bluuebunny/crossref_metadata_2025 huggingface dataset:
Improving the Utilization of Nuclear Resources
Authors: Kunze Jay F. , Sandquist Gary M. , Sentell D. Shannon
Published: November 2008

Currently, less than one percent of the latent nuclear energy in uranium mined from the earth is eventually utilized. Nearly 90% of the uranium is discarded as “tails” from the enrichment process, and less than 7% of the nuclear energy in the fuel assemblies is actually “burned” before the assemblies are discarded for disposal in a permanent repository (i.e. Yucca Mountain). Unfortunately, there is no economic incentive in the commercial nuclear industry to remedy this wasteful utilization because the cost of the fuel assemblies consumed by the current reactor LWR fleet is only about 25% of the overall operating cost. Nuclear fuel cost represents less than 10% of the nominal average wholesale price of electricity. But, current uranium utilization and nuclear fuel economics ignore government expenditures on spent nuclear fuel disposal practices, the costs of storing both the weapons grade plutonium and the depleted uranium from the uranium enrichment operations, and time that would be required to deploy the types of reactors and facilities to effectively close the fuel cycle. This paper analyzes these issues and concludes that there must be no delay in completing needed R&D and beginning deployment of the essential new fast breeder and actinide burning reactors.
Large Eddy Simulation of Mixing in the Outlet Plenum of a High Temperature Reactor: A Benchmark Exercise
Authors: Simoneau Jan-Patrice , Champigny Julien
Published: November 2008

This paper is related to the validation of the CFD codes in the frame of High Temperature Reactor studies. A code to code benchmark involving complex unsteady flows in the outlet plenum is proposed and the Large Eddy Simulation technique is retained. The paper presents the benchmark conditions, the results obtained by the Star-cd software used in Framatome-ANP and the comparison with the Trio-U code (from CEA) and the literature. It presents the advantages of such fine unsteady calculations and mainly highlights the coherence between both analyses.
Experimental Equipment Design and Testing of the DEOX Process
Authors: Bateman K. J. , Westphal B. R. , Stawicki M. A.
Published: November 2008

Several technologies exist or are under development for treating spent oxide nuclear fuels. Foremost among these are aqueous and pyrochemical reprocessing which both involve a head-end fuel dissolution step. This dissolution step may potentially be shortened if it is combined with a fuel decladding and size reduction process. Declad and Oxidize (DEOX), an advanced head-end processing concept, is being assessed at Argonne National Laboratory to meet these decladding and size reduction needs via the oxidation of UO2 to U3O8. This work is being done in collaboration with Oak Ridge National Laboratory. The primary objectives of the DEOX process are to generate suitable feed material for these two fuel treatment processes and to collect information about the behavior of spent fuel during DEOX processing. Specifically, DEOX is intended to remove the spent fuel from its cladding, while avoiding oxidation of the cladding that would contaminate the product. An additional goal is to obtain a product particle size distribution between 45μm to 4mm. Data will be collected on the extent of fuel oxidation and on the volatilization of fission products. The experimental apparatus used to perform these experiments is described in this report along with preliminary test results.
Nuclear Steam Dryers Optimized by Computational and Experimental Fluid Dynamics
Authors: Fadda Dani , Taylor David , Greis Ingemar , Kornfeldt Hans , Sjo¨vall Heikki
Published: November 2008

Secondary separators (steam dryers) are used in boiling water reactor vessels as the final stage of moisture removal to provide the highest possible quality steam. Over the past decades many distinct designs were successfully created and put into service. New computational fluid dynamics (CFD) analysis tools and advanced test facilities have been utilized to design higher capacity dryers. These tools have also been utilized for upgrading existing steam dryers and designing new compact dryers, resulting in improved plant efficiency, and better control of radioactivity within the steam loop. Such upgrades are especially attractive to nuclear plant operators and utilities in their evaluation of plant performance, power uprates, and to justify re-licensing efforts. A recent study pertaining to a steam dryer upgrade is discussed for two identical BWRs. The challenges, the analysis tools and considerations from the operator’s perspective are highlighted.
Derivations of Correlation and Model of Transition Boiling Heat Transfer: Liquid-Solid Contact Model for Transition Boiling Heat Transfer in High Wall Superheat
Authors: Ohtake Hiroyasu , Koizumi Yasuo
Published: November 2008

The transition boiling heat transfer was examined through derivations of correlation and model. The correlation of liquid-solid contact fraction in transition boiling was derived by focusing on the dimensionless wall temperature, Θ = (Tw−TCHF) / (TMHF−TCHF); the following correlation was obtained from experimental results for water, R-113 and LN2: Γ = 1.000 − 0.9120Θ − 0.1343Θ2, where qtb = qCHFΓ+qMHF(1−Γ). In the present model, considering a pseudo liquid-solid contact right after detachment of a bubble from liquid-vapor interface resulted from Rayleigh-Taylor instability, transient heat conduction was analyzed in it three dimensionally. Liquid-solid contact time and area were given by the present correlation of the fractions of liquid-solid contact: τcontact = 0.3Γt, Awet = 15 × 15 Γa, Γ = Γt Γa = 1.000 − 0.9120Θ − 0.1343Θ2, where Θ = (Tw−TCHF)/(TMHF−TCHF). The prediction by the present model was in agreement with the present experimental data for water.
Asset Ranking Manager: Ranking Index of Components
Authors: Maloney Steven M. , Morgan Tomas A. , Engle Adam M.
Published: November 2008

