bluuebunny/crossref_metadata_2025 huggingface dataset:
Rupture Frequency of CANDU® Large-Diameter Primary Heat Transport Piping Estimated Using Probabilistic Fracture Mechanics Codes
Authors: McLean Rob , Duan Xinjian , Kozluk Michael J.
Published: April 2011

This paper presents a pilot study of using probabilistic fracture mechanics codes (PRO-LOCA 2009 and WinPRAISE 2007) to estimate the rupture frequency of CANDU® large diameter Primary Heat Transport (PHT) piping. The results of this study show that WinPRAISE 2007 and PRO-LOCA 2009 produce comparable trends for the predicted probability of leak and probability of large break leak. There is a number of sensitive leak detection methods available in CANDU plants. The materials and quality of fabrication and sensitive leak detection results in the total probability of a large break leak in the large diameter PHT piping welds being estimated to be on the order of 1E−8 breaks per plant per year. The results of the pilot study indicate that probabilistic fracture mechanics codes could be used to demonstrate that a shutdown action limit of 100 kg/h is sufficient to ensure the probability of rupture of large diameter PHT piping welds is extremely low.
Thermal Ageing Embrittlement of Casting Duplex Stainless Steels for Nuclear Power Plant
Authors: Yao Yuhong , Wei Jianfeng , Liu Jiangnan , Wang Zhengpin , Wang Yu
Published: April 2011

Cast duplex stainless steels (CSS) used for PWR pipes are degraded due to thermal ageing embrittlement during long-term service at 288 °C to 327 °C. Z3CN20-09M Cast duplex Stainless Steels (CSS) made in France for domestic nuclear power plants were thermally aged at 400 °C for 100 h, 300 h, 1000 h, 3000 h and 10000 h. The tensile properties and the impact properties at different thermal aging duration were measured and the effects of the thermal aging time on the microscopic structures and substructures of Z3CN20-09M were respectively investigated by optical microscopy and transmission electron microscopy. The results showed that the tensile strengths of Z3CN20-09M CSS increased gradually with the increment of the thermal ageing time, whereas the impact properties decreased with the prolonging of the thermal ageing time. After long thermal ageing time the dislocation configurations were greatly changed in austenite, and there were precipitates along the austenite-ferrite interface. Moreover, the iron-rich α phase and the chromium-rich α phase precipitated in ferrite aged for 10000h by nucleation and growth rather than the spinodal decomposition. All of above revealed that Z3CN20-09M CSS became brittle during thermal ageing.
General Layouts of Supercritical-Water NPPs
Authors: Pioro I. , Naidin M. , Mokry S. , Saltanov Eu. , Peiman W. , King K. , Farah A. , Thind H.
Published: April 2011

Currently, there are a number of Generation IV SuperCritical Water-cooled nuclear Reactor (SCWR) concepts under development worldwide. The main objectives for developing and utilizing SCWRs are: 1) Increase gross thermal efficiency of current Nuclear Power Plants (NPPs) from 30–35% to approximately 45–50%, and 2) Decrease capital and operational costs and, in doing so, decrease electrical-energy costs. SuperCritical Water (SCW) NPPs will have much higher operating parameters compared to current NPPs (i.e., steam pressures of about 25 MPa and steam outlet temperatures up to 625°C). Additionally, SCWRs will have a simplified flow circuit in which steam generators, steam dryers, steam separators, etc. will be eliminated. Furthermore, SCWRs operating at higher temperatures can facilitate an economical co-generation of hydrogen through thermo-chemical cycles (particularly, the copper-chlorine cycle) or direct high-temperature electrolysis. To decrease significantly the development costs of an SCW NPP, to increase its reliability, and to achieve similar high thermal efficiencies as the advanced fossil-fired steam cycles, it should be determined whether SCW NPPs can be designed with a steam-cycle arrangement that closely matches that of mature SuperCritical (SC) fossil-fired thermal power plants (including their SC-turbine technology). The state-of-the-art SC-steam cycles at fossil-fired power plants are designed with a single-steam reheat and regenerative feedwater heating. Due to this, they reach thermal steam-cycle efficiencies up to 54% (i.e., net plant efficiencies of up to 43–50% on a Higher Heating Value (HHV) basis). This paper presents several possible general layouts of SCW NPPs, which are based on a regenerative-steam cycle. To increase the thermal efficiency and to match current SC-turbine parameters, the cycle also includes a single steam-reheat stage. Since these options include a nuclear steam-reheat stage, the SCWR is based on a pressure-tube design.
Thermal Aspects of Using Uranium Mononitride Fuel in a SuperCritical Water-Cooled Reactor at Maximum Heat Flux Conditions
Authors: Milner Ashley , Pascoe Caleb , Patel Hemal , Peiman Wargha , Richards Graham , Pioro Igor
Published: April 2011

