bluuebunny/crossref_metadata_2025 huggingface dataset:
Sustained Heating Experiments for the Investigation of the Boiling Heat Removal at the Upper Surface of Metallic Layer
Authors: Kang Kyoung-Ho , Park Rae-Joon , Kim Sang-Baik , Kim Hee-Dong
Published: September 2008

Sustained heating experiments are in progress to investigate the boiling configurations and to quantify the heat removal rate at the upper surface of a metallic layer in case of late in-vessel coolant injection. Pure Zinc is used as the simulant of metallic layer and bottom heating via catridge heater plate is adopted to simulate the heat transfer from the oxidic pool. The test section is rectangular shape having 0.3m in width and 0.5m in length. The scoping test was performed. In the scoping test, the height of melt pool is 10cm and the temperature of the test section bottom was intended to be controlled for maintaining the uniform temperature of 500°C. The scoping test results indicate that the melt pool effectively cooled down via boiling heat removal without energetic melt coolant interaction. The heat removal rate was calculated by 456.27kW/m2 ∼ 516.3kW/m2 which confirms the effective boiling heat removal in the scoping test.
Development of System Based Code: Case Study of Life-Cycle Margin Evaluation
Authors: Asayama Tai , Morishita Masaki , Tashimo Masanori
Published: September 2008

For a leap of progress in structural deign of nuclear plant components, The late Professor Emeritus Yasuhide Asada proposed the System Based Code. The key concepts of the System Based Code are; (1) life-cycle margin optimization, (2) expansion of technical options as well as combinations of technical options beyond the current codes and standards, and (3) designing to clearly defined target reliabilities. Those concepts are very new to most of the nuclear power plant designers who are naturally obliged to design to current codes and standards; the application of the concepts of the System Based Code to design will lead to entire change of practices that designers have long been accustomed to. On the other hand, experienced designers are supposed to have expertise that can support and accelerate the development of the System Based Code. Therefore, interfacing with experienced designers is of crucial importance for the development of the System Based Code. The authors conducted a survey on the acceptability of the System Based Code concept. The results were analyzed from the possibility of improving structural design both in terms of reliability and cost effectiveness by the introduction of the System Based Code concept. It was concluded that the System Based Code is beneficial for those purposes. Also described is the expertise elicited from the results of the survey that can be reflected to the development of the System Based Code.
Reaction, Transport and Settling Behavior of Lead-Bismuth Eutectic in Flowing Liquid Sodium
Authors: Miyahara Shinya , Ohno Shuji , Yamamoto Nobuhiro , Saito Jun-Ichi , Hirabayashi Masaru
Published: September 2008

The experimental study has been carried out to investigate reaction, transport and settling behavior of lead-bismuth eutectic (LBE) in flowing liquid sodium. In the test, 168g of LBE were poured into flowing sodium from the top of a vertical-type sodium loop which contained 23.2 kg of sodium. The initial temperature of LBE and sodium was 673K. The flow rate and the maximum velocity of sodium in the loop were controlled and measured at 20 dm3/min and 1 m/sec, respectively, using an electro-magnetic pump and an electro-magnetic flow meter. The sodium loop has a settling chamber at the lower part to investigate the concentration decrease behavior of solid particle reaction products in the sodium due to the settling effect. The concentration was measured by sodium sampling from the 11 positions of the loop during the experiment and its post-test chemical analysis. The temperature changes at the various parts of the loop were also measured during the experiment by thermo-couples attached on the outer surface of the loop. Ultrasonic detectors were attached on the outer surface of the loop below the position of a LBE pour nozzle to demonstrate the utility as a leak detector.
Characteristics of Melting for the Radioactive Aluminum Wastes From the Decommissioned Nuclear Facilities
Authors: Song Pyung Seob , Min Byung Youn , Choi Wang Kyu , Jung Chong Hun , Oh Won Zin
Published: September 2008

The characteristics of an aluminum melting and the radionuclide distribution have been investigated in a muffle furnace and a lab-scale arc furnace as part of the fundamental research for a melting of the metallic wastes generated by dismantling research reactors. The melting of aluminum was carried out with the use of surrogate nuclides such as cobalt, cesium, and strontium and fluxes such as NaCl-KCl-Na3AlF6 (Flux A), NaCl-NaF-KF (Flux B), CaF2 (Flux C), and LiF-KCl-BaCl2 (Flux D). The effects of the melting temperature and the type of flux on the melting of the aluminum and the nuclide distribution in the ingot, slag, and dust phase were investigated in a muffle furnace. The addition of the flux increased the fluidity of the aluminum melt, which has a slight difference according to the type of fluxes, and the amounts of the slag generated during the melting with the Flux types B and C were larger than those with flux types A and D. The results of the XRD (X-Ray Diffractometer) analysis showed that the surrogate nuclides move into the slag, which can be easily separated from the melt, and then they combine with the aluminum oxide to form a more stable compound. The distribution ratio of the cobalt in the ingot phase was less than 20% according to the type of fluxes. A removal efficiency of more than 99.5% for cesium and strontium from the ingot phase could be achieved. Similar results for the slag formation and the distribution of the surrogate nuclides were obtained in a direct current graphite arc melting system. Therefore, it is expected that a greater part of the aluminum wastes generated from the retired research reactors can be recycled or their volume reduced to be disposed of by a melting.
Numerical and Experimental Model Studies on Thermal Hydraulic Behavior of FBR Internal Core Catcher Assembly
Authors: Das Sanjay Kumar , Sharma Anil Kumar , Jasmin Sudha A. , Punitha G. , Lydia G. , Somayajulu P. A. , Murthy S. S. , Malarvizhi B. , Gopalakrishnan V. , Harvey J. , Kasinathan N. , Rajan M.
Published: September 2008