The Ranking Index of Components (RIC) is an Asset Reliability Manager (ARM), which itself is a Web Enabled front end where plant database information fields from several disparate databases are combined. That information is used to create a specific weighted number (Ranking Index) relating to that components health and risk to the site. The higher the number, the higher priority that any work associated with that component receives. ARM provides site Engineering, Maintenance and Work Control personnel with a composite real time / (current condition) look at the components “risk of not working” to the plant. Information is extracted from the existing Computerized Maintenance management System (CMMS) and specific site applications and processed nightly. ARM helps to ensure that the most important work is placed into the workweeks and the non value added work is either deferred, frequency changed or deleted. This information is on the Web, updated each night, and available for all employees to use. This effort assists the work management specialist when allocating limited resources to the most important work. The use of this tool has maximized resource usage, performing the most critical work with available resources. This has also helped start a review of PM’s that could be deferred or extended earlier in the work management process. The ARM numbers are valued inputs into work scoping for the workweek managers. System and Component Engineers are using ARM to identify the components that are at “risk of failure” and therefore should be place into the appropriate work week schedule. Based on the insights gained from the initial use of this tool, enhancements are being contemplated that will further increase the value of this tool for work management.
Performance Measurements of Solid-Oxide Electrolysis Cells for Hydrogen Production From Nuclear Energy
Authors: O’Brien J. E. , Stoots C. M. , Herring J. S. , Lessing P. A. , Hartvigsen J. J. , Elangovan S.
Published: November 2008

An experimental study has been completed to assess the performance of single solid-oxide electrolysis cells operating over a temperature range of 800 to 900°C. The experiments were performed over a range of steam inlet partial pressures (2.3–12.2 kPa), carrier gas flow rates (50–200 sccm), and current densities (−0.75 to 0.25 A/cm2) using single electrolyte-supported button cells of scandia-stabilized zirconia. Steam consumption rates associated with electrolysis were measured directly using inlet and outlet dewpoint instrumentation. Cell operating potentials and cell current were varied using a programmable power supply and monitored continuously. Values of area-specific resistance and thermal efficiency are presented as a function of current density. Cell performance is shown to be continuous from the fuel-cell mode to the electrolysis mode of operation. The effects of steam starvation and thermal cycling on cell performance parameters are discussed.
A Stability Investigation of Two-Dimensional Surface Waves on Evaporating, Isothermal or Condensing Liquid Films: Part II — A Universal Linear Stability Formulation
Authors: Ye Xuemin , Yan Weiping , Li Chunxi
Published: November 2008

When liquid film is under evaporating or condensing conditions, the flow stability is clearly different to that under isothermal condition due to thermal non-equilibrium effect at interface, especially under lower Reynolds number. The universal linear temporal and spatial stability formulations of the two-dimensional surface waves on evaporating or isothermal or condensing liquid films are established in present paper with the collocation method based on the boundary layer theory and complete boundary conditions. The models include the effects of Reynolds number, thermocapillarity, inclination angle, liquid property, evaporation, isothermal or condensation. The effects of above factors are investigated with the neutral stability curves at different Reynolds numbers, and stabilities characteristics are fully indicated in theory for evaporating or condensing films.
A Stability Investigation of Two-Dimensional Surface Waves on Evaporating, Isothermal or Condensing Liquid Films: Part I — Thermal Non-Equilibrium Effects on Wave Velocity
Authors: Li Chunxi , Ye Xuemin
Published: November 2008

When the liquid film is in the process of evaporation or condensation, the interfacial thermal non-equilibrium conditions are evidently different from that of isothermal film, and this difference will affect the flow stability and wave velocity of thin liquid films. The temporal stability equation of the two-dimensional traveling waves of evaporating or condensing liquid films falling down an inclined wall is established based on the Prandtl boundary layer theory and complete boundary conditions. The effects on wave velocity of evaporating, isothermal and condensing states, thermocapillarity, Reynolds number, fluid property and inclined angle are discussed, and are compared in different Reynolds numbers.
New Decladding System in the Reprocessing Process for Fast Reactor Fuel
Authors: Yamada Seiya , Washiya Tadahiro , Takeuchi Masayuki , Suganuma Takashi , Aose Shinichi
Published: November 2008

