bluuebunny/crossref_metadata_2025 huggingface dataset:
Development Trends in Nuclear Technology and Related Safety Aspects
Authors: Kuczera B. , Juhn P. E. , Fukuda K.
Published: March 2009

The IAEA Safety Standards Series include, in a hierarchical manner, the categories of Safety Fundamentals, Safety Requirements and Safety Guides, which define the elements necessary to ensure the safety of nuclear installations. In the same way as nuclear technology and scientific knowledge advance continuously, also safety requirements may change with these advances. Therefore, in the framework of the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) one important aspect among others refers to user requirements on the safety of innovative nuclear installations, which may come into operation within the next fifty years. In this respect, the major objectives of the INPRO subtask “User Requirements and Nuclear Energy Development Criteria in the Area of Safety” have been: a. to overview existing national and international requirements in the safety area, b. to define high level user requirements in the area of safety of innovative nuclear technologies, c. to compile and to analyze existing innovative reactor and fuel cycle technology enhancement concepts and approaches intended to achieve a high degree of safety, and d. to identify the general areas of safety R&D needs for the establishment of these technologies. During the discussions it became evident that the application of the defence in depth strategy will continue to be the overriding approach for achieving the general safety objective in nuclear power plants and fuel cycle facilities, where the emphasis will be shifted from mitigation of accident consequences more towards prevention of accidents. In this context, four high level user requirements have been formulated for the safety of innovative nuclear reactors and fuel cycles. On this basis safety strategies for innovative reactor designs are highlighted in each of the five levels of defence in depth and specific requirements are discussed for the individual components of the fuel cycle.
Numerical Analysis of Catalytic Recombiner Performance Considering a 3-Dimensional Gas Flow
Authors: Tahara Mika , Oikawa Hirohide , Arai Kenji
Published: March 2009

Metal-water reaction and radiolysis of water generate hydrogen during a severe accident in a light water reactor. To prevent hydrogen combustion, a flammability gas control system (FCS) is installed in reactor containment. Most of the current FCS combine hydrogen with oxygen by heating and employ active devices to maintain the gas flow through the FCS. Recently a catalytic recombiner has been developed as passive FCS. The catalytic recombiner performs its function passively and has the advantages of the robustness during an accident, easy maintenance and low cost compared with the current active FCS. The hydrogen depletion rate of the catalytic recombiner is affected by the local thermal hydraulic conditions during an accident. To evaluate hydrogen depletion by the catalytic recombiner considering these phenomena in the containment, a 3-dimensional fluid dynamics analysis is useful. A theoretical catalytic recombiner model has been developed in which the flammable gas depletion rate is estimated accounting for the gas transfer rate in porous catalyst material. The model has been incorporated with a thermal hydraulic model for the fluid dynamics in a containment that has been developed using a 3-dimensional CFD code STRA-CD. This catalytic recombiner model has been confirmed using a catalytic recombiner performance test that was carried out in the Battelle Model Containment (BMC). Further verification of the analysis model has been conducted using the test data described in NUREG/CR-6580 which addressed the wall effect on the catalytic recombiner performance. The predicted performance of the catalytic recombiner shows a good agreement with the test data, and especially the parameters effects on the recombiner performance is well described, which include the effects of the containment wall, gas flow rate to the catalytic recombiner and gas concentration distribution in the containment.
A Feasibility Study on Core Cooling of Reduced-Moderation PWR for the Large Break LOCA
Authors: Yoshida Hiroyuki , Ohnuki Akira , Akimoto Hajime
Published: March 2009

A design study of a reduced-moderation water reactor (RMWR) with tight lattice core is being carried out at the Japan Atomic Energy Research Institute (JAERI) as one candidate for future reactors. The concept is developed to achieve a conversion ratio greater than unity using the tight lattice core (volume ratio of moderator to fuel is around 0.5 and the gap spacing between the fuel rods is remarkably narrower than in a reactor currently operated). Under such tight configuration, the core thermal margin becomes smaller and should be evaluated in a normal operation and also during the reflood phase in a large break loss-of-coolant accident (LBLOCA) for PWR type reactors. In this study, we have performed a feasibility evaluation on core cooling of reduced moderation PWR for the LBLOCA (200% break). The evaluation was performed for the primary system after the break by the REFLA/TRAC code. The core thermal output of the reduced moderation PWR is 2900 MWt, the gap between adjacent fuel rods is 1 mm, and heavy water is used as the moderator and coolant. The present design adopts seed fuel assemblies (MOX fuel) and several blanket fuel assemblies. In the blanket fuel assemblies, power density is lower than that of the seed fuel assemblies. Then, we set a channel box to each fuel assembly in order to adjust the flow rate in each assembly, because the possibility that the coolant boils in the seed fuel assemblies is very high. The pressure vessel diameter is bigger in comparison with a current PWR and core height is smaller than the current one. The current 4-loop PWR system is used, and, however, to fit into the bigger pressure vessel volume (about 1.5 times), we set up the capacity of the accumulator (1.5 times of the current PWR). Although the maximum clad temperature reached at about 1200 K in the position of 0.6 m from the lower core support plate, it is sufficiently lower than the design criteria of the current PWR (1500 K). The core cooling of the reduced moderation PWR can be maintained.
Pressure Vessel and Internals of the International Reactor Innovative and Secure
Authors: Lombardi C. V. , Padovani E. , Cammi A. , Collado J. M. , Santoro R. T. , Barnes J. M.
Published: March 2009