Core Catcher is provided as an in-vessel core debris retention device to collect, support, cool and maintain in sub-critical configuration, the generated core debris from fuel melting due to certain postulated Beyond Design Basis Events (BDBE) for Fast Breeder Reactor (FBR). This also acts as a barrier to prevent settling of debris on main vessel and keeps its maximum temperature within acceptable creep range. Heat transfer by natural convection in the core catcher assembly has been assessed numerically and through water experiments using geometrically similar configuration. Resistive heating elements are used in experiment as heat source to simulate debris decay heat on core catcher. Series of experiments were carried out for two configurations referred as geometry A and geometry B. The later configuration showed enhanced natural convective heat transfer from the lower plenum of the vessel. Temperatures were monitored at critical positions and compared with numerical evaluation. Numerically evaluated flow fields and isotherms are compared with experimental data for specific steady state temperatures on heat source plate. Numerical results are found to be in good agreement with that obtained from experiments. The combined efforts of numerical and experimental work conclude core catcher assembly with geometry B to be more suitable.
Assessment of HTGR Helium Compressor Analysis Tool Based on Newton-Raphson Numerical Application to Throughflow Analysis
Authors: Kim Ji Hwan , Kim Hyeun Min , No Hee Cheon
Published: September 2008

This study describes the development of a computer program for analyzing the off-design performance of axial flow helium compressors, which is one of the major concerns for the power conversion system of a high temperature gas-cooled reactor (HTGR). The compressor performance has been predicted by the aerodynamic analysis of meridional flow with allowances for losses. The governing equations have been derived from Euler turbomachine equation and the streamline curvature method, and then they have been merged into linearized equations based on the Newton-Raphson numerical method. The effect of viscosity is considered by empirical correlations to introduce entropy rises caused by primary loss sources. Use of the method has been illustrated by applying it to a 20-stage helium compressor of the GTHTR300 plant. As a result, the flow throughout the stages of the compressor has been predicted and the compressor characteristics have been also investigated according to the design specification. The program results show much better stability and good convergence with respect to other throughflow methods, and good agreement with the compressor performance map provided by JAEA.
Development of Technologies on Innovative-Simplified Nuclear Power Plant Using High-Efficiency Steam Injectors: Part 12—Evaluations of Spatial Distributions of Flow and Heat Transfer in Steam Injector
Authors: Abe Yutaka , Kawamoto Yujiro , Iwaki Chikako , Narabayashi Tadashi , Mori Michitsugu , Ohmori Shuichi
Published: September 2008

Next-generation nuclear reactor systems have been under development aiming at simplified system and improvement of safety and credibility. One of the innovative technologies is the supersonic steam injector, which has been investigated as one of the most important component of the next-generation nuclear reactor. The steam injector has functions of a passive pump without large motor or turbo-machinery and a high efficiency heat exchanger. The performances of the supersonic steam injector as a pump and a heat exchanger are dependent on direct contact condensation phenomena between a supersonic steam and a sub-cooled water jet. In previous studies of the steam injector, there are studies about the operating characteristics of steam injector and about the direct contact condensation between static water pool and steam in atmosphere. However, there is a little study about the turbulent heat transfer and flow behavior under the great shear stress. In order to examine the heat transfer and flow behavior in supersonic steam injector, it is necessary to measure the spatial temperature distribution and velocity in detail. The present study, visible transparent supersonic steam injector is used to obtain the axial pressure distributions in the supersonic steam injector, as well as high speed visual observation of water jet and steam interface. The experiments are conducted with and without non-condensable gas. The experimental results of the interfacial flow behavior between steam and water jet are obtained. It is experimentally clarified that an entrainment exists on the water jet surface. It is also clarified that discharge pressure is depended on the steam supply pressure, the inlet water flow rate, the throat diameter and non-condensable flow rate. Finally a heat flux is estimated about 19MW/m2 without non-condensable gas condition in steam.
Uncertainty Quantification of NEPTUN Analysis Using Wilks’ Formula With an Extended Number of Calculations
Authors: Hwang Moonkyu , Lee Young-Jin , Chung Bub-Dong
Published: September 2008