As a feasibility study on commercialized fast reactor cycle systems, Japan Nuclear Cycle Development Institute (JNC) has been developing the fuel decladding system for the dry reprocessing process (Oxide electrowinning process) and aqueous reprocessing process. In the oxide electrowinning process, the spent fuel pins should be crushed and grained for quick dissolution in the molten salt at electrolyzer. For the fuel crushing method, the heat treatment method such as a voloxidation might not be applicable due to the high plutonium content such as a fast reactor fuel. Therefore, JNC proposed innovative decladding system with mechanical decladding devices. The decladding system consists of the fuel pin crushing process, the hull separation process and the fuel recovery process. In the fuel pin crushing process, a disassembled spent fuel pin is crushed and powdered by the mechanical shredding device, and in the following process, the crushed product with hull and fuel fragments is separated to fuel element and metallic element (mainly hull fragments) by the magnetic separator. Then the induction-melting device recovers small amount of fuel element contaminated in the metallic element. The recovered fuel element is fed back to the electrolyzer. In this paper, the outline of this new decladding system and its innovative devices will be descried.
Integrity of Radioactive Waste Packages at the Yucca Mountain Repository
Authors: Sandquist Gary , Biaglow Andrew , Huber Michael , Jagmin Chad
Published: November 2008

Several of the important physical and chemical processes that impact the integrity of the radioactive waste packages planned for disposal at the proposed Repository at Yucca Mountain are examined. These processes are described by the aerodynamic, thermodynamic, and chemical interactions associated with the waste packages. The effects of chemical corrosion, mechanical erosion, temperature distributions throughout the repository environs, interactions of air, water, and solid particles, and radiological and biological influences are addressed.
Development of Quantitative Analytical Procedures on Two-Phase Flow in Tight-Lattice Fuel Bundles for Reduced-Moderation Light-Water Reactors
Authors: Takae Kazuyuki , Yoshida Hiroyuki , Kureta Masatoshi , Tamai Hidesada , Ohnuki Akira , Akimoto Hajime
Published: November 2008

R&D project to investigate thermal-hydraulic performance in tight-lattice rod bundles for Reduced-Moderation Water Reactor (RMWR) has started at Japan Atomic Energy Research Institute (JAERI) in collaboration with power companies, reactor vendors, universities since 2002. The RMWR is a light water reactor which a higher conversion ratio more than one can be expected. In order to attain this higher conversion ratio, triangular tight-lattice fuel bundles which gap spacing between each fuel rod is around 1 mm are required. As for the thermal design of the RMWR core, conventional analytical methods are no good because the conventional composition equations can not predict the RMWR core with high accuracy. Then, development of new quantitative analytical procedures was carried out. Those analytical procedures are constructed by model experiments and advanced two-phase flow analysis codes. This paper describes the results of the model experiments and analytical results with the developed analysis codes.
Glovebox Dismantling Activities and Decommissioning Plan for Plutonium Fuel Fabricating Facility
Authors: Kitamura Akihiro , Okada Takashi , Asazuma Sinichiro , Uematsu Shinichi , Ishibashi Takashi
Published: November 2008

The gloveboxes and process equipment used at plutonium fuel handling facilities have had to be replaced due to deterioration or the need to make changes. So far, their removal and replacement has taken place more than 30 times in Plutonium Fuel Center, Japan Nuclear Cycle Development Institute (JNC). In most recent dismantling activities, we removed four giant gloveboxes (total size, 110 cubic meters) which possessed equipment to recover plutonium from mixed oxide (MOX) fuel scraps. We have implemented a number of procedural improvements in dismantling activities and collected various kinds of data, including type and amount of primary and secondary waste from dismantling, relation between waste volume and work force, etc. Plutonium Fuel Fabricating Facility (PFFF) is one of the three plutonium fuel handling facilities in Plutonium Fuel Center, JNC. Its final mission to produce MOX fuels for the advanced thermal reactor “Fugen” Nuclear Power Station was successfully finished in 2002. Then, we started preparatory activities to draw up a Deactivation & Decommissioning (D&D) plan for this facility and to construct a database with the experimental data of glovebox dismantling activities acquired in the past thirty years. The D&D schedule for this facility can be broken down into three phases. Phase 1 (to 2010): Stabilize all the special nuclear materials in the facility and ship them from the facility. Establish new and effective decontamination and volume reduction technologies in order to improve existing methods. Phase 2 (2010–2015): Apply the abovementioned technologies to some of the glovebox dismantling activities and confirm their adaptability for the project. Draw up a detailed D&D plan which meets to various regulations. Phase 3 (2015–2020): Dismantle all the remaining gloveboxes in the facility and promote research and development of D&D technologies for future projects. Decontaminate inner surfaces of the building in order to reuse the building as a waste storage facility.
A Review of Alloy 600 Cracking in Operating Nuclear Plants Including Alloy 82 and 182 Weld Behavior
Authors: Bamford Warren , Hall John
Published: November 2008

