bluuebunny/crossref_metadata_2025 huggingface dataset:
Twenty Years of Operation in NPP Krsko
Authors: Nemcic Kresimir , Brems Robert
Published: March 2009

This paper presents the main design and fabrication improvements that were included in the new steam generators that were installed at Nuclear Power Plant Krsko in Slovenia in 2000. These improvements were a result of an on-going R & D effort associated with steam generator technology, aimed at increased reliability and better maintainability of new steam generators. The paper also provides basic information related to 20 years of operation of NPP Krsko with an emphasis on subjects related to steam generator performance and degradation. These include inservice inspection results on steam generators up to replacement, corrective actions, corrosion-erosion counter-measures, and replacement of condenser and moisture spearator reheaters with improved material. The paper also provides data regarding changes in feedwater and steam generator water chemistry, with water chemistry results before and after steam generator replacement. This paper shows that materials issues are very important for steam generator reliability, and it gives the reasons why specific materials were selected for replaced components. Finally, the paper demonstrates how the strategy used at NPP Krsko prolonged the useful life of the old steam generators until replacement steam generators could be designed, fabricated, and installed.
Mixed Reload Design Using MOX and UOX Fuel Assemblies
Authors: Rami´rez Sa´nchez J. Ramo´n , Perry R. T.
Published: March 2009

As part of the studies involved in plutonium utilization assessment for a Boiling Water Reactor, a conceptual design of MOX fuel was developed, this design is mechanically the same design of 10×10 BWR fuel assemblies but different fisil material. Several plutonium and gadolinium concentrations were tested to match the 18 months cycle length which is the current cycle length of LVNPP, a reference UO2 assembly was modeled to have a full cycle length to compare results, an effective value of 0.97 for the multiplication factor was set as target for 470 Effective Full Power days for both cycles, here the gadolinium concentration was a key to find an average fisil plutonium content of 6.55% in the assembly. A reload of 124 fuel assemblies was assumed to simulate the complete core, several load fractions of MOX fuel mixed with UO2 fresh fuel were tested to verify the shutdown margin, the UO2 fuel meets the shutdown margin when 124 fuel assemblies are loaded into the core, but it does not happen when those 124 assemblies are replaced with MOX fuel assemblies, so the fraction of MOX was reduced step by step up to find a mixed load that meets both length cycle and shutdown margin. Finally the conclusion is that control rods losses some of their worth in presence of plutonium due to a more hardened neutron spectrum in MOX fuel and this fact limits the load of MOX fuel assemblies in the core, this results are shown in this paper.
Evaluation of Mechanical Properties of Nuclear Materials Using Non-Destructive Ball Indentation Technique
Authors: Mathew M. D. , Murty K. Linga
Published: March 2009

Integrity of structural components depends on the deformation and fracture behavior of materials. For evaluating the material condition in-service, it is generally not feasible or practical or advisable to cut samples from operating structures. Non-destructive testing (NDT) techniques are required to evaluate the mechanical properties. Although several NDT techniques such as ultrasound, magnetic strength, Barkhausen noise, microhardness etc., are employed for estimating the mechanical property degradation, these methodologies are generally empirical and indirect. Automated Ball Indentation (ABI) is a non-destructive testing technique for direct measurement of mechanical and fracture properties of metallic engineering materials. Because of the small area over which the test is carried out, it is possible to determine point to point variations in the mechanical and fracture properties, such as those that exist in weldments. Although ABI technique is non-intrusive, it is a state-of-the-art mechanical test that measures directly the current/local deformation behavior of the material. In this paper, we present results from studies on the application of ABI technique to determine tensile and fracture properties of ferritic steels, an austenitic stainless steel, a nickel base superalloy and Zircaloy in different thermo-mechanical conditions. The effects of aging and cold work on these properties were determined from the ABI tests. Gradients in mechanical properties of ferritic steel welds, particularly in the narrow heat-affected zone, were clearly established. ABI technique was found to be useful in determining the anisotropy in the tensile properties of Zircaloy cladding tubes. The technique has potential as a non-destructive method for assessing structural integrity of aged components.
Rupture and Internal Pressurization Creep of Nb-Modified Zircalloy-4 Tubing: Application to Dry Storage Feasibility
Authors: Murty K. K. , Zhou Y. , Devarajan B.
Published: March 2009