Generation IV nuclear reactor technology is increasing in popularity worldwide. One of the six Generation-IV-reactor types are SuperCritical Water-cooled Reactors (SCWRs). The main objective of SCWRs is to increase substantially thermal efficiency of Nuclear Power Plants (NPPs) and thus, to reduce electricity costs. This reactor type is developed from concepts of both Light Water Reactors (LWRs) and supercritical fossil-fired steam generators. The SCWR is similar to a LWR, but operates at a higher pressure and temperature. The coolant used in a SCWR is light water, which has supercritical pressures and temperatures during operation. Typical light water operating parameters for SCWRs are a pressure of 25 MPa, an inlet temperature of 280–350°C, and an outlet temperature up to 625°C. Currently, NPPs have thermal efficiency about of 30–35%, whereas SCW NPPs will operate with thermal efficiencies of 45–50%. Furthermore, since SCWRs have significantly higher water parameters than current water-cooled reactors, they are able to support co-generation of hydrogen. Studies conducted on fuel-channel options for SCWRs have shown that using uranium dioxide (UO2) as a fuel at supercritical-water conditions might be questionable. The industry accepted limit for the fuel centerline temperature is 1850°C and using UO2 would exceed this limit at certain conditions. Because of this problem, there have been other fuel options considered with a higher thermal conductivity. A generic 43-element bundle for an SCWR, using uranium mononitride (UN) as the fuel, is discussed in this paper. The material for the sheath is Inconel-600, because it has a high resistance to corrosion and can adhere to the maximum sheath-temperature design limit of 850°C. For the purpose of this paper, the bundle will be analyzed at its maximum heat flux. This will verify if the fuel centerline temperature does not exceed 1850°C and that the sheath temperature remains below the limit of 850°C.
The ATWS Analysis of One Control Rod Withdrawal Out of the HTR-10GT Core
Authors: Lang Minggang , Dong Yujie
Published: April 2011

The 10MW High Temperature Gas Cooled Test Reactor (HTR-10) has been built in Institute of Nuclear and New Energy Technology (INET) and has been operating successfully since the beginning of 2003. The core outlet temperature of HTR-10 is 700°C. To verify the technology of gas-turbine direct cycle, at first INET had a plan to increase its core outlet temperature to 750°C and use a helium gas turbine instead of the steam generator (then the reactor is called HTR-10GT). Though HTR-10 has good intrinsic safety, the design basic accidents and beyond design basis accidents of HTR-10GT must be analyzed according to China’s nuclear regulations due to changed operation parameters. THERMIX code system is used to study the ATWS accident of one control rod withdrawal out of the core by a mistake. After a control rod in the side reflector was withdrawn out at a speed of 1 cm/s by a mistake, a positive reactivity was inserted and the reactor power increased and the temperature of the core increased. When the neutron flux of power measuring range exceeded 123% and the core outlet temperature was greater than 800°C, the reactor should scram. It was supposed that all the control rods in the reflectors had been blocked and the reactor could not scram. Thus the accident went on and the core temperature and the system pressure increased but the reactor shutdown at last because of its natural negative temperature reactivity feedback mechanism. The residual heat would be removed out of the core by the cavity cooling system. During the accident sequence the maximum fuel temperature was 1242.4°C. It was a little higher than 1230°C–the fuel temperature limitation of HTR-10. Now the sphere fuel used in HTR-10GT will also be used in HTR-PM and the temperature limitation raised to 1620°C, so the HTR-10GT is safe during the ATWS of one control rod withdrawal out of the core. The paper also compares the analysis result of HTR10-GT to those of HTR-10. The results shows that the HTR-10GT is still safe during the accident though its operating temperature is higher than HTR-10. The analysis will be helpful to HTR-PM because they have the same outlet temperature of the core.
Discussion of the Effects of AP1000 First Core Loading Pattern Design on Nuclear Power Plant Operation
Authors: Li Xuesong
Published: April 2011

Core loading pattern design has great influence on nuclear power plant operation. An excellent core loading pattern can not only enhance operation factor, reduce operation cost, but also increase operation safety. Under the premise of nuclear safety, AP1000 first core loading pattern achieves the goal of low leakage loading by simulating the reactivity distribution of the 18-month Equilibrium Cycle design. The fuel management presented in this paper illustrates the economic performance and technical feasibility of the advanced 18-month cycle first core fuel. The advanced feature of this first core design include: the development of loading pattern that simulates the reactivity distribution of a typical low leakage reload core in order to reduced leakage, the use of radial enrichment zoning in higher enriched assemblies to lower peaking factors, and the use of axial burnable absorber zoning to improve axial power shape control and reduce axial peaking. The discussion provided in this paper demonstrates the ability of the advanced first core design operate safety and efficiently, and the core is designed with adequate peaking factor margin in both base-load and load-follow operation. Finally, this paper analyses the impact brought about by multi-enrichment on the nuclear power plant operation capacity and operation cost, and arrive at a conclusion that precise fuel cycle evaluation such that the enrich uranium and operation costs can be accurately quantified and control in a detailed economic evaluation.
Proactive Aging Management: PAM 2.0
Authors: Perdue Robert K. , Kelley Matthew H. , O’Rourke Raymond E. , Woodcock Joel , Zhang Mengyi , Zheng Hui
Published: April 2011

