bluuebunny/crossref_metadata_2025 huggingface dataset:
The Transport-Burnup Coupling System Between MCMG-II and STEP1.0 for High Fidelity Calculations
Authors: Wu Mingyu , Wang Shixi , Yang Yong , Zhang Qiang , Yang Jiayin
Published: February 2014

The three dimension multi-group neutron transport Monte-Carlo program MCMG-II and the multi-group point burnup program STEP1.0 based on linear chain method are adopted in the transport-burnup coupling system. MCMG-II has fine geometry description capability and a new collision mechanism about the material except the traditional point wise continuous cross section calculation to speed up. STEP1.0 using backtracking algorithm to build linear chains adaptively and has analytical solutions to avoid time step restriction. The coupling information between the burnup and transport program includes neutron spectrum, source strength, reaction rates, initial nuclide density and its variation and the fission products. The interpreted script language Python is used to implement the coupling system. The coupling information can be transferred accurately and smoothly between the programs by its powerful text processing functions. It can also accomplish the step by step transport-burnup calculations automatically according to the total power, radiation time and timing step numbers with its interpreted, interactive characters and perform the graphic processing of the results. The coupling system can do the high fidelity burnup calculations with more comprehensive reference to the factors such as geometry, neutron spectrum and radiation time.
A Safety Evaluation of HTTR Core Components Against 2011 Tohoku Earthquake
Authors: Iigaki Kazuhiko , Ono Masato , Shimazaki Yosuke , Tochio Daisuke , Shimizu Atsushi , Inoi Hiroyuki , Takada Shoji , Sawa Kazuhiro
Published: February 2014

On March 11th, 2011, the 2011 Tohoku Earthquake which is one of the largest earthquakes in japan occurred and the maximum acceleration in observed seismic wave in the HTTR exceeded the design value in a part of input seismic motions. Therefore, a visual inspection, a seismic analysis and a performance confirmation test of facilities were carried out in order to confirm the integrity of facility after the earthquake. The seismic analysis was carried out for the reactor core structures by using the response magnification factor method. As the results of the evaluation, the generated stress in the graphite blocks in the reactor core at the earthquake were well below the allowable values of safety criteria, and thus the structural integrity of the reactor core was confirmed. The integrity of reactor core was also supported by the visual inspections of facilities and the operation without reactor power in cold conditions of HTTR.
The Environmental Radiation Analysis on the Residential Area Around Nuclear Power Plants
Authors: Yoo Dong Han , Kim Hee Reyoung
Published: February 2014

The environmental radiation level and radioactivity were analyzed in the Ulsan which is close to Nuclear Power Plants (NPPs). Ulsan is a Metropolitan city with a population of about 1,100,000 residents, where the density of NPPs is the highest in Korea. Environmental radiation monitoring in Ulsan has been performed since the radiation monitoring facility was opened in August 2012 and the radioactivity was measured and analyzed in the environmental samples for five months in Ulsan. The measurement of environmental radiation and radioactivity was carried out periodically for various samples including airborne dust, fallout, tap water, and precipitation. Also, sampled were the soil, rice, cabbage, and indicator plants such as Artemisia and pine needles. The basic data from the radioactivity analyses brought a fundamental data for assessing the internal exposure dose for the public due to the intake of the radio nuclides included in those samples. On the other hand, the gamma radiation in the environment was measured continuously by using an environmental radiation monitor of the ion chamber type. Actually, the measurement of gamma radiation is thought to provide the base-line data on environmental radiation/radioactivity for radiological emergencies and the real-time information compared with the background radiation. The measurement analyses showed that the radiation level of the Ulsan area is kept at a normal background level in spite of the operation of many nuclear power plants near it.
The Analysis of Sedimentation Rate of Sodium Aerosol in a Hypothetical Accident of CEFR
Authors: Sun Dajie , Zhang Donghui , Ren Lixia
Published: February 2014

