bluuebunny/crossref_metadata_2025 huggingface dataset:
Design Option of Heat Exchanger for the Next Generation Nuclear Plant
Authors: Oh Chang H. , Kim Eung S.
Published: July 2009

The Next Generation Nuclear Plant (NGNP), a very High temperature Gas-Cooled Reactor (VHTR) concept, will provide the first demonstration of a closed-loop Brayton cycle at a commercial scale, producing a few hundred megawatts of power in the form of electricity and hydrogen. The power conversion unit (PCU) for the NGNP will take advantage of the significantly higher reactor outlet temperatures of the VHTRs to provide higher efficiencies than can be achieved with the current generation of light water reactors. Besides demonstrating a system design that can be used directly for subsequent commercial deployment, the NGNP will demonstrate key technology elements that can be used in subsequent advanced power conversion systems for other Generation IV reactors. In anticipation of the design, development and procurement of an advanced power conversion system for the NGNP, the system integration of the NGNP and hydrogen plant was initiated to identify the important design and technology options that must be considered in evaluating the performance of the proposed NGNP. As part of the system integration of the VHTRs and the hydrogen production plant, the intermediate heat exchanger is used to transfer the process heat from VHTRs to the hydrogen plant. Therefore, the design and configuration of the intermediate heat exchanger is very important. This paper will include analysis of one stage versus two stage heat exchanger design configurations and simple stress analyses of a printed circuit heat exchanger (PCHE), helical coil heat exchanger, and shell/tube heat exchanger.
Pu and MA Management in Thermal HTGRs: Impact at Fuel, Reactor and Fuel Cycle Levels
Authors: Kuijper J. C.
Published: July 2009

The PUMA project, a Specific Targeted Research Project (STREP) of the European Union EURATOM 6th Framework Program, is mainly aimed at providing additional key elements for the utilisation and transmutation of plutonium and minor actinides (neptunium and americium) in contemporary and future (high temperature) gas-cooled reactor design, which are promising tools for improving the sustainability of the nuclear fuel cycle. PUMA would also contribute to the reduction of Pu and MA stockpiles and to the development of safe and sustainable reactors for CO2-free energy generation. The project runs from September 1, 2006 until August 31, 2009. PUMA also contributes to technological goals of the Generation IV International Forum. It contributes to developing and maintaining the competence in reactor technology in the EU and addresses European stakeholders on key issues for the future of nuclear energy in the EU. An overview is presented of the status of the project at mid-term.
A Method to Evaluate Fission Gas Release During Irradiation Testing of Spherical Fuel
Authors: van der Merwe Hanno , Venter Johan
Published: July 2009

The evaluation of fission gas release from spherical fuel during irradiation testing is critical to understand expected fuel performance under real reactor conditions. Online measurements of Krypton and Xenon fission products explain coated particle performance and contributions from graphitic matrix materials used in fuel manufacture and irradiation rig materials. Methods that are being developed to accurately evaluate fission gas release are described here together with examples of evaluations performed on irradiation tests HFR-K5, -K6 and EU1bis.
An Infiltration Manufacturing Process for Nuclear Fuels
Authors: Ecker Lynne , Saccheri Jacopo , Bowerman Biays , Ablett James , Milian Laurence , Adams Jay , Ludwig Hans , Todosow Michael
Published: July 2009

The Infiltrated Kernel Nuclear Fuel (IKNF) process deposits nuclear fuel into the naturally occurring porosity in graphite. IKNF consists of infiltrating uranyl nitrate dissolved in an organic solvent into the graphite and then heat-treating the sample at low (<300°C) temperatures to remove the solvent and convert the uranyl nitrate to UO2. Complete conversion to UC2 can then be accomplished by heating to temperatures higher than 3000°C. IKNF is extremely flexible: it is appropriate for very high temperature applications and heating the infiltrated product to intermediate temperatures (higher than 900°C) produces nuclear fuel with a range of chemistries in the U-C-O system (similar to the current US TRISO fuel). It is probable that the process can also be used to produce fuel containing transuranics. It is believed that IKNF will be less expensive, more robust and more suitable for on-line quality monitoring than current fuel fabrication method. Graphite infiltration involves a few, easily measurable and controllable variables. It is reproducible and predictable.
Safety Related Physical Phenomena for Coupled High-Temperature Reactors and Hydrogen Production Facilities
Authors: Forsberg C. W. , Gorensek M. , Herring S. , Pickard P.
Published: July 2009