Proactive aging management or PAM is defined as the use of preventive maintenance, spares provisioning, condition monitoring and other strategies to economically optimize the availability of a plant while maintaining existing safety margins. Westinghouse has developed a set of teaching modules and supporting software, PAM 2.0, to help transfer this technology to China’s nuclear stations. This paper describes the content of the teaching modules and PAM 2.0 and how they can be used to establish an optimized total reliability and maintenance program for a new or existing nuclear plant. It also provides a perspective on the future of PAM in China provided by two Chinese engineers who took the course.
Democratization of 3D Applications in Nuclear Power Plants: Laser Scanning 3D Technology Implementation on Spanish PWRs
Authors: Jardi´ Xavier , Anga´s Jorge
Published: April 2011

On January 2008, the US NRC issued the Generic Letter 2008-01 [1], “Managing gas accumulation in emergency core cooling, decay heat removal and containment spray systems”. Among other responses, this letter requires an evaluation of locations sensitive to accumulate gases in several safety systems. In order to get accurate data related to the real slope of horizontal pipes and other geometrical parameters needed for this evaluation, laser scanning and 3D modeling techniques have been applied in Spanish Nuclear Power Plants. From October 2008 to December 2009, five Spanish units have been scanned and modeled. As a result of these activities, the plants have obtained detailed 3D models as well as 2D as-built drawings of the selected components. These models were integrated in 3D web servers which give a panoramic view of the scanned areas and permitted measurements in the local coordinate system of the plant. Moreover, the 2D elevation drawings included accurate and useful information for the plants in order to make decisions related to the GL-2008-01 requirements. The geometric information generated in the frame of the GL-2008-01 activities is being currently used for alternative applications. For instance, laser scanning technology is being used to enhance design modification procedures. A pilot project on the MSRs replacement is being currently carried out with successful results. This technology has the advantage that new components from CAD software can be updated in the as-built models obtained through laser scanning. In addition to this, it’s very easy to check fitting and interferences, and also to make accurate measurements and handling simulations. The potential applications in personnel training and radiological protection are also very important. The panoramic viewers on 3D web servers are versatile and could fit the specific requirements of each organization. Regarding staff training, virtual tours and component seekers are being currently developed. These tools provide a significant save of time and dose and also give independence for each person to get to the working place without external help or time-consuming paper consulting. Integration with existing plant databases is also possible through the panoramic viewers and is currently being developed for In-Service Inspections and Maintenance applications. The main advantage of these products is their accessibility with free visors which don’t need specific training. Therefore, the implementation of these tools doesn’t need additional investments. In conclusion, Laser 3D Technology Applications set the first step on the democratization of these powerful 3D environments among common users as integrated tools in their daily work.
Analysis of Tritium Production and Distribution in HTR-PM
Authors: Wu Yuanyu , Cao Jianzhu
Published: April 2011

The research on radiological impact of tritium is highly concerned in high temperature gas-cooled reactors. In order to better assess the environmental performance of HTR-PM (HTR demonstration project with 2 × 250MW plants), analysis of tritium behavior in HTR-PM is conducted in this paper. The main production sources of tritium are the ternary fission in the fuel and neutron capture reactions of some nuclides. Based on the tight interactions between tritium sources and sinks, differential equations are built to describe tritium behavior in primary and secondary loop. Specific analysis is conducted to tritium permeation through heat exchanger walls to secondary loop, considering the oxidation of alloys used for heat exchanger. Applied with the parameters of HTR-PM, tritium concentration in primary and secondary loop is calculated, and the amount of tritium released to the environment is evaluated. The evaluation shows that the amount of tritium released to the environment is less than the limit value prescribed by Chinese regulation on radiation protection. The calculation results can also be applied to the safety analysis and used to guide the design of relevant systems and equipments for the HTR-PM.
The Application of Uniform Design Table in Cement Solidification of Nuclear Waste Resin
Authors: Wu Minghui , Li Qiang , Ouyang Shixi
Published: April 2011

As a lower cost raw material, few demand for equipments, convenient solidification process, cement solidification for radioactive waste is widely used for several decades. Formulations of solidification are complex and diverse, involving various types of substrate and additives. Traditional approach for formulation design is single-factor test whose representation is inadequate and workload is huge. Uniform design based on the theory of Quasi-Monte-Carlo takes advantage of limited and representative tests instead of the system. In the multi-factor formulation design, it can be very quick and convenient to find the formulations required by uniform design table and direct-vision method. This article introduced the application of uniform design table for formulation in cement solidification of nuclear waste resin.
Transient Response Simulation of High Temperature Gas-Cooled Reactor With Helium Turbine for Secondary Inlet Temperature Change of Precooler and Intercooler
Authors: Li Wenlong , Zhang Zuoyi , Xie Heng , Dong Yujie
Published: April 2011

