bluuebunny/crossref_metadata_2025 huggingface dataset:
Assessment and Management of Aging in Phenix Nuclear Power Plant
Authors: Dumarcher V. , Bourrier J. L. , Chaucheprat P. , Boulegue D.
Published: September 2008

The combination of one or several processes of ruins can involve the materials failure of a nuclear power plant. These processes arise from the external agents action such as the pressure, the mechanical efforts, the heat flows and the radiations constitute the whole of the “actions” of the surrounding medium. The prolongation and the repetition of these effects can involve a deterioration of the machine. In accordance with the decree of February 26, 1974, the PWR operator must be firstly, sure that the system is controlled according to the situations considered in the file of dimensioning and secondly, be able to know anytime the life of the equipment. The physical phenomena which cause the structures ruin are less complex in the PWR than in the SFR. In the SFR, the high temperatures imposed on components for long periods can involve a significant creep. In the course of time, this deformations accelerate the release of fatigue cracks. To consider the creep, the reactor lifespan is correlated at the numbers of thermals transients envisaged initially. To realize the management of aging in Phenix power plant, it is necessary to carry out an individualized monitoring of the structures and not only on the vessel. We must ensure the good state and/or the correct operation of the significant stations for safety which are the control of the reactivity, the movement of control rods, the primary sodium containment and the decay heat removal. For that, we monitor the main vessel, the conical skirt, the IHX and the Core Cover Plug. A profound knowledge of the thermal transients of the past is necessary to carry out an effective assessment. In order to guarantee that any harmful situation is well taken into the management of aging, we monitor permanently certain measurements (primary and secondary pump speed, hot and cold pool temperatures, IHX-main vessel and reactor roof temperatures). We present in the article the scientific method used in the Physics Section. A logical diagram specific to the type of situation and the structure allows to associate the harmful transient at a identical situation which has been happened in the past. During the last two cycles, the nuclear power plant has sustained 34 startup (20 during the 51st cycle and 14 during the 52nd cycle). After two cycles of operation, there is approximately 70 to 80% of occurrences authorized for the whole of the structures. For the last 4 cycles, the number of transients to come will remain quite lower than the number dimensioned initially.
Supercritical Flow Stability in Two Parallel Channels
Authors: Chatoorgoon Vijay
Published: September 2008

An analytical study of supercritical flow stability in two parallel channels is reported here. This would be of immense value to new reactor designs that propose to use supercritical light water on the primary side. The finding is that two-phase flow instability and supercritical flow instability are not identical, as there are notable phenomenological differences as well as mathematical differences.
A Process Model for the Production of Hydrogen Using High Temperature Electrolysis
Authors: McKellar Michael G. , Harvego Edwin A. , Richards Matt , Shenoy Arkal
Published: September 2008

High temperature electrolysis (HTE) involves the splitting of steam into hydrogen and oxygen at high temperatures. The primary advantage of HTE over conventional low temperature electrolysis is that considerably higher hydrogen production efficiencies can be achieved. Performing the electrolysis process at high temperatures results in more favorable thermodynamics for electrolysis, more efficient production of electricity, and allows direct use of process heat to generate steam. This paper presents the results of process analyses performed to evaluate the hydrogen production efficiencies of an HTE plant coupled to a 600 MWt Modular Helium Reactor (MHR) that supplies both the electricity and process heat needed to drive the process. The MHR operates with a coolant outlet temperature of 950 C. Approximately 87% of the high-temperature heat is used to generate electricity at high efficiency using a direct, Brayton-cycle power conversion system. The remaining high-temperature heat is used to generate a superheated steam / hydrogen mixture that is supplied to the electrolyzers. The analyses were performed using the HYSYS process modeling software. The model used to perform the analyses consisted of three loops; a primary high temperature helium loop, a secondary helium loop and the HTE process loop. The detailed model included realistic representations of all major components in the system, including pumps, compressors, heat exchange equipment, and the electrolysis stack. The design of the hydrogen production process loop also included a steam-sweep gas system to remove oxygen from the electrolysis stack so that it can be recovered and used for other applications. Results of the process analyses showed that hydrogen production efficiencies in the range of 45% to 50% are achievable with this system.
Development of Pulse Ultrasonic Doppler Method for Flow Rate Measurements of Power Plant: Characteristics of Sound Pressure Distributions and Evaluation of the Hybrid Ultrasonic Flowmetering System Using TOF and UDM
Authors: Kikura Hiroshige , Inoue Yuto , Aritomi Masanori , Mori Michitsugu
Published: September 2008

