bluuebunny/crossref_metadata_2025 huggingface dataset:
Small and Medium Sized Reactors: Driving Forces and Technology Development
Authors: Gowin P. J. , Kupitz J.
Published: March 2009

There will be growing demands for energy in the coming decades. One aspect of particular importance is that prospects for nuclear energy will to a considerable extent be influenced by developing countries. Since population growth will occur primarily in developing countries nuclear energy cannot play a significant global role without being a viable option in these countries. Since new power plants to be built will have to be compatible with regional electricity grids, this may result in a greater focus on plants in the small and medium range, defined by the International Atomic Energy Agency (IAEA) to produce up to 700 Megawatt of electrical power. This paper first examines the driving forces that could influence the development of nuclear energy in general and of Small and Medium Sized Reactors (SMRs) in particular in the next decades and identifies key factors in that process. Concerns over climate change may to a certain extent influence the discussion on future energy options. Other factors of equal importance for the future of nuclear are a continued emphasis on maintaining high safety standards, the implementation of acceptable solutions for spent fuel and radioactive waste disposal and a globally accepted non-proliferation regime, factors that may in turn have an impact on public acceptance. Economic competitiveness of nuclear energy is an additional important factor, and without being commercially viable, no energy source can in the long run represent a major and stable component in a competitive energy sector. The introduction of SMRs in developing countries poses additional challenges, such as investment limitations. Technology development plays an important role in keeping the nuclear option open for countries wishing to use nuclear reactors to meet their energy needs, and advances in reactor design will be important to enable a significant nuclear component in developing countries. This paper considers the contribution that nuclear science and technology can make towards meeting those needs and discusses the role of the IAEA in the nuclear technology area. The activities of the IAEA encompass the broad areas of nuclear technology, safety and safeguards, together with outreach activities which seek to provide information to decision makers, civil society and the public.
Numerical Modeling of Two-Phase Flows Using Advanced Two Fluid Systems
Authors: Kumbaro Anela , Toumi Imad , Seignole Vincent
Published: March 2009

The purpose of this paper is to report on the development and assessment of approximate Riemann solver methods for the discretization of non-linear non-conservative systems arising in the simulation of two-phase flows. These methods are able to treat general two-phase flow systems with realistic state equations and are flexible enough to be applied on any mesh type, structured as well as unstructured. We will detail models that go from the basic 6 equation two-fluid model to the coupling of this system with one or more transport equations, for instance on volumetric interfacial area concentration, or on partial void fractions of groups of bubbles (MUlti-Size-Group model). This kind of transport equation is useful to predict at a finer level the interfacial patterns or bubble size distribution and takes account of coalescence or breakup rates of inclusions. We make a glimpse at the choices made regarding this aspect. Different physico-numerical benchmarks are provided in order to illustrate the numerical and physical modeling. Confrontation with experimental or analytical reference data are performed whenever possible. Computer simulations are performed using OVAP, a new multidimensional CFD code.
Validation of the Deterministic Realistic Method Applied to CATHARE on LB LOCA Experiments
Authors: Sauvage Jean-Yves , Laroche Ste´phane
Published: March 2009

Framatome-ANP and EDF have defined a generic approach for using a best-estimate code in design basis accident studies called Deterministic Realistic Method (DRM). It has been applied to elaborate a LB LOCA ECCS evaluation model based on the CATHARE code. From a prior statistical analysis of uncertainties, the DRM derives a conservative deterministic model, preserving the realistic nature of the simulation, to be used in the further applications. The conservatism of the penalized model is demonstrated comparing penalized calculations with relevant experimental data. The DRM proved to be a highly flexible tool and has been applied successfully to meet the specific French and specific Belgian requirements of Safety Authorities.
Westinghouse Owners Group’s Program to Extend the Technical Specification Completion Time for the Containment Isolation Valves
Authors: Andre´ Gerald R. , Andrachek James D. , Rawluszki April R. , Ketchum William H. , Wideman Steve G.
Published: March 2009

The Westinghouse Owners Group (WOG) has undertaken a number of programs focused on improving U.S. Nuclear Regulatory Commission (NRC) Technical Specification requirements. This particular program focused on the Completion Times (CTs) for the containment penetration isolation valves (CIVs). This paper presents and discusses the approach and analysis to extend these CTs from 4 hours to 7 days. This extension is necessary to provide sufficient time for plant personnel to address CIV inoperability and to perform preventive maintenance activities on the CIVs during power operation. Containment penetration types that were included in these analyses were for systems connected to the containment atmosphere, the reactor coolant system, and the steam generators. The approach used in this program applied both deterministic and probabilistic evaluations. A plant-specific implementation of the generic analysis is provided for the Wolf Creek Generating Station (WCGS). The results show that a CT extension to 7 days can be justified for a majority of the CIVs for the WCGS. For those CIVs where the full 7-day extension could not be justified, it was found that many could be extended to at least 12 hours. There were only a small number of CIVs where an extension could not be justified.
Numerical Simulation of Hydrodynamics of a Heavy Liquid Drop Covered by Vapor Film in a Water Pool
Authors: Ma W. M. , Yang Z. L. , Giri A. , Sehgal B. R.
Published: March 2009

