bluuebunny/crossref_metadata_2025 huggingface dataset:
A Feasibility Study on Reducing EPZ for APR-1400 in Societal Implications
Authors: Lee Young Wook , Moon Joo Hyun , Kim Suk Hoon , Kang Chang Sun
Published: November 2008

For APR-1400 in Korea, based upon a large enhancement of safety, the reduction of EPZ radius and the concept of eventual elimination of EPZ have been being seriously considered. Technical justification for the reduction of EPZ had been considered in the previous study. The reduction of an EPZ, however, must not only be justified from a technical perspective, but include recognition of public policy and societal implications. So, in this study, a on-spot survey of the public acceptances on the workers and the nearby public was conducted. The survey was performed for June 25 to July 13 in 2001, in which the respondents were requested to answer the questions in the questionnaire that are related to those for radiological emergency exercise, emergency measures, regulatory status for EPZ, criteria for determining the distance of EPZ, ways to reduce the distance of EPZ for APR-1400, and method to collect the public opinions.
Wave-Induced Streaming in Vertically-Excited, Stratified Fluids
Authors: Nakayama Hideyuki , Mori Haruhiko , Shimizu Jun , Ito Takahiro , Nakamura Hideo , Tsuji Yoshiyuki , Kukita Yutaka
Published: November 2008

Non-zero time averaged flows can be induced in fluctuating flows. Shimizu et al. found such time-averaged streaming to occur in stratified fluids excited in a vertical, stationary cylinder at a frequency close to the natural frequency of the axisymmetric interface wave [1]. PIV measurements have revealed intermittent formation of a stagnation region close the fluid-fluid interface that enables a jet-like flow departing from the center of the interface to be sustained over cycles. The jet-like flow induces a recirculating streaming flow that extends several diameters away from the interface. Difficulty still remains in quantifying experimentally the detailed flow structure and defining the driving mechanisms, since the streaming is more than one order of magnitude smaller than the forced oscillating flow. The present paper presents PIV data that show temporal evolution of the flow in each cycle, the effect of fluid viscosities and the excitation frequency and amplitude.
Possibility of Using Gamma Radiation From HTR Reactors for Processing of Food and Medical Products
Authors: Pahladsingh Remond R.
Published: November 2008

During the fission process in most of the presently operating nuclear reactors nuclear energy is converted into thermal energy and transferred to common steamcycles for powergeneration. As part of the fission process also α-, β- and neutrons particles are released from the nucleus; the release of gamma-rays is also a part of the fission process. In present nuclear reactors α-, β-, neutrons particles and particularly Gamma-rays are not gainfully used as a result of the reactor design and of the containment. These plants are built as required by regulations and international standards for safety. The inherently safe HTR reactor, by its physics and design, does not need a special reinforced containment and it is worth looking into the possibilities of this design feature to use the by-products, such as Gamma-rays, from nuclear fission. In the HTR Pebble Bed Reactors the α-, and β-particles will remain in the kernels in the pebbles. This means that only the neutron particles and gamma-rays will be available outside the reactor pressure vessel. In this report a proposal is presented to use the gamma-rays of the HTR reactor for irradiation of food and agricultural produce. For neutron shielding a reflector is placed inside the reactor while outside the reactor neutron- and thermal-shielding will be accomplished with water. The high energy gamma-rays will pass through the water-shield and could be harnessed for radiation processing of food and medical products.
Basic Design and Safety Evaluation of Metal Cask Storage Facility With New Transportation System
Authors: Satoh Takashi , Ijichi Hiroshi , Yamane Tadashi , Mannen Takeshi , Busujima Kouji , Shimizu Masashi
Published: November 2008

Transportation system utilizing fluid film technology is widely used in various industries. In the U.S. nuclear industry, it is used for transporting concrete casks. Tokyo Electric Power Company (TEPCO) is planning to adopt this design concept to transporting metal cask inside a spent fuel interim storage facility (SFISF). We call this the Metal Cask Transportation System (MCTS). By adopting this system, we can lower the ceiling height of the building, and thus reduce the amount of building materials. The MCTS has not been licensed in Japanese nuclear facility, so that we carried out operational and seismic tests with 1/3-scale model system to verify applicability of this system. As the conclusions, the MCTS has sufficient stability to maintain the safety functions of metal cask. Furthermore, we designed a SFISF introducing the MCTS and confirmed that heat from casks is removed appropriately by natural convection and the dose to the general public is enough low against the statutory dose limit.
Experimental Investigation of Mixed Convection Heat Transfer Caused by Forced-Jets in Large Enclosure
Authors: Niu Fenglei , Peterson Per F.
Published: November 2008