The present work was undertaken to evaluate the sedimentation rate of the sodium aerosol in a hypothetical sodium fire accident of China Experimental Fast Reactor (CEFR). In the sodium fire accident, sodium aerosol which was hazardous for facility and people as a result of its chemical property and radioactivity, must be carefully evaluated. In this publication, the chemical and physical characteristics of sodium aerosol particles as well as factors that influence the sedimentation rate of sodium aerosol during its transport is attempted to be analysed. Besides, a simplified model is given to calculate the sedimentation rate when aerosol getting through the sodium fire ventilation system of CEFR. It has been found that the sedimentation rate is largely determined by gravity and turbulent diffusion; other factors, such as Brownian movement and thermophoretic force, could be ignored, since they almost have little effect. Using the model, we find that the loss through the ventilation system is about 5%.
The Analysis and Design Technology Research of DAS
Authors: Ma Quan , Luo Qi , Liu Yanyang , Song Xiaoming , Wu Zhiqiang
Published: February 2014

Now the entire safety I&C system is based on one kind of software and hardware platform, the Common Cause Failure (CCF) may impact the whole safety I&C system becomes to a potential risk. How to mitigate the effect of CCF in safety system and improve the safety of the nuclear power plant is considered by the system designer. Especially after the Fukushima nuclear accident, the Defense-in-Depth and Diversity (D3) should be more concerned by all designers. The diverse actuation system (DAS) plays a very important role in the D3 system. In this paper, the related codes and standards of DAS are analyzed firstly. Then, this paper expounds the approach to demonstrate the D3 analysis for the digital I&C systems applied to the nuclear power plant in detail. In the D3 analysis, all the safety functions of the digital safety system are assumed to be disabled by a CCF. DAS provides diverse automatic reactor trip and diverse safety injection actuation which are not impaired by the postulated CCF. DAS also provides manual actuation functions and plant parameter monitoring functions which can be used to cope with CCF. Finally, the paper takes the DAS of Fujian Fuqing Nuclear Power Plant as an example, introduce how to design the structure of the DAS and calculate the suitable setpoints.
UA-AMG Methods for 2-D 1-T Radiation Diffusion Equations and Their CPU-GPU Implementations
Authors: Yue Xiaoqiang , Shu Shi , Feng Chunsheng
Published: February 2014

In this paper, we study several unsmoothed aggregation based algebraic multigrid (UA-AMG) methods with regard to different characteristics of CPUs and graphics processing units (GPUs). We propose some UA-AMG methods with lower computational complexity for CPU and CPU-GPU, and study these UA-AMG methods mixing with 4 kinds of red-black colored Gauss-Seidel smoothers for CPU-GPU since the initial mesh is structured. These UA-AMG methods are used as preconditioners for the conjugate gradient (CG) solver to solve a class of two-dimensional single-temperature radiation diffusion equations discretized by preserving symmetry finite volume element scheme. Numerical results demonstrate that, UA-NA-CG-s, which wins the best robustness and efficiency among them, is much more efficient than the default AMG preconditioned CG solvers in HYPRE, AGMG and Cusp for CPU; Under CPU-GPU, UA-W-CG-p is the most robust and efficient one, and rather more efficient than the smoothed aggregation based AMG preconditioned CG solver in Cusp.
Performance Tests After High Pressure Turbine Retrofit for Maanshan Nuclear Power Plant Unit 1
Authors: Chan Yea-Kuang , Tsai Yu-Ching , Chang Chin-Jang , Lu Chun-Chang , Hsieh Ping-Ling
Published: February 2014

The purpose of the performance test for Unit 1 of Maanshan nuclear power plant is to determine the electrical output and heat rate after the retrofit of the high pressure turbine in the latest refueling outage in 2012. The performance test was conducted in order to verify that the actual improvement in electrical output resulting from the replacement of the high pressure turbine meets the vendor’s guarantee. A total of two performance test runs was conducted in accordance with the ASME performance test code (PTC) 6. The measured electrical powers for the two test runs were 977.4 and 975.0 MWe, respectively, and the average value was 976.2 MWe. After correcting the electrical power to the rated conditions specified in the performance test procedure, the gross electric output was 983.2 MWe. The corrected heat rate for the two performance tests were 10365 and 10353 kJ/kW, respectively. The deviation between the two corrected heat rates was 0.11% and thereby satisfying the test code of 0.25% for the permitted test deviation. Moreover, the performance test results also demonstrated that the improvement in gross electrical output was 17.6 MWe comparing with the pre-retrofit performance test, which exceeded the 10.0 MWe basic performance guarantee by 7.6 MWe.
Simultaneous Fault Diagnosis of the Reactor Coolant System Based on the DSM Evidence Theory
Authors: Ren Jinshen , Jiang Botao , Zhao Fuyu
Published: February 2014

