bluuebunny/crossref_metadata_2025 huggingface dataset:
Analysis for the Case of Channel Flow and Reactivity Changes for Advanced Heavy Water Reactor
Authors: Srivastava A. , Majumdar P. , Mukhopadhyay D. , Lele H. G. , Gupta S. K.
Published: November 2008

The proposed Advanced Heavy Water Reactor (AHWR) is a vertical pressure tube type boiling light water cooled and heavy water moderated reactor. One of the important passive design features of this reactor is that the heat removal is achieved through natural circulation of primary coolant at all power level with no primary coolant pumps. Decrease in coolant flow or control rod malfunction can lead to undesirable rise in clad surface temperature depending upon severity and characteristics and response of the reactor and associated systems. In this paper safety assessment of the AHWR is made due to above events of different severity. Cause for events under category of decrease in coolant flow is mainly channel blockage of different severity at different locations. There is no other reason as it is natural circulation based reactor. Effect of flow decrease can be different in different channels and at different axial locations. In this paper channel blockages of different sizes are analysed at core inlet and using slave channel approach. Changes in reactivities can occur due to inadvertent withdrawal of one or more control rods from reactor core. In this analysis one control rod assembly is assumed to be removed from core. The event is simulated by addition of 5 mk reactivity in 120 seconds depending on the speed of withdrawal of assembly. The analysis for the above events are complex due to various complex and wide range of phenomena involved during different PIEs under this category. It involves single and two phase natural circulation at different power levels, inventories and pressures, coupled neutronics and thermal hydraulics behaviour, and coupled controller and thermal hydraulics. In this paper summary of analysis for each event is presented. In this paper, various modeling complexities are brought out; evaluation of acceptance criteria is made and design implications of each event are discussed.
Thermal Hydraulic Analysis Due to the Changes in Heat Removal for Advanced Heavy Water Reactor
Authors: Chatterjee B. , Srivastava A. , Mukhopadhyay D. , Majumdar P. , Lele H. G. , Gupta S. K.
Published: November 2008

Advanced Heavy Water Reactor is natural circulation light water cooled and heavy water moderated pressure tube type of reactor. Changes in heat removal by primary heat transport system of a reactor have significant impact on various important system parameters like pressures, qualities, reactor power and flows. Increase in heat removal leads to Cooldown of the system subsequently reducing pressure, void increase and changes in power and flows of the system. Decrease in heat removal leads to warm-up of the system subsequently raising pressure, void collapse, and changes in power and flows of the system. The behaviour is complex as system under consideration is natural circulation system. Causes for events under category of increase in heat removal are mainly malfunctioning of feed water heaters, Isolation Condensers (IC) inlet valves and controllers. These events lead to cooldown of system and addition of positive reactivity addition due to void collapse. Various events considered are Feed Water System malfunctions that result in decrease in feed water temperature, inadvertent opening of IC valve, Failure of PHT Pressure Control System and Decrease in pressure controller set point to 67 bars. Causes for events under category of decrease in heat removal are mainly malfunctioning of controllers, feedwater valves and operating events like turbine trip. Functioning of passive cooling system and different valves play important role for these events. These events lead to increase in system pressure. Various events considered are Loss of normal feed water flow (multiple trains), Turbine trip without bypass without IC, Turbine trip without bypass with IC, Turbine trip with bypass without IC, Increase in PHT pressure controller set point, Decrease in level controller set point, Turbine Trip with setback, Decrease in steam flow and Class IV power failure. Changes in the system voids and pressures as a result of change in the heat removal leads to complex reactivity feedback due to coolant temperatures, void fraction and fuel temperatures. These changes in the reactor power together with void distribution change affect two-phase natural circulation flow. This paper brings out these aspects. It discusses descretisation of the system and brings out various design aspects. In this paper summary of analysis for each event is presented, various modeling complexities are brought out, evaluation of acceptance criteria is made and design implications of each event is discussed.
Maintenance Technology Development for Alloy 600 PWSCC Issue
Authors: Taniguchi Masaru , Hori Nobuyuki
Published: November 2008

Alloy 600 aging issue within a PWR plant is one of the important aging issues especially for aged plants in terms of susceptibility of PWSCC and impact on integrity of critical components. As for Reactor vessel, alloy 600 material is used for reactor vessel head (RVH) penetrations, bottom mounted instrumentation (BMI) penetrations and nozzle safe-end weld joints. RVH replacement work has been and will be conducted at many PWR plants all over the world to manage RVH penetration aging issue, and aging countermeasures for BMI penetrations and nozzle safe-end weld joints may become still more important taking into account of leakage events at South Texas Project (2003, BMI penetrations), at V. C. Summer (2000, nozzle safe-end weld joint of Reactor vessel) and at Tsuruga Unit 2 (2003, nozzle safe-end weld joint of Pressurizer safety & release nozzle). Mitsubishi Heavy Industries, Ltd. (MHI) has developed and also is developing maintenance technology such as inspection, repair, replacement and mitigation measures in order to cope with this issue effectively and economically. This paper describes MHI’s maintenance technology as aging countermeasures for Reactor Vessel’s regions using alloy 600, especially for BMI penetrations and nozzle safe-end weld joints. As for BMI penetrations, MHI has already developed water jet peening (WJP) technology as mitigation measure and applied it to several plants since 2001, also MHI has developed repair technology in case of leakage and cracking events. As for nozzle safe-end weld joints, MHI has been developing mitigation, repair and replacement technology, and has already applied the nozzle safe-end replacement technique to Pressurizer safety & release nozzle of Tsuruga unit 2 in 2003.
Improved Lumped-Differential Models for Transient Heat Conduction in Multilayered Composite Media
Authors: Su Jian , Cerqueira Djane R.
Published: November 2008