A numerical study on the hydrodynamics of a droplet covered by vapor film in water pool is carried out. Two level set functions are used as to implicitly capture the interfaces among three immiscible fluids (melt-drop, vapor and coolant). This approach leaves only one set of conservation equations for the three phases. A high-order Navier-Stokes solver, called Cubic-Interpolated Pseudo-Particle (CIP) algorithm, is employed in combination with level set approach, which allows large density ratios (up to 1000), surface tension and jump in viscosity. By this calculation, the hydrodynamic behavior of a melt droplet falling into a volatile coolant is simulated, which is of great significance to reveal the mechanismof steam explosion during a hypothetical severe reactor accident.
Simulation of Aerosol Behavior in a Saturated Atmosphere With the CONTAIN Code
Authors: Kljenak Ivo , Mavko Borut
Published: March 2009

Experiments on aerosol behavior in an atmosphere containing saturated vapor, which were performed in the KAEVER experimental facility and proposed for the OECD International Standard Problem No. 44, were simulated with the CONTAIN thermal-hydraulic computer code. The purpose of the work was to assess the capability of the CONTAIN code to model aerosol condensation and deposition in a containment of a light-water-reactor nuclear power plant at severe accident conditions. Results of dry and wet aerosol concentrations in the test vessel atmosphere are presented and analyzed.
An Industry Program for Managing PWSCC of Closure Head Penetrations
Authors: Mathews Larry K.
Published: March 2009

The Alloy 600 Issues Task Group of the EPRI managed Materials Reliability Program initiated an industry program to address the generic aspects of the Alloy 182 weld cracking in the A hot leg nozzle weld at V. C. Summer in December 2000. The generic affects of the recent cracks in the Alloy 600 Control Rod Drive Module (CRDM) and thermocouple (T/C) nozzles at Oconee 1 in November 2000 would also be included. This need for a concerted industry effort for head penetrations was further emphasized by the discovery of similar cracking at Arkansas Nuclear One (ANO) Unit 1 and the other two Oconee units during early 2001. Prior to the experiences at Oconee and ANO, there had been only one reported case of a through wall crack in a penetration at Bugey 3 in France in 1991. The predominant form of PWSCC cracking discovered in the Alloy 600 nozzles between 1991 and the recent events had been axial cracks initiated on the inner surface of the penetration tubes. This type of cracking had been addressed through an industry program in response to Generic Letter 97-01, and included a series of lead plant inspections which were being carried out over several years. Most of the cracks at Oconee and ANO 1, however, appeared to originate on the outside surface of the stub of the penetration tube extending below the J-groove weld, or in the weld itself. Circumferential cracking above the nozzle attachment weld was discovered on four nozzles at two of the Oconee units. During inspections in the Fall 2001 outages, additional leaking or cracked nozzles were discovered at several other plants. The models that had been used to rank the susceptibility of the plants to ID initiated flaws needed revision to account for the new phenomena. Additionally, the presence of circumferential cracks above the attachment weld presented the potential safety concerns of rod ejection and small break LOCA. Also, the NDE techniques that had been developed and qualified for the ID initiated flaws would be unable to detect the OD initiated flaws, so new NDE techniques and delivery capabilities were needed. Finally the repairs required for the Oconee and ANO flaws were extremely costly and dose intensive. Therefore, new repair and/or mitigation methods and delivery techniques were needed. The MRP program was established to address these areas and has evolved significantly as more information has become available. It includes activities in assessment and management of the issue, inspection capability, and repair and mitigation. Because of the safety implications of the circumferential cracking, the Nuclear Regulatory Commission issued NRC Bulletin 2001–01 on August 3, 2001. The MRP program also included a generic submittal to assist utilities in responding to the bulletin. Long term activities to provide utilities with appropriate tools for managing the PWSCC of reactor head penetrations are planned.
Benefits of Digital Equipment Generic Qualification Activities
Authors: Thomas James E. , Steiman Samuel C.
Published: March 2009