A simulation program has been developed to study the transient characteristics of high temperature gas-cooled reactor with indirect helium turbine cycle. It is assumed that cooling water inlet temperature of precooler or intercooler has a step increase from 20°C to 25°C. Simulation result shows that generator power decreases accompanied with inlet temperature increase of cooling water of precooler or intercooler and precooler accident causes much bigger power loss for generator than intercooler accident. Both of surge margins of LPC and HPC have step decreases due to these two accidents. For this precooler accident, physical parameters of primary loop keep basically unchanged. For this intercooler accident, it causes reactor power loss, temperature increase of reactor inlet and temperature decrease of reactor outlet.
Numerical Analysis of Dual-Flow Fields in Mini-TBM for Chinese LiPb Experimental Loop Dragon-IV
Authors: Wang Wei-hua , Zhu Zhi-qiang , Huang Qun-ying , Ling Xin-zhen , Li Jin-ling
Published: April 2011

Mini-TBM will be tested in chinese LiPb experimenttal loop Dragon-IV to validate the thermal-hydraulic effect of DFLL-TBM, such as dual-flow fields heat transfer, temperature fields, velocity fields, flux distribution of liquid lithium lead and helium gas. It is difficult to measure the detailed dual-flow fields of liquid metal LiPb and helium gas in mini-TBM. Three dimensions numerical analysis of the LiPb and helium gas flow and heat transfer in Mini-Test Blanket Module (TBM) therefore has been carried out using the CFD code FLUENT. The detailed dual-flow fields, which include temperature, velocity, pressure and heat transfer of liquid LiPb and helium gas, are presented to support for the test of mini-TBM, and to supply more robust database and make a significant joint contribution to the future TBM testing in EAST and ITER, and also optimize and improve the design of DFLL-TBM system for ITER.
Thermal Integration Options for Hydrogen Cogeneration With Molten Salt Nuclear Reactors
Authors: Wang Z. , Naterer G. F. , Gabriel K. S.
Published: April 2011

Thermochemical hydrogen cogeneration using heat of molten salt nuclear reactors (MSRs) is discussed in this paper. Sulfur-iodine and copper-chlorine cycles are taken as typical examples for analysis and discussion. It is found that the heat exchanger design is predominately determined by the maximum and range of temperatures of themochemical hydrogen production cycles with MSRs. Copper-chlorine (Cu-Cl) thermochemical cycles can link with most MSRs, but sulfur-iodine (S-I) cycles can only link with very high temperature MSRs. The location of extracted heat from MSRs to S-I and Cu-Cl cycles is investigated, and its influence on the layout of nuclear reactor coolant loop is discussed. Some conceptual designs of heat exchangers are proposed to transfer heat from MSRs to Cu-Cl and S-I cycles. The available heat quantity at different hours of a day and corresponding hydrogen production scales are determined. It is found that the available heat at most hours of power demand in a day is equivalent to the hydrogen cogeneration capacity of an industrial scale steam methane reforming plant, if an MSR power station is operating at an invariable maximum power, independent of an electrical load throughout a day or year.
Recent GEH Criticality Safety Activities in Support of the Nuclear Fuel Cycle
Authors: Paulson Lon , Zino John , Ao Qi
Published: April 2011

The GE Hitachi Nuclear Energy (GEH) nuclear criticality safety (NCS) function remains actively engaged in advancements to the nuclear fuel cycle. In addition to its traditional BWR fuel manufacturing, recent GEH emphasis to become more vertically integrated into front end (enrichment) and back end (reprocessing) fuel cycle technologies has had a dramatic impact on the NCS function. Required fundamental and practical research in various fields, such as general physics, computational methods, validation methodology, cross-section data processing, criticality safety assessments, risk-informed integrated safety analyses, and domestic and international nuclear packaging licensing, collectively present significant challenges to NCS staff. As the landscape of the GEH business growth opportunities continues to evolve over time, so does the required depth of NCS knowledge and technical expertise. This paper provides an overview of select NCS design, licensing, methods, and packaging activities in support of GEH nuclear fuel cycle business subsidiaries and concludes with some insight to technical and regulatory challenges.
Measurement of Gamma Radiation Profile Inside a CANDU® Reactor
Authors: Yue S. , Smith B. H. , Jonkmans G. , Sur B. , Comeau D. , Taylor D.
Published: April 2011

An instrument based on a small and inexpensive silicon diode was developed and used to measure, for the first time, the gamma radiation exposure rate within a start-up instrumentation guide tube in a shut down CANDU® reactor undergoing refurbishment. The shape of the measured profile agreed with the expectation; however, the maximum measured exposure rate was about four times higher than the calculation. Two adjuster rods located above the core and adjacent to the guide tube, could be clearly identified from the features of the radiation profile measurement. These measurements provide confirmatory information for the creation of safe and effective radiation work plans and selection of appropriate in-core instrumentation. This technology can be used to accurately measure the positions of irradiated reactor components, such as adjuster rods, pressure tubes, calandria tubes, reactivity control absorbers and other sources of gamma radiation inside a shut down reactor.
Development of a Standard for Verification and Validation of Software Used to Calculate Nuclear System Thermal Fluids Behavior
Authors: Harvego Edwin A. , Schultz Richard R. , Crane Ryan L.
Published: April 2011