IRIS (International Reactor Innovative and Secure) is a modular, integral light water cooled, low-to-medium power reactor, which addresses the requirements defined by the US DOE for Generation IV reactors. Its integrated layout features a pressure vessel containing all the main primary circuit components: the internals and the biological shield, here described together with the pressure vessel, plus the steam generators, the pressurizer, and the main coolant pumps described in companion papers. For this reason the pressure vessel is a crucial component of the plant, which deserves the most demanding design effort. The wide inner annulus around the core is exploited to insert steel plates, in order to improve the inner shielding capability up to the elimination of the external biological shielding and to simplify decommissioning activities by having all the irradiated components inside the vessel.
The Application of Best Estimate and Uncertainty Analysis Methodology to Large LOCA Power Pulse in a CANDU 6 Reactor
Authors: Abdul-Razzak A. , Zhang J. , Sills H. E. , Flatt L. , Jenkins D. , Wallace D. J. , Popov N.
Published: March 2009

The paper describes briefly a best estimate plus uncertainty analysis (BE+UA) methodology and presents its prototyping application to the power pulse phase of a limiting large Loss-of-Coolant Accident (LOCA) for a CANDU 6 reactor fuelled with CANFLEX® fuel. The methodology is consistent with and builds on world practice [1], [2]. The analysis is divided into two phases to focus on the dominant parameters for each phase and to allow for the consideration of all identified highly ranked parameters in the statistical analysis and response surface fits for margin parameters. The objective of this analysis is to quantify improvements in predicted safety margins under best estimate conditions.
Development of New Constraction Method for LSBWR
Authors: Heki H. , Nakamaru M. , Maruyama T. , Hirai H. , Aritomi M.
Published: March 2009

LSBWR (Long operating cycle Simplified BWR) is a modular, direct cycle, light water cooled, and small power (100–300MWe) reactor. The design considers requirements from foreign utilities as well as from Japanese. LSBWR is currently being developed by Toshiba Corporation and Tokyo Institute of Technology. Major characteristics of the LSBWR are: 1) Long operating cycle (target: over 15 years), 2) Simplified systems and building, 3) Factory fabrication in module. From the perspective of economic improvement of nuclear power plant, it is needed to shorten the plant construction period and to reduce building volume. In designing LSBWR building, a new building structure, where the hull structure of a ship is applied to floors and walls of LSBWR has been studied. Since the hull structure is manufactured at a shipyard, building module that includes plant equipment becomes possible. The application of the hull structure, which can make large modules at a shipyard, is an effective solution to the lack of laborer and economic improvement. LSBWR is a small size BWR, turbine is smaller size and lighter weight than medium or larger size plant. Then, it has been studied to install a reactor and a turbine in the same building for decreasing building volume. From the view point of standardization, whole building is supported by three dimensional seismic isolation mechanism.
Seismic Re-Evaluation of the Armenian NPP Unit 2: Issues and Engineering Approach
Authors: Zadoyan Paruyr , Simos Nikolaos
Published: March 2009

As part of the overall effort of seismically upgrading the Armenian Nuclear Power Plant (ANPP) - Unit 2, the Armenian Nuclear Regulatory Authority (ANRA) with the technical support of the US NRC is performing a host of independent studies on key seismic issues associated with the nuclear facility. Specifically, by taking into account new information that has been generated by a number of sources on (a) the seismic hazards at the site, (b) the soil condition and layering, (c) the particularities of the reactor foundation design and its interface with the surrounding soil, and (d) structural details of the reactor and generator buildings, a comprehensive effort has been undertaken to independently establish floor response spectra that, in turn, better reflect the seismic reality at the plant. To accomplish the task of generating floor response spectra, that will be used in the seismic upgrade, detailed analyses focusing on (a) the site response using the capabilities of the SHAKE program, (b) the Soil-Structure-Interaction (SSI) aspect of the problem utilizing the SASSI and POROSLAM codes, and (c) the dynamic response of the reactor building through the use of the STARDYNE code were performed. The combined output of the various aspects of the analysis lead to a set of floor spectra at key locations on the structure so that the existing capacity of safety systems can be assessed as well as the need for upgrade to meet their revised seismic margins. While the current effort is work-in-progress, the paper presents preliminary results of this independent study and it attempts to make engineering comparisons with other on-going studies performed on behalf of the nuclear power utility in its effort to seismically upgrade the plant.
Short Intergranular Cracks in Elasto-Plastic Polycrystalline Aggregate
Authors: Cizelj Leon , Riesch-Oppermann Heinz
Published: March 2009