The two-phase system analysis code MARS [1] has been used for the uncertainty quantification of NEPTUN reflood test [2] analysis. By performing 10,000 calculations based on a random variation of the MARS model parameters and measured data, a mean value, and the 95% upper bounds are traced along the number of calculations. The CPU-intensive calculations were performed using the 11 node PC-cluster under Linux platform. The Monte-Carlo calculation results suggest a total number of 2,000 calculations would be sufficient to determine the stable mean and 95% upper bound values. The peak temperatures predictions are also used to find the 95% bounding values by using the Wilks’ method. For the 1st order one-sided formula, every 59 peak temperatures are examined to locate the bounding temperature, with a 95% confidence. The 2nd and 3rd order values were found in a similar way. The uncertainty band by the Wilks’ formula, when compared with the true 95% bounding value, is observed to be quite broad, especially in the case of the 1st order. The 2nd or 3rd orders or a full Monte-Carlo method would be necessary to demonstrate that the safety of the plant is ensured with a sufficient margin. A supplementary sensitivity study, for the nine uncertain parameters selected for the NEPTUN analysis, is also performed to find the degree of influence of each parameter on the peak rod temperature.
Study Plan for Material Corrosion Test in Lead and Bismuth Eutectic at High Temperature
Authors: Nakazima Toru , Rivai Abu Khalid , Hata Koji , Dostal Vaclav , Takahashi Minoru
Published: September 2008

A concept of steam lift pump type lead-bismuth cooled fast reactor (SLPLFR) is proposed as high temperature and high efficiency Pb-Bi cooled fast reactors. Fe-Al alloy-surface treated steels and ceramics of SiC, Si3N4 and SiC/SiC composites have been chosen as candidates of cladding and structural materials, respectively. A corrosion test plan is proposed, where compatibility of steels applied with Al-Fe alloy-surface treatment and the ceramics will be tested in high temperature Pb-Bi flow. A conceptual design of a test apparatus for the corrosion test is provided.
RELAP5/MOD3 Analysis of Transient Steam-Generator Behavior During Turbine Trip Test of a Prototype Fast Breeder Reactor MONJU
Authors: Shindo Yoshihisa , Endo Hiroshi , Ishizu Tomoko , Haga Kazuo
Published: September 2008

In order to develop a thermal-hydraulic model of the steam-generator (SG) to simulate transient phenomena in the sodium cooled fast breeder reactor (FBR) MONJU, Japan Nuclear Energy Safety Organization (JNES) verified the SG model using the RELAP5/MOD3 code against the results of the turbine trip test at a 40% power load of MONJU. The modeling by using RELAP5 was considered to explain the significant observed behaviors of the pressure and the temperature of the EV steam outlet, and the temperature of water supply distributing piping till 600 seconds after the turbine trip. The analysis results of these behaviors showed good agreement with the test results based on results of parameter study as the blow efficiency (release coef.) and heat transferred from the helical coil region to the down-comer (temperature heating down-comer tubes). It was found that the RELAP5/MOD3 code with a two-fluids model can predict well the physical situation: the gas-phase of steam generated by the decompression boiling moves upward in the down-comer tubes accompanied by the enthalpy increase of the water supply chambers; and that the pressure change of a “shoulder” like shape is induced by the mass balance between the steam mass generated in the down-comer tubes and the steam mass blown from the SG. The applicability of RELAP5/MOD3 to SG modeling was confirmed by simulating the actual FBR system.
Three-Dimensional CFD Analysis on Gas Flow in Corrugated Wall Channel
Authors: Tak Nam-il , Lee Won-Jae , Jang Jonghwa
Published: September 2008

A printed circuit heat exchanger (PCHE) is known as one of the promising types for an intermediate heat exchanger (IHX) of a nuclear hydrogen production system. This paper presents fundamental numerical results on gas flow behaviors in a typical PCHE geometry. Laminar and turbulent flows were analyzed based on a computational fluid dynamics (CFD) analysis. Local friction coefficient and local Nusselt number were evaluated and compared with those by typical correlations for tubes. In the case of a turbulent flow, various turbulence models were applied. The results clearly show the significance of a careful selection of a turbulence model.
Simultaneous Measurement of Temperature and Velocity in Turbulent Buoyant Plume by Combined LIF and PIV Technique
Authors: Watanabe Yoshie , Hashizume Yuji , Fujisawa Nobuyuki
Published: September 2008

An experimental technique for simultaneous measurement of temperature and velocity in a thermal flow is described. This technique is based on the two-color laser-induced fluorescence technique combined with the particle image velocimetry. Illumination is provided from Nd:YAG laser and the fluorescent dyes are chosen as Rhodamine B and Fluorescent Sodium, which combination allows the accurate velocity measurement in a wide range of flow velocity and high temperature sensitivity in temperature measurement. The measurement of temperature and velocity in turbulent buoyant plume is carried out by this method, and the structure of the plume is studied in connection with the entrainment of surrounding fluid at the interface.
Vitrification Technology Development Plan in Tokai Reprocessing Plant
Authors: Aoshima Atsushi , Tanaka Kazuhiko
Published: September 2008

