bluuebunny/crossref_metadata_2025 huggingface dataset:
Pu-Rich MOX Agglomerate-by-Agglomerate Model for Fuel Pellet Burnup Analysis
Authors: Chang G. S.
Published: November 2008

In support of potential licensing of the mixed oxide (MOX) fuel made from weapons-grade (WG) plutonium and depleted uranium for use in United States reactors, an experiment containing WG-MOX fuel is being irradiated in the Advanced Test Reactor (ATR) at the Idaho National Engineering and Environmental Laboratory (INEEL). The WG-MOX comprises five percent PuO2 and 95% depleted UO2. Based on the Post Irradiation Examination (PIE) observation, the volume fraction (VF) of MOX agglomerates in the fuel pellet is about 16.67%, and PuO2 concentration of 30.0 = (5 / 16.67 × 100) wt% in the agglomerate. A pressurized water reactor (PWR) unit WG-MOX lattice with Agglomerate-by-Agglomerate Fuel (AbAF) modeling has been developed. The effect of the irregular agglomerate distribution can be addressed through the use of the Monte Carlo AbAF model. The AbAF-calculated cumulative ratio of AGglomerate burnup to U-MAtrix burnup (AG/MA) is 9.17 at the beginning of life (BOL), and decreases to 2.88 at 50 GWd/t. The MCNP-AbAF-calculated results can be used to adjust the parameters in the MOX fuel fission gas release modeling.
Quantification Frameworks and Their Application for Evaluating the Software Quality Factor Using Quality Characteristic Value
Authors: Kim Chansoo , Chung Chang-Hyun , Ahn Kwang Won
Published: November 2008

Many problems, related with safety, frequently occur because Digital Instrument & Control Systems are widely used and expanding their ranges to many applications in Nuclear Power Plants. It, however, does not hold a general position to estimate an appropriate software quality. Thus, the Quality Characteristic Value, a software quality factor through each software life cycle, is suggested in this paper. The Quality Characteristic Value is obtained as following procedure: 1) Scoring Quality Characteristic Factors (especially correctness, traceability, completeness, and understandability) onto Software Verification and Validation results, 2) Deriving the diamond-shaped graphs by setting values of Factors at each axis and lining every points, and lastly 3) Measuring the area of the graph for Quality Characteristic Value. In this paper, this methodology is applied to Plant Control System. In addition, the series of quantification frameworks exhibit some good characteristics in the view of software quality factor. More than any thing else, it is believed that introduced framework may be applicable to regulatory guide, software approval procedures, due to its soundness and simple characteristics.
Finding Megawatts in Nuclear Power Plants With Process Data Reconciliation
Authors: Langenstein Magnus , Jansky Josef , Laipple Bernd , Eitschberger Horst , Grauf Eberhard , Schalk Hermann
Published: November 2008

Process data reconciliation with VALI III is a method for monitoring and optimising industrial processes as well as for component diagnosis, condition-based maintenance and online calibration of instrumentation. Employing process data reconciliation in nuclear power plants enables thermal reactor power to be determined with an uncertainty of less than ± 0.5%, without having to install additional precision instrumentation to measure as for example the final feed-water mass flow. This is equivalent to a measurement uncertainty recapture power uprate potential of about 1.5% (maximum allowed potential is 2.0%). In addition, process data reconciliation is able to detect any drift in the measured values at an early stage, yet allowing for the reconciled variables (such as thermal reactor power) to be calculated with consistently high precision. Without process data reconciliation • drift in measured values and • systematic errors for the feed-water temperature or the feed-water mass flow could remain undetected. With such measurements the thermal reactor power calculation may incorporate an unacceptably large deviation, which has a negative impact on both, safety and economical aspects. This paper describes, how process data reconciliation works and shows examples of the finding and gain of more than 30 MW electrical power in PWR and BWR units in Germany and Switzerland.
Characteristics of Polonium Contamination From Neutron Irradiated Lead-Bismuth Eutectic
Authors: Miura Terumitsu , Obara Toru , Sekimoto Hiroshi
Published: November 2008

After neutron capture, bismuth-209 changes to polonium-210 that emits α-particles. Lead-Bismuth eutectic (LBE) in reactor system contaminates the system by polonium. We analyzed adsorbed materials from melted LBE on quartz glass plate. Lead, bismuth and their oxides were confirmed in adsorbed materials. And, we evaluated the baking method in vacuum for removal of polonium and adsorbed materials on quartz glass plate. It was evaluated that it is possible to remove almost all of polonium from the quartz glass plate by baking at temperature more than 300 °C. Unfolding method was applied to calculate polonium distribution in LBE ingot. From measured α-particle pulse height distribution, the polonium distribution in depth of LBE ingot was calculated using quadratic programming code, where response functions are calculated by Monte Carlo method.
A Vibration Analysis and Test of a 5×5 Rod Bundle
Authors: Kang Hueng Seok , Choi M. H. , Yoon K. H. , Song K. N. , Jung Y. H.
Published: November 2008