In this paper we present improved lumped-differential formulations for one-dimensional transient heat conduction in multilayered composite media. Hermite approximations for integrals are used to obtain the average temperatures and heat fluxes in each layer. Average temperatures calculated with improved lumped parameter formulation agree well with reference finite difference solutions. The proposed heat conduction models can be used in fuel dynamics calculation for stability analysis of BWR, simplified model of PWR or real-time simulator of nuclear power plants.
The REBUS Experimental Programme for Burn-Up Credit
Authors: D’hondt Pierre , Baeten Peter , Sannen Leo , Marloye Daniel , Lance Benoit , Renard Alfred , Basselier Jacques
Published: November 2008

An international programme called REBUS for the investigation of the burn-up credit has been initiated by the Belgian Nuclear Research Centre SCK-CEN and Belgonucle´aire with the support of EdF and IRSN from France and VGB, representing German nuclear utilities and NUPEC, representing the Japanese industry. Recently also ORNL from the U.S. joined the programme. The programme aims to establish a neutronic benchmark for reactor physics codes in order to qualify the codes for calculations of the burn-up credit. The benchmark exercise investigates the following fuel types with associated burn-up: reference fresh 3.3% enriched UO2 fuel, fresh commercial PWR UO2 fuel and irradiated commercial PWR UO2 fuel (54 GWd/tM), fresh PWR MOX fuel and irradiated PWR MOX fuel (20 GWd/tM). The experiments on the three configurations with fresh fuel have been completed. The experiments show a good agreement between calculation and experiments for the different measured parameters: critical water level, reactivity effect of the water level and fission-rate and flux distributions. In 2003 the irradiated BR3 MOX fuel bundle was loaded into the VENUS reactor and the associated experimental programme was carried out. The reactivity measurements in this configuration with irradiated fuel show a good agreement between experimental and preliminary calculated values.
Effects of Contact Line Condition on Excitation of Internal Waves Confined by a Cylindrical Wall
Authors: Ito Takahiro , Shimizu Jun , Nakayama Hideyuki , Kukita Yutaka
Published: November 2008

Standing internal waves can be excited when fluids overlaid in a stationary enclosure are subjected to harmonic, vertical oscillations. The oscillatory deformation of the fluid-fluid interface near the wall, due to the limited mobility of the fluid-fluid-wall contact line, can play an important role in the excitation of waves. The contact line may or may not move depending on the amplitude of fluid excitation. When it moves, it may do so only in a limited portion of each cycle. We analyze numerically the wave excitation associated with such nonlinear, intermittent motions of the contact line. The analytical results consistently reproduce the experimental results, and give insight into the interactions between the contact line motions and the waves on the interface.
Visualization System of Swirl Motion
Authors: Nakayama Katsuyuki , Umeda Kenji , Ichikawa Toshio , Nagano Teruyuki , Sakata Hideyuki
Published: November 2008

An instrumentation of system composed of experimental device and numerical analysis is presented to visualize flow and identify swirling motion. Experiment is performed with transparent material and PIV (Particle Image Velocimetry) instrumentation, by which velocity vector field is obtained. This vector field is then analyzed numerically by “swirling flow analysis”, which estimate its velocity gradient tensor and the corresponding eigenvalue (swirling function). As an instantaneous flow field in steady/unsteady states is captured by PIV, the flow field is analyzed, and existence of vortices or swirling motions and their locations are identified in spite of their size. In addition, intensity of swirling is evaluated. The analysis enables swirling motion to emerge, even though it is hidden in uniform flow and velocity filed does not indicate any swirling. This visualization system can be applied to investigate condition to control flow or design flow.
Experiments on Performance of the Multi-Layered In-Vessel Core Catcher
Authors: Kang Kyoung-Ho , Park Rae-Joon , Kim Sang-Baik , Suh K. Y. , Cheung F. B. , Rempe J. L.
Published: November 2008