As a result of nuclear power plant instrumentation and control obsolescence issues, there have been numerous activities during recent years relating to the qualification of digital equipment. Some of these activities have been “generic” in nature in that the qualification was not limited to plant specific applications, but was intended to cover a broad base of potential applications of the digital equipment. These generic qualifications have been funded by equipment manufacturers and by utility groups and organizations. The generic activities sponsored by the Electric Power Research Institute (EPRI) have been pilot projects for an overall generic qualification approach. The primary benefit resulting from the generic qualification work to date is that a number of digital platforms and digital devices are now available for use in various nuclear safety-related applications. Many of the tests and evaluations necessary to support plant specific applications have been completed. The amount of data and documentation that each utility must develop on a case by case basis has been significantly reduced. There are also a number of additional benefits resulting from these industry efforts. The challenges and difficulties in qualifying digital equipment for safety-related applications are now more clearly understood. EPRI has published a lessons learned document (EPRI Report 1001452, Generic Qualification of Commercial Grade Digital Devices: Lessons Learned from Initial Pilots, which covers several different qualification areas, including device selection, project planning, vendor surveys and design reviews, and electromagnetic compatibility (EMC) qualification. Application of the experience and lessons learned from the EPRI pilot activities should help reduce the effort and cost required for future qualification work. Most generic qualification activities for commercial equipment have been conducted using the approach of EPRI TR-106439, Guideline on Evaluation and Acceptance of Commercial Grade Digital Equipment for Nuclear Safety Applications, and the supplemental guidance of EPRI TR-107339, Evaluating Commercial Digital Equipment for High Integrity Applications. The U. S. Nuclear Regulatory Commission (NRC) has reviewed some of the qualification activities and reports and issued a number of Safety Evaluation Reports (SERs). Although some additional work may be required to support specific applications, the generic qualification activities and NRC reviews have provided significant clarification and further stabilized the regulatory environment for use of digital equipment in safety-related applications.
Direct Contact Heat Exchange Interfacial Phenomena for Liquid Metal Reactors: Part I — Heat Transfer
Authors: Cho D. H. , Page R. J. , Hurtault D. , Abdulla S. , Liu X. , Anderson M. H. , Bonazza R. , Corradini M. L.
Published: March 2009

Experiments on direct-contact heat exchange between molten metal and water for steam production were conducted. These experiments involved the injection of water into molten lead-bismuth eutectic for heat transfer measurements in a 1-D geometry. Based on the initial results of the experiments, the effects of the water flow rate and the molten metal superheat (temperature difference between molten metal and saturated water) on the volumetric heat transfer coefficient were discussed.
Thermal Analysis for Geologic Codisposal of SNF Packages
Authors: Lee Si Y.
Published: March 2009

The engineering viability of disposal of aluminum-clad, aluminum-based spent nuclear fuel (Al-SNF) in a geologic repository requires a thermal analysis to provide the temperature history of the waste form. Calculated temperatures are used to demonstrate compliance with criteria for waste acceptance into the geologic disposal system and as input to assess the chemical and physical behavior of the waste form within the Waste Package (WP). The leading codisposal WP design proposes that a central DOE Al-SNF canister be surrounded by five Defense Waste Process Facility (DWPF) glass log canisters, that is, High-level Waste Glass Logs (HWGL’s), and placed into a WP in a geologic disposal system. A DOE SNF canister having about 0.4318m diameter is placed along the central horizontal axis of the WP. The five HWGL’s will be located around the peripheral region of the DOE SNF canister within the cylindrical WP container. The codisposal WP will be laid down horizontally in a drift repository. In this situation, two waste form options for Al-SNF disposition are considered using the codisposal WP design configurations. They are the direct Al-SNF form and the melt-dilute ingot. In the present work, the reference geologic and design conditions are assumed for the analysis even though the detailed package design is continuously evolved. This paper primarily dealt with the thermal performance internal to the codisposal WP for the qualification study of the WP containing Al-SNF. Thermal analysis methodology and decay heat source terms have been developed to calculate peak temperatures and temperature profiles of Al-SNF package in the DOE spent nuclear fuel canister within the geologic codisposal WP.
Roll Wave Effects on Annular Condensing Heat Transfer in Horizontal PCCS Condenser Tube
Authors: Kondo Masaya , Nakamura Hideo , Anoda Yoshinari , Saishu Sadanori , Obata Hiroyuki , Shimada Rumi , Kawamura Shinichi
Published: March 2009

A horizontal in-tube condensation heat exchanger is under investigation to be used for a passive containment cooling system (PCCS) of a next generation-type BWR. The flow conditions in the horizontal condenser tube were observed both visually and by local void fraction fluctuation. The observed flow regimes at a rated condition were annular flow at the tube inlet, and turned gradually into wavy flow and smooth stratified flow along the length of the tube. It was found further that frequency of the roll waves that appear on the liquid film in the annular flow is closely related to the measured local condensation heat transfer coefficient. Based on the flow observation, the roll wave frequency and measured condensation heat transfer coefficient, a model is proposed which predicts the condensation heat transfer coefficient particularly for annular flows around the tube inlet region. The proposed heat transfer model predicts well the influences of pressure, local gas-phase velocity and film thickness.
How Good Is Good: Improved Tracking and Managing of Safety Goals, Performance Indicators, Production Targets and Significant Events Using Learning Curves
Authors: Duffey Romney B. , Saull John W.
Published: March 2009

