bluuebunny/crossref_metadata_2025 huggingface dataset:
Fabrication Methods Development for the Korea TBM First Wall in the ITER
Authors: Lee Dong Won , Bae Young Dug , Kim Suk Kwon , Yoon Jae Sung , Hong Bong Guen , Jung Yang Il , Park Jung Yong , Choi Byung Kwon , Jeong Yong Hwan
Published: April 2011

The Republic of Korea has proposed and designed a helium cooled molten lithium (HCML) test blanket module (TBM) to be tested in the ITER. Ferritic Martensitic Steel (FMS) and beryllium are designed to be used as structural material and armor for the TBM first wall (FW) in this design, respectively. In order to develop the fabrication method for the TBM FW, the joining methods of FMS to FMS and Be to FMS have been developed with hot istotatic pressing (HIP). For joining FMS to FMS, mock-ups were fabricated with an HIP (1050 °C, 100 MPa, 2 hours). For joining Be to FMS, two mock-ups were fabricated with the same method (580 °C, 100 MPa, 2 hours) using different interlayers. Then, in order to evaluate the integrity of the fabricated mock-ups, they were tested at the high heat flux (HHF) test facilities, KoHLTs (Korea Heat Load Test) under 1.0 MW/m2 and 0.5 MW/m2 heat fluxes of up to 1000 cycles.
US-APWR Human System Interface System Verification and Validation Program for Digital I&C Design
Authors: Hanada Satoshi , Ito Koji , Mashio Kenji
Published: April 2011

The US-APWR, currently under Design Certification review by the U.S. Nuclear Regulatory Commission, is a four loop evolutionary pressurized water reactor with a four train active safety system applied by Mitsubishi Heavy Industries. The digital Instrumentation and Control (I&C) System and Human Systems Interface (HSI) system are to be applied to the US-APWR. This design is currently being applied to the latest Japanese PWR plant and to nuclear power plant I&C modernization program in Japan. The US-APWR digital I&C and HSI system (HSIS) utilizes computerized systems, including computer-based procedures and alarm prioritization, relying principally on an HSIS with soft controls, console based visual display units (VDUs) and a large, heads up, overview display panel. Conventional hard-wired controls are limited to system level manual actions and a Diverse Actuation System (DAS). The overall design philosophy of the US-APWR is based on the concept that operator performance will be enhanced through the integration of safety and non-safety display and control systems in a robust digital environment. This philosophy is augmented, for diversity, by the application of independent safety soft displays and controls. In addition, non-digital diverse automatic and manual actuation system is introduced. As with all the advanced designs, the digital systems open as many questions as they answer. To address these new questions, for an eight week period during the months of July and August 2008, an extensive verification and validation (V&V) program was completed with the objective of assessing US operators’ performance in this digital design environment. (Robert E. Hall et al., 2008, “US-APWR Human Systems Interface System V&V Results: Impact on Digital I&C Design”, 17th International Conference on Nuclear Engineering, ICONE17-75176) [1] Over this time period, U.S. operating crews were subjected to exercise in Mitsubishi dynamic simulator. To follow up above mentioned V&V activities, additional test during the months of this spring in 2009 has been carried out to resolve human engineering discrepancies (HEDs) induced from the previous evaluation and the participants’ comments and performance. Subjective and objective data were collected on each crew for each scenario and an extensive convergent measures analysis was performed, resulting in the identification of both specific design as well as generic conclusions. This paper discusses the digital HSIS of the US-APWR design, the V&V program data collection and analysis, and the study results related to the ongoing discussion of the impacts of digital systems on human performance, such as workload, navigation, situation awareness, operator training and licensing.
Investigations on Optimal Design of a U-Tube Steam Generator
Authors: Qin Hui-min , Yan Chang-qi , Wang Meng , He Shi-jing
Published: April 2011

Steam generator is one of the key equipments in the pressurized water reactor, from the performance point of view, it is necessary to apply optimization techniques to the design of the steam generator. In this work, the optimal designs of a U-tube steam generator (UTSG), taking minimization of the total volume and net weight as objective respectively, are carried out by considering thermohydraulic and geometric constraints using a complex-genetic algorithm (CGA). And the sensitivities of some parameters which influence the total volume and net weight of UTSG are also analyzed. Under the condition of constant secondary thermalhydraulic parameters of the steam generator, the optimal design indicates an obvious effect taking either the overall volume or the total weight of the steam generator as the objective. The optimization results show that the proposed optimal method is feasible and effective. And the results of optimal designs and sensitivity analysis would provide guidance in the engineering design of UTSG.
Researches on the Key Technologies for Tomographic Gamma Scanning
Authors: Zhang Quanhu , Hui Weihua , Li Feng , Gu Zhongmao , Qian Shaojun
Published: April 2011