A multi-beam pulse ultrasonic Doppler method has been developed for a new type of flow metering system. This new system is a hybrid of the time-of-flight type ultrasonic flowmeter and the ultrasonic velocity profile type flowmeter, having the advantages of these two types. Our final purpose is to apply the hybrid ultrasonic flow metering system to an accurate flow rate measurement of feed- or recirculation- water in nuclear power plants. The pulse ultrasonic Doppler method (UDM) has the capability to obtain instantaneous velocity profiles along an ultrasonic beam. The principle of the UDM flowmeter, which is one of the ultrasonic velocity profile type flowmeters, is based on the integration of an instantaneous velocity profile over a pipe diameter. The multi-beam system is expected to eliminate installation problems such as those of entry length, and also to follow transient flow rate more precisely by increasing the number of ultrasonic transducers. However, it needs reflectors for receiving ultrasonic Doppler signals. On the other hand, the time-of-flight (TOF) ultrasonic flow metering system does not need any reflector, but it needs profile factors (PFs) which depend on velocity profiles. PF is one of the important experimental coefficients for the accurate flow rate measurement. Therefore PFs must be corrected according to the changes in flow conditions. In the present study, we investigated to what degree the hybrid ultrasonic flow metering system can adjust the profile factors of the time-of-flight ultrasonic flow meters by using the multi-beam pulse ultrasonic Doppler method in metallic wall piping.
Detecting Structural Damage of Nuclear Power Plant by Interactive Data Mining Approach
Authors: Shu Yufei
Published: September 2008

This paper presents a nonlinear structural damage identification technique, based on an interactive data mining approach, which integrates a human cognitive model in a data mining loop. A mining control agent emulating human analysts is developed, which directly interacts with the data miner, analyzing and verifying the output of the data miner and controlling the data mining process. Additionally, an artificial neural network method, which is adopted as a core component of the proposed interactive data mining method, is evolved by adding a novelty detecting and retraining function for handling complicated nuclear power plant quake-proof data. Plant quake-proof testing data has been applied to the system to show the validation of the proposed method.
Conditioning Operation by an Encapsulation and Shielding of Spent Sealed Radium Sources
Authors: Kang Il-Sik , Hong Dae-Seok , Kim Kil-Jeong , Shon Jong-Sik , Hong Kwon-Pyo
Published: September 2008

In order to avoid accidents that could result from an improper storage of spent radium sources, it is necessary to condition and store them safely. The program for a safe conditioning of spent radium sources by the IAEA (International Atomic Energy Agency) has been established to assist the developing countries. The main object of this paper is to summarize the technology that was adapted by the IAEA for the conditioning of spent radium sources in the national inventory of Ra-226 sources in member states and the actions performed by the Korean expert team as a part of the IAEA’s project titled as ‘Radium Conditioning in Southeastern Asia’. The whole inventory of the spent radium sources of 8,671.13mCi was safely conditioned by the Korean expert team according to the guidelines under the supervision of the IAEA’s technical officer. The 1,821 sources of spent radium were encapsulated, welded, and conditioned into 18 concrete-shielded drums.
Fluid-Elastic Instability in Tube Bundles and Effect of Flow Regime Transition
Authors: Chu In-Cheol , Chung Heung June , Yun Young Jung
Published: September 2008

Fluid-elastic instability characteristics in an air-water two-phase cross-flow have been experimentally investigated using two different arrangements of cantilevered straight tube bundles. Rotated triangular array tube bundle is for the supplementary test of the existing work, and normal square array tube bundle is for the investigation of fluid-elastic instability in higher p/d condition. The present paper provides the experimental results of the tube vibration response, hydrodynamic mass, damping ratio, and fluid-elastic instability. As the two-phase gap velocity increased, the fluidic-elastic instability occurred in the lift direction and a strongly coupled tube motion was found. The damping ratio was very dependent on the void fraction, as in the previous works. For a low void fraction flow, the fluid-elastic instability could be predicted by using Connors’ equation. However, the fluid-elastic instability in a high void fraction flow was quite different. The transition between the two fluid-elastic instability regions almost coincided with the flow regime transition criteria from a continuous bubbly flow to an intermittent flow.
Validation of the TRACE Code Against Post-Dryout Experiments in Tubes and Annuli
Authors: Anghel Ionut Gheorghe , Anglart Henryk
Published: September 2008

The present paper presents results of the TRACE code assessment against post dryout experimental data obtained in tubes and annuli. The investigations have been focused on the experiments carried out at 70 bar pressure, 10 K inlet subcooling and the mass flux variation between 500–2000 kg/(m2s). Various axial power distributions (uniform, inlet peaked, middle peaked, and outlet peaked) have been used in the tube geometry. Uniform power distributions, with various ratios between the inner and the outer power have been used in the annulus geometry. The validation results indicated that the discrepancies between measured and calculated data are in a range of ±20%. In addition a sensitivity study has been performed in which the influence of several parameters (including cell size and boundary conditions) on computational results have been investigated.
Temperature Dependency of Beremin’s Parameters for 20MnMoNi5-5 Material
Authors: Dutta B. K. , Guin S. , Sahu M. K. , Samal M. K.
Published: September 2008