With the resurgence of nuclear power and increased interest in advanced nuclear reactors as an option to supply abundant energy without the associated greenhouse gas emissions of the more conventional fossil fuel energy sources, there is a need to establish internationally recognized standards for the verification and validation (V&V) of software used to calculate the thermal-hydraulic behavior of advanced reactor designs for both normal operation and hypothetical accident conditions. To address this need, ASME (American Society of Mechanical Engineers) Standards and Certification has established the V&V 30 Committee, under the jurisdiction of the V&V Standards Committee, to develop a consensus standard for verification and validation of software used for design and analysis of advanced reactor systems. The initial focus of this committee will be on the V&V of system analysis and computational fluid dynamics (CFD) software for nuclear applications. To limit the scope of the effort, the committee will further limit its focus to software to be used in the licensing of High-Temperature Gas-Cooled Reactors. In this framework, the Standard should conform to Nuclear Regulatory Commission (NRC) and other regulatory practices, procedures and methods for licensing of nuclear power plants as embodied in the United States (U.S.) Code of Federal Regulations and other pertinent documents such as Regulatory Guide 1.203, “Transient and Accident Analysis Methods” and NUREG-0800, “NRC Standard Review Plan”. In addition, the Standard should be consistent with applicable sections of ASME NQA-1-2008 “Quality Assurance Requirements for Nuclear Facility Applications (QA)”. This paper describes the general requirements for the proposed V&V 30 Standard, which includes; (a) applicable NRC and other regulatory requirements for defining the operational and accident domain of a nuclear system that must be considered if the system is to be licensed, (b) the corresponding calculation domain of the software that should encompass the nuclear operational and accident domain to be used to study the system behavior for licensing purposes, (c) the definition of the scaled experimental data set required to provide the basis for validating the software, (d) the ensemble of experimental data sets required to populate the validation matrix for the software in question, and (e) the practices and procedures to be used when applying a validation standard. Although this initial effort will focus on software for licensing of High-Temperature Gas-Cooled Reactors, it is anticipated that the practices and procedures developed for this Standard can eventually be extended to other nuclear and non-nuclear applications.
VERLIFE - Unified Procedure for Lifetime Assessment of Components and Piping in WWER NPPs During Operation: Updating and Further Development
Authors: Brumovsky Milan
Published: April 2011

VERLIFE — “Unified Procedure for Lifetime Assessment of Components and Piping in WWER NPPs during Operation” was developed within the 5th Framework Program of the European Union in 2003 and later upgraded within the 6th Framework Program “COVERS – Safety of WWER NPPs” of the European Union in 2008. This Procedure had to fill the gap in original Soviet/Russian Codes and Rules for WWER type NPPs, as these codes were developed only for design and manufacturing and were not changed since their second edition in 1989. VERLIFE Procedure is based on these Russian codes but incorporates also new developments in research, mainly in fracture mechanics, and also some principal approaches used in PWR codes. To assure that VERLIFE Procedure will remain a living document, new 3-years IAEA project (in close co-operation with the another project 6th Framework Program of the European Union “NULIFE – Plant Life Management of NPPs”) has started in 2009. Within this project, upgrading/updating of the VERLIFE procedure is prepared together with the extension by (at least) following procedures: - Leak-before-break concept for WWER NPPs; - Reduction of Probability of Break procedure for evaluation of integrity of high-energy piping in NPPs of WWER-440 and WWER-1000 types; - Lifetime of reactor pressure vessel internals; - Risk informed In-service inspection implementation process and organization; - Methodology for Qualification of In-Service Inspection Systems for WWER Nuclear Power Plants; - Component and piping supports; - Monitoring and evaluation of erosion-corrosion damage in piping materials. Final document, after its approval by expert groups of the IAEA and NULIFE, will be issued as “IAEA/NULIFE Guidelines for Integrity and Lifetime Assessment of Components and Piping in WWER NPPs”. The paper will describe these main principles and also future plans.
Surveillance Specimen Program for Lifetime Extension of RPV
Authors: Brumovsky Milan , Kytka Milos
Published: April 2011

Plant life extension (as well as Long Term Operation) to 60 or 80 years of operation also requires a reliable information about the potential irradiation embrittlement (and also thermal ageing) of reactor pressure vessel materials. Such information is usually obtained from testing specimens within the surveillance specimen program that is designed for the design reactor pressure vessel (RPV) life, regularly for 40 years only. Life extension requires modification of such program (if there is still time to perform it) or a design of a new – extended one. Such program should have to contain RPV archive materials that are not in every case available. Thus, combination of archive materials and possible surrogate materials must be taken into account for this program. Some complication can be expected with thermal ageing data as some laboratory tests at higher temperatures must be realized. The paper describes such program for Nuclear Power Plant (NPP) Dukovany, Czech Republic with WWER-440 type reactors.
Modeling Improvements for System Code Evaluation of Inlet Plenum Mixing Under Severe Accident Conditions Using CFD Predictions
Authors: Boyd Christopher , Armstrong Kenneth
Published: April 2011