LAVA-GAP experiments are in progress to investigate the performance of the in-vessel core catcher using alumina melt as a corium simulant. The hemispherical in-vessel core catcher made of carbon steel was installed inside the lower head vessel with a uniform gap of 10mm. Until now, two types of the in-vessel core catcher were used in this study. The first one is a single layered in-vessel core catcher without an internal coating of the LAVA-GAP-2 test, and the other one is a two layered in-vessel core catcher with a 0.5mm-thick ZrO2 internal coating of the LAVA-GAP-3 test. Current LAVA-GAP experimental results indicate that an internally coated in-vessel core catcher has better thermal performance compared with an uncoated in-vessel core catcher. Metallurgical inspections on the test specimens of the LAVA-GAP-3 test have been performed to examine the performance of the coating material and the base carbon steel. Although the base carbon steel had experienced severe thermal attack to the extent that the microstructures were changed and re-crystallization occurred, the carbon steel showed stable and pure chemical compositions without any oxidation and interaction with the coating layer. In terms of the material aspects, these metallurgical inspection results suggest that the ZrO2 coating performed well.
Analysis of VVER-1000 Large and Small Break LOCA Experiments With RELAP5/MOD3.2
Authors: Rychkov Maxym , Chikkanagoudar Utkarsh , Sehgal Bal Raj
Published: November 2008

A RELAP5 model for the analysis of the PSB-VVER test facility was developed by the EREC in Russia. The PSB-VVER is a large-scale integral test facility to model the VVER-1000 type NPP. The volume and power scale in this test facility is 1:300 and the elevation scale is 1:1, which corresponds to the elevation mark of the reactor prototype. At the Division of Nuclear Safety, RIT, Sweden, we have modified the PSB-VVER facility’s RELAP5 model in order to analyze two of the transient tests performed on the PSB-VVER facility, which serve as the validation matrix described by NEA/CSNI. The objective of the work conducted was to validate the results obtained from RELAP5’s calculation with the supplied experimental data from the PSB-VVER test facility. Two accident scenarios have been calculated and analyzed. After being verified against the “11% UP LOCA” test data, the RELAP5/MOD3.2 model was used for a so-called “blind” transient calculation of the test “2×25% HL LOCA” and the results obtained were compared with the experimental data provided after the calculation.
The Use of Level 3 PSA for the Integrated Leakage Rate Test Interval Relaxation in Korea
Authors: Oh Hae-Cheol , Seo Mi-Ro , Kim Myung-Ki , Chung Bag-Soon , Hong Sung-Yull
Published: November 2008

Containment structure testing is intended to assure the leak-tight integrity of the containment structure under all design basis conditions. In Korea, ILRT testing frequency for PWR-type plants had been 1 every 5 years following the requirements of operating license. However, the utility had performed more frequently ILRT than the requirements due to the refueling outage of 18 months. The feasibility study of reducing the frequency of ILRT from 1 every 5 years to 1 every 10 years was performed with the method of performance-based containment leak-test program (NUREG-1493). To verify the justification of relaxation of test frequency, two activities were conducted. First, the recent 3 periods experiences of containment leakage-rate test were reviewed. And, Level 3 PSA for Ulchin Units 3 &4 as pilot plants was performed.
Cracking in Alloy 600 Penetration Nozzles: A Regulatory Perspective
Authors: Hiser Allen L.
Published: November 2008

Cracking in Alloy 600 penetration nozzles in the reactor pressure vessel upper heads of pressurized water reactors (PWRs) has been an issue in the U.S. nuclear industry for more than 10 years. This paper provides a regulatory perspective on how the U.S. industry and the U. S. Nuclear Regulatory Commission (NRC) have addressed new findings over the last 10 years and responded to ensure safety of U.S. PWRs. This paper provides a summary going back to the early 1990’s and emphasizes activities over the last three years since the discovery of circumferential cracking at Oconee, and provides a perspective regarding on-going NRC activities.
Hydrogen Generation Using the Modular Helium Reactor
Authors: Richards Matt , Shenoy Arkal
Published: November 2008

Process heat from a high-temperature nuclear reactor can be used to drive a set of chemical reactions, with the net result of splitting water into hydrogen and oxygen. For example, process heat at temperatures in the range 850°C to 950°C can drive the sulfur-iodine (SI) thermochemical process to produce hydrogen with high efficiency. Electricity can also be used to split water, using conventional, low-temperature electrolysis (LTE). An example of a hybrid process is high-temperature electrolysis (HTE), in which process heat is used to generate steam, which is then supplied to an electrolyzer to generate hydrogen. In this paper we investigate the coupling of the Modular Helium Reactor (MHR) to the SI process and HTE. These concepts are referred to as the H2-MHR. Optimization of the MHR core design to produce higher coolant outlet temperatures is also discussed.
Study on Drag Coefficients for Two Groups of Bubbles
Authors: Sun Xiaodong , Liu Yang , Ozar Basar , Ishii Mamoru , Kelly Joseph M.
Published: November 2008

To apply the two-fluid model to a wide range of flow regimes in gas-liquid two-phase flows, the gas phase is categorized into two groups: small spherical/distorted bubbles as Group 1 and large cap/slug/churn-turbulent bubbles as Group 2 in the modeling of interfacial area transport. The interfacial transfer terms of momentum and energy for the gas phase are then divided into two groups accordingly in the implementation of the two-group interfacial area transport equation to the two-fluid model. Thus, the drag coefficients and the interfacial heat transfer for each group bubbles need to be developed. An approach has been sought for evaluating the drag coefficients of each bubble group based on a comprehensive experimental data base obtained in air-water upward flows in various size round pipes. Comparisons have been made with the theory of the drag coefficients and it was found that the agreement is not very satisfactory although the general trends can be predicted by the current approach.
Latest Technologies Applied to the Advanced Boiling Water Reactor (ABWR): A History of the Development and Manufacturing Stage
Authors: Yamamoto Y. , Yanase G. , Yoshida H. , Ukai M. , Takahashi M. , Naito M. , Yasuda M.
Published: November 2008