The mechanism of cleavage fracture in reactor structural material is significantly different from ductile fracture. The cleavage fracture in a material usually originates from micro-cracks, which are formed by different mechanisms. The micro-cracks are formed due to non-homogeneous distribution of plastic deformation within the grains, called slip-initiated cleavage. The cracked grain boundary carbides also are the sources to originate the micro-cracks. This occurs when the stress normal to the planes of carbide particles is sufficiently high. The fracture takes place by the formation of micro-cracks and their extension with little global plastic deformation. The cleavage process is stress controlled and consumes little deformation energy and hence, the crack grows unstably fast. In this case the local fracture criterion is generally based on a critical cleavage stress, which is statistical in nature. Based on the weakest link assumption and Weibull statistics, Beremin developed a model for analyzing brittle fracture process by local approach. The two parameters of the model are material properties and can be determined from notched tensile tests at low temperatures and their finite element analysis. Once these parameters are determined, these can be used to predict probability of cleavage fracture initiation in a component. In the present work, temperature dependency of Beremin’s parameters ‘m’ and ‘σu’ has been investigated for 20MnMoNi55 structural material. Tensile tests have been conducted on specimens having central groove at two different −50 °C and −100 °C temperatures. Thirty specimens have been tested in each set. Image processing technique has been used for measuring diametrical contraction from on-line images. The analytical study has been carried out to determine Beremin’s parameters using the statistical variation in load v/s diametrical contraction. For this purpose, finite element analysis has been carried out to determine Weibull stress during fracture. The experimental probability of failure of each specimen has been calculated by ranking them based on the diametrical contraction at the time of fracture. Such a table along with the corresponding Weibull stresses have been used to find out Beremin’s parameters using “maximum likelihood method”. The temperature dependency of these parameters is then investigated for the present material.
An Integration of the Restructured MELCOR for the MIDAS Computer Code
Authors: Park Sunhee , Kim Dong Ha , Kim Ko Ryu , Chol Song-Won
Published: September 2008

The developmental need for a localized severe accident analysis code is on the rise. KAERI is developing a severe accident code called MIDAS, which is based on MELCOR. In order to develop the localized code (MIDAS) which simulates a severe accident in a nuclear power plant, the existing data structure is reconstructed for all the packages in MELCOR, which uses pointer variables for data transfer between the packages. During this process, new features in FORTRAN90 such as a dynamic allocation are used for an improved data saving and transferring method. Hence the readability, maintainability and portability of the MIDAS code have been enhanced. After the package-wise restructuring, the newly converted packages are integrated together. Depending on the data usage in the package, two types of packages can be defined: some use their own data within the package (let’s call them independent packages) and the others share their data with other packages (dependent packages). For the independent packages, the integration process is simple to link the already converted packages together. That is, the package-wise structuring does not require further conversion of variables for the integration process. For the dependent packages, extra conversion is necessary to link them together. As the package-wise restructuring converts only the corresponding package’s variables, other variables defined from other packages are not touched and remain as it is. These variables are to be converted into the new types of variables simultaneously as well as the main variables in the corresponding package. Then these dependent packages are ready for integration. In order to check whether the integration process is working well, the results from the integrated version are verified against the package-wise restructured results. Steady state runs and station blackout sequences are tested and the major variables are found to be the same each other. In order to verify the results, the integrated results were compared with the restructured results for each package. Some sequences were calculated such as a steady state and SBO (Station Blackout) accident. The major variables were the same as well as the graph trends. Through out the integrating process, the base was constructed for a code improvement and an addition of new models. The integrating process proposed in this paper will be extended to the T/H and F/P packages for the MIDAS development program.
Inherent Boron Dilution Safety Issue in the French Pressurized Water Reactor: CFD Approach
Authors: Graffard Estelle , Goux Fre´de´ric
Published: September 2008

Inherent boron dilution can occur in case of a Small Break LOCA when low borated water is mainly accumulated in the U-legs due to reflux boiling in the Steam Generator tubes after the loss of natural circulation. The restart of the natural circulation may lead to criticality because of the injection of these low borated slugs towards the core. To evaluate this potential risk, the boron concentration at the core inlet has to be known which makes necessary to estimate the mixing phenomena in the cold leg, in the downcomer and in the lower plenum: CFD calculations are required. First of all the validation of CFX5 CFD code on the relevant phenomena of inherent boron dilution has been established (UPTF TRAM C3 test). Then, an application to the 900 MW French Pressurized Water Reactor series has been performed.
Entrainment of Water Around a Single Rod Immersed in Water Pool with Gas Jet Impingement
Authors: Soga Kazuo , Niikura Hideto , Sugiyama Ken-ichiro , Narabayashi Tadashi , Ohshima Hiroyuki , Suda Kazunori
Published: September 2008