This research investigates experimentally mixed convection and heat transfer augmentation by forced jets in a large enclosure, at conditions simulating those of actual passive containment cooling systems and scales approaching those of actual containment buildings or compartments. The experiment was designed to measure the key parameters governing the heat transfer augmentation by forced jets and investigate the effects of geometric factors, including the jet diameter, jet injection orientation, interior structures, and enclosure aspect ratio. The tests cover a variety of injection modes leading to flow configurations of interest that contribute to reveal the nature of mixing and stratification phenomena in the containment under accident conditions of interest. By nondimensionalizing the governing equations, the heat transfer of mixed convection can be predicted to be controlled by jet Archimedes number and geometric factors. Using a combining rule for mixed convection and appropriate forced and natural convection models, the correlations of heat transfer augmentation by forced jets are developed and then tested by experimental data. The effects of jet diameter, injection orientation, interior structures, and enclosure aspect ratio on heat transfer augmentation are illustrated with analysis of experimental results.
Detailed CFD Analysis of Coolant Mixing in VVER-440 Fuel Assembly Heads Performed With the Code CFX-5.5
Authors: Le´gra´di Ga´bor , Aszo´di Attila
Published: November 2008

3D modeling of the thermal hydraulical processes in a fuel assembly head means a great challenge for the CFD technique due to the complexity of its structure and the flow domain. On the other hand, this field is of great importance since detailed knowledge on mixing processes in the assembly heads and calculations on the signals of the thermocouples positioned just above the heads would give very significant information for the safety analyses connected to the power upgrading of nuclear power plants. Therefore development of a complex fuel assembly model was started in the Institute of Nuclear Techniques of the Budapest University of Technology and Economics in the near past. In this paper, the fuel assembly head model, the sensitivity study of it, calculations and results are presented. The main goal of our work is investigating the signal of the thermocouples which are placed just above the fuel assemblies. The calculations were performed with consideration of four kinds of different fuel assemblies. The inlet velocity and temperature fields were calculated by the COBRA subchannel code of the Paks Nuclear Power Plant of Hungary. With all kind of fuel assemblies, calculations were performed with assumptions of normal symmetrical and highly asymmetrical heat source profiles of inner assemblies and assemblies positioned beside absorber elements.
Monitoring Risk to Generation
Authors: Meyer Stephen , Julius Jeffrey
Published: November 2008

The development of tools and methods to help reduce the risk of unplanned shutdowns and power reductions is becoming a high priority objective of utility management. By having the capability to monitor the risk to power generation, management can provide additional guidance to the plant staff during work planning and daily work activities concerning conditions that place the plant at a heightened risk with respect to power generation. The Safety Monitor™ can perform evaluations of plant configurations and schedules to function in a forward looking mode to evaluate and avoid conditions that lead to a high trip/power reduction frequency.
Risk-Informed Hazard Barrier Assessment
Authors: Meyer Stephen , Krantz Eddie A.
Published: November 2008

There are a number of hazard barriers installed in power plants. The removal of hazard barriers from service can currently require substantial plant resources to evaluate. These evaluations are usually performed to assess the impact of an open barrier on a plant’s licensing basis and fire protection program and are deterministic in nature. The risk-informed approach to hazard barrier assessment helps alleviate the constraints and manpower requirements imposed by the deterministic approaches currently used. A two-step approach has been developed for performing a risk-informed hazard barrier assessment. The objectives of the first step “Data Gathering” are to review the previously performed analyses and to perform the plant walkdowns and data collection needed for the study. In the second step “Model Development and Quantification” using the most recent Probabilistic Safety Assessment (PSA) system models and the results of the “Data Gathering” walkdowns, a set of fault tree/event tree models are prepared for each unique scenario or hazard to be evaluated. The overall set of barrier impairment impacts is binned into a smaller number of generic categories for the PSA analysis. The quantification results will enable utility management to manage the risk of taking barriers out of service and to make a determination of which barriers could be candidates for a risk-informed impairment out-of-service limit if desired.
Status of Pre-Operating Tests for the Latest ABWR in Hamaoka Nuclear Power Unit No. 5
Authors: Yukinori S. , Ohta Y. , Yanase G. , Yoshida H. , Yamamoto Y. , Takahashi M. , Iwata S. , Yamada K. , Sakamoto M.
Published: November 2008

Kashiwazaki-Kariwa Nuclear Power Station Unit 6 and Unit 7 (K6 & K7) are the first advanced boiling water reactors (ABWR) in the world. Toshiba has been engaged in the design and construction of boiling water reactors (BWR) for Hamaoka Nuclear Power Station Unit 5 (H5) of Chubu Electric Power Co., Inc. and Higashidori Nuclear Power Station Unit 1 of Tohoku Electric Power Co., Inc. H5 has features of the simplified reactor coolant recirculation system including a reactor internal pump (RIP) and reduction in plant startup time using a fine-motion control rod drive (FMCRD). The H5 has been improved through additional modifications such as a reduction in the number of adjustable speed drives (ASDs) required by RIPs. This was managed through multi drive and the elimination of a shaft seal from FMCRDs by applying magnetic coupling. The basic performance of the modified FMCRD (S-FMCRD) has been verified by the joint study of Japanese operating utilities and vendors. In addition, various other tests are performed on the H5 to validate its actual component design during the manufacturing stage and pre-operation stage. Construction of H5 began in March 1999. Bedrock inspection was completed in May 2000, the reactor pressure vessel (RPV) installation was finished in July 2002, and power receiving was completed in December 2002. Currently, H5 is scheduled to start commercial operation in January 2005. Pre-operating tests began at the end of 2002. The utility systems test, like the MUWP (makeup water system), HVAC (heating, ventilation and air conditioning), etc, began two months prior to power-receiving, followed by both the RCCW (reactor building closed cooling water system) and RCWS (reactor building cooling sea water system) tests. The primary system pre-operating test, which is usually performed at the peak of the pre-operating test period, was completed by mid-February, 2004. Each test schedule was controlled and each test followed the predefined test plan. On the other hand, S-FMCRDs and control rods (CRs) were installed, followed by an RPV overpressure integrity inspection test conducted in September 2003. Following this test, RIPs and related control system confirmation tests was performed Subsequently, FMCRDs pre-operation test was planned to confirm performance levels. The modified design concepts, such as S-FMCRDs and multi drive ASDs for RIPs etc, were scheduled for verification during this phase.
Technology Road Mapping to Guide Development Planning
Authors: Goossen John E. , Congedo Thomas V.
Published: November 2008