Burst rupture properties of Zircaloy-4 and Nb-Modified Zircaloy-4 are investigated at temperatures ranging from 638K to 843K by internal pressurization of closed-end tubing samples. Various pressurization levels were imposed and rupture times were noted. The data enabled an evaluation of Larson-Miller parameter and the present experimental results on Zircaloy-4 were in excellent agreement with those reported in the literature. The effect of 1% Nb addition to Zircaloy-4 was studied since these materials are now known to resist long-time corrosion in water reactors and are thus considered for high burn-up applications. Negligible differences were noted in the rupture characteristics between the standard Zircaloy-4 and Nb-added Zircaloy-4. In addition, a limited amount of testing was performed to characterize the hoop creep behavior of these materials using a creep tester with hoop strain monitored in-situ by a Laser telemetric extensometer. Both the alloys followed an exponential stress variation of the creep-rate at high (>10−3E) stresses. At very low stresses, viscous creep (n = 1) was noted identifiable with Coble creep corresponding to small grain sized materials. These results have significance on the feasibility of surface storage of spent fuel where the creep deformation of the cladding could be a failure mode because of the residual heat and fission products following exposure to neutron irradiation in commercial nuclear plants. Blind extrapolation of short-term creep results to low stresses encountered during dry storage lead to nonconservative estimates of the creep-rates and creep-strains due to contributions of viscous creep to the total strain.
e-chem page: A Support System for Remote Diagnosis of Water Quality in Boiling Water Reactors
Authors: Kusumi Naohiro , Kasahara Takayasu , Akamine Kazuhiko , Tada Kenji , Usui Naoshi , Oota Nobuyuki
Published: March 2009

It is important to control and maintain water quality for nuclear power plants. Chemical engineers sample and monitor reactor water from various subsystems and analyze the chemical quality as routine operations. With regard to controlling water quality, new technologies have been developed and introduced to improve the water quality from both operation and material viewpoints. To maintain the quality, it is important to support chemical engineers in evaluating the water quality and realizing effective retrieval of stored data and documents. We have developed a remote support system using the Internet to diagnose BWR water quality, which we call e-chem page. The e-chem page integrates distributed data and information in a Web server, and makes it easy to evaluate the data on BWR water chemistry. This system is composed of four functions: data transmission, water quality evaluation, inquiry and history retrieval system, and reference to documents on BWR water chemistry. The developed system is now being evaluated in trial operations by Hitachi, Ltd. and an electric power company. In addition diagnosis technology applying independent component analysis (ICA) is being developed to improve predictive capability of the system. This paper describes the structure and function of the e-chem page and presents results of obtained with the proposed system for the prediction of chemistry conditions in reactor water.
The Economics of IRIS
Authors: Miller K. , Paramonov D.
Published: March 2009

IRIS (International Reactor Innovative and Secure) is a small to medium advanced light water cooled modular reactor being developed by an international consortium led by Westinghouse/BNFL. This reactor design is specifically aimed at utilities looking to install new (or replacement) nuclear capacity to match market demands, or at developing countries for their distributed power needs. To determine the optimal configuration for IRIS, analysis was undertaken to establish Generation Costs ($/MWh) and Internal Rate of Return (IRR %) to the Utility at alternative power ratings. This was then combined with global market projections for electricity demand out to 2030, segmented into key geographical regions. Finally this information is brought together to form insights, conclusions and recommendations regarding the optimal design. The resultant analysis reveals a single module sized at 335 MWe, with a construction period of 3 years and a 60-year plant life. Individual modules can be installed in a staggered fashion (3 equivalent to 1005 MWe) or built in pairs (2 sets of twin units’ equivalent to 1340 MWe). Uncertainty in Market Clearing Price for electricity, Annual Operating Costs and Construction Costs primarily influence lifetime Net Present Values (NPV) and hence IRR % for Utilities. Generation Costs in addition are also influenced by Fuel Costs, Plant Output, Plant Availability and Plant Capacity Factor. Therefore for a site based on 3 single modules, located in North America, Generations Costs of 28.5 $/MWh are required to achieve an IRR of 20%, a level which enables IRIS to compete with all other forms of electricity production. Plant size is critical to commercial success. Sustained (lifetime) high factors for Plant Output, Availability and Capacity Factor are required to achieve a competitive advantage. Modularity offers Utilities the option to match their investments with market conditions, adding additional capacity as and when the circumstances are right. Construction schedule needs to be controlled. There is a clear trade-off between reducing financing charges and optimising revenue streams.
Design of a Spacer Grid Using Axiomatic Design
Authors: Song K. N. , Kang B. S. , Yoon K. H. , Choi S. K. , Park G. J.
Published: March 2009

Recently, much attention has been focused on the design of the fuel assemblies in the Pressurized Light Water Reactor (PLWR). The spacer grid is one of the main structural components in a fuel assembly. It supports fuel rods, guides cooling water, and maintains geometry from the external impact loads. In this research, a new shape of the spacer grid is designed by the axiomatic approach. The Independence Axiom is utilized for the design. For the conceptual design, functional requirements (FRs) are defined and corresponding design parameters (DPs) are found to satisfy FRs in sequence. Overall configuration and shapes are determined in this process. Detailed design is carried out based on the result of the axiomatic design. For the detailed design, the system performances are evaluated by using linear and nonlinear finite element analysis. The dimensions are determined by optimization. Some commercial codes are utilized for the analysis and design.
Interfacial Area Transport of Bubbly Flow in a Small Diameter Pipe Under Microgravity Environment
Authors: Hazuku Tatsuya , Takamasa Tomoji , Hibiki Takashi , Ishii Mamoru
Published: March 2009