Tomographic Gamma Scanning (TGS) method is one of the most advanced non-destructive assay (NDA) methods. But for measuring heterogeneously distributed media with medium- and high-density, there are still three main problems: experiment’s method of the calibration of detection efficiency of TGS is more difficult and complicated because of large voxels, “point-to-point” model and average model can’t calculate high-density samples accurately in transmission image reconstruction and computational cost is very large for correction factor in emission image. Calibration of detection efficiency using Monte Carlo method shorten calibration cycle greatly, a new Monte Carlo statistical iteration in TGS transmission image reconstruction method which is based on MC calculation and Numerical Analysis is presented, give a chance for measuring high-density samples; the division method and pre-calculation method in reconstructing TGS emission image is used which saves a great lot of computation time and provide a fast reconstruction algorithm for emission image. Above methods apply to TGS experiment device, the relative errors between experiment and MC calibration were less 5%; the relative errors between reconstructed values and the reference values were less than 4% in transmission image; the corrected experimental results were compared to the standard values and the relative deviation was found to be 7%. It took no more than one hour to complete the reconstruction of TGS emission image for a sample model with 3×3×3 voxels using a 2.0G computer.
Alternative Laser Driven Fusion Reactions for Nuclear Energy Without Radioactivity
Authors: Ghoranneviss Mahmoud , Malekynia Babak , Azizi Nader , Hora Henrich , Miley George H.
Published: April 2011

Following the first result of generating nuclear fusion energy without dangerous radioactive radiation by laser ignition of the proton-11Boron reaction (HB11), we applied this method to evaluate other fusion reactions with no primary neutron production as the proton-7Lithium reaction (HLi7) and of the burning of solid density helium isotope 3He (He3-He3). The new method is a combination of now available laser pulses of 10 petawatt (PW) power and duration in the range of picoseconds (ps) or less. The new mechanism follows the initial theory of Chu and of Bobin for side-on ignition of solid state density fusion fuel developed in about 1972 where some later known physics phenomena had to be added. The essential innovation is the use of the discovery of a predicted anomaly when the mentioned laser pulses are sufficiently clean, i.e. free from prepulses by at least a contrast ratio 108 where acceleration by the nonlinear (ponderomotive) force is dominating.
RBMK Fuel Rod Modelling and Investigation of Process During Operation Cycle
Authors: Kaliatka Tadas , Marao Ausˇra , Karalevicˇius Renatas , Usˇpuras Eugenijus
Published: April 2011

This paper presents the analysis determining the status of fuel rods after whole normal operation. The FEMAXI–6 code was selected for such analysis. Evaluating the specifics of RBMK fuel rods, the adaptation of code was provided. After the adaptation of FEMAXI-6 code, the single fuel rod model of RBMK-1500 was developed and the processes, which occur during whole life of fuel rods, were analyzed. For this analysis the fuel rod from fuel channel with average initial power (2.5 MW) was selected. After (normal) operation the fuel rods from the reactor are transferred to the spent fuel pool and the state of the fuel rods (intactness of cladding, residual stresses in the cladding and fuel pellets, gap between cladding and pellets and etc.) is very important, because fuel rod cladding is one of the safety barriers. In this paper the stresses in cladding, plastic deformation of cladding and other parameters were calculated using FEMAXI-6 and method of final elements. The performed analysis demonstrates possibility to identify state of fuel rods after normal operation that is necessary for long-term fuel storage in spent fuel pools.
Thermal Hydraulic Test of Advanced BWR Fuel Bundle With Spectral Shift Rod (SSR): Overview and Pre-Test Analysis by TRACG Code
Authors: Kondo Takao , Chaki Masao , Ohga Yukiharu , Abe Moriyasu
Published: April 2011

Japanese national project of next generation light water reactor (LWR) development started in 2008. As one of its development items, the thermal-hydraulic test of spectral shift rod (SSR) is planned. A new component called SSR, which replaces conventional water rod (WR) of boiling water reactor (BWR) fuel bundle, was invented to enhance the BWR’s merit, spectral shift effect for uranium saving. In SSR, water boils by neutron and gamma-ray direct heating and water level is formed as a boundary of the upper steam region and the lower water region. This SSR water level can be controlled by core flow rate, which amplifies core void fraction change, resulting in the amplified spectral shift effect. In this paper, its test plan overview and pre-test analysis by TRACG code is presented. The test plan was developed with the purpose of evaluating SSR thermal-hydraulic characteristics at the actual BWR operating condition (7MPa), such as the controllability of SSR water level, and obtaining data for the validation of calculation method. In the test plan, several types of SSR simulation which covers SSR design in both next generation BWR and conventional BWR were designed. Also test operating conditions such as thermal-hydraulic parameters are determined. In order to evaluate these test specifications, pre-test analysis by TRACG code was conducted. Analysis results of each parameter’s effect on SSR characteristics are consistent with SSR mechanism, which shows that the actual operating condition for SSR fuel is simulated well.
Natural and Gas Enhanced Circulation Tests in the NACIE Heavy Liquid Metal Loop
Authors: Tarantino Mariano , Bernardi Davide , Coccoluto Giovanni , Gaggini Piero , Labanti Valerio , Forgione Nicola , Napoli Andrea
Published: April 2011