We show a new way to track and measure safety and performance using learning curves derived on a mathematical basis. When unusual or abnormal events occur in plants and equipment, the regulator and good management practice requires they be reported, investigated, understood and rectified. In addition to reporting so-called “significant events”, both management and the regulator often set targets for individual and collective performance, which are used for both reward and criticism. For almost completely safe systems, like nuclear power plants, commercial aircraft and chemical facilities, many parameters are tracked and measured. Continuous improvement has to be demonstrated, as well as meeting reduced occurrence rates, which are set as management goals or targets. This process usually takes the form of statistics for availability of plant and equipment, forced or unplanned maintenance outage, loss of safety function, safety or procedural violations, etc. These are often rolled up into a set of so-called “Performance Indicators” as measures of how well safety and operation is being managed at a given facility. The overall operating standards of an industry are also measured. A whole discipline is formed of tracking, measuring, reporting, managing and understanding the plethora of indicators and data. Decreasing occurrence rates and meeting or exceeding goals are seen and rewarded as virtues. Managers and operators need to know how good is their safety management system that has been adopted and used (and paid for), and whether it can itself be improved. We show the importance of accumulated experience in correctly measuring and tracking the decreasing event and error rates speculating a finite minimum rate. We show that the rate of improvement constitutes a measurable “learning curve”, and the attainment of the goals and targets can be affected by the adopted measures. We examine some of the available data on significant events, reportable occurrences, and loss of availability. We suggest the use of learning curves as a means of accurately tracking progress; and stress the importance of a sustained learning environment in performance improvement.
Thermo-Economic Assessment of Advanced, High-Temperature CANDU Reactors
Authors: Spinks Norman J. , Pontikakis Nikos , Duffey Romney B.
Published: March 2009

Research underway on the advanced CANDU examines new, innovative, reactor concepts with the aim of significant cost reduction and resource sustainability through improved thermodynamic efficiency and plant simplification. The so-called CANDU-X concept retains the key elements of the current CANDU designs, including heavy-water moderator that provides a passive heat sink and horizontal pressure tubes. Improvement in thermodynamic efficiency is sought via substantial increases in both pressure and temperature of the reactor coolant. Following on from the new Next Generation (NG) CANDU, which is ready for markets in 2005 and beyond, the reactor coolant is chosen to be light water but at supercritical operating conditions. Two different temperature regimes are being studied, Mark 1 and Mark 2, based respectively on continued use of zirconium or on stainless-steel-based fuel cladding. Three distinct cycle options have been proposed for Mark 1: the High-Pressure Steam Generator (HPSG) cycle, the Dual cycle, and the Direct cycle. For Mark 2, the focus is on simplification via a Direct cycle. This paper presents comparative thermo-economic assessments of the CANDU-X cycle options, with the ultimate goal of ascertaining which particular cycle option is the best overall in terms of thermodynamics and economics. A similar assessment was already performed for the NG CANDU. The economic analyses entail obtaining cost estimates of major plant components, such as heat exchangers, turbines and pumps.
Conceptual Analysis of the Economic Feasibility of Fission Electric Cell Reactors
Authors: Morrow C. , Rochau G. , Cash J. , King D.
Published: March 2009

The United States Department of Energy, Nuclear Energy Research Initiative (NERI) Direct Energy Conversion (DEC) project began in August of 1998 with the goal of developing a direct energy conversion process suitable for commercial development. With roughly two thirds of the project completed, we believe a viable direct energy device could be economic. This paper describes the financial basis behind that belief for one proposed DEC reactor, the magnetically insulated fission electric cell (FEC). It also illustrates the value of economic analysis even in these early phases of a research project. The financial basis consists of a conceptual level Economic Model comprised of five modules. The Design Model provides technical specification to other modules. The Fuel Cost Model estimates fuel expenses based on current spot market prices applied over a wide range of fuel enrichment. The Operating Cost Model uses published correlations to provide rough order of magnitude non-fuel operating costs. The Capital Cost model uses analogy and parametric estimating techniques to generate capital cost estimates for a DEC power plant. Finally, the financial model combines output from the other models to produce a Net Present Value analysis with cost of generation as the independent variable. Model results indicate that several FEC geometric configurations could be economic. Within these configurations, optimums exist. Finally, the model demonstrates that the most efficient design is not necessarily the most economic.
APR1400: Evolutionary Korean Next Generation Reactor
Authors: Kim Insik , Kim Dong-Su
Published: March 2009

Following the Korean Standard Nuclear Power Plant (KSNP) program which has shown its success through operation of its first units, Ulchin 3&4, Korean nuclear industry, with the support of Korean government, has launched its next program to develop 4000MWt (1400MWe) Advanced Power Reactor, APR1400, as a Korean next generation reactor. It is to complete an advanced standardized plant design, with increased power and with higher confidence in safety, performance, construction and operation as well as economics relative to the currently operating KSNPs. Three-phase development program was ended in December 2001, and it is our next task to implement the developed design in actual construction and to proceed commercial operation. The key to the APR1400 design development is an evolutionary approach that relies on use of proven KSNP design concepts and proven components to get high confidence in design itself and thus to make use of earned credits in safety and performance. Furthermore, advanced features and design improvements are implemented to provide additional benefits. These features are to provide additional thermal margins; to better accommodate transients and reduce unnecessary challenges to the plant safety system; to mitigate severe accident conditions; to simplify structures and to improve operability. This paper describes the development approach of the APR1400, and introduces major design features including design improvements and advanced design features focusing on the nuclear steam supply system (NSSS).
RELAP5/MOD3.2 Investigation of a VVER-1000 MCP Switching on Problem
Authors: Groudev Pavlin , Pavlova Malinka
Published: March 2009