Fault diagnosis of nuclear power plant (NPP) has become a major research topic in ensuring the reliability and safety for the NPP operation. In the light of complexity and information uncertainty of nuclear reactor coolant system, some diagnosis techniques based on the multi-sensor information fusion theories have been widely used. As one part of the those theories, the Dempster-Shafer (D-S) theory excelled in demonstrating and combining uncertain information has witnessed a wide range of applications in the fault diagnosis fields. It can be inferred from the previous studies that the fault diagnosis method based on the D-S theory is efficient; However, this method is mainly intended for single fault diagnosis. In fact, in practical application, simultaneous faults often occur. Based on the DSm theory (Dezert-Smarandache theory of plausible and paradoxical reasoning), this paper proposes a simultaneous fault diagnosis method, which is intended to deal with the simultaneous faults in the reactor coolant system and adequately cope with the sensor information uncertainty. This method can be divided into four steps. Firstly, with the features of the simultaneous fault considered, a diagnosis model of the simultaneous fault is designed, which is based on the DSm theory. Secondly, for the instant information of each senor, the fuzzy membership degree is used to set the corresponding basic probability assignment (BPA) function for each sensor. Thirdly, the above-mentioned BPA functions are combined by utilizing the classic DSm rules. Finally, a decision on the basis of maximum support rule and absolute support rule is made to determine the real fault modes. After that, the experiments on the test-bed are conducted to prove the efficiency of this method.
Study of Station Blackout Accident Process in Taishan Nuclear Power Plant
Authors: Li Yu , Zhang Huiyong , Liao Yehong , Lin Jiming , Zhan Dekui
Published: February 2014

According to the design features of the nuclear power plant in Taishan, a station blackout (SBO) at full power is a complex sequence, induced by the loss of offsite power (LOOP) combined with the loss of the emergency diesel generators (EDGs). This paper shall be performed the deterministic safety analysis of SBO without PSA study, and it mainly analyzes an overheating event of primary coolant resulting in the temporary unavailability of the steam generators (SGs) feeding systems during the station blackout. The analysis of this accident is carried out using the CATHARE2 thermal hydraulic code. The analysis pointed out that the final state can be reached after the operator performs the main manual actions such as startup of the SBO diesel generators from the main control room; startup of the two ASG pumps of trains 1 and 4; Local opening of the ASG header downstream the ASG pumps to enable the two power supplied ASG pumps to feed all the four SGs; initiation of a manual cooldown via the VDA in order to ensure the long term protection of the RCP pumps seals with respect to the thermal and mechanical loads. It corresponds to the achievement of a stable heat removal conditions by the emergency feedwater system (ASG) and the atmospheric steam dump system (VDA).
Investigation of Conditions for Zinc Particle Release and Transport in the Reactor Core
Authors: Renger Stefan , Alt Sören , Kästner Wolfgang , Seeliger André , Fiss Daniel
Published: February 2014

Background of the experimental and methodical work is a loss of coolant accident (LOCA) in a Pressurized Water Reactor (PWR). During LOCA, both the leakage water jet as well as the water in the containment can lead to corrosion of galvanized installations (e.g. grates). That means a change of consistency and chemical properties of the coolant water during the sump recirculation operation, in particular an increase of the zinc concentration (Zn2+ ions in dissolution) and the pH-value. Furthermore, it leads to separation of layer-forming zinc compounds (e.g. zinc borate) at hotspots (fouling at fuel assemblies) and/or the crystallization of zinc corrosion products out of the coolant in hot areas [HOF10]. This in turn causes a decrease of pH-value and, consequently, an increase of the corrosiveness of the coolant flow. This can come along with a chemical transformation of zinc borate to nearly indissoluble zinc oxide by thermo hydrolysis. Spalling parts of the coating on the surface of the fuel rods could lead to the release of particles into coolant water flow. In combination with the release of fibrous insulation material, which can be transported into the reactor containment, containment building sump and connected systems, this process causes pressure build-ups at the strainers and the spacers of the fuel assemblies and a reduction of the pump performance. In this case, efficient heat dissipation cannot be guaranteed.