For the past five years, Westinghouse Electric Company, PLC, has made ever increasing use of Technology Road Mapping, to direct company development efforts to achieve maximum benefits for our customers and ourselves. Comprised of business units in Nuclear Fuels, Nuclear Services and Nuclear Power Plants, including domestic and international business segments, Westinghouse must pay particular attention to coordinating development to satisfy the diverse needs of our growing international customer base. We must develop products which both benefit the individual Business Unit customer base, and which create synergy to produce the best possible offerings to the broader marketplace. The knowledge we gain through customer contacts and direct customer participation provides the basis from which we develop the Technology Road Map. This Road Map development process can be compared to painting a picture, where the background colors and features correspond to drivers related to the Customer and the prevailing features of the market environment. The subsequent layers of detail include broad Technical Objectives and then specific Technical Goals which will support achieving those Objectives. The process is described in detail, and examples are provided.
A Preliminary Safety Analysis of the Whole Assembly Seed-Blanket Heterogeneous PWR Core
Authors: Lee Jeongik , Saha Pradip , Kazimi Mujid S. , Lee Won-Jae
Published: November 2008

The “Whole Assembly Seed and Blanket” (WASB) design, which utilizes mostly thorium in the blanket, consists of 84 seed and 109 blanket assemblies which may be backfitted into existing Pressurized Water Reactors (PWRs). Since the seed assemblies produce significantly more power than the blanket assemblies, a preliminary safety analysis of this design has been performed. Three accidents/transients (Large Break Loss of Coolant Accident (LBLOCA), Complete Loss of Primary Flow (LOPF) and Loss of Off-site Power (LOSP)), have been analyzed for both the WASB design and a typical all UO2 design for a typical 4-Loop Westinghouse PWR plant. LBLOCA results show that the peak cladding temperature (PCT) for the WASB design is approximately 260 K higher than that for a typical PWR design. However, this higher PCT for the WASB design is still about 200 K lower than the present regulatory safety limit. The response of the WASB and all UO2 core for LOPF and LOSP transients are very similar, and no post-DNB type rapid cladding temperature rise was observed in either of the two calculations.
Assessment of Potential Negative Impacts of the Main Gate Valve Closure Under Loss of Coolant Accident at VVER-440 Model 213 Nuclear Power Plant
Authors: Everline Chester , Gromov Gregory , Lola Igor , Sholomitsky Stanislav , Mukoid Victor , Meyer Steve , Sevbo Alexander
Published: November 2008

The paper presents the results of the special screening assessment, conducted under the Level 1 Probabilistic Risk Assessment (PRA) study for the Rivne Nuclear Power Plant (RNPP) Unit 1, Ukraine. This analysis has been intended to investigate potential for negative impacts of the closure of the Main Gate Valves (MGV) during loss of coolant accident (LOCA). RNPP Unit 1 is a VVER-440 Model 213 plant featuring six operating primary loops with MGVs installed at the hot and cold legs of each loop. Emergency operating instructions (EOIs) direct operators to isolate a leaking primary loop using the MGVs during a LOCA. From the experience of accident analyses for some Westinghouse plants it is known that closing loop stop valves (LSV, which are analogous to the VVER-440 MGVs) during accidents may result in more severe consequences. In particular, closing a LSV (or MGV) during a LOCA could inhibit steam venting through the break, and thus depress core levels. This would hasten the onset of core damage. Considering this experience, there has been a need to address such potential impacts in the RNPP Unit 1 PRA, taking into account design specifics of the VVER-440/213 plants. The analysis, which has combined qualitative considerations and thermal-hydraulic calculations with RELAP5/Mod3.2 model of the RNPP Unit 1 nuclear steam supply system, has revealed that such impacts are unimportant for this plant. The paper presents the approach and results of the assessment, and describes the supporting reactor thermal hydraulic evaluations.
Experimental Study of Stable Droplet in Blowdown Phenomenon of Large Break LOCA
Authors: Kondo Yoshiyuki , Tanimoto Koichi , Hori Keiichi , Ishiguro Tatsuo , Ogata Junji , Umezawa Shigemitsu , Teramae Tetsuya , Furukawa Yuji
Published: November 2008