In a steam generator of liquid sodium cooled fast breeder reactor, the occurrence of secondary heat transfer tube failure has been considered due to overheating in sodium-water reaction. A computer code SERAPHIM has been developed to analyze this kind of secondary heat transfer tube failure due to steam jet with heat generation and chemical reaction by Japan Atomic Energy Agency (JAEA). The detailed experimental data that verify the adequacy of SERAPHIM code and upgrade it, are now required. In ICONE13, the local and mean heat transfer coefficients around a horizontal heated rod (φ 15mm) immersed in a water pool and in a sodium pool with gas jet impingement without chemical reaction, were reported as the first step. It was confirmed in the water pool experiment that the local and mean heat transfer coefficients slowly increase with increasing the Ar jet velocity from 8.7m/s to 78m/s The local and mean Nusselt numbers almost keeps same values independent on the Ar gas velocity in the sodium pool experiment. From these results, we inferred that the ambient liquid is entrained inside and contributes to high heat transfer rates. A series of experiments that investigate the entrainment process of ambient liquid toward jet interior are carried out by using a laser-sheet visualization and a void meter in water pool in the present work. It was observed that the entrainment of water into Ar gas jet is constantly caused in two regions just above the nozzle and just below the single rod. In the region just above the nozzle, negative pressure causes the entrainment of water. In the region below the rod, the entrainment of water is caused because the preceding Ar gas jet is caught up by the succeeding gas jet. The basic behavior of Ar gas jet causing the entrainment of water was confirmed to be almost same over the Reynolds number range of Ar gas jet, 2.17×103 to 2.17×104, in the present study. In addition, the measurement of void fraction was performed to investigate the entrainment of water quantitatively and the behavior of local heat transfer coefficients around a single rod with gas jet impingement in water pool was made clear.
Consensus Based Nuclear Public-Hearing System Model
Authors: Lee Young Wook , Kim Suk Hoon , Cho Young Ho , Ko Hyun Seok , Shin Dong Hoon , Moon Joo Hyun , Kang Chang Sun
Published: September 2008

Although the government admit the benefit of construction of a nuclear facility for national electric source, related policy could be developed and carried out only if the public, especially who have some stake on it, recognize the benefit and accept the policy. For public participation, Korea has a system of public-hearing in accordance with the law. Because of the absence of the detailed way for public opinion aggregation and for the reflection of the aggregated opinion, Korean public-hearing system is only a conceptual model. Therefore, some specific system for Korean Public-Hearing should be developed and applied. In this study, to share the right of decision making, which is an ultimate concept for public participation, decision making components and the characteristics of each phase are analyzed. The criteria weight for assessment and comparison with alternatives are founded as a valuation factor of the decision making components, which should be based on the social consensus. On these foundations, a system for aggregation and reflection of the public opinion was proposed. The system named “CPDM” (Consensus based Participatory Decision Making) has three authority groups for decision making. At first, “advisory experts group” play a role for the technical assessment and the serve utility value on the criteria for each alternatives. Next, “participatory deliberation group” play a role for consensus building on the relative-importance (weight) between the criteria by feedback to promote degree of consensus. Lastly including gentlemen of the long robe, “expert group for decision making” paly a role to reflect the utility and weight and make a decision with agreement for performance of it. Also, in this study, a mathematical model for the quantification of the degree of consensus was conceptualized using Ordered Weighted Averaging (OWA) aggregation operator and fuzzy similarity theory, which is a comparison concept. Since this model enables influence of each criteria and each participant on collective consensus to be analyzed, a direction to promote consensus building can be derived. That is to say, this model can support consensus building and promote public acceptance for the nuclear industry and related policy.
Robustness of a Neural Network Model for Power Peak Factor Estimation in Protection Systems
Authors: Souza Rose Mary G. P. , Moreira Joa˜o M. L.
Published: September 2008