High-temperature reactors are a potential low-carbon source of high-temperature heat for chemical plants—including hydrogen production plants and refineries. Unlike electricity, high temperature heat can only be transported limited distances; thus, the reactor and chemical plants will be close to each other. A critical issue is to understand potential safety challenges to the reactor from the associated chemical plant events to assure nuclear plant safety. The U.S. Nuclear Regulatory Commission (NRC) recently sponsored a Phenomena Identification and Ranking Table (PIRT) exercise to identify potential safety-related physical phenomena for high-temperature reactors coupled to a hydrogen production or similar chemical plant. The ranking process determines what types of chemical plant transients and accidents could present the greatest risks to the nuclear plant and thus the priorities for safety assessments. The assessment yielded four major observations. Because the safety philosophy for most chemical plants (dilution) is different than the safety philosophy for nuclear power plants (containment), this difference must be recognized and understood when considering safety challenges to a nuclear reactor from coupled chemical plants or refineries. Accidental releases of hydrogen from a hydrogen production facility are unlikely to be a major hazard for the nuclear plant assuming some minimum separation distances. Many chemical plants under accident conditions can produce heavy ground-hugging gases such as oxygen, corrosive gases, and toxic gases that can have major off-site consequences because of the ease of transport from the chemical plant to off-site locations. Oxygen presents a special concern because most proposed nuclear hydrogen processes convert water into hydrogen and oxygen; thus, oxygen is the primary byproduct. These types of potential accidents must be carefully accessed. Last, the potential consequences of the failure of the intermediate heat transport loop that moves heat from the reactor to the chemical plant must be carefully assessed.
An Integrated Laboratory-Scale Experiment on the Sulfur-Iodine Thermochemical Cycle for Hydrogen Production
Authors: Moore Robert , Parma Ed , Russ Ben , Sweet Wendi , Helie Max , Pons Nicolas , Pickard Paul
Published: July 2009

Sandia National Laboratories (SNL), General Atomics Corporation (GA) and the French Commissariat a` l’Energie Atomique (CEA) have been conducting laboratory-scale experiments to investigate the thermochemical production of hydrogen using the Sulfur-Iodine (S-I) process. This project is being conducted as an International Nuclear Energy Research Initiative (INERI) project supported by the CEA and US DOE Nuclear Hydrogen Initiative. In the S-I process, 1) H2SO4 is catalytically decomposed at high temperature to produce SO2, O2 and H2O. 2) The SO2 is reacted with H2O and I2 to produce HI and H2SO4. The H2SO4 is returned to the acid decomposer. 3) The HI is decomposed to H2 and I2. The I2 is returned to the HI production process. Each participant in this work is developing one of the three primary reaction sections. SNL is responsible for the H2SO4 decomposition section, CEA, the primary HI production section and General Atomics, the HI decomposition section. The objective of initial testing of the S-I laboratory-scale experiment was to establish the capability for integrated operations and demonstrate H2 production from the S-I cycle. The first phase of these objectives was achieved with the successful integrated operation of the SNL acid decomposition and CEA Bunsen reactor sections and the subsequent generation of H2 in the GA HI decomposition section. This is the first time the S-I cycle has been realized using engineering materials and operated at prototypic temperature and pressure to produce hydrogen.
CYNOD: A Neutronics Code for Pebble Bed Modular Reactor Coupled Transient Analysis
Authors: Hiruta Hikaru , Ougouag Abderrafi M. , Gougar Hans D. , Ortensi Javier , Nigg David W. , Davis Cliff B. , Weaver Walter L.
Published: July 2009

In this paper, a new neutron kinetics solver for cylindrical R-Z geometry, CYNOD, is presented for the simulation of coupled transient problems for pebble bed reactors. The code utilizes the Direct Coarse Mesh Finite Difference method, in which a set of one-dimensional equations in each transverse direction is solved by means of the analytic Green’s function method. A method that deals with control rod cusping problems is also presented. A heterogeneous fuel kernel model is implemented in order to accurately take into account Doppler feedback effects. Numerical results that demonstrate the accuracy of the code are also presented.
A New Impetus for Developing Industrial Process Heat Applications of HTR in Europe
Authors: Hittner Dominique , de Groot Sander , Griffay Gerard , Yvon Pascal , Pienkowski Ludwik , Ruer Jacques , Angulo Carmen , Iaquaniello Gaetano
Published: July 2009

Due to its high operating temperature (up to 850°C with present technologies, possibly higher in the longer term), and its power range (a few hundred MW), the modular HTR could address a larger scope of industrial process heat needs than other present nuclear systems. Even if HTR can contribute to competitive electricity generation, this potential for industrial heat applications is the main incentive for developing this type of reactor, as it could open to nuclear energy a large non-electricity market. However several issues must be addressed and solved successfully for HTR to actually enter the market of industrial process heat: 1) as an absolute prerequisite, to develop a strategic alliance of nuclear industry and R&D with process heat user industries. 2) to solve some key technical issues, as for instance the design of a reactor and of a coupling system flexible enough to reconcile a single reactor design with multiple applications and versatile requirements for the heat source, and the development of special adaptations of the application processes or even of new processes to fit with the assets and constraints of HTR heat supply, 3) to solve critical industrial issues such as economic competitiveness, availability and 4) to address the licensing issues raised by the conjunction of nuclear and industrial risks. In line with IAEA initiatives for supporting non-electric applications of nuclear energy and with the orientations of the SET-Plan of the European Commission, the (European) HTR Technology Network (HTR-TN) proposes a new project, together with industrial process heat user partners, to provide a first impetus to the strategic alliance between nuclear and non-nuclear industries. End user requirements will be expressed systematically on the basis of inputs from industrial partners on various types of process heat applications. These requirements will be confronted with the capabilities of the HTR heat source, in order to point out possible discrepancies and issues, to assess the feasibility of different coupling schemes and to identify development needs. Partners from nuclear regulatory organisations will also address the feasibility of licensing such coupling schemes. The issues they will raise will be taken into consideration for defining coupling design bases and identifying R&D needs. A detailed roadmap for designing an industrial demonstrator of a HTR coupled with process heat applications will be inferred from this analysis, as well as R&D actions required for supporting the development of the reactor, of the coupling system and of possible adaptations or innovations in industrial processes.
The Preliminary Analysis of the Loss of Primary Coolant Flow Test in the HTTR
Authors: Nakagawa Shigeaki , Takamatsu Kuniyoshi , Goto Minoru , Takeda Tetsuaki , Nakao Yasuyuki
Published: July 2009