An updated mixing model is developed for application to system codes used for predicting severe accident-induced failures of steam generator (SG) U-tubes in a pressurized-water reactor. Computational fluid dynamics is used to predict the natural circulation flows between a simplified reactor vessel and the primary side of an SG during a hypothesized severe accident scenario. The results from this analysis are used to extend earlier experimental results and predictions. These new predictions benefit from the inclusion of the entire natural circulation loop between the reactor vessel upper plenum and the SG. Tube leakage and mass flow into the pressurizer surge line also are considered. The predictions are utilized as a numerical experiment to improve the basis for simplified models applied in one-dimensional system codes that are used during the prediction of severe accident natural circulation flows. An updated inlet plenum mixing model is proposed that accounts for mixing in the hot leg as well as the inlet plenum region. The new model is consistent with the predicted behavior and can account for flow into a side-mounted pressurizer surge line if present. Sensitivity studies demonstrate the applicability of the approach over a range of conditions. The predictions are most sensitive to changes in the SG secondary side temperatures or heat-transfer rates at the SG tubes. Grid independence is demonstrated through comparisons with previous models and by increasing the number of cells in the model. This work supports the U.S. Nuclear Regulatory Commission (NRC) studies of SG tube integrity under severe accident conditions.
Power Uprate Evaluation of Main Steam Isolation and Main Steam Check Valve Using Advanced Engineering Models
Authors: Kalsi M. S. , Estep Neal , Leutwyler Zachary
Published: April 2011

Analyses were performed by Kalsi Engineering, Inc. (KEI) on the main steam isolation valves (MSIV) and main steam check valves (MSCV) at 4 pressurized water reactors to evaluate their capability to perform design basis functions at Extended Power Uprate (EPU) conditions. Disc closing velocities and actuator pressures (for MSIV) were calculated by using RELAP™. These inputs included valve specific flow and torque coefficients obtained by detailed Computational Fluid Dynamics (CFD) analyses to accurately predict hydrodynamic torque imposed on the disc throughout the stroke. Stresses, strains and displacements in all of the critical MSCV and MSIV components were determined by performing appropriate 2-D and 3-D elasto-plastic, transient dynamic finite element analysis (FEA). Fatigue life of the components under spurious closure was determined in accordance with applicable ASME Section III Code. The evaluation also included the propensity for excessive wear of the MSCV due to disc flutter, vibration and erosion.
Removal of Uranium From Aqueous Solution by Chitosan and Ferrous Ions
Authors: Wang Jing-song , Bao Zheng-lei , Chen Si-guang , Yang Jin-hui
Published: April 2011

This study focuses on developing a new method to remove uranium from aqueous solution. Chitosan and ferrous ions were used together to remove uranium ions from aqueous solution. Through two-step pH adjustment, the uptake behavior of chitosan and ferrous ions toward uranium in aqueous solution using batch systems were studied in different experimental conditions. The experimental results indicated that the removal of uranium by synergetic effect of chitosan and ferrous ions was more effective than the way of adsorbing uranium ions by chitosan alone. Under the given experimental conditions, the concentration of the residual uranium in the effluent after chitosan and ferrous ions treatment could meet the discharge standard (< 0.05mg·l−1) when initial concentration of uranium ions was 10 mg·l−1 or 100 mg·l−1. The synergetic effect of chitosan and ferrous ions including adsorption, coacervation and coprecipitation, are responsible for the high removal rate of uranium.
Safety System and Control System Separation Requirements for ACR-1000™ and Operating CANDU® Reactors
Authors: Tikku Sunil , Raiskums Gilbert , Harber John , Foster Phil
Published: April 2011

Digital control and safety plus the complete functional and physical separation between control and safety and also between the safety systems have been key long standing principles of CANDU® nuclear reactor technology. This paper presents a historical evolution of these principles that make CANDU reactors one of the safest technologies in the world today. The original Generation II CANDU 6 reactors started with complete separation of control from safety and the division of safety systems into two groups having strong physical separation such as opposite sides of the reactor or reactor building. Within each group a more moderate distance separation was employed. With the advent of distributed computer technology for control and display functions, key processing equipment is now moved out remote from the control rooms and distributed into channelized field equipment rooms around the reactor building as in the Four-Quadrant concept for ACR-1000™. This new approach is immune to total unavailability of any control room or equipment room due to events such as fire with minimal impact to any of the safety systems regardless of their grouping. In addition to physical separation, appropriate functional partitioning, design rules to avoid communication cross links, and diversity principles are applied to computer based I&C systems as defences against common cause faults.
Plutonium Rock-Like Oxide Fuel (ROXf) System: Their Once-Through Burning and Usage
Authors: Mohamed Ashraf Elsayed Mohamed
Published: April 2011