Service induced cracking in Alloy 600 has been known for a long time, having been first observed in the 1980’s in steam generator tubing and small bore piping, and later, in 1991, in reactor vessel control rod drive mechanism (CRDM) head penetrations. Other than steam generator tubing, which cracked within a few years of operation, the first Alloy 600 cracking was in base metal of Combustion Engineering small bore piping, followed closely by CE pressurizer heater sleeves. The first reactor vessel CRDM penetrations (base metal) to crack were in France, US plants found CRDM cracking several years later. Three plants have discovered weld metal cracking at the outlet nozzle to pipe weld region. This was the first known weld metal cracking. This paper will chronicle the development of service-induced cracking in these components, and compare the behavior of welds as opposed to base metal, from the standpoint of time to crack initiation, growth rate of cracks, and their impact on structural integrity. In addition, a discussion of potential future trends will be provided.
PWSCC of Bottom Mounted Instrument Nozzles at South Texas Project
Authors: Thomas Steven
Published: November 2008

Primary water stress corrosion cracking (PWSCC) susceptibility of bottom mounted instrument (BMI) nozzles in the South Texas Project reactor vessels was considered low due to the Tcold temperature of the bottom head. During a routine bare metal visual inspection of the Unit 1 bottom head in April 2003 small boric acid deposits were identified at the location where two of the BMI nozzles penetrated the bottom surface of the vessel head. Subsequent nondestructive examination showed three cracks in one of the nozzles two crack in the second nozzle and no cracks in the other 56 nozzles. The nondestructive examinations showed that both leaking nozzles had an axial crack in the nozzle base metal extending from above the J-weld to below the J-weld. One of these cracks extended nearly through-wall while the other did not extend through wall. A boat sample removed from one of the leaking nozzles confirmed the presence of a PWSCC crack in the nozzle base metal and the presence of welding defects at the tube to J-weld interface which could be responsible for initiating the cracks. This paper summarizes the field inspections and laboratory investigations and briefly descries the repair method. See References 1 and 2 for additional information.
Program Outline of Seismic Fragility Capacity Tests on Nuclear Power Plant Equipment
Authors: Iijima Toru , Abe Hiroshi , Fujita Takafumi
Published: November 2008

A seismic probabilistic safety assessment (PSA) is an available method to evaluate residual risk of nuclear plant that is designed with definitive seismic design condition. Seismic fragility capacity data are necessary for seismic PSA, but we don’t have sufficient data of active components of nuclear plants in Japan. This paper describes a plan of seismic fragility capacity tests on nuclear power plant equipment. The purpose of those tests is to obtain seismic fragility capacity of important equipment from a safety design point of view. And the equipment for the fragility capacity tests were selected considering effect on core damage frequency (CDF) that was evaluated by our preliminary seismic PSA. Consequently horizontal shaft pump, electric cabinets, Control Rod Drive system (CRD system) of BWR and PWR plant and vertical shaft pump were selected. The seismic fragility capacity tests are conducted from phase-1 to phase-3, and horizontal shaft pump and electric cabinets are tested on phase-1. The fragility capacity test consists of two types of tests. One is actual equipment test and another is element test. On actual equipment test, a real size model is tested with high-level seismic motion, and critical acceleration and failure mode are investigated. Regarding fragility test phase-1, we selected typical type horizontal shaft pump and electric cabinets for the actual equipment test. Those were Reactor Building Closed Cooling Water (RCW) Pump and eight kinds of electric cabinets such as relay cabinet, motor control center. On the test phase-1, maximum input acceleration for the actual equipment test is intended to be 6G-force. Since the shaking table of TADOTSU facility didn’t have capability for high acceleration, we made vibration amplifying system. In this system, amplifying device is mounted on original shaking table and it moves in synchronization with original table. The element test is conducted with many samples and critical acceleration, median and deviation are evaluated. For the element test we selected bearings and liner rings of horizontal shaft pump and about thirty kinds of electric parts. The actual equipment test of phase-1 started from September 2003, and it will finish March 2004. The test phase-2 for CRD system and phase-3 for vertical shaft pump will start following test phase-1.
Study on Simulation, Control and Online Assistance Integrated System of 10MW High Temperature Gas-Cooled Test Reactor
Authors: Luo Shaojie , Shi Lei , Zhu Shutang
Published: November 2008

In order to provide a convenient tool for engineering designed, safety analysis, operator training and control system design of the high temperature gas-cooled test reactor (HTR), an integrated system for simulation, control and online assistance of the HTR-10 has been designed and is still under development by the Institute of Nuclear Energy Technology (INET) of Tsinghua University in China. The whole system is based on a network environment and includes three subsystems: the simulation subsystem (SIMUSUB), the visualized control designed subsystem (VCDSUB) and the online assistance subsystem (OASUB). The SIMUSUB consists of four parts: the simulation calculating server (SCS), the main control client (MCC), the data disposal client (DDC) and the results graphic display client (RGDC), all of which can communicate each other via network. The SIMUSUB is intended to analyze and calculate the physical processes of the reactor core, the main loop system and the stream generator, etc., as well as to simulate the normal operation and transient accidents, and the result data can be graphically displayed through the RGDC dynamically. The VCDSUB provides a platform for control system modeling where the control flow systems can be automatically generated and graphically simulated. Based on the data from the field bus, the OASUB provides some of the reactor core parameter, which are difficult to measure. This whole system can be used as an educational tool to understand the design and operational characteristics of the HTR-10, and can also provide online supports for operators in the main control room, or as a convenient powerful tool for the control system design.
The Design Features of Integrated Modular Water Reactor (IMR)
Authors: Kanagawa Takashi , Goto Masashi , Usui Shuji , Suzuta Tadahiko , Serizawa Akimi , Kunugi Tomoaki , Matsumura Tetsuo , Yamauchi Toyoaki , Itoh Gen
Published: November 2008