Loss of primary coolant flow test is under planning by using the High Temperature engineering Test Reactor (HTTR). In this test, all the gas circulators are tripped and the position of all control rods keeps its initial one. The new calculation model was developed to perform the preliminary analysis for the test. This model is so improved that an equivalent fuel channel model based on one point kinetics code and a whole reactor model based on two-dimensional thermal analysis code are coupled to simulate the reactor performance during the loss of coolant flow. Both calculation codes were used in the safety evaluation of the HTTR licensing. The improved calculation model was validated by comparison between the calculated result and the experimental one obtained from the coolant flow reduction test in the HTTR. The loss of primary coolant flow test simulates the depressurization accident and the data obtained from the test is useful for the validation and improvement of the calculation code applied to the safety analysis in the future HTGR such as Very High Temperature Reactor which is selected as one of candidates of the generation IV reactor system.
Development of Local Heat Transfer Models for Safety Assessment of Pebble Bed High Temperature Gas-Cooled Reactor Cores
Authors: Stainsby Richard , Worsley Matthew , Grief Andrew , Dennier Ana , Dawson Frances , Davies Mike , Coddington Paul , Baker Jo
Published: July 2009

This paper presents a model developed for determining fuel particle and fuel pebble temperatures in normal operation and transient conditions based on multi-scale modelling techniques. This model is qualified by comparison with an analytical solution in a one-dimensional linear steady state test problem. Comparison is made with finite element simulations of an idealised “two-dimensional” pebble in transient conditions and with a steady state analytical solution in a spherical pebble geometry. A method is presented for determining the fuel temperatures in the individual batches of a multi-batch recycle refuelling regime. Implementation of the multi-scale and multibatch fuel models in a whole-core CFD model is discussed together with the future intentions of the research programme.
The Analysis of the Thorium-Fueled Modular Helium-Cooled Reactor
Authors: Venneri Francesco , Kim Yonghee , McEachern Donald , Persson Carl-Magnus , Jo Chang-Keun
Published: July 2009

The potential of thorium fertile fuel has been evaluated for a graphite-moderated MHR (Modular Helium Reactor) from the perspective of a self-sustainable U233-Th fuel cycle in the MHR. The 3-D core analyses have been performed with the thermal-hydraulic-coupled computer code systems (HELIOS-MASTER). The feasibility of a self-sustainable U233-Th fuel cycle in MHR was evaluated for a simplified equilibrium fuel cycle. A mixed oxide fuel (ThO2-UO2) was used. Whole-core analysis was performed with the MASTER code for various core configurations. In the core analysis, a 3-batch radial fuel shuffling scheme was adopted to find an equilibrium fuel cycle. Three types of fuel blocks were considered: a homogeneous fuel arrangement and two seed-blanket arrangements. It was found that a near self-sustainable U233-Th fuel cycle (conversion ratio = 0.95∼0.97) is feasible for the MHRs with the appropriate U-233 and Th-232 loadings. To achieve a high conversion ratio while maintaining a long cycle length, it is essential to maximize the thorium loading (∼30 tons) and at the same time soften the neutron spectrum to achieve sufficient reactivity. In order to achieve conversion ratios over 0.95 and an 18-month cycle length, the moderator volume needs to be increased with respect to the regular MHR design. Also, removing the inner graphite reflectors increases noticeably the core performance in terms of the conversion ratio and cycle length. A special seed-blanket block configuration (ISB, with seed and blanket fuels in the inner and outer regions of a block, respectively) provides a superior conversion ratio with respect to a homogeneously fueled block, whereas reversing the placement of seed and blanket in the ISB block configuration (with seed fuels in the outer zone) results in a worse performance. In the case of the U233-Th fuel cycle, the fissile (U233+U235) fraction in the discharged fuel is almost 90%. Denaturing of the uranium vector in the self-sustainable U233-Th fuel was investigated by adding 10% LEU in the fuel, to make the initial fissile fraction ∼20.5%. Neutronic analysis of the operation with denatured fuel reveals that the conversion ratio is substantially reduced and the available cycle length is much shorter.
Simulation and Analysis of Helium Circulator Trip ATWS Test at Full Power on the HTR-10
Authors: Dong Yujie , Chen Fubing , Zhang Zuoyi , Hu Shouyin , Shi Lei , Zheng Yanhua , Zhou Yangping
Published: July 2009