This paper provides a discussion of various RELAP5 parameters calculated for the investigation of the nuclear power reactor parameter behavior in case of switching on one main coolant pump (MCP) when the other three MCPs are in operation. The reference power plant for this analysis is Unit 6 at the Kozloduy Nuclear Power Plant (NPP) site. Operational data from Kozloduy NPP have been used for the purpose of assessing how the RELAP5 model compares against plant data. During the plant-commissioning phase at Kozloduy NPP Unit 6 a number of experiments have been performed. One of them is switching on MCP when the other three MCPs are in operation. The event is characterized by rapid increase in the flow through the core resulting in a coolant temperature decrease, which leads to insertion of positive reactivity due to the modeled feedback mechanisms. This investigation has been conducted by Bulgarian and Russian specialists on the stage when the reactor power was at 75% of the nominal level. The purpose of the experiment was the complete testing of reliability of all power plant equipment, testing the reliability of the main regulators and defining a jump of the neutron reactor power in case of switching on of one main coolant pump. The Institute for Nuclear Research and Nuclear Energy - Bulgarian Academy of Sciences (INRNE-BAS), Sofia, and Kozloduy NPP have been developing a RELAP5/MOD3.2 model for Kozloduy NPP VVER-1000 for investigation of operational occurrences, abnormal events, and design basis scenarios. This investigation is a process that compares the analytical results obtained by the RELAP5 computer model of the VVER-1000 against the experimental transient data received from the Kozloduy NPP Unit 6. The comparisons between the RELAP5 results and the test data indicate good agreement. This report was possible through the participation of leading specialists from Kozloduy NPP and with the support of Argonne National Laboratory, under the International Nuclear Safety Program (INSP) of the United States Department of Energy.
Numerical Simulation and Analyses of the Loss of Feedwater Transient at the Unit 4 of Kola NPP
Authors: Stevanovic Vladimir D. , Stosic Zoran V. , Kiera Michael , Stoll Uwe
Published: March 2009

A three-dimensional numerical simulation of the loss-of-feed water transient at the horizontal steam generator of the Kola nuclear power plant is performed. Presented numerical results show transient change of integral steam generator parameters, such as steam generation rate, water mass inventory, outlet reactor coolant temperature, as well as detailed distribution of shell side thermal-hydraulic parameters: swell and collapsed levels, void fraction distributions, mass flux vectors, etc. Numerical results are compared with measurements at the Kola NPP. The agreement is satisfactory, while differences are close to or below the measurement uncertainties. Obtained numerical results are the first ones that give complete insight into the three-dimensional and transient horizontal steam generator thermal-hydraulics. Also, the presented results serve as benchmark tests for the assessment and further improvement of one-dimensional models of horizontal steam generator built with safety codes.
Regulatory Activities Related to Circumferential Cracking of Reactor Pressure Vessel Head Penetration Nozzles
Authors: Hiser Allen L.
Published: March 2009

The recent discoveries of cracked and leaking Alloy 600 vessel head penetration (VHP) nozzles, including control rod drive mechanism (CRDM) and thermocouple nozzles, at four pressurized water reactors (PWRs) have raised concerns about the structural integrity of VHP nozzles throughout the PWR industry. Nozzle cracking at Oconee Nuclear Station Unit 1 in November 2000 and Arkansas Nuclear One Unit 1 in February 2001 was limited to axial cracking, an occurrence deemed to be of limited safety concern in the NRC staff’s generic safety evaluation on the cracking of VHP nozzles dated November 19, 1993. However, the discovery of circumferential cracking at Oconee Nuclear Station Unit 3 in February 2001 and Oconee Nuclear Station Unit 2 in April 2001 particularly the large circumferential cracking identified in two CRDM nozzles at ONS3 has raised concerns about the potential safety implications and prevalence of cracking in VHP nozzles in PWRs. In response to the circumferential cracking identified at the Oconee units, the NRC issued Bulletin 2001–01, “Circumferential Cracking of Reactor Pressure Vessel Head Penetration Nozzles,” on August 3, 2001. This bulletin requests information from licensees related to the structural integrity of the reactor pressure VHP nozzles for their facilities, including the extent of VHP nozzle leakage and cracking that has been found to date, the inspections and repairs that have been undertaken to satisfy applicable regulatory requirements, future plans to inspect VHP nozzles, and a description of how future inspection plans will ensure compliance with applicable regulatory requirements. This paper summarizes the staff’s review and assessment of licensee responses to NRC Bulletin 2001–01.
Results of Overpressurization Test of a 1:4-Scale Prestressed Concrete Containment Vessel Model
Authors: Hessheimer Michael F. , Shibata Satoru , Costello James F.
Published: March 2009