Small-to-medium-sized (300–600MWe) reactors are required for the electric power market in the near future (2010–2030). The main theme in the development of small-to-medium-sized reactor is how to realize competitive cost against other energy sources. As measures to this disadvantage, greatly simplified and downsized design is needed. From such point of view, Integrated Modular Water Reactor (IMR), which electric output power is 350 MWe, adopts integrated and high temperature two-phase natural circulation system for the primary system. In this design, main coolant pipes, a pressurizer, and reactor coolant pumps are not needed, and the sizes of a reactor vessel and steam generators are minimized. Additionally, to enhance the economy of the whole plant, fluid system, and Instrumentation & Control system of IMR have also been reviewed to make them simplest and smallest taking the advantage of the IMR concept and the state of the art technologies. For example, the integrated primary system and the stand-alone direct heat removal system make the safety system very simple, i.e., no injection, no containment spray, no emergency AC power, etc. The chemical and volume control system is also simplified by eliminating the boron control system and the seal water system of reactor coolant pumps. In this paper, the status of the IMR development and the outline of the IMR design efforts to achieve the simplest and smallest plant are presented.
Simulation of Containment Thermal-Hydraulic Transients in Different Experimental Facilities With the CONTAIN Code
Authors: Kljenak Ivo , Mavko Borut
Published: November 2008

Three experiments on containment phenomena, which were performed in the TOSQAN, PHEBUS and KAEVER integral experimental facilities, were simulated with the lumped-parameter computer code CONTAIN. In the experiments, which were performed at similar pressure and temperature conditions, steam condensation played a dominant role. The simulation results were used to assess the ability of the CONTAIN code to simulate the transient thermal-hydraulic behavior of containment vessels with different arrangements of heat structures.
Analysis of Justification for Extension of Operation of Russian NPP With VVER in Aspect of Pressure Vessel Radiation Embrittlement
Authors: Borodkin Gennady I. , Kaliberda Inna V. , Khrennikov Nikolay N.
Published: November 2008

The reactor pressure vessel (RPV) is one of the most key components of equipment of Nuclear Power Plant (NPP) with VVER type reactors. The issue of the plant lifetime extension is actual especially in aspect of RPV embrittlement. An analysis of influence of RPV embrittlement on the evaluation of RPV lifetime and by means of that on an opportunity of operation extension beyond the design life within the framework of the license renewal is presented. The Regulatory Standards Base, existing in Russia and guiding a process of justification of RPV lifetime, is surveyed. SEC NRS, realizing a scientific and engineering support of The Federal Nuclear and Radiation Safety Authority of Russia (Gosatomnadzor of Russia), takes part in such Regulatory Standard development. The role of SEC NRS in this process has been shown. Besides, the activity of SEC NRS in analysis of justification of RPV service lifetime prolongation by expertise of the utility documentation within the framework of license process is demonstrated. Different measures aimed at the management of RPV steel degradation and thus providing the prolongation of RPV operation of the first generation of VVER-440 are considered. Among such measures the thermal annealing and regulation of radiation load of pressure vessels are reviewed. A special attention is drawn to a task of RPV embrittlement prediction in the extended operation period. SEC NRS approach to the analysis of efficiency of such measures and their influence on justification of RPV lifetime is demonstrated. The conservative approach based on safety margins concept is the key aspect in the analysis. The results of expert analysis performed in SEC NRS particularly towards the application of the first generation of VVER-440 reactors are presented. On the basis of such analyses, the priority issues, solution of which considerably increases a confidence of justification of VVER-440 and VVER-1000 RPV lifetime, are assigned. Evaluation of RPV surveillance specimen and template programs is presented. The topicality of RPV weld metal embrittlement tests of specimens, obtained from templates cut out from operated reactors and trepans of RPV of shut-down VVER (for example, Greifswald NPP) is discussed. The results of such tests are indispensable both for replenishing database with reliable experimental data, and for justifications of RPV service life prolongation.
Do All RPV Head Penetration Leaks Have the Potential to Cause Head Wastage?
Authors: Roy Chandra M. , Fessler John R. , Foulds Jude R. , Latanision Ronald M. , Taylor David E.
Published: November 2008