The ROXf is a kind of inert matrix fuel U-238-free matrices, it has a high plutonium transmutation capability. The ROXf consists of chemically stable phases of fluorite ‘stabilized’ ZrO2 or ThO2, and spinel MgAl2O4. In this fuel, PuO2 is solidified in a fluorite phase. With U-238-Free matrices, a large part of the plutonium can be burned after irradiation in conventional LWRs. The spent ROXf consists of natural analogous geologically stable phases, and is disposed directly as high level wastes ‘HLWs’ after about 50 years cooling. From the high plutonium burn up rate and the high stability of the fuel, the ROXf-LWRs system has proliferation resistance and environmental safety. Characteristics of two types of ROXf, Zr-ROX and Th –ROX with weapons-Pu, in an LWR core arrangement are evaluated by cell burn up calculations and 2-D core calculations using the SRAC code system and JENDL-3 nuclear library. In an LWR of moderator to fuel volume ratio = 1.9, which corresponds to current PWRs, Pu transmutation rates with the two types of ROXf are large enough and more than 80% and 99% of Pu and Pu-239, respectively, can be burned. The calculated kinetic parameters indicate less moderate characteristics of ROX cores, especially with Zr-ROX. The fertile Th-232 in Th-ROX works like U-238 in the UO2 fuel, making kinetic parameters more moderate and reactivity drop due to burn up smaller than that Zr-ROX. The neutron capture of Th-232 to generate U-233, causes the safeguards problem. Thus, the characteristics of Zr-ROX as a typical example, were investigated in this study.
Particle Behaviors Under Acceleration Condition in Solid-Air Two Phase Flow
Authors: Zhao Tong , Takei Masahiro
Published: April 2011

This paper presents a numerical study of the particle behaviors under acceleration conditions in the solid-air two-phase flow by means of a combined two-dimensional model of computational fluid dynamics and discrete element method (CFD-DEM). The simulation model provides the information regarding the particle distribution behaviors within the calculation region and the particle run-out rate from the calculation region under different parameter conditions, such as particle size, initial particle loading and particle acceleration condition. The results demonstrate that the particle run-out rate is directly influenced by the particle size and the initial loading condition. The particle acceleration in the horizontal direction adversely affects the particle run-out rate when the initial particle loading condition is dispersed and uniform. However, this adverse effect disappears when the initial particle loading condition becomes concentrate and partial.
Advances in the Application of Remote Weld Repairs
Authors: Gunn Robert
Published: April 2011

Weld repair sites in nuclear systems are very often located in high radiation areas in difficult to reach sites. Additionally it is usually a pre-requisite to shut down the reactor while the repair is performed, ideally during a scheduled maintenance outage. To minimize the risk of an extended outage it is of paramount importance to ensure the repair time is optimized and the technique is reliable. Technical solutions to complex remote weld repairs are rarely available “off the shelf”. While some commercially available products can be adapted to suit an application, the constraints of space, distance and radiation usually require novel and unique designs involving a variety of engineering skills. Of equal importance to the weld repair is the requirement to perform pre and post repair nondestructive examinations (NDE). The application of ultrasonic, eddy current and liquid penetrate examinations requires a similar level of engineering. This paper outlines some of the welding and NDE techniques developed to address unique repairs to nuclear systems.
Wavelets Scaling Function Expansion Resonance Self-Shielding Calculation Based on the DDPM
Authors: Zu Tiejun , Wu Hongchun , Cao Liangzhi , Liu Qingjie , Yang Weiyan
Published: April 2011

The continuous-energy resonance self-shielding calculation method based on wavelets scaling function expansion method is a valuable potential method to solve the complex resonance problem. Within the fast and thermal energy ranges, the standard multi-group treatment is applied, while Daubechies’ wavelets scaling function expansion method is used to discretize the energy variable of neutron flux within the resonant energy range. In this method, the neutron transport equation is transformed to a set of expansion coefficients equations of wavelets scaling functions. Calculation of the coefficients is very time consuming so that a powerful neutron transport calculation method is needed for better calculation efficiency. In this paper, the discrete direction probability method (DDPM) is employed as a tool for solving the wavelet scaling function expansion coefficients. The DDPM combines the desirable features of interface current method as well as the method of characteristics.
Thermal Hydraulics Analysis Under Natural Circulation Condition of the 4S
Authors: Watanabe Hisao , Matsuda Atsuko , Ohno Jun
Published: April 2011

Thermal hydraulics analysis under natural circulation condition of the 4S was performed with a 3 D total integrated model which contains the primary equipment in the reactor vessel (R/V), the intermediate equipment, the reactor auxiliary cooling system (RVACS), and the intermediate reactor auxiliary cooling system (IRACS). In the long term after the reactor trip, the primary coolant, the intermediate coolant, the RVACS air flow, and the IRACS air flow were circulated by natural circulation force. As the result, it has been confirmed that the core temperature was passively decreased gradually in a safe manner. The natural circulation force in the R/V is affected by temperature distribution in the R/V. In order to comprehend the effect of radial heat transfer in the R/V, it was shown the parameter analysis changed the radial heat transfer rate. As the result, the natural circulation flow decreases, as the radial heat transfer increase. On the other hand the core mean temperature decreases because the generated heat in the core was released to the core outer region. Regarding to the inner core temperature, it has tendency to increase.
Development of a Large Diameter Electromagnetic Pump for 4S
Authors: Aizawa Rie , Fujiwara Seiji , Oyamatsu Yasuko , Katsuki Kenji , Funato Minoru , Ota Hiroyuki
Published: April 2011

