This study extended the capabilities of PCTRAN by including the Gaussian puff model and the basic numerical techniques, to efficiently calculate the radioactive effluent dispersion. PCTRAN is a PC based nuclear power plant simulation code, and capable of running faster than real time. Accident initiation events can be activated in the simulation software, and the radioactive materials are assumed to be released from the plant to the atmosphere at every time of interval. During a transient or an accident simulation, puffs are sequentially generated and dispersed in all directions governed by the Pasquill stability category, wind velocity and wind direction. The thyroid dose rate and whole body dose rate (as well as their accumulations) at every spatial location in the emergency planning zone (EPZ) are shown as a color-shaded plot. A postulated scenario was given in PCTRAN to demonstrate that the code predict the time-varying distributions of thyroid and whole body dose rates (and their accumulated dosages) efficiently. Therefore, the urgent decision-making on protective actions can be determined ahead of the accident time.

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A steam jet flow injected into a large bulk space has been simulated by using the MARS code, an integrated best-estimate nuclear thermo-hydraulics analysis code. MARS code has been constructed by KAERI by a coupling and modernization of the RELAP5 and COBRA-TF codes. The multi-dimensional analysis capability of the MARS code has been assessed by comparing it's results to the results of the OECD-SETH PANDA experiment. PANDA facility consists of two large vessels connected by a horizontal pipe. The volume of a vessel is approximately 90 m^3. A steam jet is injected into the vessel through a 0.153m diameter nozzle located at a middle height of the vessel. Among the various series of PANDA tests, the condensation phenomena involved tests are selected and simulated by the MARS code. The whole vessel and the connecting pipe are modeled as a Cartesian multi-dimensional grid. In spite of a few discrepancies and a restriction of the grid size, the MARS code results show a promising performance for predicting the temperature and steam concentration distributions in a large bulk space.

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The lack of experimental data concerning the long-term 2-dimensional concrete ablation by a prototypic core oxide melt initiated the OECD-sponsored series of CCI experiments at ANL (USA). Complementary experiments (COMET) which were sponsored by the European Commission were conducted within the LACOMERA project at FZK (Germany). The CCI-2 test investigated the long-term interaction of a heated (direct electrical heating) core oxide melt within a rectangular limestone/common sand (LCS) concrete crucible. In absence of crusts the lateral/axial erosion ratio for oxide melts is approximately 1. However, crusts which can more easily develop in oxide melts may in case of a local crust failure lead to an irregular cavity shape as observed in the CCI-1 and CCI-3 experiment. The experiment COMET-L2 was designed to investigate a long-term MCCI of metallic corium in cylindrical siliceous concrete cavity under dry conditions with decay heat simulation of intermediate power, and subsequently at reduced power. During the first 100 s, i. e. until the end of the melt overheat, the ratio of the axial to the lateral erosion is approx. 1. The long-term axial erosion became more pronounced and was a factor of 2-3 higher than the lateral ablation. This is in agreement with results obtained in the former BETA experiments. Both series of experiments provided valuable data for code (WECHSL) validation and model improvements. The current version of WECHSL code is applied to the COMET-L2 and CCI-2 experiment.

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Most of strategies yet implemented to optimal fuel core loading pattern design in nuclear power reactors, are based on maximizing the core effective multiplication factor (K_<eff>) to extract maximum energy and lowering the local power peaking factor (P_q) from a predetermined value. However, a new optimization criterion could be of interest, aiming maximum burn up of the plutonium content in nuclear fuel assemblies, i.e, minimization of remaining plutonium in spent fuel at the end of cycle (EOC). In this research, we developed a new strategy for optimal fuel core loading pattern of a VVER-1000 reactor, based on multi-objective optimization: lowering the P_q, maximization of the K_<eff> and minimization of remaining plutonium (P_u) in fuels at EOC condition. This strategy has been implemented via exact calculations of fuel burn up during the equilibrium cycle using WIMS and CITATION calculation codes. We used genetic algorithm to find the optimum fuel loading pattern.

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Steady-state and transient boiling CHFs in a pool of highly wetting liquid were studied. Boiling heat transfer processes on a platinum cylinder in a pool of ethanol due to exponentially increasing heat generation rates, Q=Q_0exp^<t/τ>, ranging from quasi-steadily increasing one to rapidly increasing one with periods, τ, were measured for a 1.0-mm diameter horizontal cylinder with different surface conditions of commercially-available and roughly-finished surface cylinders for saturated and subcooled liquid at various pressures. Steady-state CHFs at various pressures gradually increased with an increase in subcooling. It was confirmed that the CHFs for lower subcoolings at pressures almost show little dependence on cylinder surface condition. However the CHF data for the roughly-finished cylinder in many cases were increased for higher subcoolings at pressures. Typical trends of the CHFs were clarified to three groups corresponding to periods; the first, second and third groups of CHF were for longer periods, for shorter and for intermediate ones. It was clarified that the CHFs for the shorter periods were significantly affected by the surface roughness of cylinders. It appears that more study on the multi-parametric surface conditions and a wider range of experimental conditions need to be included in the study.

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A system level PHA using sequence tree method was developed to perform Safety Related digital I&C system SSA. The conventional PHA is a brainstorming session among experts on various portions of the system to identify hazards through discussions. However, this conventional PHA is not a systematic technique, the analysis results strongly depend on the experts' subjective opinions. The analysis quality cannot be appropriately controlled. Thereby, this research developed a system level sequence tree based PHA, which can clarify the relationship among the major digital I&C systems. Two major phases are included in this sequence tree based technique. The first phase uses a table to analyze each event in SAR Chapter 15 for a specific safety related I&C system, such as RPS. The second phase uses sequence tree to recognize what I&C systems are involved in the event, how the safety related systems work, and how the backup systems can be activated to mitigate the consequence if the primary safety systems fail. In the sequence tree, the defense-in-depth echelons, including Control echelon, Reactor trip echelon, ESFAS echelon, and Indication and display echelon, are arranged to construct the sequence tree structure. All the related I&C systems, include digital system and the analog back-up systems are allocated in their specific echelon. By this system centric sequence tree based analysis, not only preliminary hazard can be identified systematically, the vulnerability of the nuclear power plant can also be recognized. Therefore, an effective simplified D3 evaluation can be performed as well.