The identification of the PWSCC (Primary Water Stress Corrosion Cracking) mechanism responsible for leakage from an Alloy 600 nozzle tube of a PWR RPV (pressurized water reactor reactor pressure vessel) head more than a decade ago led to a significant body of research into understanding the phenomenon and to development of bases for safely managing primary pressure boundary integrity. However, the relatively recent experience at Davis-Besse, wherein penetration leakage resulted in significant vessel head material wastage, led to the heretofore unconsidered issue of vessel failure risk due to head rupture. This paper addresses, in preliminary fashion, one key input to determining the risk associated with head material wastage and potential rupture — the local environmental and fluid conditions associated with a range of leak paths. The results indicate a need for rigorous prediction of fluid conditions for a range of leak situations to help establish criteria for addressing penetration leaks.
SAMA Analysis in Support of License Renewal Application for the Fort Calhoun Station
Authors: Schneider Raymond E. , Hackerott Harold A. , Jacob Mathew C.
Published: November 2008

The Omaha Public Power District (OPPD) owns and operates the Fort Calhoun Station Unit 1 (FCS), a single unit Pressurized Water Reactor (PWR) plant. This plant was licensed by the U. S. Nuclear Regulatory Commission (NRC) to operate at a core power level of 1500 MWt for a period of 40 years. In 2002 OPPD submitted a license renewal application to the NRC and recently, this application for a license renewal for a period of 20 more years was approved by the NRC. In order to secure the license renewal, the utility evaluated Severe Accident Mitigation Alternatives (SAMAs) and identified potential risk significant and cost-beneficial plant changes. This process consisted of a review of past industry changes obtained via past SAMA analyses and advanced reactor design efforts as well as a review of FCS PSA insights. While most of the industry-based SAMAs were screened out, the internally focused changes were mostly found to be risk significant and cost effective. This paper presents the SAMA analysis performed in support of the FCS license renewal application and discusses the insights from this analysis.
The Development of Mobile Melt-Dilute Technology for the Treatment of Former Soviet Union Research Reactor Fuel
Authors: Sell D. A. , Howden E. A. , Allen K. J. , Marsden K. , Westphal B. R. , Peacock H. B. , Iyer N. C. , Fisher D. L. , Adams T. M. , Sindelar R. L.
Published: November 2008

United States Government funded national security nuclear non-proliferation projects have historically focused on power reactor spent fuel assemblies that contain weapons usable materials. More recently concern and emphasis have been focused on the spent fuel located at the many research reactor facilities spread throughout the Former Soviet Union. The need exists for a mobile system that can be deployed at these research reactors for the purpose of ensuring that the nuclear materials cannot be used for weapons development. On-site application of the Mobile Melt-Dilute (MMD) process offers an economical method for converting weapons usable Former Soviet Union high enriched uranium research reactor fuel to a safe and secure low enriched uranium ingot. The process will generate little waste and will be performed in a sealed canister that will contain all off-gas products generated during the melting process, eliminating the need for an off-gas treatment system. The process is modular, reusable, and readily portable to a desired reactor site or storage location. The storage canisters containing the melted ingot can be configured for compatibility with the fuel storage technologies currently available or returned to Russia for reprocessing under the Russian Research Reactor Fuel Return Program. The objective of the MMD Project is to develop the mobile melt and dilute technology in preparation for active deployment at Russian built and fueled research reactors. The project has just completed conceptual design and is beginning proof of principle experiments and integrated prototype design of the furnace and canister.
ARTIST: Introduction and First Results
Authors: Gu¨ntay S. , Suckow D. , Dehbi A. , Kapulla R.
Published: November 2008

ARTIST (Aerosol Trapping In a Steam Generator) is a seven-phase international project (2003–2007) which investigates aerosol and droplet retention in a model steam generator under dry, wet and accident management conditions, respectively. The test section is comprised of a scaled steam generator tube bundle consisting of 270 tubes and 3 stages, one 1:1 separator unit, and one 1:1 dryer unit. As a prelude to the ARTIST project, four tests are conducted in the ARTIST bundle within the 5th EU FWP SGTR. These first tests address aerosol deposition phenomena on two different scales: near the tube break, where the gas velocities are sonic, and far away from the break, where the flow velocities are three orders of magnitude lower. With a dry bundle and the full flow representing the break stage conditions, there is strong evidence that the TiO2 aerosols used (AMMD 2–4 μm, 32 nm primary particles) disintegrate into much smaller particles because of the sonic conditions at the break, hence promoting particle escape from the secondary and lowering the overall DF, which is found to be between 2.5 and 3. With a dry bundle and a small flow reproducing the far-field velocities, the overall bundle DF is of the order of 5, implying a DF of about 1.9 per stage. Extrapolating the results of the dry tests, it turns out that for steam generators with 9 or more stages, it is expected that substantial DF’s could be achieved when the break is located near the tube sheet region. In addition, better decontamination is expected using more representative proxies of severe accident aerosols (sticky, multicomponent particles), a topic which is yet to be investigated. When the bundle is flooded, the DF is between 45 and 5740, depending on the mass flow rate, the steam content, and the water submergence. The presence of steam in the carrier gas and subsequent condensation inside the broken tube causes aerosol deposition and blockages near the break, leading to an increase in the primary pressure. This has implications for real plant conditions, as aerosol deposits inside the broken tube will cause more flow to be diverted to the intact tubes, with a corresponding reduction in the source term to the secondary.
An Experimental Study on Coolability of Particulate Core-Metal Debris Bed With Oxidization
Authors: Aoki H. , Sugiyama K. , Su G. H. , Sakashita H. , Kojima Y.
Published: November 2008