A FE analysis and test has been performed for a 5×5 rod bundle which consists of twenty-three rods that are 2.2 m tall and 9.5 mm in diameter, with two guide tubes and five spacer grids. Two guide tubes were welded to the spacer grids through the medium of guide tube sleeves. Two out of the twenty-three rods were empty, and the rest of them were filled with lead (Pb) rods instead of UO2 pellets. The vibration test was performed for the rod bundle in air, and the test result was compared with the FE analysis that was done by ABAQUS. 6.5 Hz for the first natural frequency, 15 Hz for the second natural frequency and 24 Hz for the third natural frequency were obtained by the vibration test while 4.3 Hz for the first natural frequency, 10.4 Hz for the second natural frequency and 19.2 Hz for the third natural frequency were obtained respectively by the FE analysis. Since two of the first, second and third modes were obtained in two different directions by the FE analysis, tests were carried out in two directions. It is concluded that the lower 6 modes are the bundle vibration modes. The first vibration modes of the rod supported by the springs appeared at 26.9 Hz to 27.0 Hz after the bundle vibration modes, and then several bundle vibration modes followed.
Modeling and Simulation of Cross-Flow Induced Vibration in a Multi-Span Tube Bundle
Authors: Khushnood Shahab , Khan Zaffar M. , Malik M. Afzaal , Koreshi Zafarullah , Khan Mahmood Anwar
Published: November 2008

Flow-induced vibration in steam generator and heat exchanger tube bundles has been a source of major concern in nuclear and process industry. Tubes in a bundle are the most flexible components of the assembly. Flow induced vibration mechanisms, like fluid-elastic instability, vortex shedding, turbulence induced excitation and acoustic resonance results in failure due to mechanical wear, fretting and fatigue cracking. The general trend in heat exchanger design is towards larger exchangers with increased shell side velocities. Costly plant shutdowns have been the motivation for research in the area of cross-flow induced vibration in steam generators and process exchangers. The current paper focuses on the development of a computer code (FIVPAK) for the design (natural frequencies, variable geometry, tube pitch & pattern, mass damping parameter, reduced velocity, strouhal and damage numbers, added mass, wear work rates, void fraction for two-phase, turbulence and acoustic considerations etc.) of tube bundles with respect to cross flow-induced vibration. The code has been validated against Tubular Exchanger Manufacturers (TEMA), Flow-Induced Vibration code (FIV), and results on an actual variable geometry exchanger, specially manufactured to simulate real systems. The proposed code is expected to prove a useful tool in designing a tube bundle and to evaluate the performance of an existing system.
Linear Analysis of Density-Wave Oscillations in a Parallel Heated Channel
Authors: Abdou Hesham Nagah , Garea Veronica Beatriz , Larreteguy Axel Eduardo
Published: November 2008

A one-dimensional analytical model has been developed to be used for the linear analysis of density-wave oscillations in a parallel heated channel. The heated channel is divided into a single-phase and a two-phase region. The two-phase region is represented by the drift-flux model. The model accounts for phasic slip and subcooled boiling. The localized friction at the exit of the heated channel is treated considering the two-phase mixture. The exact equation for the total channel pressure drop is perturbed around the steady state. The stability characteristics of the heated channel are investigated using the Nyquist criterion. The marginal stability boundary (MSB) is determined in the two-dimensional thermodynamic equilibrium space parameters, the subcooled boiling number and the phase change number. The predictions of the model agree well with experimental data published in open literature. (1) The effect of thermal equilibrium, equal velocity (homogeneous) model increases the channel steam quality which leads to decrease in localized exit loss coefficient and finally stabilizes the system. (2) The effect of thermal equilibrium, non-equal velocity (drift-flux) model decreases the channel steam quality than the homogeneous model and finally destabilizes the system. (3) The effect of thermal non-equilibrium, equal velocity model (a) decreases the channel steam quality from high subcooled boiling number to a critical value and destabilizes the system, (b) increases the channel steam quality from the critical value to low subcooled boiling number and stabilizes the system. (4) The effect of thermal non-equilibrium, non-equal velocity model (a) decreases the channel steam quality from high subcooled boiling number to a critical value and destabilizes the system, (b) increases the channel steam quality from the critical value to low subcooled boiling number and stabilizes the system and the frequency of oscillation is in good agreement with the experimental data than the other models.
Development of FPGA-Based Safety-Related I&C Systems
Authors: Oda Naotaka , Tarumi Teruji , Tanaka Atsushi , Izumi Mikio , Sato Toshifumi
Published: November 2008

Toshiba has developed FPGA-based systems which perform signal processing by field programmable gate arrays (FPGA) for safety-related I&C systems. FPGA is a device which consists only of defined digital circuit: hardware, which performs defined processing. FPGA-based system solves issues existing both in the conventional systems operated by analog circuits (analog-based system) and the systems operated by central processing units (CPU-based system). The advantages of applying FPGA are to keep the long-life supply of products, improving testability (verification), and to reduce the drift which may occur in analog-based system. Considering application to safety-related systems, nonvolatile and non rewritable FPGA which is impossible to be changed after once manufactured has been adopted in Toshiba FPGA-based system. The systems which Toshiba developed this time are Power range Monitor (PRM) and Trip Module (TM). These systems are compatible with the conventional analog-based systems and the CPU-based systems. Therefore, requested cost for upgrading will be minimized. Toshiba is planning to expand application of FPGA-based technology by adopting this development method to the other safety-related systems from now on.
An Advanced Ultrasonic Technique for Flow and Void Fraction Measurements of Two-Phase Flow
Authors: Faccini J. L. H. , Su J. , Harvel G. D. , Chang J. S.
Published: November 2008