The first advanced boiling water reactor (ABWR)—Kashiwazaki-Kariwa Nuclear Power Station Unit 6 (K6)—was built in Japan by Tokyo Electric Power Corporation (TEPCO) in 1996. Subsequently, Kashiwazaki-Kariwa Nuclear Power Station Unit 7 (K7) was constructed (completed in 1997). The K6 and K7 have been operating safely for more than 6 years. Presently, four additional ABWRs are under construction (two in Japan and two in Taiwan), with an additional nine ABWRs in Japan that are currently in the planning stage. One of the ABWRs now under construction is Hamaoka Unit 5 (H5), owned by Chubu Electric Power Co., Inc. The H5 has an estimated completion date of January 2005. The latest technologies have been applied to H5, which will result in improved economics, ease of maintenance, increased reliability, and a reduction in radiation exposure. Below, you will find the latest major technologies applied to H5 in terms of the history of development and verification of manufacturing stages: (1) Sealless Fine-Motion Control Rod Drive (S-FMCRD): S-FMCRD provides fine-motion control by the electric motor for normal operation and hydraulic pressure for scram insertion. This redundancy of power supply is similar to the conventional FMCRD. Additionally, S-FMCRD experiences excellent performance due to its magnetic coupling. As a result, S-FMCRD provides improved operation, ease of maintenance, increased reliability, and reduced occupational radiation exposure. (2) Multi-drive configuration of Reactor Internal Pump (RIP) Adjustable Speed Drive (ASD): Under a conventional configuration, one ASD drives one RIP. Under a multi-drive configuration, one ASD drives two or three RIPs. As a result, four ASDs would drive ten RIPs resulting in improved economics, ease of maintenance, and a smaller footprint. (3) Debris Filter Integrated Fuel Support (DFFS): In order to eliminate the risk of debris causing a fuel leak, debris filters have been developed that will be attached at the core-support orifice. DFFS enhances reliability and improves economics.
Laser-Based Maintenance and Repair Technologies for Reactor Components
Authors: Yoda Masaki , Mukai Naruhiko , Ochiai Makoto , Tamura Masataka , Okada Satoshi , Sato Katsuhiko , Kimura Motohiko , Sano Yuji , Saito Noboru , Shima Seishi , Yamamoto Tetsuo
Published: November 2008

Stress corrosion cracking (SCC) is the major factor to reduce the reliability of aged reactor components. Toshiba has developed various laser-based maintenance and repair technologies and applied them to existing nuclear power plants. Laser-based technology is considered to be the best tool for remote processing in nuclear power plants, and particularly so for the maintenance and repair of reactor core components. Accessibility could be drastically improved by a simple handling system owing to the absence of reactive force against laser irradiation and the flexible optical fiber. For the preventive maintenance, laser peening (LP) technology was developed and applied to reactor components in operating BWR plants. LP is a novel process to improve residual stress from tensile to compressive on material surface layer by irradiating focused high-power laser pulses in water. We have developed a fiber-delivered LP system as a preventive maintenance measure against SCC. Laser ultrasonic testing (LUT) has a great potential to be applied to the remote inspection of reactor components. Laser-induced surface acoustic wave (SAW) inspection system was developed using a compact probe with a multi-mode optical fiber and an interferometer. The developed system successfully detected a micro slit of 0.5mm depth on weld metal and heat-affected zone (HAZ). An artificial SCC was also detected by the system. We are developing a new LP system combined with LUT to treat the inner surface of bottom-mounted instruments (BMI) of PWR plants. Underwater laser seal welding (LSW) technology was also developed to apply surface crack. LSW is expected to isolate the crack tip from corrosive water environment and to stop the propagation of the crack. Rapid heating and cooling of the process minimize the heat effect, which extends the applicability to neutron-irradiated material. This paper describes recent advances in the development and application of such laser-based technologies.
Development of Moisture Separator With High Performance of Steam Generator
Authors: Nishida Kenji , Mizutani Toshiyuki , Suzuta Tadahiko , Kondo Yoshiyuki , Hirao Yasuhiko
Published: November 2008

A steam generator of PWR plant has moisture separators with function to separate water from two-phase flow of water and steam. Recently, corresponding to the request for power uprating of current and/or future plants with large thermal capacity, development of high performance moisture separator, which can deal with the increasing steam flow rate, has been required. Therefore, MHI has developed the moisture separator with high performance (J Model) by conducting a verification test under actual plant operating conditions (high pressure and high temperature). The developed separator’s main features are: swirl vane hub with a small diameter, horizontal slits at riser barrel, and slots with lips at the downcomer barrel. Based on the combination of the high pressure test results and thermal hydraulic analyses, value of moisture carry over (MCO) at SG outlet was evaluated. The evaluation resulted in the MCO value of less than 0.01% (our target value: 0.1%).
A Fracture Probability Integral for HFIR Accident Analysis
Authors: Chang Shih-Jung
Published: November 2008