Computational algorithms aiming at modeling and visualization of the initiation and growth of intergranular stress corrosion cracks (e.g., in the steam generator tubes) on the grain-size scale have already been proposed. Main focus of the paper addresses the influence of randomly oriented anisotropic elasto-plastic grains on the microscopic stress fields at crack tips. The limited number of calculations indicate that the incompatibility strains, which develop along the boundaries of randomly oriented grains, influence the local stress fields (J-integrals) at crack tips significantly.
In-Situ Calibration for Feedwater Flow Measurement
Authors: Peyvan David , Gurevich Yuri , French Charles T.
Published: March 2009

With the approval by the Nuclear Regulatory Commission (NRC), of the Appendix K power uprates, it has become important to provide an accurate measurement of the feedwater flow. Failure to meet documented requirements can now more easily lead to plant operations above their analyzed safety limits. Thus, the objective of flow instrumentation used in Appendix K uprates, becomes one of providing precise measurements of the feedwater mass flow that will not allow the plant to be overpowered, but will still assure that maximum licensed thermal output is achieved. The NRC has licensed two technologies that meet these standards. Both are based on ultrasonic measurements of the flow. The first of these technologies, which is referred to as transit-time, relies on the measurement of differences in time for multiple ultrasonic beams to pass up and downstream in the fluid stream. These measurements are then coupled with a numerical integration scheme to compensate for distortions in the velocity profile due to upstream flow disturbances. This technology is implemented using a spool piece that is inserted into the feedwater pipe. The second technology relies on the measurement of the velocity of eddies within the fluid using a numerical process called cross-correlation. This technology is implemented by attaching the ultrasonic flow meter to the external surface of the pipe. Because of the ease in installation, for atypical situations, distortions in the velocity profile can be accounted for by attaching a second ultrasonic flow meter to the same pipe or multiple meters to a similar piping configuration, where the flow is fully developed. The additional meter readings are then used for the calibration of the initial set-up. Thus, it becomes possible to provide an in-situ calibration under actual operating conditions that requires no extrapolation of laboratory calibrations to compensate for distortions in the velocity profile. This paper will focus on the cross-correlation method of flow measurement, starting with the theoretical bases for the velocity profile correction factor and its reliance on only the Reynolds number to produce an accurate measurement of the flow, when the flow is fully developed. The method of laboratory calibration and the verification of these calibrations under actual plant operating conditions will be discussed. This will be followed by a discussion of how this technology is being used today to support the Appendix K uprates. Various examples will be presented of piping configurations, where in-situ calibrations have or will be used to provide an accurate measurement of the feedwater flow at a specific location.
Some Issues of Electrical Systems Modeling in Course of PSA
Authors: Lankin Michael
Published: March 2009

Electrical power supply systems are one of the essential parts of nuclear power plants. The distinctive feature of these systems from the PSA analyst’s point of view is significant amount of bi-directional dependencies present within electrical systems. This paper describes an approach that has been used for electrical systems modeling in course of Kola 4 NPP Level 1 PSA.
BWR 9×9 Fuel Assembly Thermal-Hydraulic Tests: 2 — Hydraulic Vibration Test
Authors: Tsukuda Yoshiaki , Kamimura Katsuichiro , Hattori Toshiitsu , Tanabe Akira , Saito Noboru , Warashina Masahiko , Nishino Yuji
Published: March 2009

Nuclear Power Engineering Corporation (NUPEC) conducted thermal-hydraulic projects for verification of thermal-hydraulic design reliability for BWR high-burnup 8×8 and 9×9 fuel assemblies, entrusted by the Ministry of Economy, Trade and Industry (METI). As a part of the NUPEC thermal-hydraulic projects, hydraulic vibration tests using full-scale test assemblies simulating 9×9 fuel assemblies were carried out to evaluate BWR fuel integrity. The test data were applied to development of a new correlation for the estimation of fuel rod vibration amplitude.
Safe and Simplified Boiling Water Reactor (SSBWR)
Authors: Ohtsuka Masaya , Fujimura Koji , Nagayoshi Takuji , Yamashita Jun’ichi , Kato Yasuyoshi
Published: March 2009

A safe and simplified BWR (SSBWR) has been developed as an innovative future reactor to provide a super-long life core of 20 years and to realize a passive core safety system with infinite grace period. Operability and maintainability can be largely improved by using the super-long life core, cutting the number of active components, and using a one-batch core with no exchange of fuel assemblies, which can also significantly reduce the possibility of nuclear proliferation. Np-237 of MAs (Minor Actinides) can be effectively transmuted using the very hard neutron spectrum of SSBWR and high level radioactive wastes can be reduced.
Steam Generator of the International Reactor Innovative and Secure
Authors: Cinotti L. , Bruzzone M. , Meda N. , Corsini G. , Lombardi C. V. , Ricotti M. , Conway L. E.
Published: March 2009