Safety demonstration tests on the 10 MW High Temperature Gas-cooled Reactor-Test Module (HTR-10) were conducted to verify the inherent safety characteristics of modular High Temperature Gas-cooled Reactors (HTGRs) as well as to obtain the reactor core and primary cooling system transient data for validation of HTGR safety analysis models and codes. As one of these safety demonstration tests, a simulated anticipated transient without scram (ATWS) test called loss of forced cooling by tripping the helium circulator without reactor scram was carried out at 100% rated power level in July, 2005. This paper simulates the reactor transient behaviour during the test by using the THERMIX code system. The reactor power transition and a comparison with the test result are presented. Owing to the negative temperature coefficient of reactivity, the reactor undergoes a self-shutdown after the stop of the helium circulator and keeps subcritical till the end of the test. Due to the loss of forced cooling, the residual heat is slowly transferred from the core to the Reactor Cavity Cooling System (RCCS) by conduction, radiation and natural convection. The thermal response of this heat removal process is investigated. The calculated and test temperature transients of the measuring points in the reactor internals are given and the differences are preliminarily discussed. With respect to the safety features of the HTR-10, it is of most importance that the maximum fuel center temperature is always lower than 1230 °C which is the limited value at the first phase of the HTR-10 project. The simulation and test results show that the HTR-10 has the built-in passive safety features, and the THERMIX code system is applicable and reasonable for simulating and analyzing the helium circulator trip ATWS test.
Project Deep-Burn: Development of Transuranic Fuel for High-Temperature Helium-Cooled Reactors
Authors: Versluis Robert M. , Venneri Francesco , Petti David , Snead Lance , McEachern Donald
Published: July 2009

The helium-cooled, graphite-moderated Very High Temperature Reactor (VHTR) has become the centerpiece of the U.S. Department of Energy’s (DOE) Next Generation Nuclear Plant (NGNP) program. The NGNP program aims to construct a VHTR prototype, with the participation of industry, by the year 2021.
Post Irradiation Examination of HTR Fuel at ITU Karlruhe
Authors: Freis D. , Bottomley P. D. , Hiernaut J.-P. , Colle J.-Y. , Ejton J. , de Weerd W.
Published: July 2009

In the last years considerable efforts have been made at the Institute for Transuranium Elements (ITU) in order to reestablish European knowledge and ability in safety testing of irradiated high temperature reactor (HTR) Fuel Elements. In the framework of the 6th European framework programme a cold finger apparatus (Ku¨FA) furnace, formerly installed at FZ-Ju¨lich (FzJ), has been installed in a hot cell at ITU [Freis 2008] in order to test fission product release under high temperature and non-oxidising conditions. Several analytical methods (e.g. Gamma-spectrometry, mass-spectrometry) have been applied in order to analyse different isotopes released during Ku¨FA tests. After the heating tests, examinations of the fuel elements were performed including scanning electron microscopy (SEM) and micro-hardness testing of coated particles. Individual coated particles were object of heating tests in a Knudsen cell with a coupled mass spectrometer measuring all released species. In order to cover more accident scenarios, a second furnace for oxidising-conditions (air- or water-ingress) was constructed and installed in a cold lab. Furthermore a disintegration apparatus, based on anodic oxidation, was constructed and fuel elements were dissolved obtaining thousands of individual coated particles for further examination. A fully automated irradiated microsphere gamma analyzer (IMGA) is under construction and will be used, in particular, to identify and sort out failed particles.
CFD Simulation of a Coolant Flow and a Heat Transfer in a Pebble Bed Reactor
Authors: In Wang-Kee , Lee Won-Jae , Hassan Yassin A.
Published: July 2009

This CFD study is to simulate a coolant (gas) flow and heat transfer in a PBR core during a normal operation. This study used a pebble array with direct area contacts among the pebbles which is one of the pebbles arrangements for a detailed simulation of PBR core CFD studies. A CFD model is developed to more adequately represent the pebbles randomly stacked in the PBR core. The CFD predictions showed a large variation of the temperature on the pebble surface as well as in the pebble core. The temperature drop in the outer graphite layer is smaller than that in the pebble-core region. This is because the thermal conductivity of graphite is higher than the fuel (UO2 mixture) conductivity in the pebble core. Higher pebble surface temperature is predicted downstream of the pebble contact due to a reverse flow. Multiple vortices are predicted to occur downstream of the spherical pebbles due to a flow separation. The coolant flow structure and fuel temperature in the PBR core appears to largely depend on the in-core distribution of the pebbles.
Stress Rupture Characteristics of Zirconium Alloy Cladding Under Closed-End Internal Pressurization
Authors: Marple B. , Song Y. J. , Gollapudi S. , Murty K. L. , Charit I.
Published: July 2009

