bluuebunny/crossref_metadata_2025 huggingface dataset:
Uncertainty Analysis of RBMK-Related Experimental Data
Authors: Urbonas Rolandas , Kaliatka Algirdas , Liaukonis Mindaugas
Published: March 2009

An attempt to validate state-of-the-art thermal hydraulic code ATHLET (GRS, Germany) on the basis of E-108 test facility was made. Originally this code was developed and validated for different type reactors than RBMK. Since state-of-art thermal hydraulic codes are widely used for simulation of RBMK reactors, further codes’ implementation and validation is required. The phenomena associated with channel type flow instabilities and CHF were found to be an important step in the frame of the overall effort of state-of-the-art validation and application for RBMK reactors. In the paper one-channel approach analysis is presented. Thus, the oscillatory behaviour of the system was not detected. The results show dependence on the nodalisation used in the heated channels, initial and boundary conditions and code selected models. It is shown that the code is able to predict a sudden heat structure temperature excursion, when critical heat flux is approached. GRS developed uncertainty and sensitivity methodology was employed in the analysis.
Development of Advanced Concept for Shortening Construction Period of ABWR Plant
Authors: Ijichi Hiroshi , Yamashita Toshio , Tsutagawa Masahiro , Mori Hiroya , Ooshima Nobuaki , Miura Jun , Kanechika Minoru , Miura Nobuaki
Published: March 2009

Construction of a nuclear power plant (NPP) requires a very long period because of large amount of construction materials and many issues for negotiation among multiple sections. Shortening the construction period advances the date of return on an investment, and can also result in reduced construction cost. Therefore, the study of this subject has a very high priority for utilities. We achieved a construction period of 37 months from the first concrete work to fuel loading (F/L) (51.5 months from the inspection of the foundation (I/F) to the start of commercial operation (C/O)) at the Kashiwazaki-Kariwa NPPs No. 6 and 7 (KK-6/7), which are the first ABWR plants in the world. At TEPCO’s next plant, we think that a construction period of less than 36 months (45 months from I/F to C/O) can be realized based on conventional methods such as early start of equipments installation and blocking of equipment to be brought in advance. Furthermore, we are studying the feasibility of a 21.5-month construction period (30 months from I/F to C/O) with advanced ideas and methods. The important concepts for a 21.5-month construction period are adoption of a new building structure that is the steel plate reinforced concrete (SC) structure and promotion of extensive modularization of equipment and building structure. With introducing these new concepts, we are planning the master schedule (M/S) and finding solutions to conflicts in the schedule of area release from building construction work to equipment installation work (schedule-conflicts). In this report, we present the shortest construction period and an effective method to put it into practice for the conventional general arrangement (GA) of ABWR. In the future, we will continue the study on the improvement of building configuration and arrangements, and make clear of the concept for large composite modules of building structures and equipment.
Effects of Turbulence and Containing Wall on Rise Velocities of Single Bubbles in a Vertical Circular Pipe
Authors: Okawa Tomio , Torimoto Kazuhiro , Nishiura Masanori , Kataoka Isao , Mori Michitsugu
Published: March 2009

Experiments were conducted to clarify the single bubble rise characteristics in turbulent flows in vertical flow channels. It was revealed that the rise velocity of a bubble relative to the time-averaged local liquid velocity could be much smaller in turbulent upflows than in stagnant liquids. The reduction of relative velocity was estimated to be caused by the two factors: turbulence in continuous phase and steep velocity gradient near wall; new correlations for describing these two effects were proposed. The relative velocity between the phases significantly affects the lateral phase distribution in multidimensional simulation of bubbly two-phase flow and the present correlations can give reasonable predictions for the relative velocity in turbulent flow. It is hence expected that the new correlations can contribute to the further improvement of the simulation models of bubbly two-phase flows.
Development of the Concrete Cask Horizontal Transfer System
Authors: Sakai Mikio , Fujiwara Hiroaki , Sakaya Tadatsugu , Sakai Akira
Published: March 2009