The state of the reactor vessel steel embrittlement as a result of neutron irradiation is reflected by its increase in ductile-brittle transition temperature (DBTT) in fracture toughness versus temperature curve. Higher DBTT implies a decrease in fracture toughness and an increase in the chance of vessel fracture in brittle fracture mode. The extent of degradation that the High Flux Isotope Reactor (HFIR) vessel has experienced is characterized by its probability of fracture that is defined as a probability count of the number of critical cracks in the reactor vessel based on a distribution of possible cracks. In this paper, the fracture probabilities under the accident pressure conditions against possible HFIR operating life are calculated for the safety analysis of the reactor vessel. Conventional methods of fracture probability calculation such as that adopted by the NRC-sponsored PRAISE CODE and the FAVOR CODE developed in this Laboratory are based on Monte Carlo simulation. Heavy computations are required. The present calculations are based on a new method of fracture probability calculation that was developed by applying direct probability integration [1]. This method offers simple and expedient procedure to obtain numerical values of fracture probability yet retains all possible features that a Monte Carlo simulation can possibly accomplish.
Influence of Temperature on the Strain Controlled Fatigue Behaviour of Alloy 690 Tube Material
Authors: Chai Guocai , Liu Ping , Frodigh Johan
Published: November 2008

Strain controlled fatigue tests for hot-extruded tube material of Alloy 690 (Sandvik Sanicro 69) have been performed at room temperature (RT) and at elevated temperature (204°C). The influences of temperature on the cyclic deformation behaviour and the fatigue life have been investigated. The influence of temperature on the fatigue life is relatively small in the temperature range investigated. However, the fatigue behaviour at elevated temperature is quite different from that at RT. A second cyclic strain hardening was observed at 204°C. The possible mechanisms have been investigated using transmission electron microscopy. Besides dislocation mechanism, the interactions between moving dislocations and stacking faults and between interstitial atoms and moving dislocations could also contribute to this secondary cyclic strain hardening. The formation of micro-twins during cyclic loading at 204°C and its influence on the cyclic stress-strain response were also discussed. Temperature affects both fatigue crack initiation and propagation behaviour. Increase in temperature promotes duplex slipping process, which causes the formation of striation.
Simulations of Thermal Performance for One- and Two-Dimensional Insulation and Aluminum Foil Fire Barriers
Authors: Sa´nchez Mauricio A. , Sutton William H. , Sa´nchez Carlos A.
Published: November 2008

Nonbearing walls made of concrete frequently include one or two-dimensional gaps between sections to allow the concrete exert expansion or contraction due to temperature transients. These section gaps require the use of a thermal fire barrier to stop a fire from spreading during a period of time. In some applications, such as seismic structures, fire barriers are large and form substructures and partial enclosures. These type of fire barriers are often manufactured by layering alternating blankets of ceramic fiber insulation with bounding thin metallic foil sheets. In this case, the barrier must meet the specifications and effectiveness given by the ASTM standard E-119. This effectiveness is determined by the requirement of maintaining structural integrity by allowing some heat release while not permitting the fire flame to pass through. Little data is available on the thermal interaction of 2-D corners and splicing the layers for large barriers. It is expected that spatial and angular effects might either degrade performance or even cause “hot spots” in a barrier wall. Therefore, a numerical simulation of the barrier is accomplished by utilizing the spectral/gray and directional/modeled data of each one of the components and by taking into account two common geometrical building shapes. This simulation analysis is done by coupling of the discrete ordinates method in radiation heat transfer and the energy equation to previously published thermophysical experimental data used as a validation of the properties for fire barrier materials. Some of the effects of directional and surface properties and radiative heat transfer in fire barrier materials have been included in the numerical model. The Fluent®-based numerical model is able to match thermal performance of previous test systems. Initial calculations suggest that a fire barrier consisting of a 2D corner geometry exposed to a fire from either side would be thermally less robust than a slab of the same characteristic aspect ratio. This approximation has shown a preferential orientation for the barrier to be positioned when a fire or other high energy source is postulated.
Protective Insulated Coating for IGSCC Mitigation in BWRs
Authors: Kim Young-Jin
Published: November 2008

A protective insulated coating (PIC) on 304 stainless steel (SS) surfaces as an IGSCC mitigation method was developed and investigated in high temperature water under various water chemistry conditions by measuring the electrochemical corrosion potential (ECP) and flow-assisted corrosion (FAC) rate. The ECP results clearly demonstrate that the PIC layer restricted oxidant transport to the metal surface, and the ECP remained at <−230 mV (SHE) in 288°C containing high oxygen (O2) and no hydrogen (H2). In this paper, long term durability of PIC layer prepared by various coating methods will be discussed.
Assessment of the MELCOR Code Against PHEBUS Experiment FPT-1 Performed in the Frame of ISP-46
Authors: Birchley J.
Published: November 2008