For blowdown phenomenon that would appear in primary stage of large break LOCA (Loss Of Coolant Accident), there exists downward misty flow over fuel rods. The behavior of droplets in such flow has effects on clad temperature because droplet removes heat from a reactor core by its direct contact on fuel rods and its evaporation at the surface and in steam flow. Droplets are generated from liquid film flowing with flashing on surface of super-heating clad or flow directory from upper plenum when coolant flows into fuel rods from upper plenum of a reactor vessel. Droplet flows in steam flow with integration and disintegration, and finally becomes steady droplet, i.e. droplet flows in the fuel rods without changing its diameter. For estimation of thermal hydraulic behavior of coolant, LOCA analysis code, MCOBRA/TRAC was developed. In order to estimate peak clad temperature (PCT) over LOCA phenomenon especially reflooding phase following blowdown phase, droplet diameter in actual plant condition must be clarified. We carried out the experiment to measure a diameter of stable droplet simulating actual condition of large break LOCA with two kinds of fluids: Sulfur Hexafluoride (SF6) and ethyl alcohol (C2H5OH) as gas and liquid phase, respectively. Those fluids can simulate a high pressure and temperature steam-water system of 6MPa, and 550K at the condition of the lower pressure and atmospheric temperature of 0.5MPa and 300K, respectively. The experiment was focused on investigating the Weaber number of steady droplet in the blowdown phenomenon of large break LOCA. It was clarified that the minimum distance was under 300mm, and the Weaber number was 3.1.
Experimental Study on Dryout Point in Vertical Narrow Annulus Under Low Flow Conditions
Authors: Tian Wenxi , Myint Aye , Li Zhihui , Qiu Suizheng , Su G. H. , Jia Dounan
Published: November 2008

Prediction of dryout point is experimentally investigated with deionized water upflowing through narrow annular channel with 1.0 mm and 1.5 mm gap respectively. The annular with narrow gap is bilaterally heated by AC current power supply. The experimental conditions covered a range of pressure from 0.8 to 3.5 MPa, mass flux from 26.6 to 68.8 kg/m2 s and wall heat flux from 5.0 to 50.0 kW/m2. The location of dryout is obtained by observing a sudden rise in surface temperature, Kutateladze correlations is cited, however proved to be not a proper one for annuli gap and modified to predict the location of dryout. Considering in detail the effects of the geometry of annuli, pressure, mass flux and heat flux on dryout, an empirical correction is finally developed to predict dryout point in narrow annular gap under low flow condition which has a good agreement with experimental data.
Study on Thermochemical Iodine-Sulfur Cycle at JAERI
Authors: Onuki Kaoru , Kubo Shinji , Kasahara Seiji , Ishiyama Shintaro , Nakajima Hayato , Higashi Shunichi , Masaki Tomoo , Takeda Tetsuaki
Published: November 2008

Japan Atomic Energy Research Institute (JAERI) has been conducting an R&D on thermochemical water-splitting processes of Iodine-Sulfur family, which is a promising candidate of heat-utilization process of High Temperature Gas-cooled Reactors. Present activity at JAERI covers the following three subjects, (a) closed-loop operation technique for stable and continuous hydrogen production by e.g. suppressing possible side reactions, (b) process improvement in terms of thermal efficiency of hydrogen production utilizing membrane technologies, and (c) selection of materials for constructing the large-scale plant mainly focusing on corrosion resistance in the representative process environments. Recent progress on these studies is briefly described.
Seismic Re-Qualification of an Underground RCC Duct
Authors: Bhargava K. , Ramesh P. , Ghosh A. K. , Jain R. C. , Ramanujam S.
Published: November 2008

The present paper deals with the re-evaluation of design stresses and assessment of safety margins available for an underground RCC duct. For stress analysis, a 2-D mathematical model has been adopted to represent the structure appropriately. The structure has been analyzed for both static and seismic loads. The seismic analysis has been carried out for site-specific response spectra. The design checks have been performed using available international standards and accordingly the safety margins have been evaluated. The structure has been found to possess sufficient safety margin under all the postulated design loadings. The seismic reassessment methodology conforms to the available international standards.
On the MEGAPIE Target Thermal Hydraulics: A RELAP5 Analysis
Authors: Leung W. H.
Published: November 2008