Concrete cask system is focused as the candidate one for spent fuel dry storage facilities from economic potential in Japan. Concrete cask consists of a concrete storage cask and a steel canister. A canister containing nuclear spent fuel is shipped by a transportation cask from a nuclear power plant to an interim storage facility. The canister is transferred from the transportation cask to a storage cask by a transfer cask in the storage facility. IHI has developed a concrete cask horizontal transfer system. This transfer system indicates that a canister is transferred to a storage cask horizontally. This transfer system has a merit against canister drop accident in transfer operation, i.e. spent fuel assemblies can be kept safe during the transfer operation. There are guide rails inside of the concrete cask, and the canister is installed into the storage cask with sliding on the rails. To develop the horizontal transfer system, IHI carried out a heat load test and numerical analyses by CFD. Heat load experiment was carried out by using a full-scale prototype canister, storage cask and transfer vessel. The decay heat was simulated by an electric heater installed in the canister. Assuming high burn-up spent fuel storage, heat generation was set between 20.0kW and 25.0kW. This experiment was focused on the concrete temperature distribution. We confirmed that the maximum concrete temperature in transfer operation period was lower than 40 °C (Heat generation 22.5kW). Moreover we confirmed the maximum concrete temperature passed 24 hours with horizontal orientation was below 90 °C (Heat generation 22.5kW). We analyzed the thermal performance of the concrete cask with horizontal transfer condition and normal storage condition. Thermal analyses for horizontal transfer operation were carried out based on the experimental conditions. The tendency of the analytical results was in good agreement with experimental results. The purpose of vertical thermal analysis was to estimate the concrete temperature increase in the case a canister contacts with guide rails in normal storage. It has a possibility that a canister contacts with guide rails during storage period after concrete cask is upended from transfer operation. The temperature increase due to this contact was calculated 5 °C at small local area. This result implies that the affect of the contact is very small. This paper addresses that the storage cask concrete is kept its integrity in transfer operation period and normal storage period.
Crosscutting Requirements in the International Project on Innovative Reactors and Fuel Cycles (INPRO)
Authors: Steur Ronald , Yaven Yanko Lyubenov , Gueorguiev Boris , Mahadeva Rao , Shen Wenquan
Published: March 2009

There are two categories of requirements: (i) user requirements that need to be met by the designers and manufacturers of innovative reactors and fuel cycles, and (ii) a wide spectrum of requirements that need to be met by countries, willing to successfully deploy innovative nuclear reactors for energy production. This part of the International Project on Innovative Reactors and Fuel Cycles will mainly deal with the second category of requirements. Both categories of requirements will vary depending on the institutional development, infrastructure availability and social attitude in any given country. Out of the need for sustainable development requirements will also more specific in the future. Over a 50-year time frame both categories of requirements will evolve with social and economic development as nuclear technology develops further. For example, the deployment of innovative reactors in countries with marginal or non-existing nuclear infrastructures would be possible only if the reactors are built, owned and operated by an international nuclear utility or if they are inherently safe and can be delivered as a “black box - nuclear battery”. A number of issues will need to be addressed and conditions and requirements developed if this is going to become a reality. One general requirement for wider utilization of innovative nuclear power will be the public and environmental considerations, which will play a role in the decision making processes. Five main clusters of topics will be handled: • Infrastuctural aspects, typology and consequences for nuclear development. • Industrial requirements for the different innovative concepts. • Institutional developments and requirements for future deployment of nuclear energy. (National as well as international). • Socio-political aspects, a.o. public acceptance and role of governments. • Sustainability: requirements following the need for sustainability. Analysis will be made of the evolution of national and international social, institutional and infrastructure requirements for the deployment of innovative nuclear technology through 2050 and beyond and requirements will be identified following the need for.
The Methodology for Innovative Nuclear Technology Evaluation
Authors: Kagramanian V. , Garmash A.
Published: March 2009

Other papers in this INPRO session address the possible user requirements for nuclear reactors and fuel cycles that might be applicable in the middle of this century. Each paper has dealt with a certain category of user requirements — safety, proliferation resistance, environmental impacts and material flows, and the broader infrastructure and institutional issues. These user requirements will need to be elaborated in terms of specific criteria applicable to the options and alternatives to be evaluated in different countries. A systematic approach to applying these criteria will also be required. Current technology assessment methods, however, are unlikely to be perfectly suited to the longer-term objective of establishing the directions, which will encourage the innovation we are seeking for the middle of this century. This paper will therefore identify methodological adjustments and extensions appropriate for this longer-term objective and present a possible “top down” approach for creating the necessary methodology.
Evaluation of Dynamic Behavior of Pile Foundations for Interim Storage Facilities Through Geotechnical Centrifuge Tests
Authors: Tsurumaki Shizuo , Watanabe Hiroyuki , Tateishi Akira , Horikoshi Kenichi , Suzuki Shunichi
Published: March 2009

In Japan, there is a possibility that interim storage facilities for recycled nuclear fuel resources may be constructed on quaternary layers, rather than on hard rock. In such a case, the storage facilities need to be supported by pile foundations or spread foundations to meet the required safety level. The authors have conducted a series of experimental studies on the dynamic behavior of storage facilities supported by pile foundations. A centrifuge modeling technique was used to satisfy the required similitude between the reduced size model and the prototype. The centrifuge allows a high confining stress level equivalent to prototype deep soils to be generated (which is considered necessary for examining complex pile-soil interactions) as the soil strength and the deformation are highly dependent on the confining stress. The soil conditions were set at as experimental variables, and the results are compared. Since 2000, the Nuclear Power Engineering Corporation (NUPEC) has been conducting these research tests under the auspices on the Ministry of Economy, Trade and Industry of Japan.
Application of Seismic Isolation to the STAR-LM Reactor
Authors: Yoo B. , Kulak R. F.
Published: March 2009