Calculations of PHEBUS FPT-1 are performed in the frame of CSNI International Standard Problem ISP-46. The objective of ISP-46 is to assess the capability of computer codes to provide an integral simulation of a severe accident in a Pressurised Water Reactor (PWR), from the initial stages of core heat-up to the behaviour of released fission products in the containment. The present calculations are performed using MELCOR, chosen as the main tool for assessment of Swiss nuclear plants by virtue of its whole-plant simulation capability, using modelling practices as similar as possible to those used in plant analyses. The calculations cover the bundle heat-up, degradation, the release, transport and retention of fission products and other materials, and the thermal-hydraulic and aerosol behaviour in the containment. Comparison between a best-estimate case and experiment demonstrates the code’s ability to capture most aspects of the sequence with fair to good accuracy. Uncertainties remain, particularly in regard to core degradation, and the chemistry and transport of fission products. Weaknesses of code models in these areas largely reflect limitations in current knowledge.
Study on Entrainment From High-Viscosity Falling Liquid Film of Counter-Current Two-Phase Flow in a Vertical Pipe
Authors: Koizumi Yasuo , Enari Ryou , Ohtake Hiroyasu
Published: November 2008

Behavior of a falling liquid film of highly viscous fluid in the counter-current flow condition was examined. In experiments, water and silicon oils of 500, 1000 and 3000 cSt were used as the liquid phase and air was adopted as the gas phase. A test section vertically oriented was a circular pipe of 30 mm in inner diameter and 5.4 m in length. Flooding velocities of the air-water system were well correlated with traditional correlations such as the Wallis correlation and the Kamei correlation. However, the flooding velocities of silicon films were greatly lower than the expected. When the effect of the viscosity was incorporated into the Wallis correlation, it predicted the experimental results well. The flooding in the air-silicon system was initiated by sudden growth of a wave on the film as in the air-water system although the film Reynolds number of the falling silicon film was considerably low; 0.02 ∼ 4. A considerable amount of droplets were detected a long time before the initiation of flooding in the air–silicon oil experiments as well as in the air–water experiments. The correlations tested for the onset condition of entrainment gave much higher gas velocities than the measured. Predicted velocities were rather close to the flooding velocities. The falling film thickness was predicted well by applying the universal velocity profile to the film flow over a wide range of a film Reynolds number; ranging from a water film to a 3000 cSt silicon oil film.
Analysis of a Total Flow Blockage of a Fuel Assembly in a Typical MTR Research Reactor by RELAP5/MOD3.3
Authors: Pierro Franco , Di Maro Beniamino , Adorni Martina , Bousbia Salah Anis , D’Auria Francesco
Published: November 2008

The main aim of the following study is to perform a safety analysis of the IAEA 10 MW MTR Pool Type Research Reactor [1]. The reactor is core cooled and moderated by downward forced circulation of light water. The transient herein considered is the related to partial and total obstruction of a single Fuel Assembly (FA) cooling channel. Such event constitutes a severe accident for this type of reactor since it may lead to local dryout and eventually to loss of the FA integrity. Two cases are analysed to emphasize the severity of the accident. The first one is a partial blockage of a single FA considering four different obstruction levels: 50%, 75%, 85% and 95% of nominal flow area. The second one is an extreme scenario consisting of total blockage of the same FA. This study constitutes the first step of a larger work which consists in performing a 3D simulation using the Best Estimate coupled code technique. However, as a first approach the instantaneous reactor power is derived through the point kinetic calculation.
Basic Provisions of NPP Safety Assurance in the Period of Extended Time of Russia NPP Units
Authors: Kaliberda I. V. , Soloviev L. P. , Medvedev P. G. , Slutsker V. P. , Havel R.
Published: November 2008

This paper presents an overview of Russian regulatory basis for extension of the service life of NPP units in Russia. The regulatory documents are being developed in the frames of Russian National IAEA Project. Russian regulatory approaches are in compliance with international practice, Russian principles of nuclear and radiation safety regulation conform the present Russian reality. The paper presents information about new Safety Rules and Guides. Problematic issues concerning improvement of Russian regulatory basis are presented and analyzed. Additional period of the NPP unit operation is defined by the period of operation of its critical equipment. If critical equipment has a very short additional period of operation, modernization and reconstruction of the equipment is required for extension of the NPP service life. Aspects of probabilistic safety analysis are discussed, taking into account effects of modernization. The paper also presents effective technical and administrative measures of ensuring safety during additional period of NPP units operation in Russia.
Industry-Specific Risk Models for Numerical Scoring of Hazards and Prioritization of Safety Measures
Authors: Khalil Yehia F. , Johnson Kathryn
Published: November 2008

Risk analysis consists of five cornerstones that have to be viewed holistically by risk practitioners of any organization regardless of the industry type or nature of its critical infrastructures. The cornerstones are hazard identification, risk assessment and consequence analysis, determination of risk management actions required to reduce risks to acceptable levels, communication of risk insights among the stakeholders, and continuous monitoring and verification to ensure sustained attainment of tolerable risk levels. Our primary objectives in this research are two fold: first, we compare and contrast a wide spectrum of current industry-specific and application-dependent semi-quantitative risk models. Secondly, based on the insights to be gained from the first task, we propose a framework for a robust risk-based approach for conducting security vulnerability assessment (SVA). Risk practitioners of critical infrastructures, such as commercial nuclear power plants, water utilities, chemical plants, transmission and distribution substations ... etc., could readily use this proposed approach to classify, evaluate, and prioritize risks to support allocation of resources required to ensure protection of public health and safety.
Melt Quenching by Water Injection From Below: Co-Injection of Water and Noncondensable Gas
Authors: Cho Dae H. , Page Richard J. , Abdulla Sherif H. , Anderson Mark H. , Klockow Helge B. , Corradini Michael L.
Published: November 2008

