete walls and pipelines is modeled using finite elements. The finite element code NEPTUNE used in this study enables a dynamic pipe whip structural analysis that accommodates large displacements and nonlinear material characteristics. The results of the study indicate that a whipping GDH pipe would not significantly damage adjacent walls or piping and would not result in a propagation of pipe failures.
Study on Enhancement of Sub-Cooled Flow Boiling Heat Transfer and Critical Heat Flux of Solid-Water Two-Phase Mixture
Authors: Koizumi Yasuo , Suzuki Tomoyuki , Ohtake Hiroyasu
Published: March 2009

The influence of particle introduction into a subcooled water flow on boiling heat transfer and critical heat flux (CHF) was examined. When the water velocity was low, the particles crowded on the bottom wall of the flow channel and flowed just like sliding on the wall. When the water velocity was high, the particles were well dispersed in the water flow. In the non-boiling region, the heat transfer was augmented by the introduction of the particles into the water flow. As the introduction of the particles were increased, the augmentation was also increased in the high water flow rate region. However, it was independent upon the particle introduction rate in the low water flow rate region. The onset of boiling was delayed by the particle inclusion. The boiling heat transfer was enhanced by the particles. However, it was rather decreased in the high heat flux fully-developed-boiling region. The CHF was decreased by the particle inclusion in the low water flow region and was not affected in the high water flow region.
Mass/Energy Release Influence to Thermal Hydraulics of Accident Localisation System at Ignalina NPP in Case of MDBA
Authors: Urbonavicius E. , Rimkevicius S.
Published: March 2009

The maximal loads on ALS structures could be reached during Maximum Design Basis Accident (MDBA). MDBA for Ignalina NPP is defined as guillotine rupture of the pressure header of main circulation pumps (MCP). In this paper the influence of calculated mass and energy release in case of MCP pressure header rupture on behavior of thermal hydraulic parameters in ALS compartments is presented. The performed analysis demonstrated the importance of mass/energy release for the calculated highest pressure in ALS. Nevertheless, the maximal calculated pressures in ALS compartments are below the design pressures even in case of most conservative mass/energy release through the break.
Benchmarking Computational Fluid Dynamics for Application to PWR Fuel
Authors: Smith L. D. , Conner M. E. , Liu B. , Dzodzo B. , Paramonov D. V. , Beasley D. E. , Langford H. M. , Holloway M. V.
Published: March 2009

The present study demonstrates a process used to develop confidence in Computational Fluid Dynamics (CFD) as a tool to investigate flow and temperature distributions in a PWR fuel bundle. The velocity and temperature fields produced by a mixing spacer grid of a PWR fuel assembly are quite complex. Before using CFD to evaluate these flow fields, a rigorous benchmarking effort should be performed to ensure that reasonable results are obtained. Westinghouse has developed a method to quantitatively benchmark CFD tools against data at conditions representative of the PWR. Several measurements in a 5×5 rod bundle were performed. Lateral flowfield testing employed visualization techniques and Particle Image Velocimetry (PIV). Heat transfer testing involved measurements of the single-phase heat transfer coefficient downstream of the spacer grid. These test results were used to compare with CFD predictions. Among the parameters optimized in the CFD models based on this comparison with data include computational mesh, turbulence model, and boundary conditions. As an outcome of this effort, a methodology was developed for CFD modeling that provides confidence in the numerical results.
Level 1 PSA of an ABWR Plant With Digital Protection Systems in Japan
Authors: Aida Toshihiro , Kikuchi Hiromu , Sugawara Masajiro , Hirose Masao
Published: March 2009

The purpose of this paper is to provide the quantitative evaluation method with regard to the Digital Protection systems (DPSs) that were newly adopted in the ABWR plants and to show the results of level1 PSA of the ABWR plants.
Emergency Diesel Generator System Surveillance Test Policy Optimization Through Genetic Algorithms Using Non-Periodic Intervention Frequencies and Seasonal Constraints
Authors: Lapa Celso M. F. , Pereira Cla´udio M. N. A. , Frutuoso e Melo P. F.
Published: March 2009