This paper presents findings from our initial work in developing a seismic isolation system for the STAR-LM reactor design. Research and development was carried out to determine the characteristics of the isolator device. The heavy weight and small footprint presented a challenge in bearing design and bearing placement. Results are also presented from a study on the use of three-dimensional seismic isolation devices to the full-scale reactor. Both two-dimensional (i.e., one device for horizontal isolation only) and integral (i.e., one device for horizontal and vertical) concepts were explored. The seismic analysis responses of the two-dimensional and the three-dimensional isolation systems for the STAR-LM are compared with that of the conventional fixed base system. Finally, results are presented from a study on the effects of the levels of vertical and horizontal damping on the seismic response of STAR-LM.
Numerical Assessment of Hydraulic Performances of Main Coolant Pump for Integral Reactor Smart
Authors: Kim Min-Hwan , Kim Jong-In , Park Jin-Seok
Published: March 2009

The performance prediction of SMART MCP was performed using a computational fluid dynamics code. A general capacity-head performance curve of MCP was obtained and it showed the typical type axial pump performance curve. When four MCPs operate in parallel and one of them stops while the others continue to operate, SMART is designed to operate with reduced power. A procedure for predicting the performance of this SMART operation mode was developed and verified with available experimental data. An analysis based on the developed procedure was performed for two cases; the impeller of stopped MCP is assumed to be fixed or free to rotate in the reverse direction. According to the results, 73% flow rate of normal operation enters the reactor core in the case of a fixed impeller. In the case of a free impeller, the flow rate entering the reactor core is 63.3%.
Improvement of the Damping Constants for Seismic Design of Piping System for NPP
Authors: Kobayashi Kei , Satoh Takashi , Kojima Nobuyuki , Hattori Kiyoshi , Nakagawa Masaki , Otani Akihito
Published: March 2009

The present design damping constants for nuclear power plant (NPP)’s piping system in Japan were developed through discussion among expert researchers, electric utilities and power plant manufactures. They are standardized in “Technical guidelines for seismic design of Nuclear Power Plants” (JEAG 4601-1991 Supplemental Edition). But some of the damping constants are too conservative because of a lack of experimental data. To improve this excessive conservatism, piping systems supported by U-bolts were chosen and U-bolt support element test and piping model excitation test were performed to obtain proper damping constants. The damping mechanism consists of damping due to piping materials, damping due to fluid interaction, damping due to plastic deformation of piping and supports, and damping due to friction and collision between piping and supports. Because the damping due to friction and collision was considered to be dominant, we focused our effort on formulating these phenomena by a physical model. The validity of damping estimation method was confirmed by comparing data that was obtained from the elemental tests and the actual scale piping model test. New design damping constants were decided from the damping estimations for piping systems in an actual plant. From now on, we will use the new design damping constants for U-bolt support piping systems, which were proposed from this study, as a standard in the Japanese piping seismic design.
Investigations on Catalyst Systems for Catalytic Recombiners
Authors: Reinecke Ernst-Arndt , Broeckerhoff Peter , Tragsdorf Inga M.
Published: March 2009

Passive Autocatalytic Recombiners (PARs) are used for hydrogen removal in the containments of Light Water Reactors after a severe accident. These devices make use of the fact that hydrogen and oxygen react exothermally on catalytic surfaces already at low temperatures generating steam and heat. One major concern is the fact that existing recombiners bear the risk of ignition of the gaseous mixture by overheated catalytic substrates or parts of the casing, since the heat generated is not removed by cooling systems. Overheating may occur due to insufficient heat removal. Experimental investigations on existing systems show that the highest temperatures appear near the leading edges of the catalyst sheets. Furthermore, local conversion rates are too high not allowing sufficient reaction heat removal by convection. Possible countermeasures are additional cooling or limiting local conversion rates. At FZJ investigations are made on adapting the catalyst activity according to the requirements by using electro-plating technology instead of washcoat coatings, allowing well defined coating densities. Substrates with corresponding coatings have been tested, proving their ability in mixtures up to oxygen limitation. Different substrate materials and pre-treating measures are investigated for optimizing the surface properties. SEM-studies give insight in the surface structure and allow detailed analysis of the catalyst activity.
Analysis of Piping Systems for Life Extension of Heavy Water Plants in India
Authors: Mishra Rajesh K. , Soni R. S. , Kushwaha H. S. , Raj V. Venkat
Published: March 2009

Heavy water production in India has achieved many milestones in the past. Two of the successfully running heavy water plants are on the verge of completion of their design life in the near future. One of these two plants, situated at Kota, is a hydrogen sulfide based plant and the other one at Tuticorin is an ammonia-based plant. Various exercises have been planned with an aim to assess the fatigue usage for the various components of these plants in order to extend their life. Considering the process parameters and the past history of the plant performance, critical piping systems and equipment are identified. Analyses have been carried out for these critical piping systems for mainly two kinds of loading, viz. sustained loads and the expansion loads. Static analysis has been carried out to find the induced stress levels due to sustained as well as thermal expansion loading as per the design code ANSI B31.3. Due consideration has been given to the design corrosion allowance while evaluating the stresses due to sustained loads. At the locations where the induced stresses (SL) due to the sustained loads are exceeding the allowable limits (Sh), exercises have been carried out considering the reduced corrosion allowance value. This strategy is adopted in view of the fact that the thickness measurements carried out at site at various critical locations show a very low rate of corrosion. It has been possible to qualify the system with reduced corrosion allowance values however, it is recommended to keep that location under periodic monitoring. The strategy adopted for carrying out analysis for thermal expansion loading is to qualify the system as per the code allowable value (Sa). If the stresses are more than the allowable value, credit of liberal allowable value as suggested in the code i.e., with the addition of the term (Sh-SL) to the term 0.25 Sh, has been taken. However, if at any location, it is found that thermal stress is high, fatigue analysis has been carried out. This is done using the provisions of ASME Code Section VIII, Div. 2 by evaluating the cumulative fatigue usage factor. Results of these exercises reveal that the piping systems of both of these plants are in a very healthy state. Based on these exercises, it has been concluded that the life of the plants can be safely extended further with enhanced in-service inspection provisions.
Large Break LOCA Safety Injection Sensitivity for a CE/ABB System 80+ PWR
Authors: Pottorf J. , Bajorek S. M.
Published: March 2009