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In PWR severe accident scenarios, involving a relocation of corium (core melt) into the lower head, the possible failure mode of the reactor pressure vessel (RPV), the failure time, the failure location and the final size of the breach are regarded as key elements, since they play an important part in the ex-vessel phase of the accident. In the framework of the LHF and OLHF programmes, the failure time and failure location predictions were obtained using numerical modelling and agreed reasonably well with the experimental values. However the final size of the failure is still an open issue. Analyses of both the LHF and OLHF experimental data (as well as of that from the Swedish FOREVER experiments) do not enable an assessment of the final size of the breach (in relation with the testing conditions and results). Indeed, the size of breach depends on the mode of crack propagation which is directly related to the variability in behaviour of the RPV material at high temperatures. To determine crack propagation and failure final size, 3D modelling would thus be needed with an adequate failure criterion which takes into account this material behaviour variability. This paper presents an outline of the methodology being used in a current research programme of IRSN, in partnership with CEA and INSA LYON. The aim is to model crack opening and crack propagation in French RPV lower head vessels under severe accidents conditions and to develop a new failure criterion for the 3D finite element models.

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It has identified that the pressure in the reactor coolant system (RCS) remains high in some severe accident sequences at the time of reactor vessel failure, causing direct containment heating (DCH). Intentional depressurization is an effective accident management strategy to prevent DCH or to mitigate its effects. Two strategies to mitigate DCH by depressurizing the RCS pressure are usually considered. One is called "early depressurization" and the other is called "late depressurization". The late depressurization is preferred as the severe accident management strategy because there are greater opportunities to recover plant functions prior to core damage, e.g. emergency core cooling systems (ECCS). According to the pressurizer PORVs design of the specific PWR plant, three late depressurization strategies are proposed in the analysis. During a station blackout TMLB' sequence, the phenomenological behavior is evaluated to determine how the strategies could depressurize the RCS pressure to a sufficiently low value where the effects of DCH will be mitigated.

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The Gas Reactor Test System (GRTS) is an experiment facility for examining the thermal hydraulic performance of the Generation IV, Very High Temperature Reactor (VHTR) during a Large-Break Loss of Coolant Accident (LB-LOCA). The LB-LOCA is defined as the double guillotine break of the VHTR coaxial inlet and outlet cross duct. Two system safety codes, MELCOR and RELAP5-3D were used to calculate core temperatures and flow rates during the LB-LOCA transient. Computational fluid dynamics modeling of the transient produced flow vectors and gas species distribution. The most important phenomenon during the transient is the lock-exchange process, which suppresses the onset of natural circulation until considerable molecular diffusion has occurred. The GRTS was designed based upon a hierarchical two tier scaling analysis whose primary objective was replicating the lock-exchange and natural circulation characteristics of the VHTR. The GRTS uses a scaled graphite core to represent the VHTR's graphite core. An in-depth scaling analysis was performed for the GRTS in order to ensure that it accurately simulated the VHTR thermal responses. RELAP5-3D thermal analyses, ProEngineer stress analyses, and combined FLUENT - STARCD CFD analyses have provided a system design that fulfills the GRTS mission statement. This paper discusses the design analyses and their implications on the GRTS capabilities. A discussion is also presented on the preliminary instrumentation plan. The GRTS will provide an extensive temperature map of the VHTR core outlet plenum and its core support, oxygen transport rates during the lock-exchange phenomenon, and thermal conduction rates from the core to the vessel. As a result of the GRTS using helium coolant at 950 C, the resulting experiment data is expected to considerably extend the U.S. database for high-temperature gas reactor operations. Finally, the discussion will present conclusions from the GRTS manufacturing and quality control processes that may benefit the VHTR design.

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In order to obtain the data on wall and interfacial friction forces for two-phase flows in a triangle tight lattice subchannel, adiabatic experiments were conducted for single- and two-phase flows under hydrodynamic equilibrium flow conditions. In the experiment, air was used as the test gas, while water and water with a surfactant as test liquids to know the effects of the reduced surface tension on the wall and the interfacial friction forces. The data showed that both the wall and the interfacial friction forces were higher in air-water with a surfactant system than air-water one. In the analysis, the respective data have been compared with the predicted values by existing correlations, and the existing correlations were modified to improve its prediction accuracy against the present data. The modified correlations can predict well the present data on the wall and the interfacial friction forces for both air-water and air-water with a surfactant systems.

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Nanofluids, colloidal dispersions of nanoparticles in a base fluid such as water, can afford very significant Critical Heat Flux (CHF) enhancement. Such engineered fluids potentially could be employed in reactors as advanced coolants in safety systems with significant safety and economic advantages. However, a satisfactory explanation of the CHF enhancement mechanism in nanofluids is lacking. To close this gap, we have identified the important boiling parameters to be measured and have deployed a pool boiling facility to measure them. The facility is equipped with a thin indium-tin-oxide heater deposited over a sapphire substrate. An infra-red high-speed camera and an optical probe are used to measure the temperature distribution on the heater and the hydrodynamics above the heater, respectively. The first data generated with this facility already provide some clue on the CHF enhancement mechanism in nanofluids.

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Spacer grids usually generate turbulence and the structure after the grids is a non developed structure. In transients we have overlaying on several effects. Therefore we investigate the effect of the grid spacer by investigating in general the effect of the not developed- or even the effect of the transient single- and multiphase flow. Many flow boiling processes in the nuclear technology are described still by correlations developed for steady developed flows. We propose in this paper how to extend the validity of such type of correlation for non developed flow or even to transient flows. The main idea is to look for those processes which are depending on turbulence and find the appropriate functional dependency. Then we will simply compare the dependences for both states: steady developed and steady- not developed or transient, and derive similarity lows. Analyzed are heat transfer at single steady developed and steady non developed and transient flows, heat transfer in multiphase steady developed and non developed and transient flows and droplet deposition in steady and transient flows. The relation between each couple of the processes is derived and proposed for practical use. The analysis presented above lead me to the following conclusions: 1. Increasing the frequency of the turbulence in single phase flow with respect to the steady developed flow increases the heat transfer depending on a square root function. 2. The increase of the turbulence leading to increase of the friction pressure drop is responsible for the increased heat transfer from or to the wall in the two phase flow region depending on a fourth root function of the Martinelli-Nelson multiplier. 3. Increasing the frequency of the turbulence in multiphase flow with respect to the steady developed flow increases the heat transfer depending on a square root function. 4. Increasing the turbulent kinetic energy of the continuum with respect to the steady developed flow increases the droplet deposition depending on a square root function.