IRIS (International Reactor Innovative and Secure) is a light water cooled, 335 MWe power reactor which is being designed by an international consortium as part of the US DOE NERI Program. IRIS features an integral reactor vessel that contains all the main reactor coolant system components including the reactor core, the coolant pumps, the steam generators and the pressurizer. This integral design approach eliminates the large coolant loop piping, and thus eliminates large loss-of-coolant accidents (LOCAs) as well as the individual component pressure vessels and supports. In addition, IRIS is being designed with a long-life core and enhanced safety to address the requirements defined by the US DOE for Generation IV reactors. The design of the steam generators, which are internally contained within the reactor vessel, is a major design effort in the development of the integral IRIS concept. The ongoing design activity about the steam generator is the subject of this paper.
U.S. Department of Energy Sponsored In-Depth Safety Assessments of VVER and RBMK Reactors
Authors: Petri Mark C. , Pasedag Walter F.
Published: March 2009

Throughout the 1990s the National Nuclear Security Administration of the U.S. Department of Energy has worked to build capability in countries of the former Soviet Union to assess the safety of their VVER and RBMK reactors. Through this Plant Safety Evaluation Program, deterministic and probabilistic analyses have been used to provide a documented plant risk profile to support safe plant operation and to set priorities for safety upgrades. Work has been sponsored at fourteen nuclear power plant sites in eight countries. The Plant Safety Evaluation Program has resulted in immediate and long-term safety benefits for the Soviet-designed nuclear plants.
History of Fire Events in the U.S. Commercial Nuclear Industry
Authors: Najafi Bijan , Kassawara Robert P. , Joglar-Biloch Francisco , Khalil Yehia
Published: March 2009

Over the past decade, interest in performance-based fire protection has increased within the nuclear industry. In support of this growing interest, in 1997 the Electric Power Research Institute (EPRI) developed a long-range plan to develop/improve data and tools needed to support Risk-Informed/Performance-Based fire protection. This plan calls for continued improvement in collection and use of information obtained from fire events at nuclear plants. The data collection process has the objectives of improving the insights gained from such data and reducing the uncertainty in fire risk and fire modeling methods in order to make them a more reliable basis for performance based fire protection programs. In keeping with these objectives, EPRI continues to collect, review and analyze fire events in support of the nuclear industry. EPRI collects these records in cooperation with the Nuclear Electric Insurance Limited (NEIL), by compiling public fire event reports and by direct solicitation of U.S. nuclear facilities. EPRI fire data collection project is based on the principle that the understanding of history is one of the cornerstones of improving fire protection technology and practice. Therefore, the goal has been to develop and maintain a comprehensive database of fire events with flexibility to support various aspects of fire protection engineering. With more than 1850 fire records over a period of three decades and 2400 reactor years, this is the most comprehensive database of nuclear power industry fire events in existence today. In general, the frequency of fires in the U.S. commercial nuclear industry remains constant. In few cases, e.g., transient fires and fires in BWR offgas/recombiner systems, where either increasing or decreasing trends are observed, these trends tend to slow after 1980. The key issues in improving quality of the data remain to be consistency of the recording and reporting of fire events and difficulties in collection of records. EPRI has made significant progress towards improving the quality of the fire events data through use of multiple collection methods as well as its review and verification. To date EPRI has used this data to develop a generic fire ignition frequency model for U.S. nuclear power industry (Ref. 1, 4 and 5) as well as to support other models in support of EPRI Fire Risk Methods such as a cable fire manual suppression model. EPRI will continue its effort to collect and analyze operating data to support risk informed/performance based fire safety engineering, including collection and analysis of impairment data for fire protection systems and features. This paper provides details on the collection and application of fire events to risk informed/performance based fire protection. The paper also provides valuable insights into improving both collection and use of fire events data.
DOE’s Stewardship of Government-Owned Uranium Materials
Authors: Jackson J. Dale , Donaldson Dale E.
Published: March 2009

Beginning in the 1980’s, a significant number of Department of Energy facilities have been shut down and are in the decommissioning process. The shutdown of additional facilities is planned. In addition, during the past several decades, the Department of Energy has loaned nuclear material to a wide variety of private and governmental institutions for research and educational purposes. Subsequent changes in the Department’s priorities have reduced the need for nuclear materials to support the Department’s programs. Similarly, there has been a reduction in the need for borrowed nuclear materials by organizations and institutions using nuclear materials “on loan” from the Department. As a result, inventories of uranium material from the Department’s facilities and “on loan” must be removed and returned to the Department. This material is in the form of low enriched uranium (LEU), normal uranium (NU), and depleted uranium (DU) in various forms. This uranium material is located at over one hundred sites within the United States and overseas, including universities and laboratories. Much of this uranium is not needed to support national priorities and programs. The Department of Energy has assumed a stewardship role in managing nuclear materials throughout their life cycle, from acquisition to storage. Surplus uranium has created challenges for DOE in managing and storing the material as well as identifying opportunities for its further use. On behalf of the Department, the Oak Ridge Operations Office has been given the responsibility to implement the Department responsibilities in meeting these challenges and managing the Department’s uranium materials. To support this effort, the Office of Nuclear Fuel Security and Uranium Technology within the ORO complex coordinates uranium management functions across the Department of Energy. This coordination provides DOE with a number of important benefits, among which are: consolidated management and storage of uranium; improved security; a reduction of operating costs; effective use and reuse of DOT certified shipping containers; and accelerated site closure.
Impact of Flow Velocity Profile on Ultrasonic Measurement Accuracy for Feedwater Flow in Nuclear Power Plants
Authors: Hauser E. , Regan J. , Estrada H.
Published: March 2009