Axial developments of one-dimensional void fraction, bubble number density, interfacial area concentration, and Sauter mean diameter of adiabatic nitrogen-water bubbly flows in a 9-mm-diameter pipe were measured under a microgravity environment using an image-processing method. The interfacial area transport mechanism was determined based on visual observation. Marked bubble coalescence occurred when fast-moving bubbles near the channel center overtook and swept up slower-moving bubbles in the vicinity of the channel wall (velocity profile entrainment). Negligible bubble breakup was observed because of weak turbulence under tested flow conditions. Axial changes of measured interfacial area concentrations were compared with the interfacial area transport equation considering the bubble expansion and wake entrainment as observed under a normal gravity environment. The velocity profile entrainment effect under microgravity was likely to be comparable to the wake entrainment effect under normal gravity in the tested flow conditions. This apparently led to insignificant differences between measured interfacial area concentrations and those predicted by the interfacial area transport equation with the wake entrainment model under normal gravity. Possible bubble coalescence mechanisms would differ, however, between normal gravity and microgravity conditions.
Leasing of Nuclear Power Plants With Using Floating Technologies
Authors: Kuznetsov Yu. N. , Gabaraev B. A. , Reshetov V. A. , Moskin V. A.
Published: March 2009

The proposal to organize and realize the international program on leasing of Nuclear Power Plant (NPP) reactor compartments is brought to the notice of potential partners. The proposal is oriented to the construction of new NPPs or to replacement of worked-out reactor units of the NPPs in operation on the sites situated near water area and to the use of afloat technologies for construction, mounting and transportation of reactor units as a Reactor Compartment Block Module (RCBM). According to the offered project the RCBM is fabricated in factory conditions at the largest Russian defense shipbuilding plant — State Unitary Enterprise “Industrial Association SEVMASHPREDPRIYATIE” (SEVMASH) in the city of Severodvinsk of the Arkhangelsk region. After completion of assembling, testing and preliminary licensing the RCBM is given buoyancy by means of hermetic sealing and using pontoons and barges. The RCBM delivery to the NPP site situated near water area is performed by sea route. The RCBM is brought to the place of its installation with the use of appropriate hydraulic structures (canals, shipping locks), then is lowered on the basement constructed beforehand and incorporated into NPP scheme, of which the components are installed in advance. Floating means can be detached from the RCBM and used repeatedly for other RCBMs. Further procedure of NPP commissioning and its operation is carried out according to traditional method by power company in the framework of RCBM leasing with enlisting the services of firm-manufacturer’s specialists either to provide reactor plant operation and concomitant processes or to perform author’s supervision of operation. After completion of lifetime and reactor unloading the RCBM is dismantled with using the same afloat technology and taken away from NPP site to sea area entirely, together with its structures (reactor vessel, heat exchangers, pumps, pipelines and other equipment). Then RCBM is transported by shipping route to a firm-manufacturer, for subsequent reprocessing, utilization and storage. Nuclear fuel and radioactive wastes are removed from NPP site also. Use of leasing method removes legal problems connected with the transportation of radioactive materials through state borders as the RCBM remains a property of the state-producer at all stages of its life cycle.
Underwater Remote Handling Equipment for Reactor Internals Maintenance
Authors: Kimura Motohiko , Shimamura Mitsuaki , Itoh Tomoyuki , Tanaka Nobuhiko , Yuguchi Yasuhiro , Naruse Katsuhiko
Published: March 2009

More than fifty nuclear reactors generate about thirty-five percent of electricity in Japan. The need to operate these reactors safely and in a stable manner constitutes a very important issue. On the other hand, aged reactors are increasing and they are not necessarily designed and constructed using the latest technology. Stress Corrosion Cracking (SCC) on reactor internal components has become a major concern regarding aged reactors in recent years. Usually maintenance work such as inspection, repair, and preventive maintenance for core components is done by using underwater remote handling and robotic technology. It becomes very important to develop not only new efficient technology for inspection, repair, and preventive maintenance for all suspect components and but also the associated application technology for execution in a reactor. We have been developing several kind of remote handling equipment for underwater maintenance work. This paper describes some results obtained in the area of underwater remote handling that can contribute to the progress of plant reliability.
Advanced MCCI Modelling Based on Stringent Coupling of Thermal Hydraulics and Real Solution Thermochemistry in COSACO
Authors: Nie M. , Fischer M. , Lohnert G.
Published: March 2009