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This work summarizes the system of partial differential equations describing multiphase, multicomponent flows in arbitrary geometry including porous structures with arbitrary thermal and mechanical interactions among the fields and between each field and the structure. Each of the fluids is designed as a universal mixture of miscible and immiscible component. The system contains the rigorously derived entropy equations which are used instead of the primitive form of the energy conservation. Based on well established mathematical theorems the equations are local volume and time averaged. The so called volume conservation equation allowing establishing close coupling between pressure and density changes of all of the participating velocity fields is presented. It replaces one of the mass conservation equations. The system is solved within the computer code system IVA together with large number of constitutive relationships for closing it in arbitrary geometry. The extensive validation on many hundreds of simple- and complex experiments, including the many industrial applications, demonstrates the versatility and the power of this analytical tool for designing complex processes in the industry and analyzing complex processes in the nature.

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This paper presents the results of POLCA-T code validation against an analytical solution at flow oscillating conditions. The POLCA-T code is a coupled computer code with 3D neutron-kinetics and thermal-hydraulics. Best estimate reactor analysis codes of this type incorporate a full 3D model of a reactor core into a system transient code. The coupled code provides a means to simulate interactions between reactor core behavior and plant dynamics. This validation paves the way for further code validation against cases with flow oscillating phenomena. Furthermore, the analytical test case provides a check for both the appropriate implementations of numerical methods and that the solution is not affected by their implementation. Moreover, a justification to minimize the error is also made by investigating the theta factor, Θ, in the time integration of mass conservation, energy and momentum equations. The comparison shows that the POLCA-T results and the analytical solution are almost identical. The effect from different time steps is very limited, when the time step changes a factor of 10. The results are most sensitive to the choice of parameter, Θ, for the numerical time integration method

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The turbulent heat transfer coefficients for the flow velocities (u=4.0 to 21 m/s), the inlet liquid temperatures (T_<in>=296.5 to 353.4 K), the inlet pressures (P_<in>=810 to 1014 kPa) and the increasing heat inputs (Q_0 exp(t/τ), τ=10, 20 and 33.3 s) are systematically measured by the experimental water loop. The Platinum test tubes of test tube inner diameters (d=3, 6 and 9 mm), heated lengths (L=32.7 to 100 mm), ratios of heated length to inner diameter (L/d=5.51 to 33.3) and wall thicknesses (δ=0.3, 0.4 and 0.5 mm) with surface roughness (Ra=0.40 to 0.78 μm) are used in this work. The turbulent heat transfer data for Platinum test tubes were compared with the values calculated by other workers' correlations for the turbulent heat transfer. The influences of Reynolds number (Re), Prandtl number (Pr), Dynamic viscosity (μ) and L/d on the turbulent heat transfer are investigated into details and, the widely and precisely predictable correlation of the turbulent heat transfer for heating of water in a short vertical tube is given based on the experimental data. The correlation can describe the turbulent heat transfer coefficients obtained in this work for wide range of the temperature difference between heater inner surface temperature and average bulk liquid temperature (ΔT_L=5 to 140 K) with d=3, 6 and 9 mm, L=32.7 to 100 mm and u=4.0 to 21 m/s within ±15% difference.

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The bypass phenomenon of safety injection flow in LBLOCA accident is an important factor in evaluating the function of safety injection system. In order to study this phenomenon, the LBLOCA accident of a PWR NPP with 600 MWe in China is simulated using SCDAP/RELAP5 code. The bypass phenomenon is analyzed quantificationally in different sequences. Meanwhile, the inventory of safety injection system is also evaluated.

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After the 1979 TMI-2 accident, the researches on the severe accidents of Nuclear Power Plants in the United States and other countries have made extensive progresses in both understanding the phenomena and developing the mitigation actions of the severe accidents. The U.S. Nuclear Regulatory Commission has indicated that the development of a plant specific accident management program is required to close the severe accident regulatory issue. In response to the policy, plant specific Severe Accident Guidelines (SAGs), which delineates the mitigation actions of core melt down accidents, is developed to support the operators and staffs in Technical Support Center (TSC) in dealing with those misfortunes. In SAGs of Boiling Water Reactor (BWR) with Mark III containment, specific mitigation actions are delineated to prevent the accumulation of hydrogen to a concentration that is high enough to threat the integrity of the containment. The mitigation actions include switch on igniters, and/or hydrogen recombiners, containment purge, and containment venting. The execution of these mitigation actions also has a potential of changing the course of accident progression. These impacts together with the effectiveness of these mitigation actions under different accident scenarios will be addressed in the paper based on the results of MAAP4 simulations. The surrogate plant analyzed is Kuosheng nuclear power plant of Taiwan Power Company. The plant employed a BWR VI reactor with Mark III containment. The rated thermal power of the plant is 2,894 MWth.

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Cernavoda NPP, the first CANDU in Europe, is one of the original CANDU 6 plants and the first CANDU 6 producing over 706 MWe. While the first series of CANDU 6 plants (which entered service in the early 1980's) have now reached the 2/3 of their 30 years design life, the Cernavoda NPP was put into service on the 2^<nd> December 1996. After 10 years of operation the Plant Life Management (PLiM) Program for Cernavoda should be an increasingly important program to Utility ("CNE - Prod") in order to protect the investment and the continued success of plant operation. The goal of the paper is to present some considerations related to Cernavoda NPP lifetime management. The Plant Life Management Program, known as PLiM Program is concerned with the analysis of technical limits of the safe operation - from the point of view of nuclear safety - in NPP units, aiming at attaining the planned 30 years life duration and its extension to 40 or even 50 years of safe and economical operation. For the CANDU[○!R] reactors the so-called PLiM and PLEX Programs are just applied. These are applied research programs that approach with priority the current practices for assessing the capability of safe operation within the limits of nuclear safety (fitness-for-service assessment). These programs also approach inspection, monitoring and prevention of degrading due to the ageing of critical systems, structures and components (CSSCs). As each nuclear plant is somewhat different in its components and systems, materials composition, procurement, construction, and operational history, directed research and development programs into materials behavior, monitoring techniques, and methods to mitigate ageing are required to support the lifetime management. Over the past 6 years, INR Pitesti (Institute for Nuclear Research - Romania) has been working on R&D Programs to support a comprehensive and integrated Cernavoda NPP Life Management Program (PLiM) that will see the Cernavoda NPP successfully and reliably through the design life and beyond. A comprehensive and integrated Plant Life Management (PLiM) program applicable to Cernavoda NPP has been proposed. A time scheduling of the multiphase Cernavoda PLiM program has been suggested.