In a joint research project between the Institute of Process Technology, Process Automation and Measuring Technology (IPM) of University Zittau/Goerlitz and the Helmholtz-Zentrum Dresden-Rossendorf (HZDR), the influence of different boundary conditions like the flow structures in the reactor pressure vessel, temperature distributions and the temperature gradient in the central core regions will be investigated. Therefore, in a final step a test facility containing a 3×3 heated fuel rod dummy configuration was designed and constructed. The design and development of the test facility was accompanied by pre-test calculations and CFD simulations to get information about local surface and fluid temperatures under different flow conditions.
Stress Analysis and Assess for Nuclear Pipes of Class 1 Based on PIPESTRESS
Authors: Liu Rui , Li Tie-ping , Wang Ming-yu
Published: February 2014

In recent years, nuclear pipes play a very important role in the nuclear power equipment. Nuclear pipes especially class 1 usually contain cooling coolant of high temperature and high pressure in the operating nuclear power plant. The integrality of the nuclear pipes is very important to the nuclear reactor. This paper discusses the mechanical analysis and fatigue analysis of Class 1 pipes using finite element software. The soft ware PIPESTRESS is used to analysis the Class 1 pipe system to introduce the procedure for mechanical analysis and safety evaluation. The objective is to satisfy the criteria RCC-M. The results show that the pipe can guarantee the integrity of the pipe under the conditions. The method introduced in this paper can be used as a practical method for mechanical analysis of a nuclear Class 1 pipes.
Using a Hybrid FPGA-CPU Platform for a User-Friendly Environment in Control Systems Design: A Possible Application to Nuclear Plants
Authors: Cappelli Mauro , Di Gennaro Stefano , Memmi Fabrizio , Sepielli Massimo
Published: February 2014

In this work, in-depth considerations about FPGA-CPU architectures design methods are presented together with some possible plant applications. FPGAs are used because of their potentials in guaranteeing parallelism, synchronism and high reliability tasks; on the other hand CPUs are crucial to higher precision computations and to reproduce data and results on a graphical interface, thus resulting in an improvement in the human-machine-interface (HMI). Design choices are here reported to justify the benefits from the use of both CPU-FPGA approaches. Taking advantage of these features, the LabVIEW environment and its digital platform is expanded in order to develop a novel tool able to support designers during the fundamental phases of a supervision and control tool realization: from the model development to the digital hardware implementation and testing. LabVIEW is able to communicate with several languages like Matlab or Scilab for a theoretical model study and simulation, and VHDL for a digital implementation, taking the advantages from both development environments.

Main results have been reported concerning studies of control systems in nuclear plant facilities, showing how a full-digital platform can be very useful in developing monitoring and control instrumentation. This new tool is thought to improve the so-called hardware in the loop (HIL) simulations and to accelerate the process of prototypes implementation and testing, using a more accurate and reliable environment in terms of performance and safety.
The Analysis of Influence of Fuel Pellet and Coolant Temperature Distributions on the VVER-1000 Active Core Characteristics
Authors: Mishin A. A. , Galchenko V. V.
Published: February 2014

The accuracy and quality of neutron-physical calculations of the active core characteristics depend heavily on the few-group constant preparation procedure. The method, based on using average in the fuel assembly fuel and coolant parameters is currently used for preparing macroscopic cross-sections. The question is what impact would considering the uneven distribution of those parameters, made on the few-group constant preparation stage exert on further analysis of the reactor facility behavior during steady-state and transients operation. The study carries out comparative analysis of the neutron-physical characteristics of the VVER-1000 core using the standard approach and using distributed in the fuel assembly fuel and coolant parameters while preparing few-group constants. It’s revealed that the fuel pellet and coolant radial temperature distributions affect the multiplication factor and temperature reactivity effect values.
Heat Transfer to Water at Supercritical Parameters in Vertical Tubes, Annular Channels, 3- and 7-Rod Bundles
Authors: Razumovskiy V. G. , Mayevskiy E. M. , Koloskov A. E. , Pis’mennyi E. N. , Pioro I. L.
Published: February 2014