The coolability characteristics of a eutectic metal debris bed, which has a low melting point, have been assessed from the viewpoint of in-vessel retention. Ag-50wt%Zircaloy eutectic alloy, constituents of which are principal metals constituting a reactor core, was chosen as the particulate core-metal debris in the present experimental study. Ag particulates and shortly chopped Zircaloy tubes were melted by induction heating, and then the molten Ag-50wt%Zircaloy was dropped into a water pool with 80cm in depth, resulting in debris particulates. The upper interface temperature of the particulate metal debris, which was electrically heated to simulate decay heat, ranged from 500°C to 900°C, and the temperature of a water layer at the bottom side was kept at 100°C. The heat flux and the temperature at the upper interface were measured for 30 minutes. Under the wet condition where heat conducted from debris particulates to a water layer produces steam, it is confirmed that the particulate eutectic-alloy debris bed is oxidized and the perfectly oxidized parts with thin cross-section are cracked into pieces. The mass median diameter measured after each run clearly decreases compared with that measured before the run. Sieving after each run shows that an amount of small particulates less than 1mm, which is expected to produce a high capillary force, drastically increases due to oxidation. The present experimental results therefore show that the particulate eutectic-alloy debris bed exposed to a vapor atmosphere is oxidized in a short time period and consequently could be cooled because of a capillary force of small particulates produced by oxidization.
Visualization and Measurement of Steady Streaming Induced in Oscillating Flow
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.
“OMIRIS®” Educational Tool
Authors: Dumont Xavier , Artus Jean-Claude , Gonin Miche`le , Bidard Franc¸oise , Hickman Bethleem
Published: November 2008

The biological effects of ionizing radiations are one of the most important issues for workers exposed to these radiations in nuclear plants. To deliver information to workers about this topic, the “Utility medical work officers” wanted an educational tool. This tool named “OMIRIS” was prepared under the authority of the FE3R (Professors federation in radiology, radiobiology and radioprotection) and the industrial nuclear partners in France (ANDRA, AREVA/COGEMA & FRAMATOME-ANP, CEA & EDF). The use of animation techniques helps to present in a simple way this complex topic using an interactive, pleasant and comprehensible form.
Application of a Piping Failure Database to Some Nuclear Safety Issues in Korea
Authors: Choi Sun Yeong , Choi Young Hwan , Chung Yeon Ki
Published: November 2008

The purpose of this paper is to introduce some experience from the application of a piping failure database to some nuclear safety issues in Korea. We used both a Korean piping failure database and the OPDE (OECD Piping Data Failure Exchange) database for the following issues in Korea, the inservice inspection (ISI) including the augmented inspection and the risk-informed ISI (RI-ISI), the screening criteria of the leak before break (LBB) application, the plant aging management in the periodic safety review (PSR), and the re-evaluation of the initiating event frequencies for the probabilistic safety assessment (PSA). The following are the results obtained from this study. (1) The rupture frequency of the main feedwater system for the PSA purpose was estimated as 4.54E−3/cr.yr, while that in the United States was 3.43E−3/cr.yr. (2) The piping failure frequencies of the five failed piping systems such as the CVCS, the primary sampling system, the main feedwater system, the essential service water system, and the CANDU purification system with the pipe population data in Korea were estimated in the range from 6.08E−5/cr.yr to 1.15E−3/cr.yr. (3) The evaluation of the screening criteria for the LBB application showed that detailed analyses are required because the pressurizer surge line, the safety injection system, and the shutdown cooling system were susceptible to thermal fatigue, and the main feedwater system was susceptible to flow accelerated corrosion (FAC) and thermal fatigue. (4) Based on the piping failure event data, augmented inspections are required for the piping systems susceptible to thermal fatigue and FAC in Korea. (5) The insight obtained from the piping failure database was used in the determination of the ranking for the plant aging management in the PSR of the Kori Unit 1.
Vorticity Modification in a Turbulent Channel Flow by Microbubble Injection
Authors: Gutierrez-Torres Claudia del C. , Jimenez-Bernal Jose A. , Dominguez-Ontiveros Elvis E. , Hassan Yassin A.
Published: November 2008

Investigation of the drag reduction phenomenon has been carried out for several years. Several techniques to reduce the drag have been applied and researched for a number of years. Microbubbles injection within a turbulent boundary layer is one method utilized to achieve reduction of drag. In this work, the effects of the presence of microbubbles in the boundary layer of a turbulent channel flow are discussed.
Thermal Fatigue in Mixing Tees: Status and Justification of French Assessment Method
Authors: Faidy Claude
Published: November 2008