In this paper, we present a hybrid type contrapropagating transmission ultrasonic technique (CPTU) for flow and time averaging ultrasonic transmission intensity void fraction measurements (TATIU) of air-water two-phase flow, which is tested in the new two-phase flow test section mounted recently onto an existing single phase flow rig at the Nuclear Engineering Institute (IEN)/CNEN, Brazil. The circular pipe test section is made of 51.2 mm stainless steel, followed by a transparent extruded acrylic pipe aimed at flow visualization. The two-phase flow rig operates in several flow regimes: bubbly, smooth stratified, wavy stratified and slug flow. The observed flow patterns are compared with the Mandhane et al.’s experimental and Lightstone et al.’s numerical flow regime map for horizontal two phase flows. These flow patterns will be identified by time averaging transmission intensity ultrasonic techniques which have been developed to meet this particular application. A contrapropagating transmission ultrasonic flowmeter is used to measure the flow rate of liquid phase. A pulse-echo TATIU ultrasonic technique used to measure the void fraction of the horizontal test section assembling at IEN is presented. Other flow parameters can be deduced by processing the signals obtained by the CPTU ultrasonic flowmeter and the pulse-echo generator-receiver (TATIU system).
Evaluation of Single-Phase CFD Applicability to Estimate Fluid Hot Spot Locations by Comparing Water DNB Data
Authors: Ikeda Kazuo , Makino Yasushi , Hoshi Masaya
Published: November 2008

For development of PWR spacer grids, it is necessary to confirm mechanical strength and thermal-hydraulic characteristics. Mitsubishi has applied Computational Fluid Dynamics (CFD) to thermal-hydraulic design of spacer grids. To satisfy the requirement of high thermal performance spacer grid, the compatibility of low-pressure loss and high Critical Heat Flux (CHF) performance is necessary; therefore, parametric CFD analyses have been carried out for new spacer grid designs. Related to spacer grid design, i.e., strap structures and mixing vane, the parametric analyses have been carried out to estimate pressure-loss of each spacer grid. Moreover, thermal analyses, where heat generation from fuel rods was taken into account, have been carried out to evaluate coolant mixing capability, which is assumed to relate the CHF performance, by comparing fluid peak temperature of each grid design. In our previous study (ICONE11-36087), the rod type Laser Doppler Velocimetry (rod LDV) and Particle Imaging Velocimetry (PIV) technique were applied to cross-flow and axial flow measurements in rod gaps and sub-channels to obtain reference data for verification of CFD estimation. Estimated velocity fields at the downstream of the grid were quantitatively compared with the measurements. As a result, it was confirmed that the CFD modeling estimated flow behavior in the rod bundle appropriately. In this study, CFD under single-phase condition, which took into account heat generation from rods, was performed to simulate flow conditions of water DNB test with the same design grid of the previous study. The correlation between estimated enthalpy distribution around fuel rods by the CFD and the CHF rods in the DNB test was examined. This study was performed in collaboration with Westinghouse Electric Company and Mitsubishi Heavy Industries, Ltd.
An Experimental Investigation of Sub-Cooled Flow Boiling in Hypervapotron Cooling Configuration
Authors: Chen Peipei , Jones Barclay G. , Newell Ty A.
Published: November 2008

This work reports on experimental studies to visualize nucleate boiling on the enhanced heat transfer surface of the hypervapotron for with application in the International Thermonuclear Experiment Reactor [ITER]. This research uses the simulant fluid Freon (R134A) instead of prototypic water to model the system performance. This results in much lower thermophysical conditions to represent the prototypic phenomena. By using reduced pressure, temperatures, etc, based on the critical physical properties of both working fluids, Freon and water, the dramatic drop in the level of these quantities with Freon allows the use of modest test conditions. The experiment was conducted for both saturated and subcooled boiling with different heat fluxes (from 50 to 300 kW/m2). A comparison of the heat transfer performance of finned structures and flat surfaces were examined under particular fluid conditions. The uniqueness of this work is the visualization method that allows direct observation of the subcooled boiling process of the Hypervapotron surfaces. Working with a high speed (12,000 frames per second), high fidelity digital camera with variable magnifications (from 1×–25×), the sub-cooled boiling phenomena was observed in detail. A major conclusion of this work is the existence of two separate zones linked to different energy removal efficiency in hypervapotron. Under high heat flux condition, enhanced boiling heat transfer (about 20–30% higher than flat surface) was observed for hypervapotron effect, while saturated boiling happened in the cavity, and a large portion of the region was vapor filled. The process of vapor bubble rotation in the slot appeared to be helpful to enhance energy transfer, as evidenced by an improved wetting condition on the heating surfaces.
Conceptual Design Study on Fast Reactor Fuel Reprocessing System Using Super-DIREX Process
Authors: Takata Takeshi , Koma Yoshikazu , Sato Koji , Shimada Takashi , Mori Yukihide , Ogumo Shinya , Ishida Yasuhiro
Published: November 2008