A new type of liquid-metal target is designed for the Spallation Neutron Source in PSI. LBE is selected to be the target material and the primary coolant as well. RELAP5/MOD 3.2 is used to analyze the system thermal hydraulics. The nominal conditions are chosen based on temperature constraints from the design assessments. The steady state results are in the proximity of the design specifications and the heat removal capacity is adequately deployed. The normal thermal hydraulic transients, namely the proton beam and beam interrupt, are studied. A basic PID (Proportional, Integral, Derivative) control is implemented in the RELAP5 for regulating the target temperature. It is found that the control chain works very well for the beam trip in limiting the temperature fluctuations. In a beam interrupt, the proton beam is completely turned off without recovering. The transition from full power to hot-standby is quite smooth, but it becomes oscillatory in the long run due to the timelags in the cooling loops’ responses. An off normal case of target main coolant trip has also been studied. Without the main pump, the target can still be operated in the natural circulation mode, and the control can cope with the normal beam transients and restarting the target from hot standby.
A Computational Study of H2/N2 Jet Diffusion Flame With Radiation Heat Loss
Authors: Alim M. A. , Malalasekera W.
Published: November 2008

In this work simulation of a turbulent H2/N2 jet diffusion flame with flamelet modeling has been presented. The favre averaged mixture fraction has been employed to model the combustion. Favre-averaged scalar quantities have been calculated from flamelet libraries by making use of a presumed Probability Density Function (PDF) method. To incorporate the effect of radiation heat transfer the combustion model has been extended using the concept of enthalpy defect. The predicted flame temperature profiles and chemical species concentrations with and without radiation heat loss are compared with experimental data. Predictions considering the radiation heat loss found to be in good agreement with temperature and chemical species measurements whereas the adiabatic model significantly overestimates temperatures in the downstream regions of flames where the significant heat loss occurs. This study shows that the combustion simulation using flamelet models considering radiation heat loss are effective for predicting the flow, temperature and chemical kinetics of H2/N2 jet diffusion flame. To account for fluctuations of mixture fraction, its distribution is presumed to have the shape of a beta-function.
Sensitivity Analysis of 200 mm LOCA at Armenian NPP Unit 2
Authors: Amirjanyan Armen , Malakyan Tsolak , Jo Jae
Published: November 2008

This paper presents the results of sensitivity analyses of a loss-of-coolant accident (LOCA) with equivalent diameter of 200 mm on a cold leg at Armenian Nuclear Power Plant (ANPP) Unit 2. ANPP is a six-loop WWER-440/270 model of Russian design. Analyses performed for the Technical Assignment for WWER-440/270 reactors showed that this reactor had high safety margins with 100 mm LOCA, which may allow increasing the safety level for the maximum design base LOCA. To demonstrate the adequacy of the safety margin with equivalent diameter of 200 mm using the existing high-pressure injection (HPI) system, analyses of LOCAs with break sizes of the equivalent diameter of 200mm in the primary side were previously performed with highly conservative assumptions using the RELAP5 code (Mod3.2.2β). Results of this study were reported at ICONE 11. The results of LOCA analyses with highly conservative assumptions showed that the acceptance criteria were not exceeded in the cases of a surge line break and equivalent break on a hot leg. However, calculations for a break with equivalent diameter of 200 mm on a cold leg showed that the cladding temperature reached 1235 °C, narrowly exceeding the safety limit mentioned in acceptance criteria (1204 °C), and prompted to conduct additional sensitivity studies with slightly relaxed conservative assumptions. At ANPP, there are two trains of HPI systems with two diesel generators on each train and one additional generator in the reserve state. Each train of diesel generators supplies power to two HPI pumps. However, in the current ANPP configuration, only one HPI pump in each train is automatically switched on to diesel generators in case of a station blackout. Therefore, the previous analysis of LOCA with highly conservative assumptions took credit for only two HPI pumps available after station blackout (i.e., two of diesel generators and reserve diesel generator are not available). The current ANPP modernization plan includes automatic starts of four HPI pumps. For this sensitivity analysis we took credit for one additional HPI pump (namely, three HPI pumps all together). As in the previous calculations, a loss of off-site power was assumed at the moment of accident beginning and simultaneous reactor scram: diesel generators were assumed to be available 20 seconds after the off-site power loss. The results of this sensitivity calculation of LOCA with equivalent diameter of 200 mm on a cold leg show that the cladding temperature stayed far below the safety limit. The peak cladding temperature was 850°C, well below that of the original study (1235°C) and the safety limit in the acceptance criteria (1204°C).
The Application of the Welding Technique at Fillet Groove by the YAG-Laser Repair-Welding Robot for Underwater Environment
Authors: Makihara Yasuhiro , Miwa Yashuhiro , Hirose Naoya , Satoh Syuichi , Ohwaki Katura
Published: November 2008

Generally, when we need repair welding due to the reduction of the thickness generated by corrosion in the inside of tanks, usually people enter and repair after the inner liquid is drained. However, the direction that performed repair work, without draining intension liquid may be desirable on work. Especially, for the radioactive waste (RW) tanks in operating nuclear plants, since there are danger that be accompanied by the risk of exposure, it is desired that to shorten their work time (abbreviation of drainage) and to use the robot system. As the welding repair method that can be performed underwater, there is underwater welding technology by YAG laser. So far, we developed the remote underwater YAG laser repair-welding robot. A repair welding work to the robot is performed as following steps: 1. Degradation area detection. 2. Crack removal and edge preparation. 3. Repair welding, and inspection. And we found that the application of the repair welding to use the robot at the base metal and butt joint. Furthermore, we will indicate the wide range application of the remote underwater YAG laser repair-welding robot to make sure the appropriate work of welding at fillet groove joint, in this time.
Intentional Back Flow Effects on Ruptured Steam Generator Cooldown During a SGTR Event for KSNP
Authors: Kim Cheol Woo , Park Seok Jeong , Choi Chul Jin , Seo Jong Tae
Published: November 2008