This work presents results of robustness verification of artificial neural network correlations that improve the real time prediction of the power peak factor for reactor protection systems. The input variables considered in the correlation are those available in the reactor protection systems, namely, the axial power differences obtained from measured ex-core detectors, and the position of control rods. The correlations, based on radial basis function (RBF) and multilayer perceptron (MLP) neural networks, estimate the power peak factor, without faulty signals, with average errors between 0.13%, 0.19% and 0.15%, and maximum relative error of 2.35%. The robustness verification was performed for three different neural network correlations. The results show that they are robust against signal degradation, producing results with faulty signals with a maximum error of 6.90%. The average error associated to faulty signals for the MLP network is about half of that of the RBF network, and the maximum error is about 1% smaller. These results demonstrate that MLP neural network correlation is more robust than the RBF neural network correlation. The results also show that the input variables present redundant information. The axial power difference signals compensate the faulty signal for the position of a given control rod, and improves the results by about 10%. The results show that the errors in the power peak factor estimation by these neural network correlations, even in faulty conditions, are smaller than the current PWR schemes which may have uncertainties as high as 8%. Considering the maximum relative error of 2.35%, these neural network correlations would allow decreasing the power peak factor safety margin by about 5%. Such a reduction could be used for operating the reactor with a higher power level or with more flexibility. The neural network correlation has to meet requirements of high integrity software that performs safety grade actions. It is shown that the correlation is a very simple algorithm that can be easily codified in software. Due to its simplicity, it facilitates the necessary process of validation and verification.
Effects of Inner Surface Roughness and Asymmetric Pipe Flow on Accuracy of Profile Factor for Ultrasonic Flow Meter
Authors: Mori Michitsugu , Tezuka Kenichi , Takeda Yasushi
Published: September 2008

Flow profile factors (PFs), which adjust measurand to real flow rates, also strongly depend on flow profiles. To determine profile factors for actual power plants, manufactures of flowmeters usually conduct factory calibration tests under ambient flow conditions. Indeed, flow measurements with high accuracy for reactor feedwater require them to conduct calibration tests under real conditions, such as liquid conditions and piping layouts. On the contrary, as nuclear power plants are highly aging, readings of flowmeters for reactor feedwater systems drift due to the changes of flow profiles. The causes of those deviations are affected by the change of wall roughness of inner surface of pipings. We have conducted experiments to quantify the effects of flow patterns on the PFs due to pipe roughness and asymmetric flow, and the results of our experiments have shown the effects of elbows and pipe inner roughness, which strongly affect to the creation of the flow patterns. Those changes of flow patterns lead to large errors in measurements with transit time (time-of-flight: TOF) ultrasonic flow meters. In those experiments, changes of pipe roughness result in the changes of PFs with certain errors. Therefore, we must take into account those effects in order to measure the flow rates of feedwater with better accuracy in actual power plants.
Calibration Tests of New Type Flow-Metering System by Ultrasonic Pulse-Doppler Profile-Velocimetry at National Standard Loops
Authors: Mori Michitsugu , Tezuka Kenichi , Suzuki Takeshi , Sapia Mark , Schrull Edward , Nichols Paul , Takeda Yasushi
Published: September 2008

To verify high accuracy of a new type flow-metering system based on the measurements of line velocity profiles, eliminating Profile Factors, calibration tests of “UDF”, the flow-metering system by ultrasonic pulse-Doppler profile-velocimetry, were conducted at national standard loops worldwide, including the National Institute of Standard Technology (NIST) of the U.S. Department of Commerce, the National Metrology Institute of Japan (NMIJ) of the National Institute of Advanced Industrial Science and Technology (AIST) in Japan, the Nederlands Meetinstituut (NMI) in Netherlands, and the Centro National de Metrologia (CENAM) in Mexico. The deviations of UDF to the standard loops in the calibration tests for water were found between −0.23% and +0.26% at NIST, 0.1% and 0.4% at NMIJ, and −0.52% and +0.59% at NMI in terms of the average values of each measurement. Following improvements to the UDF System, the calibration tests at CENAM exhibited the deviations between −0.18% and +0.23% and the expanded uncertainty with ±0.21%.
Application of MELCOR Code to a French PWR 900 MWe Severe Accident Sequence and Evaluation of Models Performance Focusing on In-Vessel Thermal Hydraulic Results
Authors: De Rosa Felice
Published: September 2008