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This study is about the safety of nuclear reactor core submitted to seismic loading. In order to reduce the incertitude margin of the present day codes we propose to develop a numerical code including the non linear behavior of the fluid/structure coupling. The challenge of this work is to find out a tractable model taking the structure complexity into account. In this paper we model the nuclear reactor core mechanical behavior including the dynamics of both fuel assemblies and fluid. Each rod bundle is considered as a deformable porous media, so the velocity field of the fluid and the displacement field of the structure are defined in the whole domain space. Fluid part and structure part are in a first time considered separately, and in second time, the two parts are coupled. The motion equations of the structure are obtained by a Lagrangian formulation, and to allow the fluid structure coupling, the motion equations of the fluid are obtained by an Arbitrary Lagragian Eulerian formulation. The finite elements method is applied to spatially discretize the equations. Simulations have been performed to analyze the influence of the fluid and structure characteristics, phenomena observed by the experience have been reproduced qualitatively.

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A three-dimensional one-way bubble tracking method is a promising numerical method for calculation of time-spatial evolution of gas-liquid interfacial configuration with use of a little computing resource. Since the method has been applied to only an adiabatic air-water bubble flow, the method was developed for analysis of a boiling flow in this study. One-dimentional Eulerian equation of energy conservation for a continuous liquid phase and interface heat transfer equation for dispersed bubbles were introduced. Then, radial liquid temperature distribution, wall heat transfer between a heated wall and subcooled liquid, bubble generation on a heated wall and expansion or condensation of bubbles were taken into account. The developed method was applied to the boiling flow experiment and radial void fraction distribution was compared. It was confirmed that the method could give good prediction of tendency of the void fraction distribution in the boiling flow.

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The skyshine dose from the turbine building (T/B) at a BWR commercial plant in Japan was calculated using a Monte Carlo code, a parallel version of MCNP5. The inner and outer walls of the T/B, shielding material around a high pressure turbine (HP-T), the HP-T itself, low pressure turbines (LP-Ts), and the pipes installed on the operating floor in the T/B were considered and modeled in detail. Much computing time is required to perform the MCNP calculation with the detailed calculation model mentioned above. In this work, a 10-node cluster with 3.2-GHz CPUs was applied to the skyshine calculation for reducing the computing time. On the other hand, the skyshine doses were measured with a NaI scintillator as the pulse-height spectra at a few points around the same T/B that was analyzed in the above calculation. In order to validate the MCNP calculations, the results were compared with the measured data, and the skyshine dose distributions calculated by MCNP were also compared with those obtained by other calculation codes--SKY-H (a code similar to G-33) and SKYSHINE code. It was found that MCNP yields a sufficiently precise skyshine dose distribution from the T/B at a BWR plant and is suitable for skyshine estimation.

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A power uprate of Forsmark's Unit 3 from 109 % to 125 % will be implemented during the 2010 refuelling outage. This uprate implies an increase in gamma heating of the core shroud which could lead to temperatures higher than the design thickness-mean temperature, 300℃, according to ASME regulations. To estimate the temperature distribution in the core shroud, a CFD model of the core bypass has been developed using the commercial code CFX 5. The model consists of the core bypass, from the lower core support plate up to the core grid, and the upper plenum, limited from above by the core shroud cover including the steam separator inlets. The bypass flow enters the computational domain at the level of the core support plate. The two-phase core flow enters the computational domain at the level of the core grid where it entrains and merges with the bypass flow. Both flows have been estimated through the POLCA code. The conjugate heat transfer at the core shroud inner wall comprises the gamma heating from the core, considered as volume distributed heat sources, and subcooled boiling of the bypass flow. The effect of subcooled boiling has been taken care of by using the model by Kurul and Podowski. Using a conservative gamma heat source distribution, leads to local temperatures slightly higher than the design temperature, with a maximum thickness-mean temperature that exceeds the temperature limit by approximately 4℃. If a higher temperature limit is accepted, the ASME regulations are not fulfilled, but the consequence is a minor change in the design stress intensity value Sm according to ASME. Using a somewhat realistic gamma heat source distribution, the results show that the maximum thickness-mean temperature is well below the design temperature.

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Assessment of integrity of reactor pressure vessels during Pressurized Thermal Shocks (PTS) are usually determining whole reactor pressure vessel lifetime. WWER (PWR type reactors of Russian design) reactor pressure vessels are characterized by relatively thick austenitic cladding on their inner surface. Existence of cladding results in high residual stresses in inner surface area and thus in high stress intensity coefficients of postulated/calculated under clad type defects. To evaluate a proper and experimentally supported procedure for assessment of WWER reactor pressure vessel integrity during PTS regimes, during years 2005 - 2006, a series of semi-large scale experiments on specimens containing underclad (embedded) crack were performed in NRI Rez. The experiments were performed within EU PHARE project EUROPAID/116529/D/SV/CZ, in cooperation with VTT, FNS, TVONS (Finland) and Tecnatom (Spain). The aim of the project was to investigate fracture mechanics properties of cladding, in particular, to establish the role of cladding in the fracture/failure process of the specimen, and to exploit the obtained knowledge in the procedure for evaluation of integrity of WWER reactor pressure vessels.

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Reactor pressure vessels are components that are practically not replaceable and thus they determine lifetime of the whole nuclear power plant (NPP). Lifetime of reactor pressure vessels are determined by level of material degradation as well as by existence of defects found during in-service inspection and their size and location. According to some codes, it is necessary to have a procedure for repair of potential defects found during in-service inspection that can be non-acceptable from point of view of assurance of required RPV lifetime. A wide experimental program was realized to propose, check and approve a procedure for repair of potential defects in underclad region of WWER-440 MWe (PWR type reactors of Russian design) reactor pressure vessels during operation. Experimental program was realized on as-received materials as well as on materials subjected to simulated ageing to level representing expected material embrittlement after design reactor lifetime. Main requirements for this procedure were: RPV will not be pre-heated during repair welding, and no post-weld heat treatment will be applied. Maximum repair depth will be 40 mm including austenitic cladding thickness (8 to 10 mm). Such procedure was developed and checked by series of mechanical testing (hardness, fracture toughness, impact notch toughness as well as by detailed metallography studies).