For an optimum recovery from a steam generator tube rupture (SGTR) event, the operators are directed to isolate the steam generator (SG) with ruptured tube as early as possible to minimize the radioactive material release. However, the reactor coolant system (RCS) cooldown and depressurization to the shutdown cooling system (SCS) operation conditions using the intact SG only are hard to achieve unless the ruptured SG is properly cooled since the ruptured SG, which is isolated by operator, remains at high temperature even though the RCS has been cooled down. The effects of intentional back flow from the SG secondary side to the RCS through the ruptured U-tube on the the ruptured SG cooldown were evaluated for the pressurized light water reactor, especially for the Korean standard nuclear power plant (KSNP). In order to evaluate the back flow effect, a series of analyses was conducted using the RELAP5/MOD3 computer code. For the first stage of the analysis, the cooldown process by natural circulation in the SG secondary side was simulated for the initial conditions of the ruptured SG cooldown. In the next analysis stage, two methods of the ruptured SG cooldown by using back flow after RCS cooldown were evaluated. The one is a tube uncovery, which utilizes the steam condensation on the uncovered U-tube surface, and the other is a SG drain and fill. In the former method, SG tubes are exposed to the steam space by draining SG secondary water into the RCS in order to condense the steam directly onto the uncovered tubes. This method showed that the steam condensation decreased SG secondary pressure and temperature rapidly, demonstrating its effectiveness for cooling. However, this process has a limited applicability in cases of that the rupture is located at the lower region. The latter method, draining by back flow and filling using the feedwater system was also found to be effective in ruptured SG cooldown and depressurization even if the rupture occurred at the top of the U-tube. However, the actuation of the feedwater system is a burden to operator since the makeup of cold feedwater was required to complete cooldown by one cycle of the draining and filling. It is concluded that utilization of the intentional back flow from the SG secondary side to the RCS is very effective for rapid cooling of the RCS to the SCS entry conditions.
Investigation of Polonium Removal Systems for Lead-Bismuth Cooled Fast Reactors Using a Tellurium Surrogate: Part II
Authors: Loewen Eric , Ohno Shuji
Published: November 2008

The alkaline extraction method is being investigated as a method to remove polonium from a lead-bismuth cooled reactor. Tellurium was used as a surrogate for polonium for these experiments to quantify the migration from lead-bismuth to molten alkaline (NaOH). Experiments provided direct evidence that molten NaOH can effectively remove tellurium from lead-bismuth. A reduction of three to four orders of magnitude in the tellurium concentration in the metals was measured in the experiments. The experimental results also showed a higher than expected concentration of NaOH in the lead or lead-bismuth. A comparison between experiments operated at 427°C and 500°C respectively indicates that the higher operating temperature produces a higher removal rate of Te. Experiments where hot NaOH was injected into the crucible at the same temperature as the lead-bismuth eutectic (LBE) resulted in the rapid removal rate of Te and remained constant until the experiment was completed. This paper is a continuation of the results reported in the previous ICONE paper, (ICONE11-36614) providing an experimental update.
The Influence of Initial Geometric Imperfection on the Localized Buckling of Pressure Vessel Under Internal Pressure
Authors: Wan Li , Tao Wei-ming , Wu Xin-xin , He Shu-yan
Published: November 2008

Pressure vessels are widely used in nuclear engineering and buckling is a common mechanical phenomenon in structure. The buckling problem of pressure vessels under external pressure has been researched for many years. This paper focuses on the influence of initial geometric imperfection on the localized elastic-plastic buckling of pressure vessel under internal pressure. The localized plastic buckling occurred in the transition region in the torispherical end closure of a pressure vessel is analyzed by FEM. By introducing two types of initial geometrical imperfections, the arc-length method of modified Riks/Ramm procedure is performed to simulate the buckling process during loading. The first type of imperfection is displacement, into the region where it is circumferentially compressed. The second type of imperfection is the irregular thickness of the vessel, also into the region where it is circumferentially compressed. The initial critical point is captured within the buckled region, and the corresponding initial buckling load is calculated. The results show that both artificial geometric imperfections can seduce the buckling. Furthermore, after the first buckling initiated, the succeeding loading will lead to more wrinkles within the compressive transition region. And then the case that with two distributed imperfections is also analyzed. It can be seen that the interaction between the imperfections is very weak before or even after the first buckling occurred, which means the buckling is fairly localized.
Safety of Nuclear Reactors: Part B — Unsteady State Strength Mathematical Model
Authors: El-Shayeb Mohamed , Yusoff Mohd. Zamri , Bondok Ali , Roseli Adnan , Ideris Fazril , Abu Hassan Saiful Hasmady
Published: November 2008