In the ambit of the Severe Accident Network of Excellence Project (SARNET), funded by the European Union, 6th FISA (Fission Safety) Programme, one of the main tasks is the development and validation of the European Accident Source Term Evaluation Code (ASTEC Code). One of the reference codes used to compare ASTEC results, coming from experimental and Reactor Plant applications, is MELCOR. ENEA is a SARNET member and also an ASTEC and MELCOR user. During the first 18 months of this project, we performed a series of MELCOR and ASTEC calculations referring to a French PWR 900 MWe and to the accident sequence of “Loss of Steam Generator (SG) Feedwater” (known as H2 sequence in the French classification). H2 is an accident sequence substantially equivalent to a Station Blackout scenario, like a TMLB accident, with the only difference that in H2 sequence the scram is forced to occur with a delay of 28 seconds. The main events during the accident sequence are a loss of normal and auxiliary SG feedwater (0 s), followed by a scram when the water level in SG is equal or less than 0.7 m (after 28 seconds). There is also a main coolant pumps trip when ΔTsat < 10 °C, a total opening of the three relief valves when Tric (core maximal outlet temperature) is above 603 K (330 °C) and accumulators isolation when primary pressure goes below 1.5 MPa (15 bar). Among many other points, it is worth noting that this was the first time that a MELCOR 1.8.5 input deck was available for a French PWR 900. The main ENEA effort in this period was devoted to prepare the MELCOR input deck using the code version v.1.8.5 (build QZ Oct 2000 with the latest patch 185003 Oct 2001). The input deck, completely new, was prepared taking into account structure, data and same conditions as those found inside ASTEC input decks. The main goal of the work presented in this paper is to put in evidence where and when MELCOR provides good enough results and why, in some cases mainly referring to its specific models (candling, corium pool behaviour, etc.) they were less good. A future work will be the preparation of an input deck for the new MELCOR 1.8.6. and to perform a code-to-code comparison with ASTEC v1.2 rev. 1.
Near Term Application of Supercritical Water Technologies
Authors: Vogt Bastian , Starflinger Joerg , Schulenberg Thomas
Published: September 2008

A pressurized water reactor with a supercritical water primary loop is analyzed (PWR-SC) within this paper. It will be shown that the PWR-SC offers considerable advantages in the fields of safety, economy and efficiency compared with a conventional PWR design. A cycle analysis shows that the net plant efficiency increases by 2% compared to currently operated or built systems. In addition, the mass flow rate of the primary side is strongly decreased, which enables a reduction of the primary pump power by a factor of 4. In the secondary loop, the mass flow rate can be decreased by about 15%, which allows downscaling of all secondary side components such as turbines, condensers and feed-water preheat systems as a consequence of the high core exit temperature. A coupled core analysis and a hot channel factor analysis are performed to demonstrate the promising safety features of the PWR-SC and to show the technical feasibility of such a system.
Stability Analysis of a Uniformly Heated Channel with Supercritical Water
Authors: Ortega Go´mez T. , Class A. , Lahey R. T. , Schulenberg T.
Published: September 2008

The thermal-hydraulic stability of a uniformly heated channel at supercritical water pressure has been investigated to help understand the system instability phenomena which may occur in Supercritical Water Nuclear Reactors (SCWR). We have extended the modeling approach often used for Boiling Water Nuclear Reactor (BWR) stability analysis to supercritical pressure operation conditions. We have shown that Ledinegg excursive instabilities and pressure-drop oscillations (PDO) will not occur in supercritical water systems. The linear stability characteristics of a typical uniformly heated channel were computed by evaluating the eigenvalues of the model. An analysis of non-linear instability phenomena was also performed in the time domain and the dynamic bifurcations were evaluated.
A Fully Implicit, Second Order in Time, Simulation of a Nuclear Reactor Core
Authors: Mousseau Vincent A.
Published: September 2008

This paper will present a high fidelity solution algorithm for a model of a nuclear reactor core barrel. This model consists of a system of nine nonlinearly coupled partial differential equations. The coolant is modeled with the 1-D six-equation two-phase flow model of RELAP5. Nonlinear heat conduction is modeled with a single 2-D equation. The fission power comes from two 2-D equations for neutron diffusion and precursor concentration. The solution algorithm presented will be the physics-based preconditioned Jacobian-free Newton-Krylov (JFNK) method. In this approach all nine equations are discretized and then solved in a single nonlinear system. Newtons method is used to iterate the nonlinear system to convergence. The Krylov linear solution method is used to solve the matrices in the linear steps of the Newton iterations. The physics-based preconditioner provides an approximation to the solution of the linear system that accelerates the Krylov iterations. Results will be presented for two algorithms. The first algorithm will be the traditional approach used by RELAP5. Here the two-phase flow equations are solved separately from the nonlinear conduction and neutron diffusion. Because of this splitting of the physics, and the linearizations employed this method is first order accurate in time. A second algorithm will be the JFNK method solved second order in time accurate. Results will be presented which compare these two algorithms in terms of accuracy and efficiency.
Lessons Learned From Implementation of Westinghouse Owners Group Risk-Informed Inservice Inspection Methodology for Piping
Authors: Stevenson Paul R. , Haessler Richard L. , McNeill Alex , Pyne Mark A. , West Raymond A.
Published: September 2008