The Tokai Vitrification Facility (TVF) is the only operating vitrification plant in Japan, constructed and operated by JAEA, to vitrify concentrated high radioactive liquid waste (HALW) in the Tokai Reprocessing Plant (TRP). JAEA started TVF hot operation in 1995 and produced 218 canisters as of March, 2006. An existing melter is the second melter, which was installed from 2002 to 2004 in place of the first melter stopped its operation by damage of a main electrode. JAEA has estimated that the damage was caused by accumulation of noble metal. Therefore, melter bottom structure was improved to get better drain ability of glass containing noble metal. Completing the melter replacement, vitrification operation was restarted in October 2004 and produced 88 canisters successfully until the end of March 2006. Through these experiences, JAEA made basic strategy to achieve stable TVF operation: keeping stable operation of the existing melter preventing adverse effect by noble metal accumulation and developing a new advanced melter with long lifetime preparing for future exchange as the third melter. Based on the basic strategy, JAEA made a decade development plan of necessary key technologies and has started the development since 2005.
Experiment and Numerical Simulation of Bubble Behaviors in Argon Gas Injection Into Lead-Bismuth Pool
Authors: Yamada Yumi , Akashi Toyou , Takahashi Minoru
Published: September 2008

In a lead-bismuth alloy (45%Pb-55%Bi) cooled direct contact boiling water fast reactor (PBWFR), steam can be produced by direct contact of feed water with primary Pb-Bi coolant in the upper core plenum, and Pb-Bi coolant can be circulated by buoyancy forces of steam bubbles. As a basic study to investigate the two-phase flow characteristics in the chimneys of PBWFR, a two-dimensional two-phase flow was simulated by injecting argon gas into Pb-Bi pool in a rectangular vessel (400mm in length, 1500mm in height, 50mm in width), and bubble behaviors were investigated experimentally. Bubble sizes, bubble rising velocities and void fractions were measured using void probes. Argon gas was injected through five nozzles of 4mm in diameter into Pb-Bi at two locations. The experimental conditions are the pressure of atmospheric pressure, Pb-Bi temperatures of 443K, and the flow rate of injection Ar gas is 10, 20, and 30 NL/min. The measured bubble rising velocities were distributed in the range from 1 to 3 m/s. The average velocity was about 0.6 m/s. The measured bubble chord lengths were distributed from 1mm up to 30mm. The average chord length was about 7mm. An analysis was performed by two-dimensional and two-fluid model. The experimental results were compared with the analytical results to evaluate the validity of the analytical model. Although large diameter bubbles were observed in the experiment, the drag force model for spherical bubbles performed better for simulation of the experimental result because of high surface tension force of Pb-Bi.
Electroslag Strip Cladding of Steam Generators With Alloy 690
Authors: Consonni M. , Maggioni F. , Brioschi F.
Published: September 2008

The present paper details the results of electroslag cladding and tube-to-tubesheet welding qualification tests conducted by Ansaldo-Camozzi ESC with Alloy 690 (Alloy 52 filler metal) on steel for nuclear power stations’ steam generators shell, tubesheet and head; the possibility of submerged arc cladding on first layer was also considered. Test results, in terms of chemical analysis, mechanical properties and microstructure are reproducible and confidently applicable to production cladding and show that electroslag process can be used for Alloy 52 cladding with exceptionally stable and regular operation and high productivity. The application of submerged arc cladding process to the first layer leads to a higher base metal dilution, which should be avoided. Moreover, though the heat affected zone is deeper with electroslag cladding, in both cases no coarsened grain zone is found due to recrystallisation effect of second cladding layer. Finally, the application of electroslag process to cladding of Alloy 52 with modified chemical composition, was proved to be highly beneficial as it strongly reduces hot cracking sensitivity, which is typical of submerged arc cladded Alloy 52, both during tube-to-tubesheet welding and first re-welding.
Analytical Study of Pb-Bi-Water Direct Contact Boiling Two-Phase Flow
Authors: Novitrian  , Takahashi Minoru
Published: September 2008

Experimental studies of Pb-Bi-Water direct contact boiling two-phase flow have been conducted (in the chimney 29.9mm in diameter and 2,367mm in length). The rated operating conditions of system pressure was 7MPa, the liquid metal temperature was 460°C and injected water temperature was 220°C. Pb-Bi flow rates were measured and analytically estimated using balance condition between buoyancy force and pressure losses. The drift-flux model was used to determine the void fraction with assumption that bubbles sizes were dependent on the fluid surface tension and the density ratio of Pb-Bi to steam and water mixture. The result of analysis shows that the deviation from the experimental data was less than 20%.
Two-Dimensional Optical Measurement of Waves on Liquid Lithium Jet Simulating IFMIF Target Flow
Authors: Itoh Kazuhiro , Itoh Taro , Kukita Yutaka , Koterazawa Hiroyuki , Kondo Hiroo , Yamaoka Nobuo , Horiike Hiroshi , Ida Mizuho , Nakamura Hideo , Nakamura Hiroo , Muroga Takeo
Published: September 2008

Waves on a liquid-lithium jet flow, simulating a proposed high-energy beam target design, have been measured using an optical technique based on specular reflection of a single laser beam on the jet surface. The streamwise and spanwise fluctuations of the local free-surface slope were least-square fitted with a sinusoidal curve to makeup the signals lost due to the constriction in the optical arrangement. The waveform was estimated with an assumption that wave phase speed can be calculated using the dispersion relation for linear capillary-gravity waves. The direction of propagation on the jet surface was also evaluated so that the wave amplitudes, calculated by integral of slope angle signal, agree consistently in streamwise and spanwise direction. These measurements and analyses show that the waves at the measurement location for a jet velocity of 1.2 m/s can best be represented by oblique waves with an inclination of 1.23 rad, a wavelength of 3.8 mm and a wave amplitude of about 0.05 mm.
The Development of the Evolutionary BWR
Authors: Murase A. , Nakamaru M. , Kuroki M. , Kojima Y. , Yokoyama S.
Published: September 2008