Following the CIVAUX 1 incident of a leak on RHR system, EDF has developed a step by step procedure to screen and analyse similar locations: mixing tees with long duration at high ΔT between the 2 fluids. The paper presents the procedure, the background of the methodology and few R&D work that support this procedure. The procedure is based on: - screening criteria on maximum DT and minimum duration; - screening criteria without any duration consideration, only DT and material; - a simplified and conservative estimation of a usage factor; - a detailed analysis of usage factor and crack growth rate, based on specific data collection of operating transients. Around that procedure EDF launched an R&D program on fatigue curves and fatigue reduction factors for high cycle fatigue. The procedure is compared with field experience and recent R&D fatigue tests.
Wavelet Analysis on Eddy Structure in Micro Bubble Two Phase Flow Using PIV
Authors: Zhen Ling , Gutierrez-Torres Claudia del Carmen
Published: November 2008

The question of “where and how the turbulent drag arises” is one of the most fundamental problems unsolved in fluid mechanics. However, the physical mechanism responsible for the friction drag reduction is still not well understood. Over decades, it is found that the turbulence production and self-containment in a boundary layer are organized phenomena and not random processes as the turbulence looks like. The further study in the boundary layer should be able to help us know more about the mechanisms of drag reduction. The wavelet-based vector multi-resolution technique was proposed and applied to the two dimensional PIV velocities for identifying the multi-scale turbulent structures. The intermediate and small scale vortices embedded within the large-scale vortices were separated and visualized. By analyzing the fluctuating velocities at different scales, coherent eddy structures were obtained and this help us obtain the important information on the multi-scale flow structures in the turbulent flow. By comparing the eddy structures in different operating conditions, the mechanism to explain the drag reduction caused by micro bubbles in turbulent flow was proposed.
Effect of Non-Condensable Gas on Steam Injector
Authors: Kawamoto Yujiro , Abe Yutaka , Iwaki Chikako , Narabayashi Tadashi , Mori Michitsugu , Ohmori Shuichi
Published: November 2008

Next-generation reactor systems have been under development aiming at simplified system and improvement of safety and credibility. A steam injector has a function of a passive pump without large motor or turbo-machinery, and has been investigated as one of the most important component of the next-generation reactor. Its performance as a pump is depends on direct contact condensation phenomena between a supersonic steam and a sub-cooled water jet. Although non-condensable gases are well known for reducing heat transfer, the effect of the non-condensable gas on the condensation of supersonic steam on high-speed water jet has not been cleared. The present paper reports the results of an experimental study of condensation of supersonic steam around turbulent water jet with model steam injector made by transparent plastic. The visual observation carried out by using high-speed camera. The non-condensable gas effect on the pump performance and flow characteristics are clarified by the image processing technique for the jet shape and gas-liquid interface behavior.
Status of the ANS Low Power and Shutdown PRA Standard
Authors: Kiper Kenneth L.
Published: November 2008

The Low Power & Shutdown (LPSD) PRA Standard (ANS-58.22) is currently being drafted by a Writing Group under the auspices of the Risk Informed Standards Committee of the American Nuclear Society. The Writing Group includes representatives from nuclear utilities, US Nuclear Regulatory Commission, national laboratory, university, and consultants with substantial experience producing LPSD PRAs. This draft standard is scheduled to be released for public comment in the second quarter of 2004, with publication by the end of 2004. This paper presents the current status of this standard in preparation for its public release.
Qualitative Risk Assessment in the ANS LPSD PRA Standard
Authors: Burchill William E. , Julius Jeffrey A. , Drouin Mary , Kiper Kenneth L.
Published: November 2008

This paper describes a Qualitative/Screening Methodology included in the draft ANS Low Power and Shutdown (LPSD) PRA (Probabilistic Risk Assessment) Standard. The screening methodology can be used to eliminate certain specified shutdown POSs (Plant Operating States) from requiring further quantitative risk assessment in the context of a specified application based on demonstrating that their risk is lower than some predetermined limiting value. The paper also describes a methodology for qualitative risk assessment (QRA) tools that are used to support the screening process. The paper outlines the bases of both methodologies.
Upgrading the Operational Safety of Nuclear Power Plants Through the TACIS Nuclear Safety Assistance Programme
Authors: Bie`th M. , Ahlstrand R. , Rieg C. , Trampus P.
Published: November 2008

The European Union’ TACIS programme was established for the New Independent States since 1991. One priority for TACIS funding is nuclear safety. The European Commission has made available a total of € 944 million for nuclear safety programmes covering the period 1991–2003. The TACIS nuclear safety programme is devoted to the improvement of the safety of Soviet designed nuclear installations in providing technology and safety culture transfer. The Joint Research Center (JRC) of the European Commission is carrying out works in the following areas: • On-Site Assistance for TACIS Nuclear Power Plants; • Design Safety and Dissemination of TACIS results; • Reactor Pressure Vessel Embrittlement for VVER in Russia and Ukraine; • Regulatory Assistance; • Industrial Waste Management and Nuclear Safeguards. This paper gives an overview of the Scientific and Technical support that JRC is providing for the programming and the implementation of the TACIS nuclear safety programmes. In particular, two new projects are being implemented to get an extensive understanding of the VVER reactor pressure vessel embritttlement and integrity assessment.