A conceptual design study for some alternatives as well as the advanced aqueous reprocessing system has been progressed as a feasibility study on commercialized fast reactor cycle systems. Among these alternatives, the Supercritical fluid Direct Extraction (Super-DIREX) process is expected to minimize the reprocessing cost because the heavy metals (U, Pu, Np, Am and Cm) are directly extracted from the spent fuel powder in a column covering the dissolution, clarification and extraction process of the advanced aqueous process. The conceptual design for the reprocessing plant using Super-DIREX process was conducted considering the flowsheet, mass balance, specification and number of the components and layout of the components. From the result of the evaluation for the reprocessing cost of the plant with the capacity of 50tHM/y, it is found out that about 30% of the reprocessing cost is reduced compared with the advanced aqueous process (the NEXT process). In addition, using the Super-DIREX process, about 30% of wastes generated from the plant are reduced.
Application of Identification Method of Swirling Motion by Swirl Function
Authors: Nakayama Katsuyuki , Umeda Kenji
Published: November 2008

A method of identification and estimation of swirling motion in complicated flow and its application are presented. Classification of flow can be performed with velocity gradient tensor and its eigenvalue, and complex eigenvalue indicates that flow is swirling motion. Here the complex number of the eigenvalue is defined as swirling function, and the local maximum point of swirling function is assumed to be the axis of swirling. This method enables to identify the swirling motion hidden in complicated flow, which is impossible to identify with velocity field or streamline. This definition prevents misunderstanding between swirling flow and non-swirling flow with (non-zero) vorticity, and the intensity of swirling can be estimated.
Numerical Study on Thermal Hydraulics in Coated-Particle-Type Fuel Assembly of Helium Gas Cooled Fast Reactor
Authors: Ohshima Hiroyuki , Nagata Takemitsu
Published: November 2008

A helium-gas-cooled fast reactor with coated-particle-type fuel is being examined as a candidate of commercialized fast reactors at Japan Nuclear Cycle Development Institute. As one of characteristics of this reactor design, coolant passes through coated particle fuel regions of fuel assemblies “laterally” to reduce pressure loss. From the viewpoint of coolability and structure integrity of the fuel assembly, it should be confirmed that the coolant keeps flowing through the fuel regions laterally and uniformly to remove heat from the fuel particles properly under various operating conditions. In this study, two kinds of numerical simulations were carried out: A whole fuel assembly simulation using a multi-dimensional thermal-hydraulic analysis code AQUA and a local detailed simulation using a commercial computational fluid dynamics code FLUENT. From these results, thermal hydraulic characteristics in the fuel region were clarified and coolability was confirmed for several operating modes as rated operation, decay heat removal operation and depressurization accident.
Flow Analyses Using RELAP/MOD3.3 Code for the Simulant Melt Experiments (LAVA-ERVC) Under the External Reactor Vessel Cooling
Authors: Kang Kyoung-Ho , Park Rae-Joon , Kim Sang-Baik , Kim Hee-Dong
Published: November 2008

Flow analyses using RELAP5/ MOD3.3 code have been performed to investigate the occurrence and the effects of steam binding for the LAVA-ERVC experiments. The main objectives of the LAVA-ERVC experiments are investigations of coolability through external reactor vessel cooling according to RPV insulation design. It could be found from the sensitivity studies for the flow path in the annulus of insulation that steam binding could occur in case of the limited steam venting capacity, which is definitely coincident with the LAVA-ERVC experimental results. In case of sufficient flow path for the steam venting, the vessel experienced effective cooling by nucleate boiling heat transfer. And existence of the upper free volume had little effect on occurrence of steam binding in the LAVA-ERVC experiments.
Replacement of Secondary Heat Transport System Components in the Experimental Fast Reactor JOYO
Authors: Kawahara Hirotaka , Ichige Satoshi , Isozaki Kazunori , Nakai Satoru
Published: November 2008

A recently completed major upgrade of the JOYO experimental sodium-cooled fast reactor, to the MK-III design, increased its irradiation capability. One major change was a 40% increase in thermal power to 140 MWt, which necessitated the replacement of the cooling system. Major challenges in the replacement of secondary components were control of impurity ingress and assurance of welding integrity. Damage to existing systems was avoided during replacement operations by taking measures to prevent ingress of air into the sodium systems. The long exposure of the used pipes made of ferritic low-alloy steel to hot sodium was a concern because previous research showed that this material changes its mechanical property in sodium hotter than 673 K. Used pipe, heat transfer tubes and welds were subjected to material tests. These tests did not show notable material problems. The replacement of components was completed without major troubles, demonstrating the effectiveness of the methods used.
Influence of Dissolved Gas Concentration on Subcooled Flow Boiling Critical Heat Flux in Short Vertical Tube
Authors: Hata Koichi , Komori Hirokazu , Shiotsu Masahiro , Noda Nobuaki
Published: November 2008