The data on deteriorated transfer to supercritical water in vertical tubes and channels simulating coolant flow in fuel assemblies obtained at the same experimental setup during more than dozen of years are considered and compared with some known results of the experimental studies performed by other authors. They involve the data for vast ranges of geometry, mass velocity, heat flux rate, and pressure, in some cases for up- and downward flow, for flow with and without thermoacoustic oscillations. For the first time the data illustrating deterioration of heat transfer in the bundles of fuel elements are presented. An attempt to explain the phenomena of “inlet” peak of wall temperature is made. It is shown that temperature regimes of the tubes cooled with supercritical water in a gaseous state (i.e., at bulk temperature above the pseudocritical temperature) are close to linear, stable and easily reproducible within a wide range of mass and heat fluxes.

Some requirements to the experimental setup, coolant quality, test sections etc. that should be followed in studying thermal and hydraulic parameters of supercritical coolant are analyzed.
ITER CC Conductor Pressure Drop Test and Result Analysis
Authors: Guo Liang , Liu Huajun , Han Peng , Cai Denggang , Xu Chengcheng
Published: February 2014

This paper introduces the pressure drop test background of ITER magnet and explains the structure and the compacted process of the correction coil (CC). According to ITER CC technology requirements, the pressure drop test was carried out with nitrogen at room temperature. Used the experimental formula to calculate in theory, and then the measured test data were obtained to compare with the theoretical values to prove that the experimental formula is well applied. CC conductor parameters were taken to calculate in low temperature.
Experimental Investigation of Steam Bubble Condensation in Flowing Subcooled Water With Two Different Injection Nozzle Geometries
Authors: Al Issa Suleiman , Weisensee Patricia B.
Published: February 2014

A multiphase flow test facility was built at the Department of Nuclear Engineering at the Technical University Munich. The main goal of this facility is to investigate the condensation of steam bubbles injected into a vertical large diameter pipe (104 mm) with flowing subcooled water (6–15 K) at low pressure conditions (1.1–1.45 bar). Current experimental investigations will contribute to a better understanding of subcooled boiling at low pressures, accidental conditions in nuclear power plants and low-pressure research reactors and correlations for the validation of CFD codes. The test section is a 1 m long transparent pipe that is surrounded by an 18×18 cm rectangular “aquarium” filled with distilled water for refraction correction. High-speed camera (HSC) recording was used to gather data about condensing bubbles including bubble diameter, shape and rising velocity. Steam was injected via two different vertical injection nozzles with an inner diameter of 4 and 6 mm, respectively, directly into the center of the test section. The present experiments were carried out at three different steam superficial velocities, water superficial velocities and water temperatures leading to bubble diameters up to 50 mm and bubble relative velocities around 1 m/s. The measurements enabled the calculation of bubble Reynolds and Nusselt numbers and comparison with correlations given in literature. Even though significant differences could be observed between the two injection nozzles with respect to the bubble’s diameter and velocity, the Nusselt and Reynolds numbers are in the same range of values. The bigger bubbles of the 6 mm with respect to the 4 mm nozzle are almost neutralized by the lower rising velocities.
Status of the Combined Third and Fourth NGNP Fuel Irradiations in the Advanced Test Reactor
Authors: Grover S. Blaine , Petti David A. , Davenport Michael E.
Published: February 2014

The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program is irradiating up to seven low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The experiments will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of several independent capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation.