An understanding of stress rupture behavior of zirconium alloy cladding tubes is of paramount importance for applications in nuclear reactors. Stress rupture properties of recrystallized Zircaloy-4 alloy were evaluated using burst testing of closed-end thin-walled tubing at varied test temperatures and internal pressures. The rupture data are correlated using the Larson-Miller parameter. The uniform circumferential elongations were also measured from which the hoop creep rates were calculated. These results were fitted to Monkman-Grant relationship with the aim of extrapolating the data to in-service stress levels. Furthermore, the creep data were plotted according to the Dorn equation where a transition in deformation mechanism from ‘power-law’ to ‘power law breakdown’ for Zircaloy-4 was noted. TEM studies corroborated the transition in mechanism from the power-law regime to a power-law breakdown regime.
A Novel Radiation Dosimetry Based on Optically Stimulated Luminescence
Authors: Liu Yanping , Chen Zhaoyang , Fan Yanwei , Ba Weizhen , Pan Shilie
Published: July 2009

A new generation of Alkaline earth sulfides (MgS, CaS, and BaS) doped with rare-earth ions have been identified by the University of Montpellier as the very high sensitivity of these phosphors, the short time constant of the luminescence and the perfectly separated spectra enable many applications in real time and online dosimetry. The online detecting technology of optically stimulated luminescent (OSL) radiation dosimeter main make use of the OSL characteristics of doping the alkaline-earth metal sulphides, makes the material into the thin films for storing energy from Ionizing radiation, the excitation light through optical fibers reached the where under radiation-field, with a sensitive detection device to read out the radiation dose from storing the OSL material, obtains a novel technology of radiation dose measurement. In the previous works, the dosimeter benefits from a printed circuit board mount. Both the sensor and the electronics are exposed to radiation, the problem of the radiation induced damage is supposedly being addressed. In both cases, the use of optical fibers can provide an elegant solution. Optical fibers offer a unique capability for remote monitoring of radiation in difficult-to-access and hazardous locations. Optical fiber can be located in radiation hazardous areas and optically interrogated from a safe distance. Hence, optical fiber dosemeters are immune to electrical and radio-frequency interference that can seriously degrade the performance of remote electronic dosimeters. In this paper, a novel remote optical fiber radiation dosimeter is described. The optical fiber dosimeter takes advantage of the charge trapping materials CaS:Ce, Sm and SrS:Eu, Sm that exhibit optically stimulated luminescence (OSL). The range of the dosimeter is from 0.01 to 1000Gy. The optically stimulated luminescent (OSL) radiation dosimeter technically surveys a wide dynamic measurement range and a high sensitivity. The equipment is relatively simple and small in size, and has low power consumption. This device is suitable for measuring the space radiation dose; it also can be used in high radiation dose condition and other dangerous radiation occasions.
Assessment of the Design and Safety Properties of PWR Cores With Reduced Soluble Boron Utilization
Authors: Papukchiev Angel Aleksandrov , Schaefer Anselm
Published: July 2009

In conventional pressurized water reactor (PWR) designs, soluble boron is used for reactivity control over core fuel cycle. As high boron concentrations have significant impact on reactivity feedback properties and core transient behaviour, design changes to reduce boron concentration in the reactor coolant are of general interest in view of improving PWR inherent safety. In the framework of an investigation into the feasibility of low boron design, two PWR core configurations based on fuel with higher gadolinium and erbium utilization have been developed permitting to reduce the natural boron concentration at begin of cycle by approx. 50% resp. 30% compared to current PWR technology. Their performance was compared with the one of a standard German PWR core in a loss-of-feedwater anticipated transient without scram, boron dilution scenario and a hypothetical beyond design basis accident. Moreover, this paper deals not only with the potential advantages, but also addresses the drawbacks of the new PWR configurations — complex core design, increased power peaking, significant amount of non-burnable gadolinium and erbium isotopes, reduced cycle length, etc. The analyses presented here were supported by detailed cell, diffusion and thermal-hydraulic simulations, performed with the well known codes HELIOS, QUABOX/CUBBOX and ATHLET.
Estimation of Pressure Loads During Ex-Vessel Steam Explosion
Authors: Leskovar Matjazˇ
Published: July 2009

An ex-vessel steam explosion may occur when, during a severe reactor accident, the reactor vessel fails and the molten core pours into the water in the reactor cavity. A steam explosion is a fuel coolant interaction process where the heat transfer from the melt to water is so intense and rapid that the timescale for heat transfer is shorter than the timescale for pressure relief. This can lead to the formation of shock waves and production of missiles that may endanger surrounding structures. A strong enough steam explosion in a nuclear power plant could jeopardize the containment integrity and so lead to a direct release of radioactive material to the environment. In the paper, different scenarios of ex-vessel steam explosions in a typical pressurized water reactor cavity are analyzed with the code MC3D, which is being developed for the simulation of fuel-coolant interactions. A comprehensive parametric study was performed varying the location of the melt release (central and side melt pours), the cavity water sub-cooling, the primary system overpressure at vessel failure and the triggering time for explosion calculations. The main purpose of the study was to determine the most challenging ex-vessel steam explosion cases in a typical pressurized water reactor and to estimate the expected pressure loadings on the cavity walls. Special attention was given to melt droplets freezing, which may significantly influence the outcome of the fuel-coolant interaction process. The performed analysis shows that for some ex-vessel steam explosion scenarios much higher pressure loads are predicted than obtained in the OECD program SERENA Phase 1.
CFD Investigation of Heat Transfer in Supercritical Water-Cooled Flow Through 3×3 Fuel Rod Bundles
Authors: Shang Zhi , Yao Yufeng
Published: July 2009