Considering the delay of the fast breeding reactor (FBR) development, it is expected that the light water reactor will still play the main role of the electric power generation in the 2030’s. Accordingly, Toshiba has been developing a new conceptual ABWR as the near-term BWR. We tentatively call it AB1600. The AB1600 has introduced the hybrid active/passive safety system in order to improve countermeasure against severe accident (SA). At the same time, we have made the simplification of the overall plant systems in order to improve economy. The simplification of the AB1600 is based on the proven technologies. To retain the safety performance superior or equivalent to the current ABWR and to strengthen the countermeasure against SA, the AB1600 has introduced the passive systems such as the passive containment cooling system (PCCS), the gravity driven core cooling system (GDCS) and the isolation condenser (IC). While we retain the safety performance superior or equivalent to the current ABWR, we have made the simplification of the safety systems. We could eliminate the high pressure core flooder system (HPCF) and the reactor core isolation system (RCIC) by extending the height of reactor pressure vessel (RPV) two meters. To achieve simplification of reactor systems, we have reduced the number of fuel bundles and the number of control rods by adopting large bundle that has a bundle pitch 1.2 times wider than that of the current ABWR. In the 1600MWe class, the number of fuel bundles could be reduced to 600 from 872 of the current ABWR, and the number of control rods could be reduced to 137 from 205 of the current ABWR. Because the reactor internal pump (RIP) of the current ABWR has sufficient performance capacity and the improvement of fuel characteristics from the current fuel enables the operation at lower core flow, the number of RIPs could be decreased from ten to eight. Furthermore, we have reduced the number of divisions of emergency core cooling system (ECCS)/heat removal system to two from three of the current ABWR. This configuration change contributes to reduce the amount of resources of not only reactor systems but also auxiliary systems. In the previous paper, the AB1600 had four low pressure flooder systems (LPFLs). We have studied about the possibility of reduction of LPFLs to two from four by providing the LPFL with alternative injection lines. This change is expected to contribute to reduce the total number of ECCS pumps and the capacity of emergency AC power.
Development of the Analysis Method for the Radioactivity Concentration Estimation of Soils for a Regulatory Clearance
Authors: Hong Dae-Seok , Shon Jong-Sik , Kim Tae-Kuk , Cho Han-Seok
Published: September 2008

In this study, for the analysis of a radioactivity concentration of soil, a sampling and an analyzing method were developed. To create homogeneity of the contents for each drum, big particles such as pebbles, rocks and scraps of concrete were removed after pouring the soil into a tray. And then the soil in the tray was mixed thoroughly. A 10 × 10 grid was used to partition the soil into 100 sections. 2 liter of soil was sampled out of 30 randomly pre-selected sections. Only 1 liter of the soil out of 2 liters was used for the γ-spectrometry analysis. The remaining 1 liter of soil was stored for a validation purpose. For a verification of the sampling process, 3 samples were taken from each drum and analyzed. 5 drums were used for the verification. And the results show that this sampling method has about a 9% sampling error. Also, the analysis results of the 865 drums of soil showed that the major nuclides in the soils were Co-60 and Cs-137 while a small amount of Mn-54, Fe-59, I-131, Cs-134 and Eu-152 were detected as γ-emitters. About 73% of the soils had a total radioactivity concentration below 0.1 Bq/g, while the soils with more than 0.4 Bq/g of a radioactivity concentration were only 3%. Based on the guide for the regulatory clearance criteria recommended by the IAEA (International Atomic Energy Agency), about 73% of the soils can be regulatory cleared without any treatment. Also, the remaining soils can be regulatory cleared after a further storage. Only 3% of the soils are considered to be necessary for a decontamination treatment. The results of this study can be applied to the treatment of radioactive soils generated in a large amount during the decommissioning of a nuclear facility.
New Design of a Digital CRCS by Introducing LOM
Authors: Nam Jung-Han , Nam Chae-Ho , Lee Jong-Moo , Jung Hae-Won , Yook Sim-Kyun , Cho Chang-Ho
Published: September 2008

A Control Rod Control System (CRCS) is one of the most important pieces of equipment in a nuclear power plant because it controls the nuclear reaction by moving the Control Rod Drive Mechanism (CRDM) in the reactor with speed and direction signals from the Reactor Regulating System (RRS). This paper introduces a CRCS with full-duplex configuration and a Local Operator Module (LOM) computer to enhance reliability in comparison to existing simplex systems. The duplex configuration, LOM program, and maintenance test program of the CRCS are explained in detail. This duplex system can perform a failsafe changeover in only a few milliseconds, which results in ‘bumpless’ CRDM coil current control. In particular, because all the control cards are hot-swappable and power converter modules are drawer type, the operator can replace a malfunctioning module without a system shutdown. For ease of operation, the CRCS has its own panel computer called Local Operator Module (LOM) which has a monitoring program installed. The LOM shows all the CRCS operating information such as urgent/non-urgent alarm status, controller operation status, CRDM coil voltage/current, data logs and detailed event alarms. In the CRCS power cabinet, there is an extra panel for easy and simple module maintenance. If the operator suspects that any module of a power cabinet such as a power control card or a power converter module is abnormal, he has only to place the module in the maintenance panel and run the test program to ascertain the module’s sanity. The operator also can utilize the maintenance panel for spare module testing before the spare module is installed in the system and thus avoid installing a defective module.
Numerical Experiments of Coolant Mixing in a Lower Plenum of PWR Under Asymmetric Thermal- Hydraulics Conditions
Authors: Ohtani Masanori , Kozuru Akito , Kashimoto Yasuyuki , Montani Mitsuto , Takeda Koutaro , Makino Yasushi
Published: September 2008