A WCOBRA/TRAC model of an evolutionary pressurized water reactor with direct vessel injection was constructed using publicly available information and a hypothetical double-ended guillotine break of a cold leg pipe was simulated. The model is an approximation of a ABB/Combustion Engineering System 80+ pressurized water reactor (PWR). WCOBRA/TRAC is the thermal-hydraulics code approved by the U.S. Nuclear Regulatory Commission for use in realistic large break LOCA analyses of Westinghouse 3- and 4-loop PWRs and the AP600 passive design. The AP600 design uses direct vessel injection, and the applicability of WCOBRA/TRAC to such designs is supported by comparisons to appropriate test data. This study extends the application of WCOBRA/TRAC to the investigation of the predicted behavior of direct vessel injection in an evolutionary design. A series of large break LOCA simulations were performed assuming a core power of 3914 MWt, and Technical Specification limits of 2.5 on total peaking factor and 1.7 on hot channel enthalpy rise factor. Two cladding temperature peaks were predicted during reflood, one following bottom of core recovery and a second caused by saturated boiling of water in the downcomer. This behavior is consistent with prior WCOBRA/TRAC calculations for some Westinghouse PWRs. The simulation results are described, and the sensitivity to failure assumptions for the safety injection system is presented.
A Main Steam Safety Valve (MSSV) With “Fixed Blowdown” According to ASME Section III, Part NC-7512
Authors: Fo¨llmer Bernhard , Schnettler Armin
Published: March 2009

In 1986, the NRC issued the Information Notice (IN) 86-05 “Main Steam Safety Valve test failures and ring setting adjustments”. Shortly after this IN was issued, the Code was revised to require that a full flow test has to be performed on each CL.2 MSSV by the manufacturer to verify that the valve was adjusted so that it would reach full lift and thus full relieving capacity and would reclose at a pressure as specified in the valve Design Specification. In response to the concern discussed in the IN, the Westinghouse Owners Group (WOG) performed extensive full flow testing on PWR MSSVs and found that each valve required a unique setting of a combination of two rings in order to achieve full lift at accumulation of 3% and reclosing at a blowdown of 5%. The Bopp & Reuther MSSV type SiZ 2507 has a “fixed blowdown” i.e. without any adjusting rings to adjust the “blowdown” so that the blowdown is “fixed”. More than 1000 pieces of this type are successfully in nuclear power plants in operation. Many of them since about 25 years. Therefore it can be considered as a proven design. It is new that an optimization of this MSSV type SiZ 2507 fulfill the requirements of part NC-7512 of the ASME Section III although there are still no adjusting rings in the flow part. In 2000, for the Qinshan Candu unit 1&2 full flow tests were performed with 32 MSSV type SiZ 2507 size 8” × 12” at 51 bar saturated steam in only 6 days. In all tests the functional performance was very stable. It was demonstrated by recording the signals lift and system pressure that all valves had acceptable results to achieve full lift at accumulation of 3% and to reclose at blowdown of 5% . This is an advantage which gives a reduction in cost for flow tests and which gives more reliability after maintenance work during outage compared to the common MSSV design with an individual required setting of the combination of the two rings. The design of the type SiZ 2507 without any adjusting rings in the flow path is presented. The stable performance depends on the interaction of flow force and spring force. The optimization of the flow path to create a suitable flow-force-curve was managed by Computational Fluid Dynamics (CFD) and flow-force-characteristic-measurements at a model 1: 2.5 ! The method of the flow-force-characteristic-measurement permits systematic dimensioning of valve spring forces by means of measurement of the fluid mechanical forces occurring on the valve spindle during flow [1], [2]. A special procedure was established to verify a spring force versus lift curve with an accuracy of 1% for each production valve. This gives high reliability at required stable performance and this can not be influenced by wrong setting of any adjusting ring during maintenance work.
Mechanical Non-Equilibrium Effect on Choking Flow at Low Pressure in Air-Water Experiment
Authors: Yoon H. J. , Ishii M. , Revankar S. T. , Wang W.
Published: March 2009