Experiments on melt quenching by the injection of water from below were conducted. The test section represented one-dimensional flow-channel simulation of the bottom injection of water into a core melt in the reactor cavity. The melt simulant was lead-bismuth alloy. For the experimental conditions employed (i.e., melt depth and water flow rates), it was found that: 1) the volumetric heat removal rate increased with increasing water mass flow rate and 2) the non-condensable gas mixed with the injected water had no impairing effect on the overall heat removal rate. Implications of these findings for ALWR ex-vessel coolability are discussed.
The Probability and Management of Human Error
Authors: Duffey Romney D. , Saull John W.
Published: November 2008

Embedded within modern technological systems, human error is the largest, and indeed dominant contributor to accident cause. The consequences dominate the risk profiles for nuclear power and for many other technologies. We need to quantify the probability of human error for the system as an integral contribution within the overall system failure, as it is generally not separable or predictable for actual events. We also need to provide a means to manage and effectively reduce the failure (error) rate. The fact that humans learn from their mistakes allows a new determination of the dynamic probability and human failure (error) rate in technological systems. The result is consistent with and derived from the available world data for modern technological systems. Comparisons are made to actual data from large technological systems and recent catastrophes. Best estimate values and relationships can be derived for both the human error rate, and for the probability. We describe the potential for new approaches to the management of human error and safety indicators, based on the principles of error state exclusion and of the systematic effect of learning.
Next Generation Nuclear Plant: High-Level Functions and Requirements
Authors: Ryskamp John M. , Harvego Edwin A. , Khericha Soli T. , Gorski Edward J. , Beitel George A. , Harrell Dennis J.
Published: November 2008

The Idaho National Engineering and Environmental Laboratory (INEEL) prepared a functions and requirements (F&R) document for the Next Generation Nuclear Plant (NGNP) Project [1] The highest-level functions and requirements for the NGNP Project design are identified in the F&R document, which establishes performance definitions to be achieved by the NGNP. The requirements for the NGNP are based on the Generation IV roadmap [2] goals. Based on these requirements, NGNP designs will be developed by commercial vendor(s). Of the six most promising Generation IV nuclear energy systems, the Very High Temperature Reactor (VHTR) is the nearest-term reactor concept that also has the capability to efficiently produce hydrogen. The U.S. Department of Energy (DOE) has selected the VHTR as the concept to demonstrate the use of nuclear power for electricity and hydrogen production without greenhouse gas emissions. This paper reviews the NGNP Project and the selection of the VHTR, then presents the NGNP functions and requirements.
Impact Analysis of a Full Model Doublet-Type Spacer Grid in the Fuel Assembly of a PWR
Authors: Park K. J. , Kwon Y. D. , Park G. J. , Song K. N. , Yoon K. H.
Published: November 2008

A spacer grid is one of the main structural components in the fuel assembly of a Pressurized Light Water Reactor (PWR). It supports fuel rods and maintains geometry from external impact loads. In ICONE10, a paper was published for impact analysis of a partial model spacer grid. A five by five grid model was established and the analysis results were discussed. At this moment, a full model with a sixteen by sixteen grid is analyzed for impact analysis. This grid has a new doublet-type shape. The finite element model with 16×16 grid cells is established for nonlinear analysis. It is composed of inner straps, outer straps, guide thimbles, etc. The model is considered the aspects of welding and the contacts between the components. The critical impact load that leads to plastic deformation is identified. Nonlinear finite element analysis is carried out by a software system called ABAQUS/EXPLICIT. The results are discussed in the context of the previous results from the partial model and incorporation of the analysis results into the design change is discussed.
Analysis of Power Spectrum Density in the PWR Fuel Assembly Using the 3-D LES Turbulent Model of Fluent 6
Authors: Kim Hyeun Min , No Hee Cheon
Published: November 2008

Trublence-induced vibration is an important concern in the design of the spacer grids of nuclear power plants. This study addresses numerically and statistically the effects of random pressures due to turbulent flows upon the fluctuating responses to the power spectrum density in one-dimensional nuclear fuel rod supported simply by the spacer grids. The dynamic forces produced by the pressure fluctuation on the rod surface are calculated by the 3-dimensional large eddy simulation turbulent model in Fluent 6 to simulate the flow field in the same as being measured empirically via pressure transducers. To acquire response to fluctuating pressure, the mode response equation of vibration is used in case of a cylindrical rod in one-dimensional case. The first modal longitudinal joint acceptance integral including a coherence function is also an important parameter affecting the displacement in the form of the r.m.s. of modal responses along with the damping ratio. The root mean square of the lateral displacement in addition to the natural frequency is studied using the PSD and the longitudinal joint acceptance integral in a fundamental mode. The random pressure PSD on the middle point of the rod shows the typical turbulence pattern: the PSD energy decreases slightly in a low frequency region, but decreases rapidly and linearly with frequency as the frequency exceeds a certain value. The PSD in a very high frequency region is obtained assuming the slope is constant in a logarithmic graph after smoothing the PSD. It turns out that the r.m.s. displacement ranges from 15 to 40 micro-meter at the maximum value using the mode response equation under the modal damping ration ranging from 0.01 to 0.05.
High Efficiency Combined Aggregate – Submerged Combustion Melter–Electric Furnace for Vitrification of High-Level Radioactive Wastes
Authors: Pioro L. S. , Pioro I. L.
Published: November 2008