Nuclear standby safety systems must frequently, be submitted to periodic surveillance tests. The main reason is to detect, as soon as possible, the occurrence of unrevealed failure states. Such interventions may, however, affect the overall system availability due to component outages. Besides, as the components are demanded, deterioration by aging may occur, penalizing again the system performance. By these reasons, planning a good surveillance test policy implies in a trade-off between gains and overheads due to the surveillance test interventions. In order maximize the systems average availability during a given period of time, it has recently been developed a non-periodic surveillance test optimization methodology based on genetic algorithms (GA). The fact of allowing non-periodic tests turns the solution space much more flexible and schedules can be better adjusted, providing gains in the overall system average availability, when compared to those obtained by an optimized periodic tests scheme. The optimization problem becomes, however, more complex. Hence, the use of a powerful optimization technique, such as GAs, is required. Some particular features of certain systems can turn it advisable to introduce other specific constraints in the optimization problem. The Emergency Diesel Generation System (EDGS) of a Nuclear Power Plant (N-PP) is a good example for demonstrating the introduction of seasonal constraints in the optimization problem. This system is responsible for power supply during an external blackout. Therefore, it is desirable during periods of high blackout probability to maintain the system availability as high as possible. Previous applications have demonstrated the robustness and effectiveness of the methodology. However, no seasonal constraints have ever been imposed. This work aims at investigating the application of such methodology in the Angra-II Brazilian NPP EDGS surveillance test policy optimization, considering the blackout probability growth during summer, due to the electrical power demand increase. Here, the model used penalizes test interventions by a continuous modulating function, which depends on the instantaneous blackout probability. Results have demonstrated the ability of the method in adapting the surveillance tests policy to seasonal behaviors. The knowledge acquired by the GA during the searching process has lead to test schedules that drastically minimize the test interventions at periods of high blackout probability. It is compensated by more frequent tests redistributed through the periods of low blackout probability, in order to provide improvement on the overall average availability at the system level.
IRIS: Proceeding Towards the Preliminary Design
Authors: Carelli Mario D. , Miller K. , Lombardi C. , Todreas N. , Greenspan E. , Ninokata H. , Lopez F. , Cinotti L. , Collado J. , Oriolo F. , Alonso G. , Moraes M. , Boroughs R. , Barroso A. , Ingersoll D. , Cavlina N.
Published: March 2009

The IRIS (International Reactor Innovative and Secure) project has completed the conceptual design phase and is moving towards completion of the preliminary design, scheduled for the end of 2002. Several other papers presented in this conference provide details on major aspects of the IRIS design. The three most innovative features which uniquely characterize IRIS are, in descending order of impact: 1. Safety-by-design, which takes maximum advantage of the integral configuration to eliminate from consideration some accidents, greatly lessen the consequence of other accident scenarios and decrease their probability of occurring; 2. Optimized maintenance, where the interval between maintenance shutdowns is extended to 48 months; and 3. Long core life, of at least four years without shuffling or partial refueling. Regarding feature 1, design and analyses will be supplemented by an extensive testing campaign to verify and demonstrate the performance of the integral components, individually as well as interactive systems. Test planning is being initiated. Test results will be factored into PRA analyses under an overall risk informed regulation approach, which is planned to be used in the IRIS licensing. Pre-application activities with NRC are also scheduled to start in mid 2002. Regarding feature 2, effort is being focused on advanced online diagnostics for the integral components, first of all the steam generators, which are the most critical component; several techniques are being investigated. Finally, a four year long life core design is well underway and some of the IRIS team members are examining higher enrichment, eight to ten year life cores which could be considered for reloads.
Energy System Expectations for Nuclear in the 21st Century: A Plausible Range
Authors: Langlois Lucille M. , McDonald Alan , Rogner Hans-Holger , Vera Ivan
Published: March 2009

This paper outlines a range of scenarios describing what the world’s energy system might look like in the middle of the century, and what nuclear energy’s most profitable role might be. The starting point is the 40 non-greenhouse-gas-mitigation scenarios in the Special Report on Emissions Scenarios (SRES) of the Intergovernmental Panel on Climate Change (IPCC, 2000). Given their international authorship and comprehensive review by governments and scientific experts, the SRES scenarios are the state of the art in long-term energy scenarios. However, they do not present the underlying energy system structures in enough detail for specific energy technology and infrastructure analyses. This paper therefore describes initial steps within INPRO (the International Project on Innovative Nuclear Reactors and Fuel Cycles of the International Atomic Energy Agency) to translate the SRES results into a range of possible nuclear energy technology requirements for mid-century. The paper summarizes the four SRES scenarios that will be used in INPRO and the reasons for their selection. It provides illustrative examples of the sort of additional detail that is being developed about the overall energy system implied by each scenario, and about specific scenario features particularly relevant to nuclear energy. As recommended in SRES, the selected scenarios cover all four SRES “storyline families.” The energy system translations being developed in INPRO are intended to indicate how energy services may be provided in mid-century and to delineate likely technology and infrastructure implications. They will indicate answers to questions like the following. The list is illustrative, not comprehensive. • What kind of nuclear power plants will best fit the mid-century energy system? • What energy forms and other products and services provided by nuclear reactors will best fit the mid-century energy system? • What would be their market shares? • How difficult will it be to site new nuclear facilities? • Which are nuclear energy’s biggest competitors? • Which non-nuclear technologies can nuclear power complement? • What is the range of potential demand growth for new capacity? • How is demand growth distributed geographically around the world? Different scenarios imply different answers, which are then the starting point for estimating what future reactor users might require of reactor and fuel cycle designs around mid-century. These user requirements — in terms of economics, safety, proliferation resistance, waste, and environmental impacts — are intended to help establish key directions in which to encourage innovation. They are intended as a useful input to managers designing R&D strategies targeted on the anticipated energy system needs, and other relevant needs, of mid-century.