At low pressure and low flow conditions, the prediction of two-phase flow transients is much more difficult than at relatively higher pressure or at high flow due to the large density ratio and thermal and mechanical non-equilibrium between the phases. A mechanical non-equilibrium effect was studied with air-water two-phase critical flow in pipes at low pressure (< 1 MPa). Critical flow test were conducted in a well-scaled test facility with several on-line instruments. The slip ratio, which is the key factor in the mechanical non-equilibrium, is directly measured at the upstream of the critical section. The break geometry effect was investigated using the nozzle and orifice as critical flow sections. The experimental results showed that the slip ratio increased as the quality increased. The slip ratio value at low pressure was relatively higher than the slip ratio at the high pressure for the same flow quality. The measured critical mass flux for the nozzle was higher than the orifice at the low flow quality. Thus there is a geometry effect on critical mass flux at the low quality region, even though there is no difference in the slip ratio at the upstream of choking plane. Thus, it is concluded that there is a strong mechanical non-equilibrium at the choking plane.
The Destructive High Pressure Tests of Cylindrical Shells
Authors: Kitajima Yasumi , Shibata Satoru
Published: March 2009

We conducted the limit state tests of cylindrical shells to establish criteria for the occurrence of steel wall/liner tearing in the reactor containment vessels (such as Steel Containment Vessels (SCV), Prestressed Concrete Containment Vessels (PCCV) and Reinforced Concrete Containment Vessels (RCCV)) under the limit state pressure. In the tests, precisely manufactured cylindrical shell vessels (about 800 mm in height and 300 mm in diameter) were pressurized to the failure using water. We also conducted the finite element analyses. The conclusions are as follows: 1. We obtained good agreement (within 2–3%) between the tests and the analyses in structural behavior such as internal pressure loading vs. displacement and strain to the failure. However, in the case of the test piece which included weld line on the cylindrical wall, the difference between the tests and the analyses was larger (about 1.5 times) than the rest. 2. The localized strains began to increase when radial strains in general structure reached 5–10%. We are intended to apply these results to the finite element analyses and the integrity evaluation of containment vessels (SCV, PCCV and RCCV).
Flow Instabilities and Main Steam Line Vibrations in a Pressurized Water Reactor
Authors: Henriksson Mats , Westin Johan , Granha¨ll Tord , Andersson Lars , Bjerke Lars-Erik
Published: March 2009

Severe vibrational problems occurred in the main steam system of a PWR nuclear power plant, about 18 months after a steam generator replacement had been carried out. The magnitude of the vibrations reached levels at which the operators had to reduce power in order to stay within the operating limits imposed by the nuclear inspectorate. To solve the problem the following analyses methods were employed: • Testing the influence on vibration level from different modes of plant operation; • Analyses of plant measurement data; • Calculations of: – hydraulic behaviour of the system, – structural dynamic behaviour of the system, – flow at the steam generator outlet; • Scale model testing of the steam generator outlet region. Hydraulic flow disturbances in the main steam system were measured using pressure and strain gauges, which made it possible to track individual pressure pulses propagating through the main steam system. Analyses showed that the pressure pulses causing the vibration originated from the vicinity of the steam generator outlet. By using computer codes for network fluid flow analyses the pressure pulses found in the measurement traces could be generated in calculations. Careful studies of the flow at the steam generator outlet region, using model testing in a 1:3 scale model as well as transient 3D CFD calculations, gave clear indications that flow separation occurred at the steam generator outlet nozzle and at the first bend. Finally, by substituting the outlet nozzle for a different design with a multiport nozzle, the steam line vibration problem has been solved.
Conceptual Analysis of the Power Production of Fission Electric Cell Reactors
Authors: King Donald , Rochau Gary , Morrow Charles , Cash Jamie , Seidel David , Slutz Stephen
Published: March 2009

The United States Department of Energy, Nuclear Energy Research Initiative (NERI) Direct Energy Conversion project has as its goal the development of direct energy conversion (DEC) processes suitable for commercial development. DEC is defined as any fission process that returns usable energy with no intermediate thermal process. This project includes the study of the fission electric cell (FEC). In the FEC, fission fragments exit the fuel element cathode and are collected by the cell anode. Previous work [1] has shown the potential of FECs, with theoretical efficiencies up to 60%. Inspection of this work indicates the need for additional system modeling prior to any conclusions regarding the final FEC reactor configuration. This paper builds on the previous work and outlines the development of models to facilitate design decisions. The models address criticality, design life, reactor configuration, and current-voltage characteristics. In addition, this paper proposes future work to complete the design model.
Conceptual Analysis of the Thermo-Hydraulic Characteristics of Fission Electric Cell Reactors
Authors: Morrow C. , Rochau G. , Cash J. , King D.
Published: March 2009