Asymmetric thermal-hydraulic conditions among primary loops during a postulated steam line break (SLB) induce a non-uniform temperature distribution at a core inlet. When coolant of lower temperature intrudes into a part of core, it leads to a reactivity insertion and a local power increase. Therefore, an appropriate model for the core inlet temperature distribution is required for a realistic SLB analysis. In this study, numerical experiments were conducted to examine the core inlet temperature distribution under the asymmetric thermal-hydraulic coolant conditions among primary loops. 3D steady-state calculations were carried out for Japanese standard Pressurized Water Reactor (PWR) such as 2, 3, 4 loop types and an advanced PWR. Since the flow in a reactor vessel involves time-dependent velocity fluctuations due to a high Reynolds number condition and a complicated geometry of flow path, the turbulent mixing might be enhanced. Hence, the turbulent thermal diffusivity for the steady-state calculation was examined based on experimental results and another transient calculation. As a result, it was confirmed that (1) the turbulent mixing in a downcomer and a lower plenum were enhanced due to time-dependent velocity fluctuations and therefore the turbulent thermal diffusivity for steady-state calculation was specified to be greater, (2) the core inlet temperature distribution predicted by a steady-state calculation reasonably agreed with a experimental data, (3) the patterns of core inlet temperature distribution were comprehended to be dependent on the plant type, i.e. the number of primary loop and (4) under a low flow rate condition, the coolant of lower temperature appeared on the opposite side of the affected loop due to the effect of a natural convection.
Elasto-Plastic Analysis of a Miniature Circular Disk Bending Specimen
Authors: Verma Vishnu , Ghosh A. K. , Behera G. , Sharma Kamal , Singh R. K.
Published: September 2008

Miniature disk bending test is used to evaluate the mechanical behavior of irradiated materials and its properties — mainly ductility loss due to irradiation in steel. In Miniature Disk Bending Machine the specimen is firmly held between the two horizontal jaws of punch, and an indentor with spherical ball travels vertically. Researchers have observed reasonable correlations between values of the yield stress, strain hardening and ultimate tensile strength estimated from this test and mechanical properties determined from the uniaxial tensile test. Some methods for the analysis of miniature disk bending, proposed by various authors have been discussed in the paper. It is difficult to distinguish between the regimes of elastic and plastic deformation since local plastic deformation occurs for very small values of load when the magnitude of spatially averaged stress will be well below the yield stress. Also, the analytical solution for large amplitude, plastic deformation becomes rather unwieldy. Hence a finite element analysis has been carried out. The finite element model, considers contact between the indentor and test specimen, friction between various pairs of surfaces and elastic plastic behavior. The load is increased in steps and converged solution has been obtained and analysis terminated at a load beyond which a stable solution cannot be obtained. A sensitivity study has been carried out by varying the various parameters defining the material properties by ±10% around the base values. This study has been carried out to generate a data base for the load-deflection characteristics of similar materials from which the material properties can be evaluated by an inverse calculation. It is seen that the deflection obtained by analytical elastic bending theory is significantly lower than that obtained by the elasto-plastic finite element solution at relatively small values of load. The FE solution and experimental results are in reasonably good agreement.
Recent Progress in Risk Consideration and Issues on Developing Risk Informed Regulations in Japan
Authors: Hidaka Akihide
Published: September 2008

The Nuclear Safety Commission (NSC), Japan set up the taskforce on introduction of risk informed regulation (RIR) into nuclear safety regulations in April 2004. Since then the taskforce (Chairperson: Prof. Genki Yagawa, Toyo university) has reviewed the status of risk considerations at related organizations and discussed the issues for developing RIR in Japan. Recently, the taskforce prepared the interim report on the review results and discussions, and NSC approved it in December 2005. The report described that the risk consideration in related organizations in Japan has made progress in line mostly with the NSC’s basic policy for introduction of RIR expressed in 2003. However, the following topics were identified as important issues for further promotion of RIR introduction: policy for utilization of risk information considering Japanese features, usage of safety goals and performance objectives in RIR, decision-making process using risk information, pilot program, PSA quality, improvement of safety examination guidelines considering risk information, utilization of risk information in nuclear fuel cycle facilities and risk communication.
Contaminated Metal Components in Dismantling by Hot Cutting Processes
Authors: Cesari Franco G. , Rogante Massimo , Giostri Angelo , Conforti Gianmario
Published: September 2008