Risk-informed inservice inspection (ISI) programs have been in use for over seven years as an alternative to current regulatory requirements in the development and implementation of ISI programs for nuclear plant piping systems. Programs using the Westinghouse Owners Group (WOG) (now known as the Pressurized Water Reactor Owners Group - PWROG) risk-informed ISI methodology have been developed and implemented within the U.S. and several other countries. Additionally, many plants have conducted or are in the process of conducting updates to their risk-informed ISI programs. In the development and implementation of these risk-informed ISI programs and the associated updates to those programs, the following important lessons learned have been identified and are addressed. • Concepts such as “loss of inventory,” which are typically not modeled in a plant’s probabilistic risk assessment (PRA) model for all systems. • The importance of considering operator actions in the identification of consequences associated with a piping failure and the categorization of segments as high safety significant (HSS) or low safety significant (LSS). • The impact that the above considerations have had on the large early release frequency (LERF) and categorization of segments as HSS or LSS. • The importance of automation. • Making the update process more efficient to reduce costs associated with maintaining the risk-informed ISI program. The insights gained are associated with many of the steps in the risk-informed ISI process including: development of the consequences associated with piping failures, categorization of segments, structural element selection and program updates. Many of these lessons learned have impacted the results of the risk-informed ISI programs and have impacted the updates to those programs. This paper summarizes the lessons learned and insights gained from the application of the WOG risk-informed ISI methodology in the U.S., Europe and Asia.
Is Nuclear Power Also the Key to Economically Clean Coal Gasification?
Authors: Kunze Jay F. , Sandquist Gary M. , Pardo David Martinez
Published: September 2008

Reducing the amount of carbon dioxide emitted to the atmosphere is a major goal and an imperative need for most of the world’s nations, even for those nations such as the USA who are not Kyoto Treaty signatories. A response by the current USA administration is to develop a national transportation economy for automobiles based upon efficient, environmentally sound fuel cells. However, hydrogen is a secondary fuel requiring a primary energy source for production. Nuclear power (or renewables such as hydroelectric, wind or solar) must be the source of the primary energy required to produce hydrogen from water, if the overall energy system is to be free of carbon dioxide emissions to the atmosphere. The dissociation of water leaves oxygen as a major byproduct. Currently, there are no existing commercial markets for the large quantities of oxygen that would result from a US transportation economy based upon hydrogen fuel cells. However, Integrated Coal Gasification Combined Cycle (IGCC) power plants operating on pure oxygen for both gasification and combustion produce no greenhouse gas releases. This highly desirable feature results from the combustion output being only water and carbon dioxide. Pure CO2 can be relatively easily captured and delivered to a sequestration site. Also, hazardous trace metal compounds (e.g., Hg, As, Pb, Sn, Sb, Se, U, Th, etc.) that would ordinarily be emitted to the atmosphere could be captured as solids, for environmentally acceptable disposal.
Development of the Integrated IHX/Pump Component 1/4-Scale Vibration Testing
Authors: Hayafune Hiroki , Konomura Mamoru , Morita Hideyuki , Usui Yukinori , Sawa Naoki , Tsujita Yoshihiro
Published: September 2008

This report summarizes the evaluation method, the development schedule and the result of the test. A conceptual design study of the Japanese sodium-cooled fast reactor (JSFR) is in progress of “Feasibility Study on Commercialized Fast Reactor Cycle Systems.” In this study we have proposed a component which integrated IHX (Intermediate heat exchanger) with primary pump into one vessel (integrated IHX/Pump) as one of advanced concepts for the plant economy. By development this component, the amount of material (steel) of the primary cooling system can be reduced and the maintenance performance of IHX can be improved. In order to confirm its feasibility, it is important to avoid the resonant vibration caused by integrating of the pump into the static IHX and to prevent excessive wear-out of the heat exchange tubes caused by contact with the tube support plates. Therefore, the vibration evaluation method has been developed and a lot of examinations were carried out to confirm the feasibility of this concept. The pump is arranged at the center of the IHX and heat exchange tubes are arranged concentrically. This component consists of the multiple cylindrical layers. Tube support plates are installed and they support the tube bundle with 0.1–0.3mm gaps. The 1/4-scale model was selected for vibration tests. The first step test with an internal vibration exciter which simulated the pump was performed, and the test with the pump installed into IHX was done. This test model has been installed at Takasago R&D center of Mitsubishi Heavy Industries, Ltd. The vibration caused by the pump was measured. As a result of data analysis, the character of frequency response was acquired. The analysis code called as “FINAS” can directly handle the vibration behavior of complex multi-cylinders with fluid between the cylinders. The eigenvalue analysis and response analysis were performed using a detailed model and the results were compared with the test results. Vibration evaluation methods of the integrated IHX/Pump have been studied. The 1/4-scale vibration test was performed and the results were compared with the analysis.
Application of Nuclear Energy to Bitumen Upgrading and Biomass Conversion
Authors: Numata Mamoru , Fujimura Yasushi , Amaya Takayuki , Hori Masao
Published: September 2008