This paper discusses the nature of flow velocity profiles in nuclear plant feedwater system piping, and the impact of flow profile development on flow measurement. The mechanism for flow velocity profile development is described, and the influence of upstream pipe fittings and the relative roughness of the piping inside diameter on flow development is illustrated. The paper defines fully developed flow profiles and illustrates, using several practical applications, how velocity profiles in nuclear plant feedwater piping are rarely fully developed.
Rulemaking Activities to Risk-Inform Requirements in 10 CFR Part 50
Authors: McKenna Eileen M. , Reed Timothy A.
Published: March 2009

This paper presents the status of NRC activities to revise requirements in Title 10 of the Code of Federal Regulations (10 CFR Part 50) to make them more risk-informed. The activities include proposed changes to treatment requirements (requirements such as quality assurance, testing, inspection, documentation that are imposed to add confidence in the capability of structures, systems and components (SSCs) to perform their intended functions) as well as changes to certain technical requirements. The technical requirements under consideration are those relating to combustible gas control and for emergency core cooling systems (including the evaluation models).
Applications of Trajectory Solid Angle for Probabilistic Safety Assessment
Authors: Wong Po Kee , Wong Adam E. , Wong Anita
Published: March 2009

In 1974, a well-known research problem in Statistical Mechanics entitled “To determine and define the probability function P.sub.2 of a particle hitting a predetermined area, given all its parameters of generation and ejection.” was openly solicited for its solution from research and development organizations in U.S.A. One of many proposed solutions of the problem, initiated at that time, is by means of the TRAJECTORY SOLID ANGLE (TSA). TSA is defined as the integral of the dot product of the unit tangent of the particle’s trajectory to the vector area divided by the square of the position vector connecting between the point of ejection and that of the surface to be hit. The invention provides: (1) The precise and the unique solution of a previously unsolved P.sub.2 problem: (2) Impacts to the governmental NRC safety standards and DOD weapon systems and many activities in the Department of Energy; (3) Impacts to update the contents of text books of physics and mathematics of all levels; (4) Impacts to the scientific instruments with applications in high technologies. The importance of TRAJECTORY SOLID ANGLE can be quoted from a letter by the late Institute Professor P. M. Morse of MIT who reviewed the DOE proposal P7900450 (reference No. 7) in 1979 and addressed to the inventor. “If the TRAJECTORY SOLID ANGLE is correct it will provide a revolutionary concept in physics.”
Solution and Study of the Two-Dimensional Nodal Neutron Transport Equation
Authors: Panta Pazos Ruben , de Vilhena Marco Tullio , Hauser Eliete Biasotto
Published: March 2009

In the last decade Vilhena and coworkers10 reported an analytical solution to the two-dimensional nodal discrete-ordinates approximations of the neutron transport equation in a convex domain. The key feature of these works was the application of the combined collocation method of the angular variable and nodal approach in the spatial variables. By nodal approach we mean the transverse integration of the SN equations. This procedure leads to a set of one-dimensional SN equations for the average angular fluxes in the variables x and y. These equations were solved by the old version of the LTSN method9, which consists in the application of the Laplace transform to the set of nodal SN equations and solution of the resulting linear system by symbolic computation. It is important to recall that this procedure allow us to increase N the order of SN up to 16. To overcome this drawback we step forward performing a spectral painstaking analysis of the nodal SN equations for N up to 16 and we begin the convergence of the SN nodal equations defining an error for the angular flux and estimating the error in terms of the truncation error of the quadrature approximations of the integral term. Furthermore, we compare numerical results of this approach with those of other techniques used to solve the two-dimensional discrete approximations of the neutron transport equation6.
Development of Interfacial Structure in a Confined Air-Water Cap-Turbulent and Churn-Turbulent Flow
Authors: Sun Xiaodong , Kim Seungjin , Cheng Ling , Ishii Mamoru , Beus Stephen G.
Published: March 2009

The objective of the present work is to study and model the interfacial structure development of air-water two-phase flow in a confined test section. Experiments of a total of 9 flow conditions in cap-turbulent and churn-turbulent flow regimes are carried out in a vertical air-water upward two-phase flow experimental loop with a test section of 200-mm in width and 10-mm in gap. Miniaturized four-sensor conductivity probes are used to measure local two-phase parameters at three different elevations for each flow condition. The bubbles captured by the probes are categorized into two groups in view of the two-group interfacial area transport equation, i.e., spherical/distorted bubbles as Group 1 and cap/churn-turbulent bubbles as Group 2. The acquired parameters are time-averaged local void fraction, interfacial velocity, bubble number frequency, interfacial area concentration, and bubble Sauter mean diameter for both groups of bubbles. Also, the line-averaged and area-averaged data are presented and discussed. The comparisons of these parameters at different elevations demonstrate the development of interfacial structure along the flow direction due to bubble interactions.
Major Component Replacement Experience in IKATA Power Station
Authors: Asai Yoshihisa , Takagi Toshimitsu , Azuma Tatsuya
Published: March 2009