It is well known that high-level radioactive wastes (HLRAW) are usually vitrified inside electric furnaces. Disadvantages of electric furnaces are their low melting capacity and restrictions on charge preparation. Therefore, a new concept for a high efficiency combined aggregate – submerged combustion melter (SCM)–electric furnace was developed for vitrification of HLRAW. The main idea of this concept is to use the SCM as the primary high-capacity melting unit with direct melt drainage into an electric furnace. The SCM employs a single-stage method for vitrification of HLRAW. The method includes concentration (evaporation), calcination, and vitrification of HLRAW in a single-stage process inside a melting chamber of the SCM. Specific to the melting process is the use of a gas-air or gas-oxygen-air mixture with direct combustion inside a melt. Located inside the melt are high-temperature zones with increased reactivity of the gas phase, the existence of a developed interface surface, and intensive mixing, leading to intensification of the charge melting and vitrification process. The electric furnace clarifies molten glass, thus preparing the high-quality melt for subsequent melt pouring into containers for final storage.
Film Boiling on Downward Quenching Hemisphere of Varying Sizes
Authors: Kim Chan S. , Suh Kune Y. , Rempe Joy L. , Cheung Fan-Bill , Kim Sang B.
Published: November 2008

Film boiling heat transfer coefficients for a downward-facing hemispherical surface are measured from the quenching tests in DELTA (Downward-boiling Experimental Laminar Transition Apparatus). Two test sections are made of copper to maintain low Biot numbers. The outer diameters of the hemispheres are 120 mm and 294 mm, respectively. The thickness of all the test sections is 30 mm. The effect of diameter on film boiling heat transfer is quantified utilizing results obtained from the test sections. The measured data are compared with the numerical predictions from laminar film boiling analysis. The measured heat transfer coefficients are found to be greater than those predicted by the conventional laminar flow theory on account of the interfacial wavy motion incurred by the Helmholtz instability. Incorporation of the wavy motion model considerably improves the agreement between the experimental and numerical results in terms of heat transfer coefficient. In addition, the interfacial wavy motion and the quenching process are visualized through a digital camera.
Vertical Liquid Film Impinged on Advanced Reactor Downcomer Wall
Authors: Kim Won J. , Yoon Sang H. , Son Hyoung M. , Suh Kune Y.
Published: November 2008

Experiments are conducted to study the vertical film width and downcomer gap effects on impingement upon the vertical flat wall simulating the core barrel in the APR 1400 (Advanced Power Reactor 1400 MWe). This work is being carried out in ALPHA (Accelerated Liquid Phase Hydrodynamics Apparatus) at the Seoul National University. The results are to be integrated into a system code THEOS (Thermomechanics Hydrodynamics Eutectics Online Simulator). At larger distances from stagnation to the injection nozzle, the gap effect falls off. The data suggest that, for impingement proximate to the stagnation point, the flow velocities can be computed by applying a turbulent correction factor to the laminar value calculated for velocities having the same pressure distribution as that present in the impingement region. To find the correction factor, one needs to obtain the liquid film thickness and breakup factors related to the turbulent flow. This is in contrast to the laminar flow, in which the liquid film thickness is uniform and the breakup is nonexistent. This work is particularly concerned with the liquid film thickness. Previous ALPHA experiments with varying nozzle diameters advocate that the correction factor is found to be a function of the injection speed rather than the Reynolds number. This paper documents on measurement of the radial liquid film thickness fluctuations and their point wise values at the same vertical height. It additionally reports on results of numerical calculation for the liquid film profiles.
Experimental Study About ONB and Subcooled Boiling Heat Transfer
Authors: Peng Changhong , Myint Aye , Guo Yun , Jia Dounan
Published: November 2008

Water subcooled boiling heat transfer were experimentally investigated in the vertical annuli with narrow gap. Subcooled flow boiling covers the region from the location where the bubbles forms on the wall to the location where the bulk temperature reaches saturated temperature. Three locations in the subcooled flow boiling have been identified by the earlier researcher as the onset of nucleate boiling (ONB), the beginning of fully developed boiling, the location where the thermodynamic quality is zero that is attained from the enthalpy balance equation. The heat transfer regions are identified as single-phase heat transfer prior to ONB, partial boiling (PB) and fully developed boiling (FDB). In this study, the available models for predicting heat transfer in the different regions and the modified correlation can predict our experimental data.