The Nuclear Power Engineering Corporation (NUPEC) of Japan and the U.S. Nuclear Regulatory Commission (NRC) have been co-sponsoring and jointly funding a Cooperative Containment Research Program at Sandia National Laboratories. The purpose of the program is to investigate the response of representative models of nuclear containment structures to pressure loading beyond the design basis accident and to compare analytical predictions with measured behavior. This is accomplished by conducting static, pneumatic overpressurization tests of scale models at ambient temperature. The first project in this program was a test of a mixed scale steel containment vessel (SCV). Next, a 1:4-scale model of a prestressed concrete containment vessel (PCCV), representative of a pressurized water reactor (PWR) plant in Japan, was constructed by NUPEC at Sandia National Laboratories from January 1997 through June, 2000. Concurrently, Sandia instrumented the model with over 1500 transducers to measure strain, displacement and forces in the model from prestressing through the pressure testing. The limit state test of the PCCV model was conducted in September, 2000 at Sandia National Laboratories. This paper describes the conduct and some of the results of this test.
Shutdown and Closure of the Experimental Breeder Reactor–II
Authors: Michelbacher John A. , Baily Carl E. , Baird Daniel K. , Henslee S. Paul , Knight Collin J. , Rosenberg Kenneth E.
Published: March 2009

The Department of Energy mandated the termination of the Integral Fast Reactor (IFR) Program, effective October 1, 1994. To comply with this decision, Argonne National Laboratory-West (ANL-W) prepared a plan providing detailed requirements to maintain the Experimental Breeder Reactor-II (EBR-II) in a radiologically and industrially safe condition, including removal of all irradiated fuel assemblies from the reactor plant, and removal and stabilization of the primary and secondary sodium, a liquid metal used to transfer heat within the reactor plant. The EBR-II is a pool-type reactor. The primary system contained approximately 325 m3 (86,000 gallons) of sodium and the secondary system contained 50 m3 (13,000 gallons). In order to properly dispose of the sodium in compliance with the Resource Conservation and Recovery Act (RCRA), a facility was built to react the sodium to a solid sodium hydroxide monolith for burial as a low level waste in a land disposal facility. Deactivation of a liquid metal fast breeder reactor (LMFBR) presents unique concerns. Residual amounts of sodium remaining in circuits and components must be passivated, inerted, or removed to preclude future concerns with sodium-air reactions that could generate potentially explosive mixtures of hydrogen and leave corrosive compounds. The passivation process being implemented utilizes a moist carbon dioxide gas that generates a passive layer of sodium carbonate/sodium bicarbonate over any quantities of residual sodium. Tests being conducted will determine the maximum depths of sodium that can be reacted using this method, defining the amount that must be dealt with later to achieve RCRA clean closure. Deactivation of the EBR-II complex is on schedule for a March, 2002, completion. Each system associated with EBR-II has an associated layup plan defining the system end state, as well as instructions for achieving the layup condition. A goal of system-by-system layup is to minimize surveillance and maintenance requirements during the interim period between deactivation and decommissioning. The plans also establish document archival of not only all the closure documents, but also the key plant documents (P&IDs, design bases, characterization data, etc.) in a convenient location to assure the appropriate knowledge base is available for decommissioning, which could occur decades in the future.
Prediction of Parameters Distribution of Upward Boiling Two-Phase Flow With Two-Fluid Models
Authors: Yao Wei , Morel Christophe
Published: March 2009

In this paper, a multidimensional two-fluid model with additional turbulence k–ε equations is used to predict the two-phase parameters distribution in freon R12 boiling flow. The 3D module of the CATHARE code is used for numerical calculation. The DEBORA experiment has been chosen to evaluate our models. The radial profiles of the outlet parameters were measured by means of an optical probe. The comparison of the radial profiles of void fraction, liquid temperature, gas velocity and volumetric interfacial area at the end of the heated section shows that the multidimensional two-fluid model with proper constitutive relations can yield reasonably predicted results in boiling conditions. Sensitivity tests show that the turbulent dispersion force, which involves the void fraction gradient, plays an important role in determining the void fraction distribution; and the turbulence eddy viscosity is a significant factor to influence the liquid temperature distribution.
Design by Formula of Power Steam Line Equipped by Energy Dissipative Devices
Authors: Cesari F. G. , Battistella P.
Published: March 2009

To evaluate the ability of passive devices in protecting nuclear piping during earthquake a theoretical/experimental campaign has been performed. By means of numerical runs the effect of viscous dampers application on most critical points of a power plant steamline has been evaluated. The principle is to employ a local safety solution against heavy dynamic solicitations placing passive devices in crotch region of bends. The devices location corresponds to an in plane position in respect of the curve. Considerations on structural configuration and stress/strain states are also presented with the aim to respect the philosophy of design/verification requirements stated by the ASME Sct. III Cl.1 code. For experimental tests a C mock-up, whose sizes are derived by a thermal plant steamline, has been suggested and studied [1]. Comparison of numerical data on piping with/without dissipative elements are also included. The impact on the whole structure has been also taken into account. Some of the results included in the paper have been obtained in the E.U. contract named REEDS.
Structural Integrity Assessment of Steam Generator Tubes Using a New EPRI Statistical Approach
Authors: Alexandre de Jesus Miranda Carlos , Neto Miguel Mattar
Published: March 2009