Unit 1 & 2 in IKATA power station, both are two-loop PWRs, have experienced major component replacement. The components are two steam generators (SG) and a reactor pressure vessel closure head (VH) for each plant and were replaced with improved ones. This paper explains the reasons for the replacement and major improvements in the components.
The Characteristics of Steam Bubbles in Subcooled Boiling Flow
Authors: Takemoto Takatoshi , Matsuzaki Mitsuo , Kikura Hroshige , Aritomi Masanori , Bunyajitradulya Asi
Published: March 2009

In two-fluid modeling and three-fluid modeling, the accurate prediction of the interfacial area concentration, interfacial heat transfer and interfacial shear stress, were required. In this works, the axial profiles of void fraction, interfacial area concentration and interfacial heat transfer coefficient along the flow direction could be measured. For the steam bubbles whose diameter were less than 8mm, the interfacial area concentration and the mean bubble diameter had a correlation with void fraction despite the variation of liquid flow rate and subcooling. In case the steam bubble collapse occurred due to an irregular bubble condensation and a turbulence of liquid flow, interfacial heat transfer coefficient with the bubble collapse was about twice of that without a bubble collapse. And the interfacial heat transfer coefficient without bubble collapse showed a good agreement with the correlation proposed by Akiyama. In addition, the supposed image processing method could be applied to the present experimental condition.
Development of Reconstitution Technology for Surveillance Specimens
Authors: Atago Yasushi , Hatano Shunichi , Otsuka Eiichiro
Published: March 2009

The Japan Power Engineering and Inspection Corporation (JAPEIC) has been carrying out the project titled “Nuclear Power Plant Integrated Management Technology (PLIM)” consigned by Japanese Ministry of Economy, Trade and Industry (METI) since 1996FY as a 10-years project. As one of the project themes, development of reconstitution technology for reactor pressure vessel (RPV/RV) surveillance specimens, which are installed in RPVs to monitor the neutron irradiation embrittlement on RPV/RV materials, is now on being carried out to deal with the long-term operation of nuclear power plants. The target of this theme is to establish the technical standard for applicability of reconstituted surveillance specimens including the reconstitution of the Charpy specimens and Compact Tension (CT) specimens. With the Charpy specimen reconstitution, application of 10mm length inserts is used, which enables the conversion of tests from the LT-direction to the TL-direction. This paper presents the basic data from Charpy and CT specimens of RPV materials using the surveillance specimens obtained for un-irradiated materials including the following. 1) Reconstitution Technology of Charpy Specimens. a) The interaction between plastic zone and Heat Affected Zone (HAZ). b) The effects of the possible deviations from the standard specimens for the reconstituted specimens. 2) Reconstitution Technology of CT specimens. a) The correlation between fracture toughness and plastic zone width. Because the project is now in progress, this paper describes the outline of the results obtained as of the end of 2000 FY.
The Development and Application of SCDAP-3D©
Authors: Coryell E. W. , Harvego E. A. , Siefken L. J.
Published: March 2009

The SCDAP-3D© computer code (Coryell 2001) has been developed at the Idaho National Engineering & Environmental Laboratory (INEEL) for the analysis of severe reactor accidents. A prominent feature of SCDAP-3D© relative to other versions of the code is its linkage to the state-of-the-art thermal/hydraulic analysis capabilities of RELAP5-3D©. Enhancements to the severe accident models include the ability to simulate high burnup and alternative fuel, as well as modifications to support advanced reactor analyses, such as those described by the Department of Energy’s Generation IV (GenIV) initiative. Initial development of SCDAP-3D© is complete and two widely varying but successful applications of the code are summarized. The first application is to large break loss of coolant accident analysis performed for a reactor with alternative fuel, and the second is a calculation of International Standard Problem 45 (ISP-45) or the QUENCH 6 experiment.
Transient Analysis Needs for Generation IV Reactor Concepts
Authors: Siefken L. J. , Harvego E. A. , Coryell E. W. , Davis C. B.
Published: March 2009