The paper reports on the results carried out from the natural circulation and gas-injection enhanced circulation tests performed on a heavy liquid metal loop, named NACIE, and located by the Brasimone ENEA Research Centre. The work is aimed at providing information on the characterization and interpretation of the basic mechanisms proposed in the design of future reactor relying on these circulation mechanisms. The results discussed in the present work concern the experiments performed using Lead Bismuth Eutectic (LBE) as coolant. Both natural circulation and gas-injection enhanced have been addressed, drawing conclusions about the observed phenomena. Numerical simulations have been performed in collaboration with the University of Pisa, adopting the RELAP5/Mod3.3 system code modified to allow for LBE as a cooling fluid. Post-test calculations have been performed to compare the code response with the experimental results under the natural circulation and gas enhanced circulation flow regime, allowing to qualify the adopted nodalisation as well as the performance of the code when employed on HLM loop. The available data will allow to validate and qualify numerical tools for engineering application, establishing a reference experiment for the benchmark of commercial codes when employed in HLM loop.
Steam-Reheat Options for Pressure-Tube SCWRs
Authors: Saltanov Eugene , Peiman Wargha , Farah Amjad , King Krysten , Naidin Maria , Pioro Igor
Published: April 2011

Concepts of nuclear reactors cooled with water at supercritical pressures were studied as early as the 1950s and 1960s in the USA and Russia. After a 40-year break, the idea of developing nuclear reactors cooled with SuperCritical Water (SCW) became attractive again as the ultimate development path for water cooling. The main objectives of using SCW in nuclear reactors are: 1) to increase the thermal efficiency of modern Nuclear Power Plants (NPPs) from 30–35% to about 45–48%, and 2) to decrease capital and operational costs and hence decrease electrical energy costs. SCW NPPs will have much higher operating parameters compared to modern NPPs (pressure about 25 MPa and outlet temperature up to 625°C), and a simplified flow circuit, in which steam generators, steam dryers, steam separators, etc., can be eliminated. Also, higher SCW temperatures allow direct thermo-chemical production of hydrogen at low cost, due to increased reaction rates. To achieve higher thermal efficiency Nuclear Steam Reheat (NSR) has to be introduced inside a reactor. Currently, all supercritical turbines at thermal power plants have a steam-reheat option. In the 60’s and 70’s, Russia, the USA and some other countries have developed and implemented the nuclear steam reheat at subcritical-pressure experimental boiling reactors. There are some papers, mainly published in the open Russian literature, devoted to this important experience. Pressure-tube or pressure-channel SCW nuclear reactor concepts are being developed in Canada and Russia for some time. It is obvious that implementation of the nuclear steam reheat at subcritical pressures in pressure-tube reactors is easier task than that in pressure-vessel reactors. Some design features related to the NSR are discussed in this paper. The main conclusion is that the development of SCW pressure-tube nuclear reactors with the nuclear steam reheat is feasible and significant benefits can be expected over other thermal-energy systems.
Thermal Aspects of Using Uranium Dicarbide Fuel in an SCWR at Maximum Heat-Flux Conditions
Authors: Pascoe Caleb , Milner Ashley , Patel Hemal , Peiman Wargha , Richards Graham , Grande Lisa , Pioro Igor
Published: April 2011

There are 6 prospective Generation-IV nuclear reactor conceptual designs. SuperCritical Water-cooled nuclear Reactors (SCWRs) are one of these design options. The reactor coolant in SCWRs will be light water operating at 25 MPa and up to 625°C, actually at conditions above the critical point of water (22.1 MPa and 374°C, respectively). Current Nuclear Power Plants (NPPs) around the world operate at sub-critical pressures and temperatures achieving thermal efficiencies within the range of 30–35%. One of the major advantages of SCWRs is increased thermal efficiency up to 45–50% by utilizing the elevated temperatures and pressures. SuperCritical Water (SCW) behaves as a single-phase fluid. This prevents the occurrence of “dryout” phenomena. Additionally, operating at SCW conditions allows for a direct cycle to be utilized, thus simplifying the steam-flow circuit. The components required for steam generation and drying can be eliminated. Also, SCWRs have the ability to support hydrogen co-generation through thermochemical cycles. There are two main types of SCWR concepts being investigated, Pressure-Vessel (PV) and Pressure-Tube (PT) or Pressure-Channel (PCh) reactors. The current study models a single fuel channel from a 1200-MWel generic PT-type reactor with a pressure of 25 MPa, an inlet temperature of 350°C and an outlet temperature of 625°C. Since, SCWRs are presently in the design phase there are many efforts in determining fuel and sheath combinations suited for SCWRs. The design criterion to determine feasible material combinations is restricted by the following constraints: 1) The industry accepted limit for fuel centreline temperature is 1850°C, and 2) sheath-material-temperature design limit is 850°C. The primary candidate fuel is uranium dioxide. However; previous studies have shown that the fuel centreline temperature of an UO2 pellet might exceed the industry accepted limit for the fuel centreline temperature. Therefore, investigation on alternative fuels with higher thermal conductivities is required to respect the fuel centreline temperature limit. Sheath (clad) materials must be able to withstand the aggressive SCW conditions. Ideal sheath properties are a high-corrosion resistance and high-temperature mechanical strength. Uranium dicarbide (UC2) is selected as a choice fuel, because of its high thermal conductivity compared to that of conventional nuclear fuels such as UO2, Mixed OXide (MOX) and Thoria (ThO2). The chosen sheath material is Inconel-600. This Ni-based alloy has high-yield strength and maintains its integrity beyond the design limit of 850°C. This paper utilizes a generic SCWR fuel channel containing a continuous 43-element bundle string. The bulk-fluid, sheath and fuel-centreline temperature profiles together with Heat Transfer Coefficient (HTC) profile were calculated along the heated length of a fuel channel at the maximum Axial Heat Flux Profiles (AHFPs).
Thermal Aspects of Using Thorium Dioxide as Alternative Nuclear Fuel in SuperCritical Water-Cooled Reactors
Authors: Patel Hemal , Milner Ashley , Pascoe Caleb , Peiman Wargha , Richards Graham , Grande Lisa , Pioro Igor
Published: April 2011