A fundamental step in tube plugging management of a Steam Generator (SG), in a Nuclear Power Plant (NPP), is the tube structural integrity evaluation. The degradation of SG tubes may be considered one of the most serious problems found in PWRs operation, mainly when the tube material is the Inconel 600. The first repair criterion was based on the degradation mode where a uniform tube wall thickness corrosion thinning occurred. Thus, a requirement of a maximum depth of 40% of the tube wall thickness was imposed for any type of tube damage. A new approach considers different defects arising from different degradation modes, which comes from the in-service inspections (NDE) and how to consider the involved uncertainties. It is based on experimental results, using statistics to consider the involved uncertainties, to assess structural limits of PWR SG tubes. In any case, the obtained results, critical defect dimensions, are within the regulatory limits. In this paper this new approach will be discussed and it will be applied to two cases (two defects) using typical data of SG tubes of one Westinghouse NPP. The obtained results are compared with ‘historical’ approaches and some comments are addressed from the results and their comparison.
Main Results of Na-K Alloy Boiling Investigation
Authors: Sorokin G. A. , Bogoslovskaya G. P. , Ivanov E. F. , Sorokin A. P.
Published: March 2009

Boiling experiments on eutectic sodium-potassium alloy in the model of fast reactor subassembly under conditions of low-velocity circulation carried out at the IPPE call for further investigations into numerical modeling of the process. The paper presents analysis of pin bundle liquid metal boiling, stages of the process, its characteristics (wall temperature, coolant temperature, flow rate. pressure void fraction and others), that allowed the pattern map to be drawn. The problem of conversion of the data gained in Na-K mock-up experiments to in-pile sodium reactor operating conditions is analyzed here, as well as thermodynamic similarity of liquid metal coolants and eutectic Na-K alloy. Data on bundle boiling in Na-K are presented in comparison with those in different liquid metals. Analysis of data on liquid metal heat transfer in cases of pool boiling, boiling in tubes, in slots, and in pin bundles, as well as data on critical heat flux in tubes was performed and discussed in the paper. The relationship for calculation of critical heat flux in liquid metal derived by the authors is presented. Results of numerical modeling of liquid metal boiling heat transfer during accident cooling of reactor core applied to experimental conditions of going from forced to natural circulation are presented, too.
U.S. Industry Activities on Inspection of CRDM Penetrations
Authors: Alley Tom , Kietzman Kim , Ammirato Frank
Published: March 2009

The discovery of primary water stress corrosion cracking (PWSCC) in control rod drive mechanisms (CRDM) penetrations in U.S. and European plants prompted the U.S. nuclear industry to focus considerable effort on development and implementation of effective inspection methods. In particular, cracking was discovered in butt welds connecting reactor vessel nozzles to main coolant piping and in control rod drive mechanism (CRDM) head penetration base material and attachment welds. The EPRI Materials Reliability Program (EPRI-MRP) formed an Inspection Committee to address development of industry guidance for inspection of these components, development of effective non-destructive examination (NDE) methods, and demonstration of inspection processes. This paper discusses the MRP activities pertaining to inspection of CRDM penetrations. Results of demonstrations and field inspections conducted will also be summarized.
Radial Power Profile of MOX and LEU Fuel Pellet Versus Burnup
Authors: Chang Gray S. , Pedersen Robert C.
Published: March 2009

One of challenge to burn the WG-Pu in Mixed Oxide (MOX) fuel in light water reactors (LWR) is to demonstrate that the differences between WG-MOX, RG-MOX, and LWR LEU fuel are minimal, and therefore, the commercial MOX and LEU fuel experience base is applicable. The MCWO-calculated Radial Power Profile of LEU, Weapons Grade-MOX and Reactor Grade-MOX fuel pellets at various burnups are similar toward the end of life (50 GWd/t). Therefore, the LEU fuel performance evaluation code — FRAPCON-3 with modifications, such as, the detailed fission power profiles versus burnup, can be used in the MOX fuel pellet performance analysis. MCWO also calculated the 240Pu/Pu ratio in WG-MOX versus burnup, which reaches an average of 31.25% at discharged burnup of 50 GWd/t. It meets the spent fuel standard for WG-Pu disposition in LWR.
EU Funded Research Activities on NPPs Operational Safety
Authors: Manolatos P. , Van Goethem G.
Published: March 2009