The importance of nuclear energy as a vital and strategic resource in the U. S. and world’s energy supply mix has led to an initiative, termed Generation IV by the U.S. Department of Energy (DOE), to develop and demonstrate new and improved reactor technologies. These new Generation IV reactor concepts are expected to be substantially improved over the current generation of reactors with respect to economics, safety, proliferation resistance and waste characteristics. Although a number of light water reactor concepts have been proposed as Generation IV candidates, the majority of proposed designs have fundamentally different characteristics than the current generation of commercial LWRs operating in the U.S. and other countries. This paper presents the results of a review of these new reactor technologies and defines the transient analyses required to support the evaluation and future development of the Generation IV concepts. The ultimate objective of this work is to identify and develop new capabilities needed by INEEL to support DOE’s Generation IV initiative. In particular, the focus of this study is on needed extensions or enhancements to SCDAP/RELAP5/3D code. This code and the RELAP5-3D code from which it evolved are the primary analysis tools used by the INEEL and others for the analysis of design-basis and beyond-design-basis accidents in current generation light water reactors.
Collection of Spent Solid Radioactive Sources
Authors: Robbe Marie-France , Matton Sandrine , Lamadie Fabrice , Devenelle Pascal , Vincon Eric , Laloubere Laurent
Published: March 2009

In France, the main producers of radwastes are EDF, COGEMA and its subsidiaries in the frame of the fuel cycle activities, CEA for the research activities principally on reactors or the fuel cycle. The management and treatment of radwastes are mainly performed by ANDRA, COGEMA, EDF and CEA.
Next Generation CANDU Core Physics Innovations
Authors: Chan P. S. W. , Hopwood J. M. , Love J. W.
Published: March 2009

NG CANDU is the “Next Generation” CANDU® reactor, aimed at producing electrical power at a capital cost significantly less than that of the current reactor designs. A key element of cost reduction is the use of H2O as coolant and Slightly Enriched Uranium fuel in a tight D2O-moderated lattice. The innovations in the CANDU core physics result in substantial improvements in economics as well as significant enhancements in reactor licensability, controllability, and waste reduction. The full-core coolant-void reactivity in NG CANDU is about −3 mk. Power coefficient is substantially negative. Fuel burnup is about three times the current natural-uranium burnup.
STAR-H2: The Secure Transportable Autonomous Reactor for Hydrogen Production and Desalinization
Authors: Wade D. C. , Doctor R. , Peddicord K. L.
Published: March 2009

The Secure Transportable Autonomous Reactor for Hydrogen production is a modular fast reactor intended for the mid 21st century energy market wherein electricity and hydrogen are employed as complementary energy carriers and nuclear energy contributes to sustainable energy supply based on full transuranic recycle in a passively safe, environmentally friendly and proliferation-resistant manner suitable for widespread worldwide deployment.
A PVM Executive Program for Use With RELAP5-3D©
Authors: Weaver W. L. , Tomlinson E. T. , Aumiller D. L.
Published: March 2009

A PVM executive program has been developed for use with the RELAP5-3D© computer program. The PVM executive allows RELAP5-3D© to be coupled with any number of other computer programs to perform integrated analyses of nuclear power reactor systems and related experimental facilities. The executive program manages all phases of a coupled computation. It starts up and configures a virtual machine, spawns all of the coupled processes, coordinates the time step size between the coupled codes, manages the production of printed and plotable output, and shuts the virtual machine down at the end of the computation. The executive program also monitors that status of the coupled computation, repeating time steps as needed and terminating a coupled computation gracefully if one of the coupled processes is terminated by the computational node on which it is executing.
Bifurcation Analysis of Nuclear-Coupled Thermal Hydraulics of BWR Using BIFDD
Authors: Zhou Quan , Rizwan-Uddin
Published: March 2009

Stability and bifurcation analyses of nuclear-coupled thermal hydraulic instability in BWR has been performed using a semi-analytical method. The BWR model used in this study consists of three parts: neutron kinetics, fuel rod heat conduction and single and two-phase heated channel thermal hydraulics. Point reactor model is currently being used for neutron kinetics and will be extended in the future to higher order lambda or omega-mode. In the heat conduction part, a piecewise quadratic approximation to radial temperature distribution in fuel pellet and cladding is assumed. ODEs for the expansion coefficients of the quadratic spatial profiles are developed by applying variational principle. Similar to the heat conduction model, the spatial enthalpy distribution in the single phase region and steam quality in the two-phase region in the BWR core are approximated by quadratic polynomials. Two-phase flow is modeled using the homogeneous equilibrium model. A bifurcation analysis code, BIFDD, is then used to perform the analysis for the stability boundary (SB) and the nature of Poincare´-Andronov-Hopf bifurcation (PAH-B). Results in control-rod-induced-reactivity—inlet-subcooling-number space show that both super or sub-critical bifurcation can occur along the SB—the subcritical bifurcation occurs for very small or very large subcooling number values; super-critical PAH-B occurs for intermediate values of subcooling number.
Stability and Bifurcation Analyses of Two-Phase Flow Using a Drift Flux Model and Bifurcation Code BIFDD
Authors: Dokhane A. , Hennig D. , Chawla R. , Rizwan-Uddin
Published: March 2009