During the preparatory dismantling activities of Caorso’s Nuclear Power Plant (NPP), an experimental campaign using plasma and oxyacetylene metal cutting processes has been performed and applied to plates and tubes exposed to the coolant steam of the reactor. The plant (Boiling Water Reactor, 870 MWe) was designed and built in the 70s, and it was fully operating by 1981 to 1986 being shut down after 1987 Italy’s poll that abrogated nuclear power based on U235 fission. The campaign concerns no activated materials, even if the analyses have been performed of by use contaminated components under the free release level, not yet taking into account radioactivity. In this paper, the parameters related to inhalable aerosol, solid and volatile residuals production have been, studied during hot processes which applies the same characteristics of the cutting in field for the dismantling programs of Caorso NPP. The technical parameters such as cutting time and cutting rate vs. pipe diameter/thickness/schedule or plate thickness for ferritic alloys and the emissions composition coming from the sectioning are also reported. The results underline the sort of trouble that can emerge in the cutting processes, in particular focusing on the effects comparison between the two cutting processes and the chemical composition of powders captured by filtering the gaseous emission. Some preliminary considerations on methodology to be used during the dismantling have been presented.
Study on Applicability of Numerical Simulation to Evaluation of Gas Entrainment From Free Surface
Authors: Ito Kei , Sakai Takaaki , Ohshima Hiroyuki
Published: September 2008

An onset condition of gas entrainment (GE) due to free surface vortex has been studied to establish a design of fast breeder reactor with higher coolant velocity than conventional designs, because the GE might cause the reactor operation instability and therefore should be avoided. The onset condition of the GE has been investigated experimentally and theoretically, however, dependency of the vortex type GE on local geometry configuration of each experimental system and local velocity distribution has prevented researchers from formulating the universal onset condition of the vortex type GE. A real scale test is considered as an accurate method to evaluate the occurrence of the vortex type GE, but the real scale test is generally expensive and not useful in the design study of large and complicated FBR systems, because frequent displacement of inner equipments accompanied by the design change is difficult in the real scale test. Numerical simulation seems to be promising method as an alternative to the real scale test. In this research, to evaluate the applicability of the numerical simulation to the design work, numerical simulations were conducted on the basic experimental system of the vortex type GE. This basic experiment consisted of rectangular flow channel and two important equipments for vortex type GE in the channel, i.e. vortex generation and suction equipments. Generated vortex grew rapidly interacting with the suction flow and the grown vortex formed a free surface dent (gas core). When the tip of the gas core or the bubbles detached from the tip of the gas core reached the suction mouth, the gas was entrained to the suction tube. The results of numerical simulation under the experimental conditions were compared to the experiment in terms of velocity distributions and free surface shape. As a result, the numerical simulation showed qualitatively good agreement with experimental data. The numerical simulation results were similar to the experimental results in terms of the shape of free surface dent and the velocity distribution around the vortex, although the GE itself was not completely reproduced due to a lack of enough mesh partition. After confirming the applicability of the numerical simulation to the GE evaluation, several parameters, such as suction velocity and a configuration around suction mouth, were numerically examined to evaluate their influence on the GE. The tendencies of the GE occurrence enhanced by larger suction velocity or suction mouth on bottom surface enhance occurrence of the GE were obtained from the simulation results. These simulation results implied that the numerical simulation has enough potential to be used for the design work.
The Development of Severe Accident Codes at IRSN and Their Application to Support the Safety Assessment of EPR
Authors: Caroli Cataldo , Bleyer Alexandre , Bentaib Ahmed , Chatelard Patrick , Cranga Michel , Van Dorsselaere Jean-Pierre
Published: September 2008

IRSN uses a two-tier approach for development of codes analysing the course of a hypothetical severe accident (SA) in a Pressurized Water Reactor (PWR): on one hand, the integral code ASTEC, jointly developed by IRSN and GRS, for fast-running and complete analysis of a sequence; on the other hand, detailed codes for best-estimate analysis of some phenomena such as ICARE/CATHARE, MC3D (for steam explosion), CROCO and TONUS. They have been extensively used to support the level 2 Probabilistic Safety Assessment of the 900 MWe PWR and, in general, for the safety analysis of the French PWR. In particular the codes ICARE/CATHARE, CROCO, MEDICIS (module of ASTEC) and TONUS are used to support the safety assessment of the European Pressurized Reactor (EPR). The ICARE/CATHARE code system has been developed for the detailed evaluation of SA consequences in a PWR primary system. It is composed of the coupling of the core degradation IRSN code ICARE2 and of the thermalhydraulics French code CATHARE2. The CFD code CROCO describes the corium flow in the spreading compartment. Heat transfer to the surrounding atmosphere and to the basemat, leading to the possible formation of an upper and lower crust, basemat ablation and gas sparging through the flow are modelled. CROCO has been validated against a wide experimental basis, including the CORINE, KATS and VULCANO programs. MEDICIS simulates MCCI (Molten-Corium-Concrete-Interaction) using a lumped-parameter approach. Its models are being continuously improved through the interpretation of most MCCI experiments (OECD-CCI, ACE…). The TONUS code has been developed by IRSN in collaboration with CEA for the analysis of the hydrogen risk (both distribution and combustion) in the reactor containment. The analyses carried out to support the EPR safety assessment are based on a CFD formulation. At this purpose a low-Mach number multi-component Navier-Stokes solver is used to analyse the hydrogen distribution. Presence of air, steam and hydrogen is considered as well as turbulence, condensation and heat transfer in the containment walls. Passive autocatalytic recombiners are also modelled. Hydrogen combustion is afterwards analysed solving the compressible Euler equations coupled with combustion models. Examples of on-going applications of these codes to the EPR safety analysis are presented to illustrate their potentialities.
Cyclic Hardening Behaviors and Reduction in Fatigue Life of Type 316LN Austenitic Stainless Steel in 310°C Low Oxygen-Containing Water
Authors: Cho Hyunchul , Kim Byoung Koo , Jang Changheui , Kim In Sup , Hong Seung Mo
Published: September 2008