SuperCritical Water-Cooled nuclear Reactors (SCWRs) are one of six choices for Generation IV (Gen IV) reactor concepts. These reactors use light water as a coolant and operate at a pressure of 25 MPa, inlet temperatures 280–350°C and an outlet temperature up to 625°C. Operating at these elevated temperatures and pressures are beneficial due to: 1) increased gross thermal efficiency of SCW Nuclear Power Plants (NPPs) (from 30%–35% of the current NPPs to 45%–50%) and 2) decreased capital and operational costs. Use of SCW as a reactor coolant will permit a direct-cycle steam circuit. SCWRs eliminate the need for steam generators, steam separators, and steam dryers. Another advantage of SCWRs is a possibility for hydrogen co-generation through thermochemical cycles. At these extreme operating conditions we must be ensured that all fuel-channel materials, i.e., sheath (clad) and fuel, will operate below accepted temperature limits. The industry accepted limit for the fuel centerline temperature is 1850°C, and the design limit for sheath temperature is 850°C. Material investigations have begun with existing NPP fuel-channel designs. Previous studies with UO2 fuel at SCW conditions have indicated that the fuel centerline temperature may exceed the temperature limit. Zirconium alloys cannot operate at temperature beyond 350–500°C due to high corrosion rates. Therefore, Inconel-600 was chosen as a sheath material since is maintains a high yield strength and corrosion resistance at high temperatures. Uranium dioxide fuel is widely used and world resources are becoming limited. Thoria or thorium dioxide (ThO2) is considered as an alternative nuclear fuel and offers many benefits. Thorium dioxide is compliant to the Non-Proliferation Treaty, abundant in global reserves and has higher thermal conductivity than that of UO2. An objective of this paper is to determine the suitability of ThO2 fuel in an Inconel-600-sheath fuel bundle within an SCWR fuel channel. Bulk-fluid, outer-sheath and fuel centerline temperature profiles along with Heat Transfer Coefficient (HTC) profiles were computed along the heated length of a bundle string at the maximum heat flux.
Measurement of 129I Level Around a Nuclear Power Plant by Accelerator Mass Spectrometer
Authors: Wang Zhiwen , Li Huaibin , He Chaohui , Zhao Yaolin , Chen Ning , Zhang Luyuan , Liu Qi , Luo Maoyi , Liang Wangguo , Fan Yukun
Published: April 2011

129I is a long-lived (15.7M year) radioisotope of iodine. It can be used as a tracer for monitoring nuclear proliferation and the 129I/127I ratio can be used to evaluate the radiation contamination level. Nowadays a great number of nuclear power plants will be built in China, but the data of 129I concentration in environmental samples around nuclear power plants are limited. Accelerator Mass Spectrometer (AMS) whose detection limit is about 10−14, is one of the best instruments for analyzing environmental samples. The methods of making 129I target samples for AMS measurement from different type samples were studied, and the processing system for water and soil samples were established. Six surface seawater samples were collected at different distance away from a nuclear power plant in China. These samples were measured by Xi’an AMS. The ratios of 129I/127I in the seawater samples are between 0.829 × 10−10 and 9.451 × 10−10, and the average value is about 3.518 × 10−10. The ratios of 129I/127I in these samples are compared with other measurement results under different circumstances in other parts of the world. The results show that this nuclear power plant has not released superfluous 129I into environment after several years’ operation. Since the AMS and sample processing system are established, we will do much work on nuclear technology application with 129I tracer.
A Position Sensitive Detector With PMT Readout
Authors: Li Jiaxing
Published: April 2011

A CsI(T1) scintillation detector with a two-dimension position-sensitive photomultiplier readout at RIBLL is developed. The position resolution of the detector was obtained with RIB 17N, and the performance of detector was simulated by using GEANT4. The simulation results accorded with the experimental ones. The reflection factor of the aluminum foil outside the scintillator plays an important role on the position resolution.
Treatment and Disposal of the Radioactive Graphite Waste
Authors: Tian Lifang , Wen Mingfen , Chen Jing
Published: April 2011