Toshiba has been developing the Super-Safe, Small and Simple sodium-cooled fast reactor, named as the 4S. This reactor has features of passive safety, long refueling interval up to 30 years and low maintenance requirement.[1] To approach the features, the 4S adopts a high temperature sodium-immersed electromagnetic pump (EM pump) as the primary circulation pump. The rated flow rate and head are 10.6m3/min and 50kPa respectively. EM pump is static equipment with no moving parts, and then it enables the long-term maintenance to be unnecessary. However, there are some technical issues to apply an EM pump to the 4S. The outer diameter of the 4S EM pump is 1.8 times larger than that of 160m3/min EM pump which Toshiba has already developed. [2] Aspect ratio of the 4S EM pump defined by stator diameter to the length is much larger than our EM pump experiences, which may concern generation of flow instability. In order to confirm the design and fabrication procedure, the prototype full-scale EM pump for the 4S was manufactured. This EM pump characteristics were evaluated using electromagnetic flow code named as EAGLE developed by Toshiba, which combined with Maxwell’s equations and Navier-Stokes equation. To confirm characteristic of the completed EM pump, magnetic field in the annuals flow gap was measured in the air along the axial directional in condition of 250V-18Hz. The data were confirmed to agree with the analytical results of EAGLE. In order to demonstrate applicability of EM pump for the 4S, sodium test facility was constructed. The facility consists of one electromagnetic flow meter (EM flow meter), 14 sodium pressure sensors, EM flow meter calibration tank, 600kW cooler and sodium impurity control system. The pump characteristic tests were performed in sodium, it was demonstrated that the EM pump could apply to the 4S as the performance was satisfied with the design rated.
Immunization with Bacillus Spores Expressing Toxin A Peptide Repeats Protects against Infection with Clostridium difficile Strains Producing Toxins A and B
Authors: Permpoonpattana Patima , Hong Huynh A. , Phetcharaburanin Jutarop , Huang Jen-Min , Cook Jenny , Fairweather Neil F. , Cutting Simon M.
Published: April 2011

Clostridium difficile
is a leading cause of nosocomial infection in the developed world. Two toxins, A and B, produced by most strains of
C. difficile
are implicated as virulence factors, yet only recently has the requirement of these for infection been investigated by genetic manipulation. Current vaccine strategies are focused mostly on parenteral delivery of toxoids. In this work, we have used bacterial spores (
Bacillus subtilis
) as a delivery vehicle to evaluate the carboxy-terminal repeat domains of toxins A and B as protective antigens. Our findings are important and show that oral immunization of the repeat domain of toxin A is sufficient to confer protection in a hamster model of infection designed to closely mimic the human course of infection. Importantly, neutralizing antibodies to the toxin A repeat domain were shown to be cross-reactive with the analogous domain of toxin B and, being of high avidity, provided protection against challenge with a
C. difficile
strain producing toxins A and B (A
+
B
+
). Thus, although many strains produce both toxins, antibodies to only toxin A can mediate protection. Animals vaccinated with recombinant spores were fully able to survive reinfection, a property that is particularly important for a disease with which patients are prone to relapse. We show that mucosal immunization, not parenteral delivery, is required to generate secretory IgA and that production of these neutralizing polymeric antibodies correlates with protection. This work demonstrates that an effective vaccine against
C. difficile
can be designed around two attributes, mucosal delivery and the repeat domain of toxin A.
Increased Disease Severity of Parasite-Infected TLR2−/−Mice Is Correlated with Decreased Central Nervous System Inflammation and Reduced Numbers of Cells with Alternatively Activated Macrophage Phenotypes in a Murine Model of Neurocysticercosis
Authors: Gundra Uma Mahesh , Mishra Bibhuti B. , Wong Kondi , Teale Judy M.
Published: April 2011

In a murine model for neurocysticercosis (NCC), intracranial inoculation of the helminth parasiteMesocestoides cortiinduces multiple Toll-like receptors (TLRs), among which TLR2 is upregulated first and to a relatively high extent. Here, we report that TLR2−/−mice displayed significantly increased susceptibility to parasite infection accompanied by increased numbers of parasites in the brain parenchyma compared to infection in wild-type (WT) mice. This coincided with an increased display of microglial nodule formations and greater neuropathology than in the WT. Parasite-infected TLR2−/−brains exhibited a scarcity of lymphocytic cuffing and displayed reduced numbers of infiltrating leukocytes. Fluorescence-activated cell sorter (FACS) analyses revealed significantly lower numbers of CD11b+myeloid cells, γδ T cells, αβ T cells, and B cells in the brains of parasite-infected TLR2−/−mice. This correlated with significantly reduced levels of inflammatory mediators, including tumor necrosis factor alpha (TNF-α), gamma interferon (IFN-γ), CCL2, CCL3, and interleukin-6 (IL-6) in the central nervous system (CNS) of TLR2−/−mice. As TLR2 has been implicated in immune regulation of helminth infections and as alternatively activated macrophages (AAMs) are thought to play a profound regulatory role in such infections, induction of AAMs in infected TLR2−/−mice was compared with that in WT mice. Parasite-infected WT brains showed larger numbers of macrophages/microglia (CD11b+cells) expressing AAM-associated molecules such as YM1, Fizz1 (found in inflammatory zone-1 antigen), and arginase 1 than TLR2−/−brains, consistent with a protective role of AAMs during infection. Importantly, these results demonstrate that TLR2-associated responses modulate the disease severity of murine NCC.