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Lifetime of reactor pressure vessels practically depends on a level of degradation of RPV material properties during operation. The most important degradating mechanism of RPV materials is usually radiation damage, characterized by values on neutron fluence on one side and radiation embrittlement of RPV materials on the second side. WWER reactor pressure vessels in the Czech Republic are a subject of a very thorough and complex monitoring program, that includes: - Standard material surveillance program containing of WWER-440 RPV materials - base metal, weld metal, heat affected zone, but irradiated with high lead factor (13 to 18), - Supplementary surveillance program of WWER-440 RPV materials, including additionally austenitic cladding materials, IAEA reference material JRQ irradiated with low lead factor (2 to 3) with parts subjected to annealing and re-irradiation after annealing, - Modified surveillance program of WWER-1000 RPV materials - base metal, weld metal, heat affected zone, cladding materials, IAEA reference JRQ material irradiated in low lead factor (2 to 3) near RPV inner beltline region, - Integrated surveillance specimen program for WWER-1000 reactor including materials from NPP Temelin (Czech Republic), Belene (Bulgaria), Kalinin (Russia) and Ukranian NPPs, - Continous ex-vessel monitoring of neutron fluence on outer RPV surface for both WWER-440 and WWER-1000 plants, - Neutron fluence determination on inner RPV surface (austenitic cladding) using special technique for removal of specimens from cladding for Nb activity measurements, - Ex-vessel temperature measurements during RPV operation. All these programs serve for precision of operation conditions and determination of degradation of RPV materials for RPV integrity and lifetime assessment.

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The introduction of fast breeder reactors (FBRs) requires Pu be recovered from light water reactors (LWRs) spent fuel. The "Flexible Fuel Cycle Initiative (FFCI) " can supply enough Pu and holds no surplus Pu, while responding flexibly to future technical and social uncertainties. In this paper, the potential of FFCI to increase economy of the fuel cycle system was investigated. On the other hand, during the FBR introductory period, Pu from LWR spent fuel is used for startup of FBRs. But the FBR core being loaded with Pu from LWR spent fuel has a larger burnup reactivity than the core being loaded with Pu from the FBR multi-recycling core. The increased burnup reactivity may reduce the cycle length of the FBR core. In this paper, an FBR transitional core concept to handle the issues of the FBR introductory period was investigated. The results obtained through this study are as follows. (1) The FFCI has a potential to flatten the reprocessing amount of LWR spent fuel and to increase economy of the next fuel cycle system. (2) Minor actinides - mixed FBR transitional core has a potential to maintain the operation cycle length even supposing use of Pu from LWR spent fuel.

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In the present study, the FLAME large-scale hydrogen combustion tests F-8 and F-22 performed at Sandia National Laboratories, USA, were simulated using the TONUS CFD code. In test F-8 a configuration without obstacles and no top venting was used. The objective of the calculation of test F-8 was to investigate the effect of two different boundary condition specifications at the test channel exit as well as the effect of mesh refinement. The difference between the tested boundary conditions was observed to be significant. The CREBCOM combustion model was used in the simulation and attention was paid at finding an appropriate model parameter K_0 for the current test. Test F-22 which includes obstacles resulting in a 33% blockage ratio in the test channel was modeled by the k-ε turbulence model with Eddy Break-Up reaction kinetics. Both CREBCOM and EBU models were found to perform qualitatively well for both 2D and 3D models of the system. However, it was observed that the correlations presently available for the key parameter K_0 of the CREBCOM model are not suitable for the FLAME tests. In test F-22, an additional challenge was provided by the unclear conditions near the exit of the channel and it was noticed that high local pressure peaks were formed near the obstacles in the channel.

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This paper deals with the results of the preliminary scoping analyses performed to assess the transmutation capability of an Accelerator Driven System cooled by LBE and loaded with MA-dedicated fuel assemblies. It also describes the constraints taken into account in the design of such cores and reports the thermo-mechanical behaviour of the hot fuel rod along the cycle under related irradiation conditions. The selected fuel is a U-free CERCER, the fissile phase being a solid solution of plutonium and MA oxides dispersed in a magnesia matrix. The neutronic core and thermo-mechanical fuel rod analyses were performed by the ERANOS and TRANSURANUS codes, respectively. The preliminary analysis shows that there is a good compromise between transmutation and core performance for an ADT having a power ranging from 200 to 400 MWth. Of course the increase of the core size has a significant implication on the overall plant architecture (accelerator, spallation module, decay heat removal system, etc.), which is discussed in the paper. The fuel pin analysis shows that the cladding integrity is assured during in-reactor residence.

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EDF R&D branch and Basic Design department (SEPTEN) are currently working on an innovative methodology to provide an electric overpower to PWRs having feedwater tank and under frequency control operation. This paper presents the principles and the first simulation results of this new control, named "PARI", in various cases. It also presents a statistical analysis of the 50 Hz European frequency grid behavior in order to give elements to evaluate the risk of reaching extreme levels in the feedwater tank or in the condenser, and then to stop providing a part of the contractual power reserve. Thanks to this new control it appears that the power set point could be increased by 0.5 to 1% of Nominal Power, considering the few months of frequency data availability. This overpower can only be obtained from PWRs having a feedwater tank with a sufficient volume (34 among the 58 French plants) and when under frequency control operation (about one third of operation time). A financial evaluation based on the elements mentioned above comparing gains and costs, like changing the control hardware, has shown the interest of this proposal. Grid safety is the other decisive element for the wide spread use of this new methodology. Thus not only producers but also regulation authorities have to discuss on the acceptability of such an innovation. The "PARI" project is still under development and decision of testing out or not the new controller should be taken soon. Any company, authority, institute or agency interested in the project is welcome to contact EDF.

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This paper presents the recent applications of electrical impedance tomography technique for uses in nuclear fuel process and plant decommission, including, the measurement of solids distribution in pipelines for hydraulic conveying control, a remote method of measuring the vortex depth in continuous stirred tanks into which reactants are added to produce a heavy metal precipitate, monitoring the supersaturation of the selection of a suitable organic salt in batch temperature-controlled crystallisation, linear EIT sensors for observation of both sediment bed properties and of the sedimentation process in a nuclear waste storage tank. These applications demonstrate EIT technique has particular advantages for measurement and control of nuclear plant operation.