The subcooled flow boiling critical heat fluxes (CHFs) are systematically measured by using the water loop installed the pressurizer to investigate the pressurization gas effect for the flow velocities (u = 4.0 to 13.3 m/s), the inlet subcoolings (ΔTsub,in = 70.08 to 145.57 K) and the inlet pressure (Pin = 740.67 to 975.43 kPa). The SUS304 tube of the inner diameter (d = 3 mm), the heated length (L = 66.5 mm) and L/d = 22.17 with the inner surface of rough finished was used. The CHF data under the pressurization by saturated vapor in a pressurizer were compared with the values calculated from the CHF correlations against outlet and inlet subcoolings based on the experimental data under the nitrogen (N2) gas pressure previously obtained. The CHFs for the dissolved oxygen concentration of 6.59, 4, 2, 1, 0.5, 0.0496 and 0.0265 ppm were systematically measured for the flow velocities (u = 4.0 to 13.3 m/s), the inlet subcoolings (ΔTsub,in = 137.49 to 153.87 K) and the inlet pressure (Pin = 762.97 to 975.78 kPa). And the CHFs for the test tube with the smooth finished inner surface were measured at the dissolved oxygen concentration of 0.1004 and 0.0413 ppm for the same experimental conditions. Influence of the inner surfaces of rough and smooth finished on the CHFs was clarified for low dissolved gas concentration.
Shielding Analyses for Gamma Spectrometer GAMMACELL in Experiments on Neutron Generator
Authors: Serikov Arkady
Published: November 2008

The paper is devoted to computational analysis of shielding properties of spectrometer GAMMACELL irradiated by neutrons from 14.1 MeV Neutron Generator (NG). The shielding is required to protect the BaF2 crystals of the gamma spectrometer GAMMACELL from background thermal neutrons, scattered from the walls and equipment in experimental room. GAMMACELL is a scintillation spectrometer of full gamma particles absorption. It consists of 9 BaF2 crystals; each of them has 40×40×100 mm3 dimensions and is viewed by photomultiplier tube. Neutrons also cause flashes in BaF2 crystals and give spurious pulses. The ultimate aim of the work is to simulate the severe radiation load on spectrometer GAMMACELL and to find proper radiation shield to fulfill the most favorable operating conditions for the GAMMACELL. The neutrons were generated with energy 14.1 MeV form the source on titanium–tritium target on a copper substrate. Accelerator tube makes deuteron current with spot 4 mm diameter on the target. Between the neutron source and GAMMACELL crystals collimator is placed in order to separate neutron current with particular energy and angle. The collimator is pointed to the central BaF2 crystal, counter is turned on only in case when gammas hit it. The other 8 periphery BaF2 crystals are intended for registration of Compton and annihilation gamma radiation which scattered from central crystal. The periphery crystals must be covered by lateral shielding case for protection form background radiation. It is supposed that on the basis of shielding analyses provided and data obtained in this work the real experiment on the Neutron Generator (NG) will be performed. But the experiment on NG will be not at the final destination for the results application obtained in this work, diagnostics of plasma in fusion tokamak-type reactor by means of GAMMACELL gamma spectrometry is planned. Plasma diagnostics by gamma emission from nuclear reactions on particles accelerated upon high energy during Deuterium-Tritium nuclear fusion. Resonances and thresholds in nuclear reactions reveal the value of particle energy based on registered gamma energy in GAMMACELL. High-penetrating capability of gamma quantum gives an opportunity to examine the plasma on all full depth.
Study on Mixing Due to Transversal Flow in a Fuel Subassembly of Fast Reactor: Sodium Experiment Using a 37-Pin Subassembly Model
Authors: Kamide Hideki , Miyakoshi Hiroyuki , Kawamata Nobuhiro , Ohshima Hiroyuki
Published: November 2008

Estimation of the highest temperature in a fuel subassembly is needed for safety and reliable operation of fast reactors. Cross flow has a significant role for mixing phenomena in the subassembly. Especially, the cross flow will be enhanced in a deformed pin bundle of high burnup core. Such mixing due to the cross flow was investigated. Sodium experiments were carried out using a wire wrapped 37-pin subassembly model. In order to see a cross flow in a simple and clear geometry as an element of mixing phenomena, a blockage was installed in the subassembly. Mixing due to the cross flow was found in the experiment. The subchannel analysis code, ASFRE, was applied to the experimental analyses. Measured axial and transverse temperature distributions in the subassembly were correlated with the cross flow and the wire spacer. The calculated temperature profiles were in good agreement with the experimental results. ASFRE simulated the influence of the wire spacer also. Database for the mixing phenomena with the cross flow was obtained.
Results of INPRO in the Area of Economics
Authors: Depisch Frank , Kupitz Juergen
Published: November 2008