The 5th framework programme (FP-5), the pluriannual research programme of the European Union (EU), covers the period 1998–2002. Research on nuclear energy, fusion and fission, is covered by the Euratom part of the FP-5. An overview of the Euratom’s research on Nuclear Reactor Safety, managed by the DG-RTD of the European Commission (EC), is presented. This concerns 70 multi-partner projects of approximately € 82.5 million total contract value that have been selected and co-financed during the period 1999–2001. They form the three clusters of projects dealing with the “Operational Safety of Existing Installations”. “Plant Life Extension and Management” (PLEM), “Severe Accident Management” (SAM) and “Evolutionary concepts” (EVOL). Emphasis is given here to the projects of the PLEM cluster.
Completion of Experimental Breeder Reactor-II Sodium Processing at Argonne National Laboratory
Authors: McDermott Mary D. , Griffin Charles D. , Baird Daniel K. , Baily Carl E. , Michelbacher John A. , Rosenberg Kenneth E. , Henslee S. Paul
Published: March 2009

The Experimental Breeder Reactor - II (EBR-II) at Argonne National Laboratory - West (ANL-W) was shutdown in September 1994 as mandated by the United States Department of Energy. Located in eastern Idaho, this sodium-cooled reactor had been in service since 1964, and was a test facility for fuels development, materials irradiation, system and control theory tests, and hardware development. The EBR-II termination activities began in October 1994, with the reactor being maintained in an industrially and radiologically safe condition for decommissioning. With the shutdown of EBR-II, its sodium coolant became a waste necessitating its reaction to a disposal form. A Sodium Process Facility (SPF), designed to convert sodium to 50 wt% sodium hydroxide, existed at the ANL-W site, but had never been operated. The SPF was upgraded to current standards and codes, and then modified in 1998 to convert the sodium to 70 wt% sodium hydroxide, a substance that solidifies at 65°C (150°F) and is acceptable for burial as low level radioactive waste in Idaho. In December 1998, the SPF began operations. Working with sodium and highly concentrated sodium hydroxide presented some unique operating and maintenance conditions. Several lessons were learned throughout the operating period. Processing of the 330 m3 (87,000 gallons) of EBR-II primary sodium, 50 m3 (13,000 gallons) of EBR-II secondary sodium, and 290 m3 (77,000 gallons) of Fermi-1 primary sodium was successfully completed in March 2001, ahead of schedule and within budget.
Analysis of Primary Side Feed and Bleed Operation at Armenian NPP Unit 2
Authors: Amirjanyan Armen , Malakyan Tsolak , Jo Jae
Published: March 2009

Analyses were performed for the primary side feed and bleed (F&B) procedures for Armenian Nuclear Power Plant (ANPP), using the RELAP5 code. ANPP is a six-loop WWER-440/270 model of Russian design. The purpose of the analyses was to demonstrate the core cooling capability of the plant using the high pressure injection (HPI) system (feed) and pressurizer safety valves (bleed), and to evaluate temperature changes in the Borated Water Supply Tank (BWST). The Mod3.2.2β version of RELAP5 was used for the analyses. The six loops of the plant were modeled by 4 loops; three loops represented three individual loops and one loop combined the other three loops. The input described the primary and secondary loops in detail, as well as all safety systems and most of control systems. The HPI system was connected to the cold leg of each loop. The Emergency Core Cooling (ECC) mixer component was applied in the connection of the HPI system to simulate the phenomena associated with the subcooled ECC injection into a reactor coolant system. Nominal operating parameters were used to establish the steady state conditions. Some of the initiating events and boundary conditions were based on the experience and information obtained from previous transients and incidents in the plant. The incident was assumed to be initiated by a loss of off-site power supply and simultaneous reactor scram. Diesel generators were assumed not to be available for 11,000 seconds. Thus, feedwater to the steam generators, main coolant pumps, and HPI pumps were not available during this period and the core was cooled by natural circulation in the primary loops supported by the remaining water in the steam generators. During this period the primary water temperature and pressure were near or slightly higher than the steady state values, but the water remained subcooled. At 11,000 seconds, upon starting of the diesel generators, the operator switched on two HPI pumps and opened one pressurizer safety valve. At this time, all six steam generators were found to be nearly empty. Once the F&B operation began, the pressure and temperature of the primary loops declined rapidly. The results of F&B analyses show that there is no violation of the core coolability conditions when two HPI pumps are available, and the temperature of BWST tank does not reach the saturation temperature when two heat exchanger pumps are available.
Applications of Artificial Neural Network for the Prediction of Pool Boiling Curves
Authors: Su Guanghui , Fukuda K. , Morita K.
Published: March 2009

Artificial neural network (ANN) has the advantage that the best-fit correlations of experimental data will no longer be necessary for predicting unknowns from the known parameters. The ANN was applied to predict the pool boiling curves in this paper. The database of experimentel data presented by Berenson, Dhuga et al., and Bui and Dhir etc. were used in the analysis. The database is subdivided in two subsets. The first subset is used to train the network and the second one is used to test the network after the training process. The input parameters of the ANN are: wall superheat ΔTw, surface roughness, steady/transient heating/transient cooling, subcooling, Surface inclination and pressure. The output parameter is heat flux q. The proposed methodology allows us to achieve the accuracy that satisfies the user’s convergence criterion and it is suitable for pool boiling curve data processing.