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Calibration of recently installed proportional counter at the Hot Cells Facility of the Jozef Stefan Institute was performed. Instrument was calibrated for determination of total beta activity, Sr-90 and Pb-210. Detection efficiencies for K-40, Sr-90, Y-90, Pb-210 and Bi-210 were determined, allowing for more accurate determination of the particular nuclide as a single K-40 efficiency. In addition, self-absorption curves for different surface densities for the nuclides mentioned were derived. Two empirical equations for faster and more accurate determination of Sr-90 and Pb-210 were derived. These two equations consider differences in surface density and in-growth of Y-90 and Bi-210 respectively. The detection efficiencies obtained ranged from 10 to 52 %, depending on the nuclide, surface density and chemical compositions of the salts used or precipitates obtained following radiochemical separation in the experiment. As an example, total beta activities in five samples of mineral waters according to the relevant ISO standard were determined. All procedures and formulae developed include calculation of minimal detectable activities and uncertainty budgets for the determinations concerned.

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The present work is an attempt to develop a laser cutting method for cutting and removing stainless steel liners from concrete walls and floors in nuclear facilities. The effect of basic laser cutting parameters such as energy, cutting speed, assist gas flow etc. were first studied through cutting experiments on mock-up concrete specimens lined with 3mm thick stainless steel sheets using a 1kW Nd:YAG laser. These initial studies were followed by further studies on the effect of unevenness of the liner surface and on a new method of confining contamination during the cutting process using a sliding evacuation hood attached to the laser cutting head. The results showed that laser cutting is superior to other conventional cutting methods from the point of view of safety from radioactivity and work efficiency when cutting contaminated stainless steel liners.

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Thermal stratification after a reactor scram may cause significant thermal stress in the reactor vessel and components. Water experiment using an 1/10^<th> scaled upper plenum model was carried out to investigate thermal hydraulic issues in an advanced loop-type sodium cooled fast reactor, which was designed by Japan Atomic Energy Agency. A permeable upper inner structure (UIS) is adopted, which has a radial slit to simplify the fuel handling system. This slit also allows high velocity flow through the UIS. The experiments showed steep temperature gradient and large temperature fluctuation at the stratification interface near the UIS slit, which was caused by local impingement of the jet through the UIS slit. Parameter experiments for core flow rate and temperature difference indicated that the rising speed of the stratification interface was dependent on Richardson number and the temperature gradient of the stratification interface was also influenced. Configuration, i.e., a cylindrical plug in front of the slit and a perforated outer shell of the UIS were examined to mitigate the thermal stress at the stratification interface. The temperature gradient was reduced greatly in a case where the plug was located at a lower position near the core in the upper plenum.

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This paper describes a quantitative evaluation of annular linear induction pump efficiency. The evaluation was carried out through an experiment and a numerical analysis. Comparing the experimental data and the numerical result of the pump efficiency, their tendencies are in agreement, but the discrepancy becomes conspicuous with increase of two non-dimensional numbers, the magnetic Reynolds number and the interaction parameter. The discrepancy is probably due to some azimuthal non-uniformity of the magnetic field and of the sodium velocity. In fact, the experimental data reveals that the magnetic field and the sodium velocity are azimuthally non-uniform and their non-uniformity increases with increase of the magnetic Reynolds number and the interaction parameter. In addition, the numerical result indicates that a non-uniform magnetic field and sodium velocity bring about a decrease of the developed pressure and an increase of the power factor, which results in the decrease of the pump efficiency. These experimental data and numerical result suggest that high-efficiency large-scale electromagnetic pumps may be realized by optimizing the pump specifications such as the supply frequency and the pole pitch so that the magnetic Reynolds number and the interaction parameter should be as small as possible but close to the critical ones.

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In relation to the design of an innovative Water Reactor for FLexible fuel cycle (FLWR), investigation of thermal-hydraulic performance in tight-lattice rod bundles of the FLWR is being carried out at Japan Atomic Energy Agency (JAEA). The FLWR core adopts a tight triangular lattice arrangement with about 1 mm gap between adjacent fuel rods. In view of the importance of accurate prediction of cross flow between subchannels in the evaluation of the boiling transition (BT) in the FLWR core, this study numerically simulated steam-water two-phase cross flow between two modeled subchannels of tight-lattice rod bundle for the FLWR by using a detailed two-phase flow simulation code with an advanced interface tracking method (named TPFIT), statistically evaluated the simulation results, and clarified mechanisms of cross flow for developing a model. The effects of flow pattern, inlet and outlet of mixing section, and gap spacing on cross flow, and the local and general characters of cross flow were extensively investigated. It was confirmed that there exist strong correlation between differential pressure and gas/liquid mixing coefficients. Mechanistically, cross flow results mainly from differential pressure. Liquid cross flow occurs locally and its time lag is negligibly small (less than 2 ms). Gas cross flow, however, occurs across the whole mixing section, and propagates with main stream in the mixing section. The time and space lags are relatively large and can be determined from average velocities in mixing section. The local properties may be negligible as well.

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We have developed an integrated simulator for the maintenance optimization of LWRs (Light Water Reactors) based on PFM (Probabilistic Fracture Mechanics). The aim of the simulator is to provide a method to optimize maintenance activities for representative components and piping systems in nuclear power plants totally and quantitatively in terms of safety, availability and economic efficiency (both from cost and profit). The simulator has a function of visualization of the calculated results by a divided multi-dimensional visualization method. The simulator will also provide a guideline regarding social acceptance of the risk-based decision makings. This study has been being conducted under "Innovative and Viable Nuclear Energy Technology (IVNET) Development Project" financially supported by Japanese METI (Ministry of Economy, Trade and Industry).

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The precooler is very important equipment as cooling source and helium will be cooled finally down to the lowest temperature in the HTGR with direct Gas Turbine cycle. Three lumped parameter models were developed to research on dynamic performances of the precooler. A lumped parameter model (model 1) neglecting thermal capacity of precooler's metal wall was proposed to study on the precooler's dynamic characteristics and meanwhile, to find out effect of wall thermal capacity to the model. A linear model (model 2) was proposed to study on the precooler's dynamic performance, including wall temperature, but it failed to obtain correct results because of wrong characteristic temperatures. A modified model to the linear model (model 3) could achieve correct and reasonable results. Comparing the model 1 with model 3, it was found that the model 1 could forecast dynamic performance of the precooler, except for the wall temperature. Loss of cooling water accident, a severe accident related to the precooler was studied with the help of the model 3. When loss of cooling water accident happened, it was very important to maintain cooling water to some degree, which could make the cycle running safely until stop.