In the area of Economics four selected scenarios from the SRES study have been analysed within the International Project on Innovative Reactors and Fuel Cycles (INPRO) of the IAEA. They cover a range of possible future developments characterized by different degrees of globalisation and by different relative priorities on economic and environmental objectives. Four “aggressive nuclear” variants, one for each of the four selected SRES scenarios, are also analyzed. Provided innovative nuclear energy systems (INS) are economically competitive, they can play a major role in meeting future energy needs. Future economic competitiveness will depend on the speed of continuing cost reductions achieved by nuclear energy relative to competing technologies. The paper presents specific capital costs and electricity production costs at which nuclear energy is competitive in 2050 in the four selected SRES scenarios, and estimates corresponding costs for nuclear energy in the four aggressive nuclear variants. The important message is that for nuclear technology to gain and grow market share it must benefit sufficiently from learning to keep it competitive with competing energy technologies. For such learning to take place experience must be gained and to gain such experience the energy from INS must be cost competitive with energy from alternative sources and INS must represent an attractive investment to compete successfully in the capital market place. In total, INPRO defined two basic principles, five user requirements and several criteria in this area, which are presented in the full paper. To be cost competitive all component costs, e.g., capital costs, operating and maintenance costs, fuel costs, must be considered and managed to keep the total unit energy cost competitive. Limits on fuel costs in turn imply limits on the capital and operating cost of fuel cycle facilities, including mines, fuel processing and enrichment, fuel reprocessing and the decommissioning and long term management of the wastes from these facilities. Cost competitiveness of energy from INS will contribute to investor confidence, i.e. to the attractiveness of investing in INS, as will a competitive rate of return.
Calculation of Forces on Reactor Containment Fan Cooler Piping
Authors: Miller Joseph S. , Ramsden Kevin
Published: November 2008

The purpose of this paper is to present the results of the Reactor Containment Fan Cooler (RCFC) system piping load calculations. These calculations are based on piping loads calculated using the EPRI methodology (Refs. 1 & 2) and RELAP5 (Ref. 3) to simulate the hydraulic behavior of the system. The RELAP5 generated loads were compared to loads calculated using the EPRI GL 96-06 methodology. This evaluation was based on a pressurized water reactor’s RCFC coils thermal hydraulic behavior during a Loss of Offsite Power (LOOP) and a loss of coolant accident (LOCA). The RCFC consist of two banks of service water and chill water coils. There are 5 SX and 5 chill water coils per bank. Therefore, there are 4 RCFC units in the containment with 2 banks of coils per RCFC. Two Service water pumps provide coolant for the 4 RCFC units (8 banks total, 2 banks per RCFC unit and 2 RCFC units per pump). Following a LOOP/LOCA condition, the RCFC fans would coast down and upon being reenergized, would shift to low-speed operation. The fan coast down is anticipated to occur very rapidly due to the closure of the exhaust damper as a result of LOCA pressurization effects. The service water flow would also coast down and be restarted in approximately 43 seconds after the initiation of the event. The service water would drain from the RCFC coils during the pump shutdown and once the pumps restart, water is quickly forced into the RCFC coils causing hydraulic loading on the piping. Because of this scenario and the potential for over stressing the piping, an evaluation was performed by the utility using RELAP5 to assess the piping loads. Subsequent to the hydraulic loads being analyzed using RELAP5, EPRI through GL 96-06 provided another methodology to assess loads on the RCFC piping system. This paper presents the results of using the EPRI methodology and RELAP5 to perform thermal hydraulic load calculations and compares them.
Practical Procedures for Estimation of Plant Specific Common Cause Failure Quantification Parameters for Plants With Accumulated Operating Experiences
Authors: Jo Young G.
Published: November 2008

One of the major tasks in a plant specific common cause failure (CCF) analysis is to determine the overall event applicability factors of the original CCF events. The overall event applicability factor of a CCF event, which is used for specializing the impact of the original event for a target system, should be determined by considering the target system’s defenses against such CCF. However, lack of currently available detailed guidelines made it difficult to determine the overall event applicability factors in a systematic and consistent manner. In this paper, practical procedures for plant specific CCF analysis were developed. Especially, guidelines and a set of criteria were developed for more systematic and consistent determination of the overall event applicability factors. Additional set of criteria was also developed especially for crediting plants with many years of operations for their enhanced defenses against CCFs based on accumulated operating experiences. The developed procedures were successfully implemented for three plant specific CCF analyses. It is believed that the use of the developed procedures would significantly facilitate a plant CCF analysis and produce consistent results.
OECD Pipe Failure Data Exchange Project (OPDE): 2003 Status Report
Authors: Lydell Bengt , Mathet Eric , Gott Karen
Published: November 2008

Established in 2002 as a three-year multilateral cooperation in the collection and analysis of data relating to pipe failure events, the OPDE project currently is supported by eighteen organizations from twelve member countries of the Nuclear Energy Agency of the Organisation for Economic Co-operation and Development (OECD-NEA). Objectives of the OPDE are to collect and analyze pipe failure data to promote a better understanding of underlying causes of failure, observed and potential impact on plant operations and safety, and prevention. The data analysis includes characterization of pipe failure events by reliability attributes and influence factors to facilitate the estimation of piping reliability parameters. With emphasis on data validity and quality, during the first year of operation a coding format has been developed to ensure consistent interpretation and applications. This paper represents an abbreviated 2003 status report and an overview of the project organization and objectives.
Thermal Hydraulic Instability Characteristic in Natural Circulation Parallel Boiling Channels Upflow System Under Low Pressure
Authors: Watanabe Noriyuki , Subki Muhammad Hadid , Kikura Hiroshige , Aritomi Masanori
Published: November 2008