The United States Department of Energy, Nuclear Energy Research Initiative (NERI) Direct Energy Conversion (DEC) project has as its goal the development of a direct energy conversion process suitable for commercial development. Direct energy conversion is defined as any fission process that returns usable energy without using an intermediate thermal process. This project includes the study of the fission electric cell (FEC). Inherent to the FEC is a cathode insulated from the rest of the cell by a vacuum. This arrangement has the potential to trap energy on the cathode, increasing temperatures and potentially causing structural problems and gas emissions. This paper describes the project efforts to address this and similar thermal issues. This paper describes the development of a Thermal Model that integrates thermodynamic and heat transfer considerations into an overall Design Model. The Thermal Model begins with the basic energy balance. It then applies heat transfer methods to these models to develop a general relationship between temperature and system operating parameters. Finally, the model uses published correlations to relate the general parameters to specific geometric configurations. The Thermal Model demonstrates that the low energy densities typical of fission electric cells result in assemblies whose available heat transfer mechanisms efficiently transport any waste heat the systems generate. This results in a design at this conceptual level with significant opportunities to optimize system operations and economics using operating temperature. The model also demonstrates that for the same reasons, heat buildup has a weak correlation with system operating voltage.
The Global Nuclear Futures Model: A Dynamic Simulation Tool for Energy Strategies
Authors: Bixler N. E.
Published: March 2009

The Global Nuclear Futures Model (GNFM) is a dynamic simulation tool that provides an integrated framework to model key aspects of nuclear energy, nuclear materials storage and disposition, global nuclear materials management, and nuclear proliferation risk. It links nuclear energy and other energy shares dynamically to greenhouse gas emissions and twelve other measures of environmental impact. It presents historical data from 1990 to 2000 and extrapolates energy demand through the year 2050. More specifically, it contains separate modules for energy, the nuclear fuel cycle front end, the nuclear fuel cycle back end, defense nuclear materials, environmental impacts, and measures of the potential for nuclear proliferation. It is globally integrated but also breaks out five regions of the world so that environmental impacts and nuclear proliferation concerns can be evaluated on a regional basis. The five regions are the United States of America (USA), The Peoples Republic of China (China), the former Soviet Union (FSU), the OECD nations excluding the USA, and the rest of the world (ROW).
Experimental Study on Radiation Induced Boiling Enhancement for Stainless Steel Plate
Authors: Okamoto Koji , Akiyama Hiroshi , Madarame Haruki , Takamasa Tomoji
Published: March 2009

The Radiation Induced Boiling Enhancement phenomena (RIBE) were confirmed using the SUS304 foil. The SUS304 with plasma oxidized surface shows higher CHF, i.e., about 20% improvement. While, the natural and mixed gas oxidized surface does not show the boiling enhancement. The RIBE has been highly related to the surface conditions.
Evaluation of Flow-Induced Vibration for Fixed Type Guide Rods of Shroud Head and Steam Dryer in ABWR
Authors: Takahashi Shirou , Tamako Hiroaki , Kawamura Tsutomu , Shiina Kouji , Tsubaki Masaaki , Sakashita Akihiro , Yamashita Norimichi , Narabayashi Tadashi , Hagiwara Tsuyoshi , Komita Hideo
Published: March 2009

For the purpose of shortening outage schedules, fixed type guide rods are expected to be used in the ABWR. Guide rods are the component for a shroud head and a steam dryer installation. However, guide rods are located near the main steam nozzle, therefore flow-induced vibration (FIV) is concern due to the high steam velocity. In the present study, tests of the 1/1.87-scale model and computational fluid dynamics (CFD) analysis for the 1/1.87-scale test and the actual ABWR model were conducted to prove the structural integrity of fixed type guide rods against FIV. As a result of CFD calculation, reduced damping was more than 5 and reduced velocity was approximately 1.44, so resonance did not occur fixed type guide rods. The maximum fluctuating stresses were conservatively evaluated as 8MPa by the turbulence and 2MPa by the Karman vortex shedding. Both values were below allowable limit. As noted above, the structural integrity against FIV was confirmed, so it is feasible to use the fixed type guide rods in the ABWR.
Thermal Interaction Between Molten Metal and Sodium: Examination of the Fragmentation Mechanism of Molten Jet
Authors: Nishimura Satoshi , Kinshita Izumi , Sugiyama Ken-Ichiro , Okada Ryohei , Ueda Nobuyuki
Published: March 2009

In order to clarify the mechanism of thermal fragmentation of a molten jet dropped into a sodium pool at instantaneous contact interface temperatures below its freezing point, a basic experiment was carried out using molten copper and sodium. Copper was melted in a crucible with an electrical heater and was dropped through a short nozzle into a sodium pool, in the form of a jet column. Thermal fragmentation originating inside the molten copper jet with a solid crust was clearly observed in all runs. It is verified that a small quantity of sodium, which is locally entrained inside the molten jet due to the organized motion between the molten jet and sodium, is vaporized by the sensible heat and the latent heat of molten copper, and the high internal pressure causes the molten jet with a solid crust to fragment. It is also concluded that the thermal fragmentation is more dominant than the hydrodynamic fragmentation, in the present range of Weber number and superheating of molten jet. Furthermore, it can be explained that the thermal fragmentation caused by the molten copper jet - sodium interaction is severer than that caused by the molten uranium alloy jet - sodium interaction, which was reported by Gabor et al., because the latent heat and the thermal diffusivity of molten copper, which are the physical properties that dominate the degree of fragmentation, are much higher than those of molten uranium alloy jets.
Pre-Operational Seismic Walk-Through of NPPs in India
Authors: Soni R. S. , Mishra R. K. , Agrawal M. K. , Reddy G. R. , Kushwaha H. S. , Raj V. Venkat , Badrinarayan G. , Hawaldar R. V. , Ingole S. M.
Published: March 2009