CFD investigation of heat transfer in supercritical water-cooled flow through fuel rod bundles has been carried out, using commercial software STAR-CD 4.02 with specific ad hoc user routines for modeling physical property of supercritical water. The configuration considered is a typical core assembly of 3×3 fuel rod (round tube) bundles inside solid square box, as seen in the nuclear reactor. After priori mesh convergence studies, investigations are focused on key characteristics of flow and heat transfer performance, notably the wall temperature distributions, the mass flux and the secondary flow patterns in the cross-section. It is found that the rod wall temperature distributions exhibit highly non-uniform feature near the domain exit with very high wall temperatures: about 625°C observed on the corner rod and about 562.5°C on the border rod, respectively. It is believed that the appearance of the extremely wall temperature may be related to the non-uniform distributions of mass flux in the cross-section of the bundles as the low mass flux co-existing with the high wall temperature. Further analysis of the secondary flow in the cross-section reveals wider spectrum of vortex flow structures, more complicated than previously noted by the sub-channel analysis. To verify the influence of turbulence models on the secondary flow, both linear and non-linear k-ε models are applied and results are quite similar. This finding indicates that the cause of the secondary (cross) flow might not be solely due to the anisotropic property of turbulence as suggested by other researchers. The present 3D CFD study provides more complete database of 3×3 rod bundle flows and will be useful to improve the industry practice of applying the sub-channel analysis.
CONSUL Code Package Application for LMFR Core Calculations
Authors: Chibinyaev A. V. , Teplov P. S. , Frolova M. V.
Published: July 2009

CONSUL code package designed for the calculation of reactor core characteristics has been developed at the beginning of 90-th. The calculation of nuclear reactor core characteristics is carried out on the basis of correlated neutron, isotope and temperature distributions. The code package has been generally used for LWR core characteristics calculations. At present CONSUL code package was adapted to calculate liquid metal fast reactors (LMFR). The comparisons with IAEA computational test “Evaluation of benchmark calculations on a fast power reactor core with near zero sodium void effect” and BN-1800 testing calculations are presented in the paper. The IAEA benchmark core is based on the innovative core concept with sodium plenum above the core BN-800. BN-1800 core is the next development step which is foreseen for the Russian fast reactor concept. The comparison of the operational parameters has shown good agreement and confirms the possibility of CONSUL code package application for LMFR core calculations.
Steam Generator In-Service Inspections at Paks NPP
Authors: Do´czi Miklo´s
Published: July 2009

Steam Generator is one of the most critical components in nuclear power plants. It has of overriding importance from point of view of safe and reliable operation of the whole plant. Variety of degradation mechanisms affecting SG tube bundle may cause different types of material damage. In Paks NPP eddy current in-service inspection have been performed since 1988. In the year 1997 higher number of defected tubes were found in case of Unit#2, compared to results of the previous years. A medium term SG inspection program had been performed in the time period between 1998–2004. Based on the results of eddy current inspections high number of heat exchanger tubes had been plugged. Chemical cleanings of all steam generators were performed aiming to reduce the magnetite, copper deposits and corrosion agents acting on the surface of the tubes. Replacement of the main condensers had been performed to stop the uncontrolled water income caused by the relatively frequent leakages of the condenser tubes. Several tube samples had been cut from the first row of the tube bundles of different steam generators to study the effectiveness of the cleaning process and to determine the composition of deposits on the tube outside surface. Also several tubes with eddy current indications had been pulled out from the steam generators to determine the acting degradation mechanism. Examination of removed tubes can provide opportunity to check the reliability of eddy current inspection using bobbin coil. Also there were tubes pulled out form SG with existing cracks. From the year 2005 new inspection program had been started. As the first results of the new inspection program shows, there is only a few new indications had been found and there is no measurable crack propagation in case of existing indications. During the recent years feed-water collectors were replaced in case of all units of the power plant, because of material damage (erosion corrosion). The paper summarizes the results of eddy current in-service inspection of heat exchanger tubes, results of examinations of removed tubes and also deals with results of visual examination of the feed-water distributor system.
Development of PIRT and Verification of RELAP5 Void Model for Application to the Loss-of-RHR Event During Mid-Loop Operation
Authors: Nagumo Hiroichi , Sasaki Yasuhiro , Murase Michio , Yoshida Yoshitaka
Published: July 2009