The significant mechanical properties that determine the strength of concrete-filled square steel columns are compressive strength, modulus of elasticity, ultimate strain of the concrete, yield strength and modulus of elasticity of the steel. A survey of the literature shows that the variation of these properties with temperature is influenced by many factors. The compressive strength of concrete at elevated temperature is affected by the rate and duration of heating, the size and shape of the test specimen and the loading during the heating. During exposure to fire the strength of the column decreases with the duration of exposure. The strength of the column can be calculated by a method based on load-deflection analysis which in turn is based on a stress-strain analysis of cross-sections. A mathematical model describing the behavior of concrete-filled square steel column structures under load and heat has been thoroughly examined in this paper. The main goals of this model are to study the structure member characteristics (temperature, deformation and strength) under excessive heat conditions and to study the possibility of increasing the endurance time by changing the column composite structure.
Integral Solutions of Phase Change With Internal Heat Generation
Authors: Siahpush Ali , Crepeau John
Published: November 2008

This paper presents solutions to a one-dimensional solid-liquid phase change problem using the integral method for a semi-infinite material that generates internal heat. The analysis assumed a quadratic temperature profile and a constant temperature boundary condition on the exposed surface. We derived a differential equation for the solidification thickness as a function of the internal heat generation (IHG) and the Stefan number, which includes the temperature of the boundary. Plots of the numerical solutions for various values of the IHG and Stefan number show the time-dependant behavior of both the melting and solidification distances and rates. The IHG of the material opposes solidification and enhances melting. The differential equation shows that in steady-state, the thickness of the solidification band is inversely related to the square root of the IHG. The model also shows that the melting rate initially decreases and reaches a local minimum, then increases to an asymptotic value.
The Capabilities of HTRs to Burn Actinides and to Optimize Plutonium Exploitation
Authors: Cerullo Nicola , Bufalino D. , Forasassi G. , Lomonaco G. , Rocchi P. , Romanello V.
Published: November 2008

At present, the 125 GWe of nuclear power in the European Union produce about 3000 tons of spent fuel annually, containing about 25 tons of plutonium, 2.5 tons of minor actinides (MA) and about 100 tons of fission products, of which 3.1 tons are long-lived fission products. Actual reprocessing of LWR fuel and a first recycling as mixed plutonium and depleted uranium oxide fuel (MOX) in LWR already contribute to a significant reduction of waste volumes and radiotoxicity. However HTRs have some characteristics which make them particularly attractive: intrinsic safety, cost-effectiveness, reduced thermal pollution, capability of increasing energy availability (with the use of Pu-Th cycle) and of minimizing actinides radiotoxicity and volume of actinides. In this paper particularly the last item is investigated. Symbiotic fuel cycles of LWR and HTR can reach much better waste minimization performances. It happens because of the specific features of HTRs cores that leads to an ultra-high burnup and, last but not least, the ability to accommodate a wide variety of mixtures of fissile and fertile materials without any significant modification of the core design. This property is due to a decoupling between the parameters of cooling geometry and of neutronic optimization. In our calculations we considered a pebble-bed HTR using a Pu-based fuel (deriving from reprocessing of classical LWR fuel and/or weapons grade plutonium) at the maximum technological discharge burnup. As results, we find, at EOL (End Of Life), a relatively small amounts of residual Pu and MA produced in terms of quantities and of radiotoxicities. Furthermore we used in our calculations a different type of fuel based on a mixture of Pu and Th to try to optimize the previous results and to increase energy availability. Calculations have been done using MCNP-based burnup codes, capable of treating 3-D complex geometry and ultra-high burnup.
Experimental Investigation of Coolant Mixing in the RPV of a PWR During Natural Circulation Conditions
Authors: Kliem S. , Hoehne T. , Prasser H.-M. , Rohde U. , Weiss F.-P.
Published: November 2008

Partial depletion of the primary circuit during a hypothetical small break loss of coolant accident can lead to the interruption of one-phase flow natural circulation. In this case, the decay heat is removed from the core in the reflux-condenser mode. For the scenario of a hot leg side leak and hot leg safety injection thermal hydraulics analyses using the system code ATHLET showed, that weakly borated condensate can accumulate in particular in the pump loop seal of those two loops, which do not receive safety injection. According to these ATHLET-calculations, one-phase flow is maintained in the remaining two loops at high residual heat conditions because of the entrainment of safety injection coolant into the steam generators. After refilling of the primary circuit, natural circulation in the two stagnant loops simultaneously re-establishes and the de-borated slugs are shifted towards the reactor pressure vessel (RPV). Mixing in the downcomer and the lower plenum is an important phenomenon mitigating the reactivity insertion into the core in this postulated scenario. Therefore, mixing of the de-borated slugs with the ambient coolant in the RPV was investigated at the four loop 1:5 scaled ROCOM mixing test facility. Based on the ATHLET-calculations, a volume flow rate of 5% of the nominal rate was set in the loops running in one-phase flow. The volume flow rate in the two restarting loops increases from zero to 6%. In these two loops, de-borated slugs of 7.2 m3 were assumed corresponding to the volume of the whole loop seal. An experimental parameter study was carried out with different duration of the flow ramp and variation of the density difference between de-borated slug and ambient coolant due to differences in boron concentration and temperature. The variation of the density difference significantly changes the mixing behavior. With no density difference, the weakly borated coolant almost perpendicularly flows down in the downcomer and a maximum of 64% of the initial perturbation is detected in the core entry section below the loops where the slugs were formed. Increasing the density difference, a stratification is observed in the downcomer. The less dense slugs flow around the core barrel at the top of the downcomer. At the opposite side the lower borated coolant is entrained by the colder safety injection water and transported to the core. This entrainment effect leads to the admixture of boron from the safety injection to the under-borated slugs. Consequently, the maximum under-boration at the core entry is lower. For the maximum investigated density difference of 2%, a value of 31% only of the initial under-boration was measured at the core entrance.
Nuclear Research Reactor RTS-1 “G. Galilei” Decommissioning: Preliminary Operations and Cutting Facilities
Authors: Boschi A. , Cimini E. , Parracone L. , Pocai M. , Russo M.
Published: November 2008