In nuclear power plants, it is essential to design the various safety and safety related systems and components of the plant in such a manner that they maintain their structural integrity as well as serve their functional performance during a seismic event. The pre-operational seismic walk-through helps in ensuring the installation of various seismic supports as per design intent, identifying the areas where supports are inadequate, identifying the interaction concerns between the systems of various safety classes and locating the various undesired loose, untied / unanchored components, tools, etc. used during the construction activity. A detailed procedure for the pre-operational seismic walk-through of the NPPs was therefore, prepared. Since the types and locations of seismic supports for the various systems and components of the plant had been already reviewed, the major emphasis during the walk-through was laid on their proper installation.
Development of Russian Regulatory Basis for License Renewal/Life Extension of Nuclear Power Plants
Authors: Pachner Jaroslav , Kaliberda Inna V.
Published: March 2009

By 2010, fifteen Russian nuclear power plant (NPP) units will reach the end of their scheduled service life of 30 years. An extension of NPP operation beyond the scheduled service life, which is provided for by the Russian safety standard OPB-88/97, requires a renewal of the plant operating license by the regulatory organization Gosatomnadzor (GAN). This paper presents an overview of an IAEA project aimed at assisting GAN in the development of a set of regulatory guidelines for NPP license renewal/life extension. The overview includes a description of the Russian regulatory strategy for life extension of NPPs and IAEA activities to facilitate drafting and application of the regulatory guidelines using international experience.
The IRIS General Plant Arrangement
Authors: Robertson J. , Love J. , Morgan R. , Conway L. E.
Published: March 2009

IRIS (International Reactor Innovative and Secure) is a light water cooled, 335 MWe power reactor which is being designed by an international consortium as part of the US DOE NERI Program. IRIS features an integral reactor vessel that contains all the major reactor coolant system components including the reactor core, the coolant pumps, the steam generators and the pressurizer. This integral design approach eliminates the large coolant loop piping, and thus eliminates large loss-of-coolant accidents (LOCAs) as well as the individual component pressure vessels and supports. In addition, IRIS is being designed with a long life core and enhanced safety to address the requirements defined by the US DOE for Generation IV reactors. Bechtel, with Westinghouse consultation, has performed a layout study of the IRIS plant and this paper will discuss the results of this design effort.
Multiple Tier Fuel Cycle Studies for Waste Transmutation
Authors: Hill R. N. , Taiwo T. A. , Stillman J. A. , Graziano D. J. , Bennett D. R. , Trellue H. , Todosow M. , Halsey W. G. , Baxter A.
Published: March 2009

As part of the U.S. Department of Energy Advanced Accelerator Applications Program, a systems study was conducted to evaluate the transmutation performance of advanced fuel cycle strategies. Three primary fuel cycle strategies were evaluated: dual-tier systems with plutonium separation, dual-tier systems without plutonium separation, and single-tier systems without plutonium separation. For each case, the system mass flow and TRU consumption were evaluated in detail. Furthermore, the loss of materials in fuel processing was tracked including the generation of new waste streams. Based on these results, the system performance was evaluated with respect to several key transmutation parameters including TRU inventory reduction, radiotoxicity, and support ratio. The importance of clean fuel processing (∼0.1% losses) and inclusion of a final tier fast spectrum system are demonstrated. With these two features, all scenarios capably reduce the TRU and plutonium waste content, significantly reducing the radiotoxicity; however, a significant infrastructure (at least 1/10 the total nuclear capacity) is required for the dedicated transmutation system.
Analysis of a CRDM Nozzle Break LOCA Without Scram Using the U.S. NRC Coupled Code TRAC-M/PARCS
Authors: Cuadra A. , Ragusa J. , Downar T. , Ivanov K.
Published: March 2009

Recently, through-wall circumferential cracks in several control rod drive mechanism (CRDM) nozzle penetrations were detected at the Oconee-3 nuclear power plant. The presence of these cracks was seen as a potential precursor to a small break loss of coolant accident. In order to assess the impact of a postulated failure of a CRDM housing, analyses were performed using the U.S. NRC coupled thermal-hydraulics and neutronics code TRAC-M/PARCS. Although deemed highly unlikely, it was assumed that no control rods inserted in order to bound any possible reactivity transient associated with the break. The thermal-hydraulic model used to perform the study is based on an existing model of Oconee, built for PTS analysis, which models the whole plant and some of its control systems. A refined vessel model based on a TMI model was used to increase the resolution of the results and facilitate coupling to PARCS. All relevant ECCS systems were modeled and the control system allowed for, in addition to automatic actions, some assumed operator intervention. In particular, HPI throttling was modeled to maintain the hot leg subcooled at 42 +/- 7 K, and prevent an excessive amount of cold water from being injected into the system. A spatial kinetics analysis of this event was necessary because of the wide range of core conditions which occurred during the transient, from hot full power operation conditions to the cold zero power shutdown state. Analysis of the event with point kinetics and “best estimate” reactivity coefficients resulted in significant miss-prediction of the core power response. Conversely, the three-dimensional kinetics solution with cross section data generated over the entire range of the event led to a more accurate calculation of the power response and the overall analysis of the system transient response. This paper will describe the analysis of the control rod drive nozzle break event without scram using TRAC-M/PARCS.
Failure Pressure and Leak Rate of Steam Generator Tubes With Stress Corrosion Cracks
Authors: Majumdar S. , Kasza K. , Park J. Y. , Bakhtiari S.
Published: March 2009