A large number of nuclear reactors with graphite as moderator and reflector material are facing to be decommissioned now or later, and the radioactive graphite waste is a large part of the involved wastes. In addition, high temperature gas-cooled reactors being developed rapidly use a large quantity of graphite material (up to 95%) in the nuclear fuel elements, besides graphite material as their moderator and reflector material in the reactor cores. Therefore, it is very critical to manage these graphite wastes from the decommissioned and being decommissioned reactors. The part with low-level radioactive contamination that could not be reused now, may be disposed of as solid waste to reduce its volume, and the possibility of its being retrieved and reused in the future with advanced technology should be considered. The other graphite waste with high-level radioactive contamination requires much more consideration. Due to several factors, such as its large quantity, a lack of available disposal sites and public acceptance, it may not be disposed of directly in the repository any more. An option may be the transformation of the high-level radioactive graphite waste into low-level radioactive waste through physical and chemical processes. The current technologies involve, e.g., thermal treatment to release 36Cl, capture of the 14C from the gases of incineration of carbon material and decomposition of carbon dioxide into solid carbon. After these treatments the carbon material might be decontaminated and separated as low-level radioactive waste and a small amount of residual high-level waste could be disposed of ultimately. In order to achieve a sustainable development of graphite material, the maximum utility and the minimal disposal of radioactive graphite should be considered in the management of radioactive graphite waste. It is urgent to explore new technologies for decontaminating and recycling radioactive graphite.
Status of Future Reactor Technology Development in Korea
Authors: Hahn Dohee , Kim Yeong-Il , Kim Yong Wan
Published: April 2011

In order to provide a consistent direction to long-term R&D activities, the Korea Atomic Energy Commission (KAEC) approved a long-term development plan for future nuclear reactor systems which include sodium cooled fast reactor (SFR) and very high temperature reactor (VHTR) on December 22, 2008. The SFR system is regarded as a promising technology to perform actinide management. The final goal of the long-term SFR development plan is the construction of an advanced SFR demonstration plant by 2028. The nuclear hydrogen project in Korea aims at designing and constructing a nuclear hydrogen demonstration system by 2022 to demonstrate its hydrogen production capability. This paper summarizes the overall long-term project plans for SFR and VHTR technology development and explains results of detailed design studies with supporting R&D activities.
Steady-State PWR Core-Wide Analysis Using Subchannel Analysis Method
Authors: Zhang Xiaoying , Liang Zhitao
Published: April 2011

A subchannel program was designed to make a steady-state core-wide analysis for PWR core. Equations of continuity, energy, axial and transverse momentum, which are the main elements of the program, were presented in the paper. The spatial forward differentiation method was used to get the thermal hydraulic parameters, such as temperature, pressure, enthalpy of different spatial steps along the axial direction in each channel. In order to validate the program, a physical model of the reactor core, which is based on 900MW PWR nuclear plant, was established and computed in the program. In the program, the core was radially divided into 8 subchannels. Variations of the temperature, pressure and enthalpy in each coolant subchannel were computed. The exterior surfaces’ temperatures of the fuel rods in subchannels were computed as well. The computation shows that outlet pressure of the core is about 15.26 MPa; the average outlet temperature of each channel is about 325.34°. These results are within reasonable range based on comparing with published datum [1].
Thermal and Thermo-Mechanical Analysis of Lower Hybrid Wave Antenna in East Tokamak
Authors: Li Jun , Liu Fukun
Published: April 2011

Today, the tokamak has emerged as the leading approach to controlling nuclear fusion for the purpose of electrical power generation. As an important power system in experimental advanced superconducting tokamak (EAST), an advanced steady-state plasma physics experimental device, the lower hybrid current drive (LHCD) system provides a high-energy microwave for plasma heating and current drive. The microwave power is delivered to the plasma through the lower hybrid wave (LHW) antenna. In order to couple microwave energy to the plasma more efficiently, the antenna is exposed to the plasma. During the plasma operation, the LHW antenna will not only withstand the high thermal flux from the plasma but also the thermal stress owing to thermal loads. The temperature of the antenna has been analyzed using the numerical analysis method. At the same time, the thermal stresses and displacements due to thermal loads are also calculated using the finite element code. The paper is organized as follows: First, the structure of the LHW antenna is briefly described; the 3D model is given. Secondly, thermal loads and boundary conditions are shown and discussed. Finally, the temperature, thermal stresses and displacements are simulated. All the results of the simulation are presented and discussed. These could be useful for the development of the lower hybrid wave antenna.
Development of Remotely Operated Inspection Devices for BWR Internals
Authors: Shimamura Mitsuaki , Togasawa Yutaka , Hozumi Hisashi , Mukai Naruhiko , Yuguchi Yasuhiro
Published: April 2011

Two ultrasonic testing (UT) devices to inspect the internals of nuclear reactor have been developed. The one is a jet pump UT device to inspect the inner weld line of the jet pump, and the other is a shroud UT device to inspect the outer weld line of the shroud. The jet pump UT device is mainly composed of an inspection probe scanner and a wheeled platform with a telescopic guide. Since the inspection probe scanner has been designed slim enough to pass through the narrow opening of the jet pump nozzle, it can be remotely positioned inside the jet pump transported by the wheeled platform. The shroud UT device is mainly composed of a flat-type remotely operated vehicle (ROV) and a positioning mast of ROV. The ROV is installed remotely on the outer surface of the shroud using the positioning mast. And the ROV has been designed thin enough to pass through the narrow gap between the jet pump and the shroud, so that it can move horizontally on the surface of the shroud with automatic cable feeding. Consequently, the proposed remote and automatic inspection devices can perform the inspection work in short time without using fuel handling machine (FHM). Therefore, the inspections can be performed simultaneously with the refueling work, which contributes to the shortening of regular inspection periods of nuclear power plants.
Swirling and Secondary Flows in PWR Primary Loops: CFD Might Bring Some Light
Authors: Bellet Serge , Benhamadouche Sofiane
Published: April 2011