The development of the MCCI code COSACO specifically addresses the ex-vessel MCCI phase of the core melt retention concept of the EPR. The general philosophy behind COSACO is a rigerous representation of thermochemical phenomena related to the MCCI. In particular, the code incorporates a real solution database to predict the simultaneous formation of solid and liquid phases as well as chemical reactions for a significant number melt constituents. This offers a great flexibility in terms of application to MCCIs involving reactor materials and to tests conducted with simulant melts. The approach to model heat transfer in oxidic melt pools is based on the phase segregation hypothesis. Besides a brief description of the principal models incorporated in COSACO, this paper highlights specific thermochemical effects that arose as part of post-test calculations of the tests MACE M3b and MACE M4 with this new code version. Particular attention is drawn to the effect of melt ejections on the pool temperature as well as to the evolution of solid volumetric fraction and of melt front progression during the MCCI. Finally, the application to a representative EPR specific sequence indicates that the principal objectives of the MCCI in the reactor pit can be safely fulfilled.
Hot Isostatic Pressing of Ceramic Waste From Spent Nuclear Fuel
Authors: Bateman Kenneth J. , Rigg Richard H. , Wiest James D.
Published: March 2009

Argonne National Laboratory has developed a process to immobilize waste salt containing fission products, uranium, and transuranic elements as chlorides in a glass-bonded ceramic waste form. This salt was generated in the electrorefining operation used in electrometallurgical treatment of spent Experimental Breeder Reactor-II fuel. The ceramic waste process culminated with a hot isostatic pressing operation. This paper reviews the installation and operation of a hot isostatic press in a radioactive environment. Processing conditions for the hot isostatic press are presented for non-irradiated material and irradiated material. Sufficient testing was performed to demonstrate that a hot isostatic press could be used as the final step of the processing of ceramic waste for the electrometallurgical spent fuel treatment process.
Multiphasing CFD
Authors: Stosic Zoran V. , Stevanovic Vladimir D.
Published: March 2009

Computational fluid dynamics for multiphase flows is an emerging field. Due to the complexity and divergence of multiphase thermal and hydraulic problems, further development of multiphase flow modelling, closure laws and numerical methods is needed in order to achieve the general purpose and optimised CFD (Computational Fluid Dynamics) methods, which will be applicable to the wide variety of multiphase flow problems. In the paper, an original approach to the various aspects of multiphase CFD modelling is presented. It is based on the multi-fluid modelling approach, development of necessary closure laws and derivation of appropriate numerical methods for efficient governing equations solution. Velocity and pressure fields are solved with the SIMPLE (Semi-Implicit Method for Pressure-Linked Equations) type pressure-corrector method developed for the multiphase flow conditions. For the solution of scalar parameters transport equations both implicit and explicit methods are presented. The implicit method is suitable for steady state, slow transients and problems without the sharp fronts propagation. Explicit method is developed in order to predict scalar parameters fronts propagation, as well as phase interface tracking problems. The challenge towards the multiphase flow solution on both the macro and micro level is presented in order to perform multiphase CFD simulations and analyses of multiphase flows in complex geometry of nuclear power plant components, such as nuclear fuel rod bundles thermal-hydraulics. Presented methodology and obtained CFD results comprise micro-scale phenomena of phases’ separation, interface tracking, heated surfaces dry-out and critical heat flux occurrence, as well as macro-scale transport and distributions of phase volumes.
Vibration Analysis of a Multispan Tube in a Bundle
Authors: Khushnood Shahab , Khan Zaffar M. , Malik M. Afzaal , Koreshi Zafar Ullah , Khan Mehmood Anwer
Published: March 2009

Nuclear and process industry has seen tube bundle failures more frequently in recent years. There is some concern that the practice of structural configuration has not advanced as rapidly as thermal design of heat exchangers. Catastrophic vibration of tube in a bundle can be avoided if the tube natural frequency is kept well above the estimated existing turbulent buffeting frequencies. Flow distribution and partial admission in to the tube bundle due to upstream piping further complicates the system. In this paper, tube in a bundle has been modeled as a beam on multiple supports. Effects of damping and motion of boundaries have also been considered. Governing equations have been solved analytically for natural as well as forced vibrations. The model has been extended for a bundle of tubes by developing and implementing a computer code. Subsequent experimental verification of results on a research heat exchanger bundle has been carried. The proposed model is expected to prove a useful guide in predicting the vibration behavior of tube bundles.
Development of Sodium Two-Phase Flow Model for KALIMER Core Analysis
Authors: Chang W. P. , Hahn Dohee
Published: March 2009

An algorithm for sodium boiling is developed in order to extend the applicability of SSC-K, which is a main system analysis code for the KALIMER (Korea Advanced LIquid MEtal Reactor) conceptual design. As the capability of the current SSC-K version is limited to simulation of only a single-phase sodium flow, its applicable range should not be enough to assess the fuel integrity under some of HCDA (Hypothetical Core Disruptive Accident) initiating events where sodium boiling is anticipated. The two-phase flow model similar to that used for the light water system is known to be no more effective directly to liquid metal reactors, because the phenomena observed between two reactor coolant systems are definitely different. The developing algorithm is based on a multiple-bubble slug ejection model, which allows a finite number of bubbles in a channel at any time. The present work is a continuous effort following the former study to confirm a qualitative acceptance on the model. Since the model has been applied only to the active fuel region in the former study, a part of its qualification seems to have already been demonstrated. For its application to the whole KALIMER core channel, however, the model needs to be examined the applicability to the fuel regions other than the active fuel. The present study primarily focuses on that point. In a result, although the model may be improved in a sense through the present study over the previous modeling, a clear limitation is also confirmed with the validity of the model. The further development, therefore, is required for this model to achieve its goal by resolving such limitations.
Seismic Proving Test of Ultimate Piping Strength: Current Status of Preliminary Tests — II
Authors: Suzuki Kenichi , Namita Y. , Abe H. , Ichihashi I. , Suzuki Kohei , Ishiwata M. , Fujiwaka T. , Yokota H.
Published: March 2009