This paper illustrates the use of an “equivalent rectangular crack” approach to predict leak rates through laboratory generated stress corrosion cracks. A comparison between predicted and observed test data on rupture and leak rate from laboratory generated stress corrosion cracks are provided. Specimen flaws were sized by post-test fractography in addition to pre-test advanced eddy current technique. The test failure pressures and leak rates are shown to be closer to those predicted on the basis of fractography than on NDE. However, the predictions based on NDE results are encouraging, particularly because they have the potential to determine a more detailed geometry of ligamented cracks from which more accurate predictions of failure pressure and leak rate can be made in the future.
The Joint Owners’ Group Program on MOV Periodic Verification
Authors: Damerell Paul S. , Spears Todd A.
Published: March 2009

To address long-term motor operated valve (MOV) performance, the Babcock & Wilcox, Boiling Water Reactor, Combustion Engineering and Westinghouse Groups (B&WOG, BWROG, CEOG and WOG) teamed in 1997 to form the Joint Group (JOG) MOV Periodic Verification (PV) Program. This program is nearing completion, with 98 of the 103 operating U.S. reactor units participating. The goal of the program is to provide a justified approach for periodically testing MOVs, that addresses potential degradation. The program defines an interim approach that specifies periodic tests without flow and differential pressure (DP), at a frequency determined by the MOV’s risk significance and margin. To justify this approach, each participating plant is also DP testing 2 valves per unit. Each valve is tested three times over five years, with at least one year between tests. The data are evaluated jointly to confirm or adjust the initial guidance. The majority of the tests are complete and conclusions are coming into focus. For gate valves, when the valve factor is initially low, increases can occur between one test and a later test. One common way that the valve factor becomes low is disassembling and reassembling the valve. The data show that, following valve disassembly and reassembly, the valve factor tends to be reduced, and it tends to increase in subsequent service. Outside of the valves disassembled and reassembled, some gate valves have low valve factors apparently because the valves are not stroked under DP conditions in service. For butterfly valves, there have been no observations of degradation in bearing friction coefficient. A few valves with bronze bearings in raw (untreated) water service have shown significant variations in friction, but they tend to be a mixture of increases and decreases with no pattern of degradation. Globe valves, both unbalanced and balanced, tend to show a constant valve factor with no indication of degradation.
Comparison of Novovoronezh Unit 5 NPP and South Ukraine Unit 1 NPP Level 1 PRA Results
Authors: Musicki Zoran , Ginsberg Ted
Published: March 2009

This paper describes a study undertaken to explain the risk profile differences in the results of PRAs of two similar VVER-1000 nuclear power plants. The risk profile differences are particularly significant in the area of small steam/feedwater line breaks, small-small LOCAs, support system initiators and containment bypass initiators. A top level (limited depth) approach was used in which we studied design differences, major assumptions, data differences, and also compared the two PRA analyses on an element-by-element basis in order to discern the major causative factors for the risk profile differences. We conclude that the major risk profile differences are due to differences in assumptions and engineering judgment (possibly combined with some design and data differences) involved in treatment of uncertain physical phenomena (primarily sump plugging in LOCAs and turbine building steaming effects in secondary system breaks). Additional major differences are attributable to support system characteristics.
Modeling of Porous Crud Layer Based on Effects of Diffusion and Water Radiolysis on Anomalous Porous Crud Deposition on Fuel Pin Surfaces in PWRs
Authors: Joe Hyun-Jong , Jones Barclay G.
Published: March 2009

Many studies have been undertaken to understand crud formation on the upper spans of fuel pin clad surfaces, which is called axial offset anomaly (AOA), is observed in pressurized water reactors (PWR) as a result of sub-cooled nucleate boiling. Separately, researchers have considered the effect of water radiolysis in the primary coolant of PWR. This study examines the effects of radiolysis of liquid water, which aggressively participate in general cladding corrosion and solutes within the primary coolant system, in the terms of pH, temperature, and Linear Energy Transfer (LET). It also discusses the effect of mass transfer, especially diffusion, on the concentration distribution of the radiolytic products, H2 and O2, in the porous crud layer. Finally it covers the effects of chemical reactions of boric acid (H3BO3), which has a negative impact on the mechanisms of water recombination with hydrogen, lithium hydroxide (LiOH), which has a negative effect on water decomposition, dissolved hydrogen (DH), and some trace impurities.