Nonlinear boiling water reactor (BWR) studies are generally pursued by employing large system codes, like RAMONA and RETRAN. However, detailed investigations using this approach in which parametric studies are to be carried out over large regions of parameter space are very expensive and time consuming. Therefore, it is necessary to use the so-called reduced order models containing a “manageable” number of system equations incorporating the essential features of the physical phenomena. Since the thermal-hydraulic model determines the main feedback gain and the associated time delay (void feedback reactivity), the modelling of the fluid dynamics is of paramount importance for the reduced order model analyses for BWRs. For a certain set of hydraulic parameter values, the system of nonlinear differential equations describing the fluid dynamics generates the so-called self-sustained density wave oscillations (DWO), which is a typical two-phase flow instability.
A Novel Molten Salt Reactor Concept to Implement the Multi-Step Time-Scheduled Transmutation Strategy
Authors: Csom Gyula , Feher Sandor , Szieberth Mate
Published: March 2009

Nowadays the molten salt reactor (MSR) concept seems to revive as one of the most promising systems for the realization of transmutation. In the molten salt reactors and subcritical systems the fuel and material to be transmuted circulate dissolved in some molten salt. The main advantage of this reactor type is the possibility of the continuous feed and reprocessing of the fuel. In the present paper a novel molten salt reactor concept is introduced and its transmutational capabilities are studied. The goal is the development of a transmutational technique along with a device implementing it, which yield higher transmutational efficiencies than that of the known procedures and thus results in radioactive waste whose load on the environment is reduced both in magnitude and time length. The procedure is the multi-step time-scheduled transmutation, in which transformation is done in several consecutive steps of different neutron flux and spectrum. In the new MSR concept, named “multi-region” MSR (MRMSR), the primary circuit is made up of a few separate loops, in which salt-fuel mixtures of different compositions are circulated. The loop sections constituting the core region are only neutronically and thermally coupled. This new concept makes possible the utilization of the spatial dependence of spectrum as well as the advantageous features of liquid fuel such as the possibility of continuous chemical processing etc. In order to compare a “conventional” MSR and a proposed MRMSR in terms of efficiency, preliminary calculational results are shown. Further calculations in order to find the optimal implementation of this new concept and to emphasize its other advantageous features are going on.
Assessment of the Mechanical Effects of a HCDA in a Simple Scale Model of a LMFBR
Authors: Robbe M. F. , Lepareux M. , Cariou Y.
Published: March 2009

To ensure a high safety level to the French nuclear power plants safety authorities impose to the plant designers to take into consideration several cases of hypothetical severe accidents to which the plant must be able to resist. The aim is to guarantee that even a core melt accident would not require any stringent countermeasures such as evacuation or relocation of the population beyond the immediate vicinity of the plant.
Progress in the Development of J Estimation Schemes for the RSE-M Code
Authors: Le Delliou Patrick , Sermage Jean-Philippe , Cambefort Pierre , Barthelet Bruno , Gilles Philippe , Michel Bruno
Published: March 2009

The RSE-M Code provides rules and requirements for in-service inspection of French Pressurized Water Reactor power plant components. The Code gives non mandatory guidance for analytical evaluation of flaws. To calculate the stress intensity factors in pipes and shells containing semi-elliptical surface defects, influence coefficients are given for a wide range of geometrical parameters. To calculate the J integral for a circumferential surface crack in a straight pipe, simplified methods are available in the present version of the Code (2000 Addenda) for mechanical loads (in-plane bending and torsion moments, pressure), thermal loads as well as for the combination of these loads. This paper presents the recent advances in the development of J-estimation schemes for two configurations: • a longitudinal surface crack in a straight pipe, • a longitudinal surface crack in the mid-section of an elbow.
The Consideration of Entrainment in the Drift-Flux Correlation Package MDS
Authors: Hoeld Alois
Published: March 2009

A comprehensive drift-flux correlation package (MDS) based on the SONNENBURG drift-flux correlation has been established. Its aim is to support thermal-hydraulic mixture-fluid models needed for the simulation of the steady state and transient behaviour of characteristic thermal-hydraulic parameters for single- or two-phase fluids flowing up- and downwards along coolant channels of different types (e.g., channel elements of NPP-s, steam generators etc.). Hence, the resulting package MDS had to give special consideration to the two-phase behaviour at co- and counter-current flow situations, its inverse solutions needed for steady state simulations and its behaviour when approaching the lower or upper boundary of a two-phase region. Its characteristic properties, its verification and behaviour with respect to other correlations have (together with an adequate driver code MDSDRI) already be at the ICONE-9 conference at Nice (April 2000). The extension of the code package to situations where droplet entrainment can be expected is subject of this paper. It will be demonstrated that entrainment can, according to the criteria by ISHII-GROLMES (inception) and Ishii-MISHIMA (entrainment fraction), only take place if the given total mass flux exceeds a certain lower (mass flux) limit, a limit being only dependent on system pressure and geometry data. The same is the case for the appearance of counter-current flow. It can only be expected if the given total mass flux is situated within a certain window, a window lying in a low flux range. It is thus not overlapping with the entrainment region, proving that entrainment is not a pre-stadium of counter-current flow. Test calculations will demonstrate the ability of the code package MDS to calculate two-phase flow behaviour along a coolant channel within a wide range of upwards and downwards flow conditions and give an insight over the influence of entrainment at high void fractions and, at lower mass flux regions, of counter-current flow to the overall two-phase behaviour.