Low cycle fatigue tests were conducted to investigate the cyclic behavior and the fatigue life of type 316LN stainless steel (SS) at various strain rates in 310°C low oxygen-containing water. The strain rates were 0.008, 0.04, and 0.4%/s, and the applied strain amplitude was varied from 0.4 to 1.0%. The dissolved oxygen concentration of the test water was maintained below 1 ppb. The test material in 310°C low oxygen-containing water experienced a primary hardening, followed by a softening. From our data, we confirm the occurrence of the dynamic strain aging (DSA), and finally it can be considered that the primary hardening was brought about by the DSA. The secondary hardening was observed distinctly for 0.4%/s and 0.4%. The improvement of fatigue resistance and the secondary hardening occurred under the same loading condition. Therefore, the improvement of fatigue resistance may be related to the occurrence of the secondary hardening. When the secondary hardening occurs, intense slip bands are replaced by the corduroy structure. The corduroy structure can induce retardation of crack initiation, and ultimately the fatigue resistance is improved. Comparative study between the fatigue life generated in the current study and some prediction models was performed to evaluate the reliability of our data.
Containment Thermal-Hydraulic Simulations With an LP-CFD Approach: Qualification Matrix of the TONUS Code
Authors: Bentaib A. , Bleyer A. , Malet J. , Caroli C. , Vendel J. , Kudriakov S. , Dabbene F. , Studer E. , Beccantini A. , Magnaud J. P. , Paille`re H.
Published: September 2008

The French Atomic Energy Commission (CEA) and the Institute for Radiological Protection and Nuclear Safety (IRSN) are developing a hydrogen risk analysis code (safety code) which incorporates both lumped parameter (LP) and computational fluid dynamics (CFD) formulations. In this paper we present briefly the main physical models for containment thermal-hydraulics. Validation and typical numerical results will be presented for hydrogen distribution and combustion applications in small and realistic large geometries.
Analysis of Swage Autofrettage in Metal Tube
Authors: Malik Muhammad Afzaal , Khan Muddasar , Rashid Badar , Khushnood Shahab
Published: September 2008

Autofrettage (self-hooping) is used to induce advantageous residual stresses into pressure vessels to enhance their fatigue lifetime. The process is achieved by increasing elastic strength of a cylinder with various methods such as hydraulic pressurization, mechanical swaging, or by utilizing the pressure of a powder gas. This research work deals with the swage or mechanical autofrettage of metal tubes. The objective is to attain a bore size of 125mm. Normally such a bore size is achieved with hydraulic autofrettage. However, we have used two-stage mechanical autofrettage to achieve the desired bore size. At first stage the swage diameter achieved is 118mm, and in the second stage, the diameter achieved after machining is 125mm. The temperature variation for swage is 38°C to 50°C. The applied pressure varies from 85 to 180 bars inside the tube. The process was applied to a number of tubes selected randomly. The swage autofrettage process was also analyzed using numerical simulation based on finite element method. The results of numerical simulation are compared with design parameters.
Update to Risk-Informed Pressurized Water Reactor Vessel 10 to 20 Year Inspection Interval Extension
Authors: Palm Nathan A. , Bishop Bruce A. , Boggess Cheryl L.
Published: September 2008

The Pressurized Water Reactor Owners Group (formerly the Westinghouse Owners Group (WOG)) methodology for extending the inservice inspection interval for welds in pressurized water reactor (PWR) reactor pressure vessel (RPV) was introduced as ICONE12-49429. The paper presented a risk informed basis for extending the interval between inspections from the current interval of 10 years to 20 years. In the paper presented at ICONE-12, results of pilot studies on typical Westinghouse and Combustion Engineering Nuclear Steam Supply System (NSSS) designs of PWR vessels showed that the change in risk associated with the proposed inspection interval extension was within the guidelines specified in the United States Nuclear Regulatory Commission (NRC) Regulatory Guide 1.174 for an acceptably small change in risk. Since the methodology was originally presented, the evaluation has been updated to incorporate the latest changes in the NRC Pressurized Thermal Shock (PTS) Risk Reevaluation Program and expanded to include the Babcock and Wilcox NSSS RPV design. The results of these evaluations demonstrate that the proposed RPV inspection interval extension remains a viable option for the industry. The updates to the methodology and input, pilot plant evaluations, results, process for demonstrating applicability of the pilot plant analysis to non-pilot lead plants and lessons learned from the evaluations performed are summarized in this paper.