In 1998FY, the 6 year program of piping tests was initiated with the following objectives: i) to clarify the elasto-plastic response and ultimate strength of nuclear piping, ii) to ascertain the seismic safety margin of the current seismic design code for piping, and iii) to assess new allowable stress rules. In order to resolve extensive technical issues before proceeding on to the seismic proving test of a large-scale piping system, a series of preliminary tests of materials, piping components and simplified piping systems is intended. In this paper, the current status of the piping component tests and the simplified piping system tests is reported with focus on fatigue damage evaluation under large seismic loading.
Experimental Study on Reactivity Worth of Burnable Poison in Ultra-Long Life, Small LMR
Authors: Kikuchi Shigeto , Tsuji Kenji , Nishikawa Akira , Inagaki Hiromitsu , Yoshioka Kenichi , Matsumiya Hisato , Yamaoka Mitsuaki
Published: March 2009

Experimental study using an LWR-type critical assembly, NCA, has been conducted on the reactivity worth of the burnable poison (BP) that is employed in a small reactor, 4S, for achieving an extremely long core life. The experimental BP is composed of natural gadolinium oxide (Gd2O3) mixed in polyethylene. Three types of BPs are prepared: polyethylene without Gd, polyethylene with Gd/H ratio of 0.015at% and 0.943at%. Reactivity worth of each BP assembly is measured through change in critical water height. Power profile of fuel rods and flux profile are also measured. Analyses are conducted with a Monte-Carlo code, MCNP.
Computer Simulation of Diffusion of Pb-Bi Eutectic in Liquid Sodium by Molecular Dynamics Method
Authors: Qi Yingxia , Takahashi Minoru
Published: March 2009

Lead-bismuth eutectic is a potential candidate for coolant of secondary loops of sodium-cooled fast breeder reactors (FBR). The studies on the diffusion of liquid Pb-Bi in liquid Na are carried out corresponding to the case that liquid Pb-Bi leaks to liquid Na by accident. As the diffusion processes are the results of atomic motions, molecular dynamics method has been used to study the diffusion process. The self-diffusion coefficients of pure liquid Pb and Na, and liquid Pb-Bi are calculated and compared with ones by the empirical equations. The discrepancy between them could be eliminated by changing the densities of the liquids. The diffusion of lead-bismuth in sodium is simulated based on the changed densities under which the self-diffusion coefficients of individual liquid metals are close to those by the empirical equations. The simulation results show that the diffusion process of liquid Pb-Bi in liquid Na is a heat releasing process and the density of ternary liquid Na-Pb-Bi is higher than the average value of the densities of liquid Na and liquid Pb-Bi. It is also found that the diffusion coefficients of liquid Pb-Bi in liquid Na are much higher than their self-diffusion coefficients, indicating that liquid Pb-Bi are easy and quickly to diffuse in liquid Na. However, the diffusion coefficient of liquid Na is decreased due to the existence of liquid Pb-Bi, implying that liquid Na-Pb-Bi have a higher viscosity than that of pure liquid Na.
Modularization Technology in Power Plant Construction
Authors: Akagi Kenji , Murayama Kouichi , Yoshida Miki , Kawahata Junichi
Published: March 2009

Since the early 1980s, Hitachi has been developing and applying modularization technology to domestic nuclear power plant construction, and has achieved great rationalization. Modularization is one of the plant construction techniques which enables us to reduce site labor by pre-assembling components like equipment, pipes, valves and platforms in congested areas and installing them using large capacity cranes for cost reduction, better quality, safety improvement and shortening of construction time. In this paper, Hitachi’s modularization technologies are described especially from with respect to their sophisticated design capabilities. The application of 3D-CAD at the detailed layout design stage, concurrent design environment achieved by the computer network, module design quantity control and the management system are described.
Development of Reliability Function and Failure Rate Considering Information of Censored Data
Authors: Ahn Kwang-Won , Kim Yoonik , Chung Chang-Hyun , Kim Kil-Yoo
Published: March 2009