Natural circulation BWR actively equipped with passive safety features has been promoted to solve the recent challenges in BWR nuclear power and safety technology. With regard to startup stage, various thermo-hydraulic instabilities might be induced due to an elimination of re-circulation pumps. A lot of studies have been made on the instabilities in evaporated system as well as in a reactor. In the instabilities, geysering accompanied with flow reversal phenomena has been investigated in a vertical closed loop, U-shaped closed loop, twin parallel channels, and so on. However, in twin parallel study the effect of non-heated length on geysering has not been sufficiently clarified. The objective of this research is to experimentally investigate the thermo-hydraulic instabilities, particular in geysering, with a natural circulation loop consisting of parallel boiling channels and the single connection channel, which simulates the basic flow around the reactor core in the system pressure range from atmospheric to 0.7MPa. The parallel boiling channels are consisted of heated and non-heated section. The heated section forms annulus and heated from the inner wall. The input heat flux range of 0 up to 580kW/m2, and inlet subcooling temperatures of 5, 10, and 15K respectively, are imposed in the experiments. In the parallel channels with non-heated risers, three types of thermo-hydraulic instabilities were detected in the following sequence, geysering, natural circulation oscillation, and density wave oscillation. Especially in Geysering, it is induced due to rapid condensation in the non-heated risers and it is not be suppressed even at 0.7MPa though it has a tendency to be suppressed with an increase in the system pressure. On the other hand, in the parallel channels without non-heated risers, sinusoidal oscillation similar to natural circulation oscillation has been detected, and geysering had never observed. The new findings are that the sinusoidal oscillation is induced due to the hydrostatic head fluctuation in the connection channel, where the flow regime is constantly slug flow. The oscillating period is well correlated with the sum of delay time for boiling and passing time of slug bubbles in the connection channel. From the facts described above, it is found that non-heated region in a channel box should be as shorter as possible to prevent geysering from occurring, and sinusoidal oscillation similar to natural circulation oscillation is induced in any configuration of parallel channels.
Toughness and Embrittlement of RPV Steels Using Ultrasonic Measurements
Authors: Hiser Allen L.
Published: November 2008

Neutron bombardment of reactor pressure vessel (RPV) steels causes reductions in fracture toughness in these steels, termed neutron irradiation embrittlement. Currently there are no accepted methods for nondestructive determination of the extent of the irradiation embrittlement nor the actual fracture toughness of the reactor pressure vessel. This paper summarizes the findings from an effort addressing the use of ultrasonic attenuation as a suitable parameter for nondestructive determination of irradiation embrittlement in RPV steels. The materials examined in this study include one heat of RPV steel that was heat treated to induce changes in its fracture toughness, several heats of RPV steel irradiated to assess neutron embrittlement changes in fracture toughness, and a matrix of RPV steels (in the unirradiated condition) with a range of as-fabricated fracture toughness levels.
Analysis of Transient Coolant Void Formation During a Guillotine-Type HX Tube Rupture Event in the STAR-LM System Employing a Supercritical CO2 Brayton Cycle
Authors: Farmer M. T. , Sienicki J. J.
Published: November 2008

One proposed concept for the STAR-LM Lead Fast Reactor (LFR) incorporates a supercritical CO2 gas turbine Brayton cycle to achieve high cycle efficiency and reduced plant footprint. In this design, 100+% of core full power is transferred by natural circulation from the core, located at the bottom of the reactor vessel, to in-vessel heat exchangers (HXs) located at the top of the vessel in the annulus between the core shroud and vessel inner wall. Although this approach extremely simplifies the plant design, the presence of the HX’s within the vessel raises concerns regarding the potential rupture of a HX tube that would initiate a high-pressure blowdown of CO2 into the lead coolant. The principal issue is to what extent, if any, is void entrained downwards with the coolant and then upwards through the core where adverse reactivity effects or degraded heat removal could result. To address this question, a scoping analysis of transient void formation during a guillotine-type HX tube rupture event in the STAR-LM employing a supercritical CO2 Brayton cycle has been performed. The void formation process is evaluated by solving a coupled set of ordinary differential equations describing: i) the supercritical CO2 blowdown, ii) bubble center-of-mass trajectory, iii) bubble growth rate, iv) bubble gas internal energy, and v) discrete bubble formation rate due to Taylor instability at the bubble/coolant interface. The results indicate that for thermal hydraulic conditions consistent with the current STAR-LM design, the peak blowdown rate from a single tube rupture is ∼ 2.5 kg/sec. The void formation process is dominated by large coherent gas bubbles that penetrate minimally downwards into the coolant due to the large coolant density. Rather, the gas pockets are predicted to periodically rise due to buoyancy and vent to the core cover gas region, as opposed to being swept downwards with the coolant. Moreover, the total CO2 fraction that is rendered in the form of discrete bubbles during blowdown is found to be small (∼ 3%), and the bubbles are of fairly large diameter (≥ 0.7 cm). Thus, these discrete bubbles are also calculated to benignly rise to the cover gas region since the terminal rise velocity for the bubbles exceeds the average lead coolant down flow velocity below the HX.
A Regulatory Inspection Tool in SDP Context
Authors: Chao Chun-Chang , Lin Jyh-Der , Lin Tsu-Jen , Huang Shih-Hsun
Published: November 2008