Swirling and secondary flows play an important role in several locations of the primary circuit of a PWR. After a brief reminder of the experimental period during the initial designs, few examples such as the lower plenum, the upper Plenum and the hot legs, the fuel assemblies and the U-bend downstream the steam generator are given to illustrate this statement. Although experiments will remain mandatory to understand these complex flows, CFD could bring some light either to perform parametric studies or to analyze the flow structures if not possible in the experiment or in a reactor configuration. In order to have a “clean” approach, one has to first validate the CFD tools on few academic cases dominated by swirling/rotation/secondary effects. A full program for a better understanding of these kinds of flows with CFD computations has been launched at EDF. Examples of the very first calculations are given in the present paper. Although no mature conclusions can be drawn at this stage with the available simulations, one can argue that first moment closures such as k-epsilon or k-omega models can perform well in some configurations while Reynolds Stress Models performs almost always rather well. LES is clearly superior but more expensive. Even the growth of the computing power will not allow us to perform LES computations whatever the configuration, thus focusing on RSM model remains important.
Assessment of the EM Plus Advanced BE Code Platform LOCA Analysis Approach With Application in a Chinese 300MW NPP
Authors: Kuang Bo , Lu Lu , Liang Thomas K. S.
Published: April 2011

In order to develop licensing LOCA analysis program platform for nuclear power plant (NPP), a relatively realistic technique have been adopted in this paper, which is to satisfy the requirements of conservative evaluation models (EM) in related regulation (10CFR50, Appendix K) by modifying related models or correlations of the best estimate (BE) program RELAP5/MOD3, in an effort to form a licensing LOCA analysis tool. By using this “advanced program platform plus conservative EM” technique, LOCA analysis on the Chinese 300MW NPP, of which LOCA used to be conservatively evaluated, is attempted so that further potential of margins should be achieved. By comparing with Appendix K, 10 models or correlations in RELAP5/MOD3 code are identified, which need to be modified, invoked, evaluated or verified. Modifications or control have been separately introduced on such models or correlations as fission product decay model, correlations for critical heat flux (CHF), post-CHF heat transfer correlations, logic of preventing from returning nucleate boiling and transition boiling heat transfer prior to reflood, discharge model, metal-water reaction model, emergency core cooling (ECC) bypass model and related models for pressurized water reactor (PWR) refill and reflood. With the accomplishment of modification or invoking the above-mentioned models, conservatism verifications by utilizing proper standard curve or separate test data have been carried out. The standard curve or separate effect tests data includes: (1) 1971 fission product decay standard curve; (2) related ORNL THTF test data; (3) Marviken test-22 data; (4) Cathcart oxidation test data; and (5) Westinghouse FLECHT - SEASET test data, etc. Further, for the integral effect test LOFT L2–5, calculations with both original RELAP5/MOD3 program and the program of which some related models modified in compliance with the requirement of 10CFR50 Appendix K have been made. The results indicate that, on the one hand, ability of the original BE program simulating the accident progress is identified; and on the other hand, the modified models are confirmed to be conservative. Moreover, both individual contributions and combined effects of the model modification to conservatism of the integral test analysis are observed and analyzed. To assess effects of the “advanced program platform plus conservative EM” approach on LOCA analysis, finally, calculations with the modified code for both hypothetical small break and large break LOCA of the Chinese 300MWe NPP are conducted. Compared with the calculated results of original RELAP5/MOD3 program and that delivered in Final Safety Analysis Report (FSAR) of the NPP, mutual and integrated effects with different model modifications are investigated. And finally, safety margin achieved by the present calculation is preliminarily discussed.
Dynamic Analysis of Core Magnetic Circuit of Direct Action Solenoid Valve
Authors: Liu Qianfeng , Bo Hanliang , Qin Benke
Published: April 2011

Control rod hydraulic drive mechanism (CRHDM) is a newly invented patent with Institute of Nuclear and New Energy Technology Tsinghua University which owns CRHDM’s independent intellectual property rights. The combined valve which is made up of three direct action solenoid valves is the key part of this technology, so the performance of the solenoid valve directly affects the function of the integrated valve and the CRHDM. The core magnetic circuit of 8 types of the direct action solenoid valve is analyzed using the ANSYS software with various voltages, the results of which are validated by experiments. It shows that the magnetic force increases and time shortens with the increasing voltage. What’s more, under the same condition, the electromagnetic force of the solenoid valve with cone armatures increases much faster than the one of flat armatures, however, the electromagnetic force of the latter is bigger than the former when the force is steady. Furthermore, different shaped interfaces of the seal ring structure between armatures lead to different results. Especially, the electromagnetic force of the solenoid valve with the flat armatures which has seal ring increases faster than the one which does not have this structure, but the result is opposite when it is the solenoid valve with the cone armatures. At last, the subsection structure of magnetic material accelerates the magnetic force sharply.
Study on Neutronics and Fuel Cycle of Fusion-Fission Hybrid Reactor
Authors: Xu Hong , Yang Yongwei , Zhou Zhiwei
Published: April 2011