Generally, there are two methods to estimate reliability of system or component failure in engineering, or survival function of patient death in medicine: 1) parametric and 2) non-parametric methods. The parametric method has been used widely since there is few data on severe accidents in nuclear power plants. However, if we have a lot of data on system failure or patient death, the non-parametric method could be used to estimate reliability or survival function. In this paper we focus on the latter case in which one problem arises in treating the data. The problem is how to reflects efficiently the censoring information into reliability or survival function. In this study, two works have been done: 1) develop the model to reflect censoring information and 2) show how efficient the model is using the Monte-Carlo Simulation. In the WASH-1400 (NUREG-75/014), Failure to Operate of Motor Operated Valves from Nuclear Experience is suggested 1×10−3/d. The new developed method can estimate closer to the failure rate 1×10−3/d than the other existing methods. Also the failure rate using the new method satisfies the conservative limit.
State of Advancement of the International REVE Project: Computational Modelling of Irradiation-Induced Hardening in Reactor Pressure Vessel Steels and Relevant Experimental Validation Programme
Authors: Malerba Lorenzo , van Walle Eric , Domain Christophe , Jumel Stephanie , Van Duysen Jean-Claude
Published: March 2009

The REVE (REactor for Virtual Experiments) project is an international joint effort aimed at developing multiscale modelling computational toolboxes capable of simulating the behaviour of materials under irradiation at different time and length scales. Well grounded numerical techniques such as molecular dynamics (MD) and Monte Carlo (MC) algorithms, as well as rate equation (RE) and dislocation-defect interaction theory, form the basis on which the project is built. The goal is to put together a suite of integrated codes capable of deducing the changes in macroscopic properties starting from a detailed simulation of the microstructural changes produced by irradiation in materials. To achieve this objective, several European laboratories are closely collaborating, while exchanging data with American and Japanese laboratories currently pursuing similar approaches. The material chosen for the first phase of this project is reactor pressure vessel (RPV) steel, the target macrosocopic magnitude to be predicted being the yield strenght increase (Δσy) due, essentially, to irradiation-enhanced formation of intragranular solute atom precipitates or clouds, as well as irradiation induced defects in the matrix, such as point defect clusters and dislocation loops. A description of the methodological approach used in the project and its current state is given in the paper. The development of the simulation tools requires a continuous feedback from ad hoc experimental data. In the framework of the REVE project SCK·CEN has therefore performed a neutron irradiation campaign of model alloys of growing complexity (from pure Fe to binary and ternary systems and a real RPV steel) in the Belgian test reactor BR2 and is currently carrying on the subsequent materials characterisation using its hot cell facilities. The paper gives the details of this experimental programme — probably the first large-scale one devoted to the validation of numerical simulation tools — and presents and discusses the first available results, with a view to their use as feedback for the improvement of the computational modelling.
Steam Generator Replacement and Power Uprating on Tihange 2 Nuclear Plant: Safety Study Analyses
Authors: Malaval Andre´ , Marin-Lafle`che Pascale , Forgeot d’Arc Myriam , Collin Ce´line
Published: March 2009

The Tihange2 900 MWe 3-L PWR NPP, operated by the Belgian utility Electrabel, was first commissioned in 1982 with a design core power of 2775 MWth. Following an initial core power uprating by 4,5% in 1995, Electrabel has since replaced the Steam Generators which has allowed a further core power increase by roughly 5% (total 10%) in 2001. For both of each projects, licensing and implementation studies were successfully performed by Tractebel Energy Engineering and Framatome ANP. The demanding new operating conditions required a complete review of the plant design basis for which advanced methods were applied and licensed through a continuous process of discussions with the client and the Belgian Safety Authorities AVN. The licensing process required flexibility in the methods application in order to meet the specific requirements of the S.A., which was achieved within the time schedule and without jeopardising the technical objectives of the utility.
RBMK Safety Analysis in Accidents Initiated by Partial Ruptures of the Circulation Circuit
Authors: Dostov Anatoly I. , Kramerov Alexander Ja.
Published: March 2009

The paper gives an analysis of the current state of the RBMK safety evaluation in accidents initiated by partial ruptures of the delivery part of the circulating loop. It appears from this analysis that applicability and uncertainty of the international code RELAP for RBMK safety analysis could not be determined up to the present. At the same time it is shown in the paper that fuel rod cladding temperature can reach the acceptability criterion in the accidents. As a result it has been concluded that bases of the next stage of the RBMK safety analysis would be creation of a code oriented to the special features of a reactor RBMK.
Analysis of a Defected Dissimilar Metal Weld in a PWR Power Plant
Authors: Efsing P. , Lagerstro¨m J.
Published: March 2009