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The contribution describes the new instrumentation and control (I&C) system of the VR-1 training reactor at Czech Technical University in Prague. The reactor was put into operation in the 1990. The original reactor I&C seemed to be obsolete and their replacement was being carried out. The new reactor I&C consists of human-machine interface, control system and protection system. The human-machine interface was designed with respect to functional, ergonomic and aesthetic requirements. Czech and English interface versions are available. The control system is based on a high quality industrial PC mounted in a 19'' crate. The operating system of the PC is the Microsoft Windows XP with the real time support RTX of the VentureCom Company. The protection system consists of the operational power measuring (OPM) and the independent power protection (IPP) systems The OPM is a full power range system. The IPP system works in the two highest power range decades. The computers of both systems calculate the reactor power and power rate, compare them with the safety limits, and if they are exceeded, the safety action is initiated. The OPM and IPP systems are diverse; different types and location of chambers, completely different hardware, software algorithms, development tools and teems for software manufacturing.

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Most valves installed in Optimized Power Reactor 1000 (OPR1000) have been qualified using the required response spectra or zero period acceleration (ZPA) having maximum 5.0g for safety shutdown earthquake (SSE) loads. The ground acceleration for the SSE loads in Advanced Power Reactor 1400 (APR1400) has increased to 0.3g from 0.2g which has been applied to OPR1000. This increase would result in higher response acceleration in piping system as well as in-line components such as active valves. If pipe and equipment interface requirements are checked during the design stage, the equipment allowable rather than the pipe stress allowable seems to be exceeded. In order to reduce the interface loads to an acceptable level, many supports need to be added to the piping systems. It became a sensitive issue to set the appropriate acceleration value for the active valve qualification to a reasonable level since an increase in the seismic loads may cause an extra expenditure to the equipment suppliers in qualifying these components. Furthermore, in the process of compromising between piping stresses and pipe/equipment interface loads, it is a challenge to keep a good balance between adding supports and not losing flexibility of the piping system. Because of these concerns, various kinds of case studies have been performed to evaluate the effect of the increased seismic design loads associated with the qualification of in-line components (valves) of the APR1400. As a result of analyzing more than 430 cases studies, the maximum seismic response accelerations are ranged from 5.0g to 5.7g. In this regard, 6g acceleration value could be regarded as an appropriate acceleration value for the equipment qualification considering 20% margin to take account for possible deviation from the design condition that could happen during the construction stage.

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R&D project to investigate thermal-hydraulic performance in tight-lattice rod bundles for Innovative Water Reactor for Flexible Fuel Cycle (FLWR) has been progressed at Japan Atomic Energy Agency in collaboration with power utilities, reactor vendors and universities since 2002. In this series-study, we will summarize the R&D achievements using large-scale test facility (37-rod bundle with full-height and full-pressure), model experiments and advanced numerical simulation technology. This first paper described the master plan for the development of design technology and showed an executive summary for this project up to FY2005. The thermal-hydraulic characteristics in the tight-lattice configuration were investigated and the feasibility was confirmed based on the experiments. We have developed the design technology including 3-D numerical simulation one to evaluate the effects of geometry/scale on the thermal-hydraulic behaviors.

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The application of Loviisa Nuclear Power Plant (NPP) operating license extension for 20 years was submitted to the Ministry of Trade and Industry of Finland on November 1, 2006. For the application a great number of different studies, analyses, clarifications etc. were needed to show to the authorities that the plant can be operated safely for another 20 years. The main tool in the safety analysis work was APROS (Advanced PROcess Simulator) simulation software, which has been developed in Finland during the past 20 years. This paper deals with the cold leg loss-of-coolant accidents (LOCA) which were analyzed using APROS. The break size was between 140 - 260 cm2. Models with different descriptions of the core region were applied in the analyses. The first model had an average core with one flow channel and separate hot rods modeled in isolated sub channels. In the second phase the hot rod was implemented in a hot fuel assembly and so called hot channel analyses taking boundary conditions from the previous analyses were run. Finally the analysis was carried out using a special Loviisa large break LOCA analysis model in which the core is divided into seven parallel channels. All three analyses resulted in a qualitatively similar peak cladding temperature behavior but the maximum value of cladding temperature depended on the model which was used. This paper presents the results of these analyses focusing on the behavior of fuel cladding temperature. The paper also explains why the peak cladding temperature behavior depends on the core model.

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Methods to systematically analyse existing nuclear power plants (NPP) regarding the adequacy of their existing protection equipment against external hazards, e.g. flooding, can be of deterministic as well as probabilistic nature. In the past the adequacy of the protection measures has been assessed only on a deterministic basis. The German regulatory body has issued probabilistic safety assessment (PSA) guidelines, which had been elaborated for a comprehensive integrated safety review of all NPP in operation. Amongst others the guidelines imply, that probabilistic considerations regarding external flooding are required. This paper presents a newly developed graded approach for the probabilistic assessment of external flooding. Main aspects are explained such as the underlying probabilistic considerations and the mathematical procedures for the calculation of exceedance frequencies, which have recently been developed and issued as part of the German Nuclear Safety Standard. Exemplarily it has been investigated if extreme events such as tsunami waves could be a hazard for NPP at coastal sites in Germany. Here it could be shown that due to limited source mechanisms and the specific morphological conditions in the North Sea no dedicated measures for protection against tsunamis in the German Bight are necessary.

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In FCIs, the fragmentation process is a key factor to determine the ratio of heat transferred to power. So, in this paper, the fragmentation process of melt droplets triggered by boiling effect is investigated. The dynamic boiling process is modeled by writing a momentum equation for vapor film dynamics, an energy equation for each region of the droplet, coolant vapor and liquid and linking each region by the appropriate boundary conditions. The integral approach is used in each region for the energy equations where the differential equation is integrated over the region and a temperature profile is assumed. By using these developed models, the fragmented mass of the droplet triggered by boiling effect can be calculated. The result shows that the fragmentation rate is large than that given by hydrodynamic model. The parameter, which is related to the droplet condition and the acceleration exerted on the droplet surface, is a function of pressure in the vapor film.