The Institute of Nuclear Energy Research (INER) is now developing a tool to help the inspectors to perform the Phase 2 SDP (Significance Determination Process) assessment. The purpose of this tool is to simplify the process of the assessment and help the inspectors to obtain the results quickly. Risk models for the calculation of core damage frequency (CDF) of internal initiating events will be used to determine the risk increase of specific inspection findings. The risk model used for the SDP tool was converted from the original Probabilistic Risk Assessment (PRA) model of the individual plant. The risk engine, INERISKEN, is adopted to perform the calculation. After input of the inspection findings are setup, all calculations can be done in less than one minute by INERISKEN. To perform the Phase 2 assessment, the inspector has to decide the affected components/systems and the impact to initiating events for his inspection findings. When changes of plant configuration are specified, the ΔCDF can be obtained by solving the risk model with new plant configuration. Simplified P&IDs for most of the safety systems are also available for user’s reference. If any inspector cannot make sure the response of the inspection findings to the affected systems or the initiating events, experiences and examples are also provided on user’s click. The ΔCCDP (Conditional Core Damage Probability) obtained can be related to one of the color (green, white, yellow or red) by the methodology defined in the counting rule worksheet of NRC inspection manual. Meanwhile, the status of the critical safety functions is also displayed according to the new plant configuration. These additional results will help the inspector to further understand the risk insight of inspection findings.
Insights From Plant Specific Common Cause Failure Analysis
Authors: Jo Young G.
Published: November 2008

Insights from three plant specific common cause failure (CCF) analyses, which used CCF Data Base and Analysis System software by US NRC, were presented in this paper. It was found that detailed guidelines/criteria were needed for performing a plant specific CCF analysis, especially for more systematic and consistent determination of the impacts of historical CCF events on a specific target system. Information collection task, impact of using unscreened independent events, and insights from specific components CCF analyses were also discussed. Finally, the following recommendations were made: (1) Develop procedures and criteria, like those developed for the plant specific CCF analyses, for more systematic and consistent plant specific CCF analysis. In developing criteria, it should be considered to credit existing plants for the enhanced defenses against CCFs based on their accumulated operating experiences. (2) Use of generic CCF data is strongly discouraged. Instead, perform plant specific CCF analysis, at least initial screening of the original events as a minimum, in order to avoid potential over-conservatism. (3) Enhance the current internet based information data bases and develop a communication network among CCF analysts in order to facilitate the information collection task.
Variational Approach in Transport Theory
Authors: Panta Pazos Rube´n , de Vilhena Marco Tu´llio
Published: November 2008

In this work we present a variational approach to some methods to solve transport problems of neutral particles. We consider a convex domain X (for example the geometry of slab, or a convex set in the plane, or a convex bounded set in the space) and we use discrete ordinates quadrature to get a system of differential equations derived from the neutron transport equation. The boundary conditions are vacuum for a subset of the boundary, and of specular reflection for the complementary subset of the boundary. Recently some different approximation methods have been presented to solve these transport problems. We introduce in this work the adjoint equations and the conjugate functions obtained by means of the variational approach. First we consider the general formulation, and then some numerical methods such as spherical harmonics and spectral collocation method.
Thermal Analysis of a Marine Products Irradiator and Its Transportation Cask
Authors: Ali Seik Mansoor , Verma Vishnu , Markandeya S. G. , Ghosh A. K. , Kushwaha H. S.
Published: November 2008

Irradiation of sea-foods is carried out in a marine products irradiator. During the radiation processing, it is required to maintain the product temperature within very narrow temperature limits. A transient thermal analysis of the irradiator (when in use) was carried out to determine the chilled air temperature and velocity required to maintain the product temperature within the specified range. In order to transport the irradiator cask with its contents, it is enclosed in an outer enclosure. The transportation cask is required to satisfy regulations pertaining to temperature distribution in various constituents. A transient thermal analysis of the transportation cask was carried out to determine the temperature distribution under normal and accident conditions (800°C external fire).
The Status of the IAEA International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) and the Ongoing Activities of the Phase 1B of INPRO
Authors: Steur Ronald , Depisch Frank , Kupitz Juergen
Published: November 2008

The IAEA General Conference in 2000 has invited “all interested Member States to combine their efforts under the aegis of the Agency in considering the issues of the nuclear fuel cycle, in particular by examining innovative and proliferation-resistant nuclear technology”. In response to this invitation, the IAEA initiated an “International Project on Innovative Nuclear Reactors and Fuel Cycles” (INPRO). The overall objectives of INPRO are to help to ensure that nuclear energy is available to contribute in fulfilling in a sustainable manner energy needs in the 21st century, and to bring together all interested Member States, both technology holders and technology users, to consider jointly the international and national actions required to achieve desired innovations in nuclear reactors and fuel cycles that use sound and economically competitive technology. In the first phase of the project the report “Guidance for the evaluation of innovative nuclear reactors and fuel cycles” has been published. (June 2003, IAEA tecdoc 1362, Report of Phase 1A) In the following phase member states are contributing by case studies to validate the methodology for assessment and to evaluate the application of the basic principles, requirements and criteria. The paper will shortly summarize the main findings of the published report in the following fields (a) Prospects and Potentials of Nuclear Power, (b) Economics; (c) Sustainability and Environment, (d) Safety of Nuclear Installations, (e) Waste Management, (f) Proliferation Resistance, (g) Crosscutting issues and (h) the Methodology for Assessment. Further on the paper will deal with the actual phase of INPRO and the ongoing activities.