During the refueling outage 2000, inspections of the RC-loops of one of the Ringhals PWR-units, Ringhals 4, indicated surface breaking defects in the axial direction of the piping in a dissimilar weld between the Low alloy steel nozzle and the stainless safe end in the hot leg. In addition some indications were found that there were embedded defects in the weld material. These defects were judged as being insignificant to the structural integrity. The welds were inspected in 1993 with the result that no significant indications were found. The weld it self is a double U weld, where the thickness of the material is ideally 79,5 mm. Its is constructed by Inconel 182 weld material. At the nozzle a buttering was applied, also by Inconel 182. The In-service inspection, ISI, of the object indicated four axial defects, 9–16 mm deep. During fabrication, the areas where the defects are found were repaired at least three times, onto a maximum depth of 32 mm. To evaluate the defects, 6 boat samples from the four axial defects were cut from the perimeter and shipped to the hot-cell laboratory for further examination. This examination revealed that the two deep defects had been under sized by the ISI outside the requirement set by the inspection tolerances, while the two shallow defects were over sized, but within the tolerances of the detection system. When studying the safety case it became evident that there were several missing elements in the way this problems is handled with respect to the Swedish safety evaluation code. Among these the most notable at the beginning was the absence of reliable fracture mechanical data such as crack growth laws and fracture toughness at elevated temperature. Both these questions were handled by the project. The fracture mechanical evaluation has focused on a fit for service principal. Thus defects both in the unaffected zones and the disturbed zones, boat sample cutouts, of the weld have been analyzed. With reference to the Swedish safety evaluation system in accordance to the regulatory demands, a safety evaluation was performed using the R6-method. The failure assessment diagram is modified by the addition of the ASME XI safety factors both for limit load analysis and fracture assessment. This results in a very high conservatism since the secondary stresses such as residual stresses are high in the area. In order to quantify this effect an analysis in accordance to ASME IWB-3640, App. C was performed. This analysis provides the decision-makers with a sensitivity study; important to have to value the real risk of any missed defects in the area.
Autonomous Load Following and Operational Aspects of the Star-LM HLMC Natural Convection Reactor
Authors: Sienicki James J. , Petkov Plamen V.
Published: March 2009

STAR-LM is a 300 to 400 MWt class modular, fully transportable, proliferation resistant, and passively safe reactor system that offers developing nations and power producers nearly autonomous operation for very long term. Autonomous load following is the ability of the reactor core power to adjust itself to match the heat removal as a consequence of inherent physical phenomena. STAR-LM achieves autonomous load following through the utilization of a fast neutron spectrum core, inert lead-bismuth eutectic primary coolant, high thermal conductivity transuranic nitride fuel, and 100+% natural circulation heat transport of the primary coolant. To investigate and demonstrate autonomous operation as well as other operational aspects, representative operational transients are analyzed for the 300 MWt STAR-LM design using a coupled thermal hydraulics-neutron kinetics plant dynamics analysis computer code. Autonomous load following without reactivity effects from control rods is demonstrated for decrease-in-turbine load and increase-in-turbine load transients. For initial startup, startup from hot standby, and normal shutdown transients effected by withdrawal or insertion of shutdown rods, the reactor transitions stably to the desired steady state.
Feasibility of Natural Circulation Heat Transport in the ENHS
Authors: Sienicki James J.
Published: March 2009

An analysis has been carried out of natural circulation thermal hydraulics in both the primary and intermediate circuits of the Encapsulated Nuclear Heat Source (ENHS). It is established that natural circulation enhanced by gas injection into the primary coolant above the core, or the intermediate coolant above the heat exchange zone, is effective in transporting the nominal core power to the steam generators without the attainment of excessive system temperatures. Uncertainties in thermophysical properties and wall friction have a relatively small effect upon the calculated best estimate primary and intermediate coolant system temperature rises.
Inaccurate Use of Boron Concentration for Reactivity Feedback
Authors: Aktas Birol , Ergun Sule
Published: March 2009

A deficiency has been identified with the use of boron concentration for formulating reactivity feedback while the power of a PWR in response to boron injection in the aftermath of a scram failure was simulated. The US NRC Consolidated Code (TRAC-M) was used to simulate this transient, in which the boron injection was expected to shutdown the reactor. The results of this study, which employed the point-kinetics model of TRAC-M, reveal that the use of core-average boron concentration for formulating the reactivity feedback is inadequate for transients when there is substantial coolant void inside the core. It is recommended that the macroscopic boron density be used for more accurate predictions of the boron feedback as the use of solute concentration is only adequate when no voiding occurs in the core.
Managing Feedwater Heater Shell Thinning
Authors: Simons John W. , Keating Robert B.
Published: March 2009