A conceptual design on neutronics of Fusion-Fission Hybrid Reactor (FFHR) blanket is presented in this paper. The code system COUPLE2.0, which couples the codes MCNP and ORIGEN2 is uesd to simulate the depletion of nucleus for FFHR blanket. By comparing different kinds of coolant, fission fuel and tritium breeding material etc, light water was selected as the coolant, the U-Zr alloy as fuel and Li4SiO4 as tritium breeding material. The volumetric ratio of showed that under that the coolant to fuel is approximately 2.0. The tritium breeding ratio (TBR) must be larger than 1.0, the keff and the energy multiplication factor of the blanket are both comparatively high, so that high breeding ratio of fuel can obtained under realistic condition of fusion plasma parameters.
Analysis of Radioactive Source Term for Modular HTGR During Normal Operation
Authors: Cao Jianzhu , Liu Tao , Wu Yuanyu , Li Hong , Liu Yuanzhong
Published: April 2011

The methods of radioactive source term analysis are introduced in detail for the modular high temperature gas cooled reactor in China. Radioactive fission products and activation products produced in the reactor are described. For fission products, the emphasis is on the process from production through release to the environment for noble gas, iodine and long-lived metallic nuclides. For activation products, it mainly introduces the behaviors of H-3 and C-14. Especially the permeation process from primary circuit to secondary circuit is described for H-3. Using the preliminary design parameters of demonstration HTGR in China, basic prediction of radioactive source term is done and the results are given.
Comparison of CCFL Experiments Performed in Two Different Models of the Hot Leg of a PWR With Rectangular Cross-Section
Authors: Valle´e Christophe , Seidel Tobias , Lucas Dirk , Tomiyama Akio , Murase Michio
Published: April 2011

In order to investigate the two-phase flow behaviour during counter-current flow limitation in the hot leg of a pressurised water reactor, two test models were built: one at the Kobe University and the other at the TOPFLOW test facility of Forschungszentrum Dresden-Rossendorf (FZD). Both test facilities are devoted to optical measurement techniques, therefore, a flat hot leg test section design was chosen. Counter-current flow limitation (CCFL) experiments were performed, simulating the reflux condenser cooling mode appearing in some accident scenarios. The fluids used were air and water, both at room temperature. The pressure conditions were varied from atmospheric at Kobe to 3.0 bar absolute at TOPFLOW. According to the presented review of the literature, very few data is available on flooding in channels with rectangular cross-section, and no experiments were performed in the past in such rectangular models of a hot leg. Usually, the macroscopic effects of CCFL are represented in a flooding diagram, where the gas flow rate is plotted versus the discharge water flow rate. Commonly, the non-dimensional superficial velocity (also known as the Wallis parameter) is used to plot the flooding diagram. However, the classical definition of the Wallis parameter contains the pipe diameter as characteristic length, which was originally defined by Wallis (1969) for counter-current flow limitation in vertical pipes and not in near horizontal channels with rectangular cross-section. In order to be able to perform comparisons with pipe experiments and to extrapolate to the power plant scale, the appropriate characteristic length should be determined. Because the experimental projects on this subject at the Kobe University and at FZD were launched independently, a detailed comparison of both test facilities is presented. With respect to the CCFL behaviour, it is shown that the essential parts of the two hot leg test sections are very similar. This geometrical analogy allows to perform meaningful comparisons. However, clear differences in the dimensions of the cross-section (H × W = 150 × 10 mm2 in Kobe, 250 × 50 mm2 at FZD) make it possible to point out the right characteristic length for hot leg models with rectangular cross-sections. The hydraulic diameter, the channel height and the Laplace critical wavelength (leading to the Kutateladze number) were tested. The experimental results obtained in the two test facilities clearly show that the channel height is the suited characteristic length. Finally, the experimental results are compared with similar experiments and empirical correlations for pipes available in the literature. In spite of the scatter of the data and of the different correlations, it was noticed that flooding is reached at slightly lower gas fluxes in the hot leg models with rectangular cross-section compared to pipes.
Radioactive Waste Water Treatment by Inorganic Membrane System
Authors: Li Junfeng , Han Fujuan , Wang Jianlong
Published: April 2011

The removal efficiencies of strontium, cesium and cobalt by inorganic membrane with cut-off 8kD were studied. In order to improve the removal efficiency of non-active nuclides, soluble macropolymers sodium poly(acrylic) acid (NaPAA) of different molecule sizes were selected as chemical assistant reagents to make comparative experimental studies. The molecule of 8000,50000 and 100000 D NaPAA were used. The flux and removal rates of different systems were mainly studied. The results show that for inorganic membrane with pore size of 8kD molecule of 50000 Da NaPAA is the best selection for assistant reagents.
Statistical Uncertainty Analyses of Void Fraction Predictions Using Two Different Sampling Strategies: Latin Hypercube and Simple Random Sampling
Authors: Hernandez-Solis Augusto , Ekberg Christian , Jensen Arvid O¨dega˚rd , Demaziere Christophe , Bredolt Ulf
Published: April 2011