The loss of RHR during mid-loop operation in PWR is relatively high risk event. More confident analysis of the event is desirable to develop better counter measures and increase plant safety. The analysis methodology with statistical method using a best estimate analysis code to increase confidence of analysis result is under development. The method employs the RELAP5/MOD3.2 code as a best estimate code and is being developed along the CSAU methodology. One of the most important steps in the CSAU methodology is development of PIRT (Phenomena Identification and Ranking Table) for the event. The PIRT is developed for the loss of RHR event during mid-loop operation with mitigation measure of reflux cooling and gravity injection from RWST and important models of the RELAP5/MOD3.2 related to high ranked phenomena are identified. Verification matrix is also developed for the important models. One of the important models identified is void model. This model affects two phase water level of the reactor vessel and how much water is transported with vapor from reactor vessel. Verification of void model is especially focused on low power and low pressure conditions which are characteristics of the loss of RHR event under mid-loop operation. Prediction error of void model was quantified for both heated rod bundle channel and non-heated channels. Experiment with rod bundle core geometry under low power and low pressure conditions used for verification analysis is the THETIS experiment. The experiment was performed under quasi-steady condition. Two phase level under specified collapsed level was measured with varying power and pressure. Analysis results with pressure 0.5 to 1.0 MPa predict two phase level within 10% error. Void prediction analyses with non-heated channels were conducted against both steam-water experiment and air-water experiment with various pressure and hydraulic diameter. Most of data are predicted within 30% error.
Investigation on Calculating Methods of Mean Temperature Differences
Authors: Lu Guangyao , Wang Jing , Li Zhan
Published: July 2009

In the calculation of thermal performance of heat exchanger, the mean temperature differences are often adopted. This study is carried out to investigate the logarithmic and arithmetic mean temperature differences. In the experiments, two calculating methods are adopted, which are measuring wall temperature (MWT) and logarithmic-mean temperature difference (LMTD). The comparisons between these two calculating methods are conducted on the basis of experimental and theoretical analysis. It is shown that the results gained by these two methods are different when calculating the thermal performance of heat transfer. The differences come mainly from the difference between LMTD and the arithmetic-mean temperature difference (AMTD). Experimental results show that there are large difference between LMTD and MWT in small Reynolds number region, whereas in turbulent flow area, the results gained by LMTD and MWT have small difference. It is shown that in small Reynolds number region, adopting LMTD may induce error.
A Point Process Model for Piping Failure Frequency Analysis Using OPDE Data
Authors: Yuan X.-X. , Pandey M. D. , Riznic J.
Published: July 2009

The accurate estimation of piping failure frequency is an important task to support the probabilistic risk assessment and risk-informed in-service inspection of nuclear power plants. Although probabilistic models have been reported in the literature to analyze the piping failure frequency, this paper proposes a stochastic point process model that incorporates both a time dependent trend and plant specific (or cohort) effects on the failure rate. A likelihood based statistical method is proposed for estimating the model parameters. A case study is presented to analyze the Class 1 pipe failure data given in the OPDE Database.
Design and Technology Development Status and Design Considerations for Innovative Small and Medium Sized Reactors
Authors: Kuznetsov Vladimir V.
Published: July 2009

There is continuing interest in Member States in the development and application of small and medium sized reactors (SMRs), i.e., the reactors with an equivalent electric power of less that 700 MW. In 2006–2007, several distinct ‘families’ of innovative SMRs comprising more than 50 innovative concepts and designs have been analyzed or developed by national or international programmes involving Argentina, Brazil, China, Croatia, France, India, Indonesia, Italy, Japan, Republic of Korea, Lithuania, Morocco, Russian Federation, South Africa, Turkey, USA, and Vietnam. Innovative SMRs are under development for all principal reactor lines. The target dates when they could be ready for deployment protrude from 2010 to 2030. The designers of innovative small and medium sized reactors pursue new design and deployment strategies making use of certain advantages provided by smaller reactor size and capacity to achieve reduced design complexity and simplified operation and maintenance requirements, and to provide for incremental capacity increase through multiple plant clustering or multi-module plant construction. Competitiveness of SMRs depends on the incorporated strategies to overcome loss of economies of scale but equally it depends on finding appropriate market niches for such reactors, which generically include markets with limited investment capability potentially benefiting from reaching the targeted nuclear power station capacity incrementally, small electrical grids, off-grid locations, and non-electrical applications of nuclear power.
Analysis of a Loss of Feedwater Case at the Ringhals-3 NPP Using RELAP5/PARCS Coupled Codes
Authors: Ba´na´ti Jo´zsef , Sta˚lek Mathias , Demazie`re Christophe , Holmgren Magnus
Published: July 2009

This paper deals with the development and validation of a coupled RELAP5/PARCS model of the Swedish Ringhals-3 pressurized water reactor against a Loss of Feedwater transient, which occurred on August 16, 2005. At first, the stand-alone RELAP5 and PARCS models are presented. All the 157 fuel assemblies are modeled in individually in both codes. The model is furthermore able to handle possible asymmetrical conditions of the flow velocity and temperature fields between the loops. On the neutronic side, the dependence of the material constants on history effects, burnup, and instantaneous conditions is accounted for, and the full heterogeneity of the core is thus taken into account. The reflectors are also explicitly represented. The coupling between the two codes is touched upon, with emphasis on the mapping between the hydrodynamic/heat structures and the neutronic nodes. The transient was initiated by a malfunction of the feedwater valve at the 2nd steam generator. Consequently, the turbines were tripped and, because of the low level in the SG-2 the reactor was scrammed. Activation of the auxiliary feedwater provided proper amount of cooling from the secondary side, resulting in safe shutdown conditions. Capabilities of the RELAP5 code were more challenged in this transient, where the influences of the feedback from the neutron kinetic side were also taken into account in the analysis. The calculated values of the parameters show good agreement with the measured data.
Best Practices in Japan of Human Resource Development for NPP O&M: Roles and Lessons From Training Centers
Authors: Yamamoto Shinji , Kitamura Toshiro
Published: July 2009