The RTS-1 “Galileo Galilei” is an open pool research reactor with light water as moderator and coolant, it has a nominal power of 5 MWth and an average thermal flux of 5 E+13 n/cm2sec. It went critical for the first time on April 1963 and it was definitively shutdown in March 1980. The reactor is situated at CISAM (Joint Centre of Studies for Military Application - Italian Ministry of Defence), S. Piero a Grado, Pisa, Italy, and its decommissioning is in progress. In this paper the preliminary activities necessary to eliminate the most part of radioactive materials present into the plant are described. Emphasis is placed on the description of the “Irradiation Channel” facility, used to manage safely all the activated materials to be conditioned and on the MASCOT robot, used in the channel. All the conditioning devices to be used are described with particular consideration to the cleanness of the cutting process and the radiological risk due to the operations. The first cutting and conditioning operation carried on in the Galilei Reactor is described. This operation regards the conditioning of some experimental equipments used during the reactor life, with particular attention to the radiation protection of the personnel and to the control of radioactive emission.
Material Studies for Pulsed High-Intensity Proton Beam Targets
Authors: Simos Nicholas , Kirk Harold , Ludewig Hans , Thieberger Peter , Weng W.-T. , McDonald Kirk , Yoshimura K.
Published: November 2008

Intense beams for muon colliders and neutrino facilities require high-performance target stations of 1–4 MW proton beams. The physics requirements for such a system push the envelope of our current knowledge as to how materials behave under high-power beams for both short and long exposure. The success of an adopted scheme that generates, captures and guides secondary particles depends on the useful life expectancy of this critical system. To address the key technical challenges around the target of these initiatives, a set of experimental studies have either been initiated or being planned that include (a) the response and survivability of target materials intercepting intense, energetic protons, (b) the integrity of beam windows for target enclosures, (c) the effects of irradiation on the long-term integrity of candidate target and focusing element materials, and (d) the performance of the integrated system and the assessment of its useful life. This paper presents an overview of what has been achieved during the various phases of the experimental effort including a tentative plan to continue the effort by expanding the material matrix. The paper also attempts to interpret what the experimental results are revealing and seeks for ways to extrapolate to the required intensities and anticipated levels of irradiation and it discusses the feasibility of the proposed approaches to achieving such high-performance systems. Further it explores the connection of accelerator target systems with reactor systems in order to utilize experience data that the nuclear reactor sector has acquired over the years.
Very High Flux Steady State Reactor and Accelerator Based Sources
Authors: Ludewig H. , Todosow M. , Simos N. , Shapiro S. , Hastings J.
Published: November 2008

With the number of steady state neutron sources in the US declining (including the demise of the BNL HFBR) the remaining intense sources are now in Europe (i.e. reactors - ILL and FMR, accelerator - PSI). The intensity of the undisturbed thermal flux for sources currently in operation ranges from1014 n/cm2-s to 1015 n/cm2-s. The proposed Advanced Neutron Source (ANS) was to be a high power reactor (∼ 350 MW) with a projected undisturbed thermal flux of 7×1015 n/cm2-s but never materialized. The objective of the current study is to explore the requirements and implications of two source concepts with an undisturbed flux of 1016 n/cm2-s. The first is a reactor based concept operating at high power density (10 MW/l–15 MW/l) and a total power of 100 MW–250 MW, depending on fissile enrichment. The second is an accelerator based concept relying on a 1 GeV–1.5 GeV proton LINAC with a total beam power of 40 MW and a liquid lead-bismuth eutectic target. In the reactor source study the effects of fissile material enrichment, coolant temperature and pressure drop, and estimates of pressure vessel stress levels will be investigated. The fuel form for the reactor will be different from all other operating source reactors in that it is proposed to use an infiltrated graphitic structure, which has been developed for nuclear thermal propulsion reactor applications. In the accelerator based source the generation of spallation products and their activation levels, and the material damage sustained by the beam window will be investigated.