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The availability of large amounts of reactor and weapons grade plutonium in the world shows the necessity of anticipating situations for the use and disposition of it. Because Light Water Reactors (LWRs) prevail on the stage of electric energy generation by nuclear power, it is important to take into account the potential of these reactors to reduce the plutonium inventory. Several studies performed in Pressurized Water Reactors (PWRs) show that reactor and weapons grade plutonium can effectively be burned in these reactors, in assemblies with fertile-free fuel, and maintaining reactivity control and other safety issues at least comparable to those related to the standard fuel normally used. The Instituto Nacional de Investigaciones Nucleares, currently carries out research on diverse alternatives to use Inert Matrix Fuel (IMF) as an option to fuel reloads for the two BWR/5 Units at the Laguna Verde Nuclear Power Plant. This work presents first the neutronic analysis of a fuel assembly conceptual design, which contains a combination of plutonium oxide (in an inert matrix) fuel rods, uranium oxide fuel rods, and uranium oxide with gadolinia fuel rods. Then, simulations for three different fuel assembly reload options were performed for Unit 1. Results of reactor operation from the different reload options are presented. The results obtained with reload fuel using inert matrix fuel assemblies observe a decrease in the length of operation cycle in the plant. However, the mass of uranium used is minor to require for make all fuel assemblies.

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The mechanism of the transition from bubble flow to slug flow with co-current gas and liquid flowing through vertical tubes under rolling condition is studied. According to the experimental observation, it is found that the dispersed bubble flow pattern is periodically changed with the rolling motion. When the test section is deviated from the vertical condition, bubbles tend to flow at the upper part of the tube, and the bubble density reaches its maximum while test section rolling to the place with maximum incline angle. Besides, through the analysis of forces acted on moving bubbles under rolling condition, it is also found that the radial components of forces keeping dispersed bubbles at the upper part of the tube reach their maximums at maximum incline angle. Therefore, the transition from bubble flow to slug flow is most likely to take place at this moment. Based on the experimental data, a new correlation for predicting the transition from bubble flow to slug flow is proposed. The results show that the model predictions coincide well with the experimental data when gas superficial velocity is larger than 0.4 m/s.

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Since many clandestine nuclear activities in the Middle East, the Korean peninsula and other areas of the world have been disclosed during the last 15 years, a series of the counter-measures have been proposed and taken. The Destructive Analysis (DA) for environmental sampling for safeguards (ESS), based on the Additional Protocol to the NPT Safeguards Agreement has played very important role since 1990s. It is designed to detect nuclear materials in the environment of a facility that may reveal the presence of an undeclared nuclear activity such as plutonium recovery from irradiated fuel or isotopic enrichment of uranium. Among the verification tools for the comprehensive Safeguards agreement, DA for nuclear material accountancy and its verification is extremely important for drawing quantitative Safeguards conclusions. In particular, DA is the best approach for detecting "bias defects", which arise when small amounts of nuclear material are diverted over a protracted length of time. Nuclear accountancy and verification based on the DA with a state-of-the-art determination technique providing highest possible measurement accuracy is the fundamental and essential technology, without which Safeguards cannot be concluded. Timely and more accurate analytical services are current challenges against getting larger scale of nuclear fuel cycle. This paper describes the role of DA for both environmental sampling and verification of nuclear material accountancy.

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Void fraction distribution and its estimation methods in tight-lattice rod bundles for R&D of Innovative Water Reactor for Flexible Fuel Cycle (FLWR) core have been investigated based on many kinds of void fraction data and summarized in this paper. Void fraction data were measured by neutron radiography technique, quick-shut-valve technique, and electro void fraction meter which were developed by JAEA using 7-, 14-, 19- and 37-rod bundle test sections under from atmospheric pressure to 7.2 MPa. The rod geometry in the test sections simulates the FLWR fuel assembly, that is, gap between rods of 1.0 or 1.3 mm, rod diameter of 12-14 mm with triangular tight-lattice arrangement. Mass velocity ranged from 121 to 2000 kg/m^2s, and mainly controlled to 400 - 600 kg/m^2s as nominal condition. And, also spacer effect test was evaluated in this project. In this paper, (1) boiling flow characteristics on void fraction in the tight-lattice rod bundle, (2) comparison of advanced numerical analysis codes with 3D void fraction data and (3) TRAC-BF1 code and drift-flux model for 1D void fraction prediction were discussed. Followings were made clear from this study; (a) lower void fraction was observed at peripheral than at center of rod bundle and higher void fraction was observed at rod gap part of lower core and at center of subchannel of upper core, (b) numerical analysis codes calculated the similar void fraction distribution to the data, and (c) TRAC-BF1 code and drift-flux model tended to overestimate the void fraction at lower quality region, on the other hand at higher quality, those methods tended to same characteristics to the data. It was confirmed throughout the void fraction study that several special features were existed in the tight-lattice rod bundle but a lot of advanced numerical analysis codes can be applied to the tight-lattice core design.

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Effect of rolling motion on natural circulation capacity was studied experimentally and theoretically. Experiments were conducted under the conditions of rolling and unrolling motions. The experimental results show that natural circulation capacity decreases under rolling motion condition. A mathematic model was developed to calculate the natural circulation capacity under rolling motion condition, considering the characteristics of natural circulation, the model was modified. The calculated results agree with experimental data well. Effect of rolling motion on natural circulation was analyzed through calculation and the following conclusions were obtained: (1) The increase of flow resistance coefficient is the main reason that the natural circulation capacity decreases under rolling motion condition; (2) Non-uniform distribution of fluid mass in the pipe has also influence on natural circulation capacity.

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Inorganic membranes exhibit greater mechanical durability in some operations than polymeric membranes. They do not suffer from the performance degradation that was resulted from compaction of the membrane structure under pressure or ageing. Membrane permeation combined with complexation was tested for radioactive wastes processing purpose. Sodium poly-acrylic acid was selected as the complexing agent, the efficiency of inorganic membrane with cut-off 1kD, 3kD, 8kD assisted by sodium poly-acrylic acid of different molecular weight were compared. The removal efficiencies of nuclides such as strontium, cesium and cobalt by were compared. The flux and retention factors of different membrane system were compared. The impacts of complexation agent concentration on permeate flux and retention factors were studied. The long term behaviours of the membrane system were also studied. Diatomite filter was selected as the pretreatment method, and the efficiency of diatomite filter for pretreatment was investigated also.

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KAERI is in the process of carrying out the Nuclear Hydrogen Development and Demonstration (NHDD) Program. The indirect cycle gas cooled reactors that produce heat at temperatures in the order of 950℃ are being considered in the NHDD program. For the indirect gas cooled reactors, the intermediate hear exchanger (IHX) is one of the key components. For the NHDD program we are in the process of establishing a conceptual design of the IHX and a hot gas duct (HGD). The conceptual design activities include a preliminary design of the IHX and the HGD, the appropriate material selection, identification of the design code criteria, and the selection of an exchanger type.

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