The use of best practices and their lateral expansion as a benchmark is one of effective methods of “knowledge management (NK)”. Best practices of human resources development were collected (selected examples are listed below) from all 11 training centers annexed to the nuclear power plants in Japan and lessons were learned for possible lateral development for improving other stakeholders’ NK. Such best practices will provide productive information for designing their own human resources development strategies. Examples of collected good practices: • Exhibition of troubles and negative legacies: The actual machineries, equipment or components, explanatory documents or news articles of the past troubles experienced by themselves are effective to maintain and refresh the awareness and preparedness of trainees and other employees for recurrence prevention. The exhibitions are open to the visitors, too. • Experience-type training facilities: Off-normal conditions of components and systems are simulated for the staff practical training by the use of the facilities which provide an off-normal environment. Examples are: water hammers, abnormal vibrations and noises of rotating machineries, pump cavitations, pinholes, plumbing airs, etc. • Advanced simulators for operators training: Each electric company has its own simulators for training their own operating staff. These simulators are annexed to the nuclear power plants and used to train the operation staff by the experienced shift managers. The operation staff use the simulator for continually confirming the operation procedures and the plant behavior, etc. specific to their plants. Training for generic plant behavior and operators’ responses are mainly outsourced to the dedicated training centers run by the Owners’ Groups (BWR, PWR). • The SAT methods: The IAEA proposed SAT (Systematic Approach to Training) approach is applied to the training of the operating staff and the maintenance staff. It is structured in a flow of Job analysis ← Training program design ← Training material development ← Training ← Evaluation. • Training in real situations: An example is a trainee actually hung with a lifeline on a harness to learn a method of putting on the lifeline. On the other hand, the efficiency (availability) of the training facilities for maintenance work is very limited, because each electric company installs the training facilities individually. Experiences of ICONE-16 participants from other countries in improving the availability are of our interest.
Preliminary Design Analysis of a Hot Gas Ducts for the Nuclear Hydrogen System
Authors: Song Kee-Nam , Kim Yong-Wan
Published: July 2009

Korea Atomic Energy Research Institute (KAERI) is in the process of carrying out a nuclear hydrogen system by considering the indirect cycle gas cooled reactors that produce heat at temperatures in the order of 950 °C. A coaxial double-tube hot gas duct (HGD) is a key component connecting the reactor pressure vessel and the intermediate heat exchanger (IHX) for the nuclear hydrogen system. Recently, a preliminary design analysis for the primary and secondary hot gas ducts of the nuclear hydrogen system was carried out. These preliminary design activities include a preliminary decision on the geometric dimensions, a preliminary strength evaluation and an appropriate material selection. In this study, a preliminary strength evaluation for the HGDs of the nuclear hydrogen system has been undertaken. Preliminary strength evaluation results for the HGDs showed that the geometric dimensions of the proposed HGDs would be acceptable for the design requirements.
Review of the Primary and Secondary Water Chemistry Operation of Paks Nuclear Power Plant From the Point of View of Lifetime Extension
Authors: Doma A´rpa´d
Published: July 2009

The review of the water regime used for the Units of Paks Nuclear Power Plant was carried out in 2005, after 18 – 23 years of operation. It was clearly concluded after processing the huge data base of the water regime, that there is nothing to prevent the life time extension of the Units. In 2006, a new water regime was developed that will be applied during the preparation for the life time extension and the extended service life as well. In connection with this work, recommendations were made for some modifications of the previously used water regime. Currently there is no uniform startup and shutdown water regime for WWER-440 Units. Therefore, special attention was paid to developing of a startup water regime, which will be applied for the outages as early as 2008. The summarised recommendations for water regime modification will be subject to international expert review in 2008, and the modifications judged to be implemented will be finalised after the review.
Droplet Diameter Measurements for the Prediction of Liquid Droplet Impingement Erosion
Authors: Morita Ryo
Published: July 2009

The characteristics of liquid droplet diameter are investigated for the prediction of liquid droplet impingement erosion (LDI) occurring in a power plant. Since the main factors causing LDI are fluid-related factors such as droplet velocity and diameter, it is possible to predict the location of LDI attack if the local flow conditions such as velocity, wetness and droplet diameter are clarified. These local flow conditions except for droplet diameter can be obtained by calculation. Thus, we conducted steam experiment to measure the droplet diameter in the steam flow and clarify the characteristics of the droplet and the correlation between droplet diameter and another state quantities of the steam. As a result, we found that a distribution of the droplet diameter could be approximated by gamma distribution, which is a general form of the exponential distribution. To obtain the correlation between the diameter and the state quantities, we considered the change of the heat energy in addition to the surface tension and shear force, and we checked the correlation between the state quantites of steam and droplet diameter. A good correlation was found between these quantities and the droplet diameter, and the proposed approximate formula for the diameter was obtained.