Thinning of feedwater heater (FWH) shells has become an important issue throughout the U.S. nuclear industry. The thinning occurs primarily in the vicinity of the extraction steam and drain inlet nozzles where two-phase steam/water mixtures enter the heater. The thinning can be general wall thinning over a large area or highly localized thinning in a limited area. Eventually, shell repairs become necessary to ensure integrity of the pressure boundary and personnel safety, and to restore compliance with the ASME pressure vessel code. Ideally, long-term monitoring of the thickness of FWH shells would allow for timely repairs before the wall thickness decreases below the ASME Code minimum wall thickness. However, since most plants did not start inspecting FWH shells until recently, the initial inspection may find shells that are thinned to near or below the ASME Code minimum wall thickness. Because the cost and manpower requirements for shell repairs can be significant, it is imperative that shell repairs provide a long-term solution and that they be implemented as a planned outage evolution. Accordingly, the approach to managing FWH shell thinning must provide sufficient advance warning of the need for repairs while ensuring safe operation until the permanent repair can be implemented. The approach outlined in this paper does both. • Establish acceptance criteria to ensure safe, interim operation with thinned shells until a permanent repair can be implemented. • Develop contingency plans for interim repairs should the inspections identify thinning below the interim acceptance criteria. • Inspect FWH shells during a refueling outage or possibly during planned system outages or power reductions. If necessary, perform interim repairs to ensure safe operation until the next refueling outage. • Develop and implement permanent repairs that will prevent shell thinning.
Passive Accident Protection Devices on the Basis of Technology Lyophobous of Capillary-Porous Systems
Authors: Sorokin A. P. , Portianoy A. G. , Serdun E. N. , Maltsev V. G.
Published: March 2009

At present the special attention is given to increase safety of technical systems, and first of all to prevent emergencies connected with heavy ecological and economic consequences. One of perspective ways of prevention of heavy accident is the use of passive accident protection devices. The most distinctly given tendency is displayed in atomic engineering. At this moment a few hundreds technical decisions on accident protection devices of various functional purpose be known. Nevertheless devices to the full satisfying to all showed requirements are not created. New types of passive accident protection devices on a level of temperature and pressure on the basis of technology by capillary-porous lyophobous system (LCPS) are considered. The Lyophobous Passive Accident Protection Device on a level of Temperature (LPAP-T) functions on the basis of high dylatometrical effect of LCPS and represents a design as sylphon, containing capillary-porous matrix and lyophobous liquid, melting at temperature of operation. At a melting the lyophobous liquid leaves from pores of a matrix under Laplas pressure and actuates the working mechanism. Carried out calculate-experimental research have shown, that considered (LPAP-T) have the following characteristics: • range of threshold operation 70°-700° C; • generation of significant efforts (up to 104N); • low dependence a testimonial from of the operational factors. Developed (LPAP-T) has a number of advantages before known, for example by the magnetic thermal passive accident protection devices (generation of efforts), as compared with device presented in Patent USA No. 5051229 (sylphon filled with liquid metal) — threshold character, multichannels of operation.
Polonium Issue in Fast Reactor Lead Coolants and One of the Ways of Its Solution
Authors: Khorasanov G. L. , Ivanov A. P. , Blokhin A. I.
Published: March 2009

One of the main issues in using materials for nuclear facilities is to minimize the production of the most hazardous radionuclides. In the ideal case, all nuclear reactor materials, except a fuel, should be low-activation. The term “low-activation material” means that this one loses its induced activity in a short time after removal from irradiation. Proposals for building a fusion reactor using low-activation materials are given in Ref.1, 2. For this purpose, low-activation structural materials based on V-Ti-Cr alloys are in the stage of R&D in several countries [3,4]. Another technique to avoid the hazardous activity is in using isotopically enriched materials [5–7]. Although isotopical tailoring option requires tremendous technical efforts and it is too expensive, its application can be first of all assumed for those structural and functional materials which generate very hazardous radionuclides under irradiation. In modern projects of next generation NPPs the preference is given to fast reactors (FRs) with a lead coolant [8]. As it known, the coolant circulating through a FR core is activated, and in the future we should have problems with handling a completed coolant after FR decommissioning or at realization of repair or emergency activities. There, it is desirable to have a low-activation coolant with the low contents of hazardous radionuclides. In papers [9,10] presented at the previous ICONE conferences it was proposed to use lead isotope, Pb-206, as a coolant instead of lead natural, Pb-nat. This paper is devoted to more detailed calculations of accumulating stable bismuth, Bi-209, and polonium radioisotopes, Po-209 (T1/2 = 102 y) and Po-210 (T1/2 = 138 d), in 1 kg of Pb-nat or Pb-206 placed in the core of the BOR-60 type FR.
Numerical Study on the Local Temperature in IRWST Pool
Authors: Kang H. S. , Bae Y. Y. , Park J. K.
Published: March 2009

The Safety Depressurization Vent System (SDVS) of APR 1400 prevents Reactor Coolant System (RCS) from over-pressurization by discharging high pressure and temperature coolant through the I-sparger into the IRWST during an accident. If IRWST water temperature rises locally, around the sparger, beyond 200 °F by the discharged coolant, unstable steam condensation may occur and cause large pressure load on the IRWST wall. To investigate whether this condition could be avoided along with the most severe accident, Total Loss of Feed Water (TLOFW) event and to find the local temperature location for the I-typed sparger in APR 1400, the flow and temperature distribution of water in the IRWST pool is calculated by using CFX 4.4 computational fluid dynamic code. According to the results, the IRWST water temperature does not exceeds temperature limit upto 15 seconds after the opening of POSRVs. The location of the local temperature is estimated as the location above and below of the sparger discharge head by examining the flow and temperature field.