Key drivers for the increasing use of nuclear energy are the need to mitigate global warming and the requirement for energy security. Nuclear energy can be applied not only to generate electricity but also as a heat source. Moreover, nuclear energy can be applied for hydrogen as well as water production. The application of nuclear energy to oil processing and biomass production is studied in this paper.
Development of Technologies on Innovative-Simplified Nuclear Power Plant Using High-Efficiency Steam Injector: Part 14—Cooling Characteristics of the Particulate Core Material Debris Accumulated in Horizontal Narrow Gap
Authors: Hirao Y. , Su G. , Sugiyama K. , Narabayashi T. , Mori M. , Ohmori S.
Published: September 2008

When LOCA occurs in proposed nuclear reactor systems, the coolability of the core would be kept by the SI core injection system and therefore the probability of the core meltdown is negligible small. In this research work, we make it clear that the coolability of the RPV bottom is secured even if a part of the core should melt and a substantial amount of debris should be deposited on the lower plenum. In this report, we examined experimentally the coolability of the RPV bottom that a Zircaloy-based loose debris layer is deposited on. We set up a heat supply section made by SUS304 on the loose debris layer and measured the heat flux released into the loose debris bed and the temperature at the lower surface of the heat supply section. In addition, we measured the temperature distribution at the bottom of the loose debris bed. It became clear in this study that the coolability depends on the amount of coolant supplied, and the hot spot would not occur when coolant is supplied. Even if a hotspot should occur in the oxidization of loose metal debris accompanied with rapid heat generation. It is found that when a small amount of coolant can be supplied, it disappears because of a high capillary force of oxidized loose debris. So it is confirmed that the soundness of RPV is basically maintained.
Development of Technologies on Innovative Simplified Nuclear Power Plant Using High-Efficiency Steam Injectors: Part 10—Application to a Small District-Heating Reactor
Authors: Narabayashi Tadashi , Shimazu Yoichiro , Murase Toshihiko , Nagai Masatoshi , Mori Michitsugu , Ohmori Shuichi
Published: September 2008

A steam injector (SI) is a simple, compact and passive pump and also acts as a high-performance direct-contact compact heater. This provides SI with capability to use as a passive ECCS pump and also as a direct-contact feedwater heater that heats up feedwater by using extracted steam from the turbine. In order to develop a high reliability passive ECCS pump and a compact feedwater heater, it is necessary to quantify the characteristics between physical properties of the flow field. We carried out experiments to observe the internal behavior of the water jet as well as measure the velocity of steam jet using a laser Doppler velocimetry. Its performance depends on the phenomena of steam condensation onto the water jet surface and heat transfer in the water jet due to turbulence on to the phase-interface. The analysis was also conducted by using a CFD code with the separate two-phase flow models. With regard to the simplified feed-water system, size of four-stage SI system is almost the same as the model SI that had done the steam and water test that pressures were same as that of current ABWR. The authors also conducted the hot water supply system test in the snow for a district heating. With regard to the SI core cooling system, the performance tests results showed that the low-pressure SI core cooling system will decrease the PCT to almost the same as the saturation temperature of the steam pressure in a pressure vessel. As it is compact equipment, SI is expected to bring about great simplification and materials-saving effects, while its simple structure ensures high reliability of its operation, thereby greatly contributing to the simplification of the power plant for not only an ABWR power plant but also a small PWR/ BWR for district heating system.
Development of Material and Structural Reliability Evaluation System MSS-REAL
Authors: Asayama Tai , Kato Takehiko , Morishita Masaki
Published: September 2008

For the development of the System Based Code, which was proposed by Asada and intends to optimize structural design of nuclear components by enabling margin exchange between various technical options, a tool for life cycle structural reliability evaluation method is necessary. For this purpose, the authors are developing a material strength and structural reliability evaluation system MSS-REAL. The system is primarily for fast breeder reactors but its methodologies can also be applied to the other types of reactors. This paper summarizes the features of the MSS-REAL system with examples and also describes a future development plan.
Development of a Thermal Fragmentation Model for Single Metal Droplet in FCIs
Authors: Lin Qian , Cao Xuewu
Published: September 2008

The fragmentation process of high-temperature molten drop is a key factor to determine the ratio heat transferred to power in FCIs, which estimated possible damage degree. But at present the fragmentation mechanism is not yet clear, the corresponding physics and mathematics model has not been set up completely, and this greatly restricts the practical application of research results. In this paper, the fragmentation process of molten drops is investigated by theoretic analysis. A thermal fragmentation model for a single melt droplet is developed, and the fragmentation correlations are verified by comparing the predictions of fragments size with experimental data.