In recent years, a more realistic safety analysis of nuclear reactors has been based on best estimate (BE) computer codes. Because their predictions are unavoidably affected by conceptual, aleatory and experimental sources of uncertainty, an uncertainty analysis is needed if useful conclusions are to be obtained from BE codes. In this paper, statistical uncertainty analyses of cross-sectional averaged void fraction calculations using the POLCA-T system code, and based on the BWR Full-Size Fine-Mesh Bundle Test (BFBT) benchmark are presented by means of two different sampling strategies: Latin Hypercube (LHS) and Simple Random Sampling (SRS). LHS has the property of densely stratifying across the range of each input probability distribution, allowing a much better coverage of the input uncertainties than SRS. The aim here is to compare both uncertainty analyses on the BWR assembly void axial profile prediction in steady-state, and on the transient void fraction prediction at a certain axial level coming from a simulated re-circulation pump trip scenario. It is shown that the replicated void fraction mean (either in steady-state or transient conditions) has less variability when using LHS than SRS for the same number of calculations (i.e. same input space sample size) even if the resulting void fraction axial profiles are non-monotonic. It is also shown that the void fraction uncertainty limits achieved with SRS by running 458 calculations (sample size required to cover 95% of 8 uncertain input parameters with a 95% confidence), result in the same uncertainty limits achieved by LHS with only 100 calculations. These are thus clear indications on the advantages of using LHS.
Applying Markov Processes for KIC Minimum Configuration Calculation
Authors: Tan Ke , Jiang Guojin , Sun Yongbin , Peng Huaqing , Shi Ji
Published: April 2011

CPR1000 plant operation is based on a centralized and computerized control mean (the KIC system), that provides all the means needed by the shift team to operate the plant in any situation. There are two main consequences regarding the definition of the KIC configuration. 1) The KIC composition shall be defined so that they are sufficient for the shift team to finish their tasks. It is linked to “Human Factors” concerns as well as organization concerns. 2) The probability of KIC OWP (Operator Work Place) unavailable shall be taken into account in order to figure out the relationship between the fault leading to KIC OWP unavailable and different KIC degraded situation. The technology and architecture choices fix limits to the flexibility and robustness that the KIC can offer regarding its own degraded situations. Therefore, in order to keep a low frequency of use of the Back-Up Panel, it might be needed to adapt and review the operating organization and needs to stay on the KIC by raising the discomfort of operation for the shift team. Nevertheless, such an approach is limited as reducing the discomfort of operation as obviously a price in term of safety notably. Therefore, the KIC degraded situation shall be studied in order to identify the criteria beyond which the safety operation of the plant cannot be ensured anymore from the KIC. After studying the features of the fault leading to KIC OWP key component unavailable, the data from DCS supplier, the data from reference NPP (Nuclear Power Plant) and the statistic during commissioning, the MTBF (Mean Time Between Failures) for each OWP has been calculated. Applying Markov Processes based on the studying above, CPR1000 operation principles, features of SOP (state oriented procedures to deal with accident situation) and PSA, CNPEC suggest the method to calculate KIC minimum configuration for reaching the best balance to maximize the availability of the KIC, to minimize frequency of switch to the Back-Up Panel (conventional control means) and to increase the plant availability (as the switch to the Back-Up Panel can lead to the fallback of the unit).
Based on 200L Barrels of Low-Level Solid Waste Encapsulates Device Design
Authors: Shang Aiguo , Qin Jin , Lu Changjie
Published: April 2011

This paper presents the design of the low-level solid waste encapsulation device for the 200L barrels. The device is used to compress and package the low-level contaminated waste such as cotton goods, rubber, plastic products, leather products and paper products.
The Progress of ICRP New Recommendations
Authors: Shang Aiguo , Lu Changjie , Qin Jin
Published: April 2011

In order to probe into the usage of the Recommendations of the ICRP, through comparative analysis of low-dose-rate radiation-induced stochastic effects of a nominal risk coefficient, radiation weighting factor, tissue weighting factor as well as the the implementation of changes on the radiological protection system, analysis of the international on Radiological Protection fundamental recommendations of the Committee on the latest changes in radiological protection and development, and that these changes can not affect the existing radiation protection of China’s basic policy and standards.
Quantifying Detection Capabilities of Detrimental Conditions in High-Density Polyethylene Butt-Fusion Joints
Authors: Frederick Caleb J.
Published: April 2011

Today, commercial nuclear power plants are installing High-Density Polyethylene (HDPE) piping in non-safety-related and safety-related systems. HDPE has been chosen in limited quantity to replace carbon steel piping as it does not support rust, rot, or biological growth. However, due to its relatively short nuclear service history, developing a way to accurately evaluate joint integrity in HDPE is crucial to utilities and the U.S. Nuclear Regulatory Commission (USNRC). This paper will investigate using ultrasonic Phased Array to quantify detection of flaws and detrimental conditions in butt-fusion joints throughout the full spectrum of applicable HDPE pipe diameters and wall-thicknesses. Currently the most concerning joint condition is that of “Cold Fusion”. A cold-fused joint is created when molecules along the fusion line do not fully entangle or co-crystallize. Once the fusion process is complete, there is the appearance of a good, quality joint. However, the fabricated joint does not have the required strength as the co-crystallization along the pipe faces has not occurred. Therefore, performing a visual examination of the bead, as required by the current revision of ASME Code Case N-755, does not provide adequate assurance of joint integrity. As a potential solution, volumetric examination is being considered by the USNRC to safeguard against this and other types of detrimental conditions. Factors addressed will include pipe diameter, wall-thickness, fusing temperature, interfacial pressure, dwell (open/close) time, and destructive correlation with ultrasonic data.