The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2) started irradiation in June 2010 and is currently scheduled to be completed in September 2013. The third and fourth experiments have been combined into a single experiment designated (AGR-3/4), which started its irradiation in December 2011 and is currently scheduled to be completed in April 2014. Since the purpose of this combined experiment is to provide data on fission product migration and retention in the NGNP reactor, the design of this experiment is significantly different from the first two experiments, though the control and monitoring systems are extremely similar. The design of the experiment will be discussed followed by its progress and status to date.
Method for Joint Configuration of Nuclear Power Plant Fuel Elements/Assemblies
Authors: Ma Zirong , Nie Lihong , Mao Yulong
Published: February 2014

The method is about a design for nuclear fuel elements/assemblies configuration conducted jointly by several units with interchangeable fuel elements/assemblies. The first core of the new units may totally use the fuel elements/assemblies with relatively high enrichments which are identical or equivalent to the equilibrium cycle enrichments. Wherein, a part of new fuel elements/assemblies with relatively high enrichments may be put into one or more operating units for combustion, such that the new fuel elements/assemblies loaded in the operating units is much more than when such method is not implemented, and then considerable more once burned fuel elements/assemblies are obtained. Meanwhile, considerable twice burned fuel elements/assemblies may be provided by the operating units additionally (taking the reactor core with 157 fuel assemblies refueling every 18 months for example). From the considerable obtained once and twice burned fuel elements/assemblies, the burned fuel elements/assemblies with the desired quantity and burnup may be selected to construct the first reactor cores of the new units together with new fuel elements/assemblies with relatively high enrichments. After the method for joint configuration design of nuclear fuel elements/assemblies has been implemented, the average discharge burnup of the operating units added with more new fuel elements/assemblies and the first reactor cores of the new units is higher than that of the equilibrium cycle.
Nuclear Steam Turbine With 60 inch Last Stage Blade
Authors: Haraguchi Motonari , Nakamura Tateki , Yoda Hideo , Kudo Takeshi , Senoo Shigeki
Published: February 2014

Nuclear steam turbines can be classified into two categories, one for BWR reactors where some countermeasures are taken for radiated steam and water, the other is for PWR reactors and PHWR (CANDU) reactors where steam and water are not radiated. As for Low Pressure section, there is some difference in LP rotor end structure, and LP last three stage blade components can be applied to all reactor types. The trend in nuclear power equipment is in a direction of larger capacity. In response to this trend, longer last stage blade is required if same number of casing is kept to make nuclear turbines reasonably compact. This paper addresses some of the key developments and new technologies to be employed focusing on longer Last Stage Blade (LSB) development with Continuous Cover Blades (CCB), and other enhancements in product reliability and performance.
Analyzing and Comparing the Floor Response Spectra of the NPP Based on Different Soil Dynamic Numerical Models
Authors: Zhu Xiu-yun , Pan Rong
Published: February 2014

The traditional soil dynamic impedance models, recommended by the main international seismic design codes of the nuclear power plant (NPP), are only expressed by a single parallel connection system of spring and dashpot which can not reflect the dynamic stiffness varying with excitation frequencies, and also can not simulate the cases of non-homogeneous site conditions. With the recent development of soil-structure interaction (SSI) analysis, based on the damping-solvent extraction method (DSEM) and the lumped parameter models recommended by seismic design codes of ASCE4-98,RCCG which are all applicable to the homogeneous site and also massless foundation model and viscous-spring artificial boundary model of especially fit for the numerical simulation of non-homogeneous site, comparative study of both the direct method and sub-structure method is carried out in this paper. Finally, by taking the analysis of floor response spectra (FRS) for a certain CPR1000 reactor building as an example, comparative analyses of homogeneous and layered site conditions using various soil dynamic numerical models above-mentioned are performed. In addition, in order to validate the accuracy, the calculated results are compared to that of SASSI program. The results show that FRS in the horizontal direction are good agreement regardless for the homogeneous and layered site conditions, the shapes of FRS in the vertical direction change obviously in the homogeneous site condition. This paper provides some guidance and reference in the aspect of evaluation the seismic suitability for the site of nuclear power plant (NPP).
Instrumentation and Control Enhancements in EC6 Reactor Design
Authors: Archer Paul , Tikku Sunil , Dinh Ngoc-Boi
Published: February 2014

The Enhanced CANDU 6® (EC6®) is a nuclear reactor based on the proven CANDU 6 reactor that uses natural uranium and heavy water as a coolant and as a moderator. The EC6 has many enhancements to improve production and to meet next generation reactor standards. A modern distributed control system is used coupled with a new plant display system to facilitate the operation of the EC6. Computerization of safety systems is extended to include neutronic trip parameters and the Emergency Core Cooling (ECC) system. Computerized systems allow for thorough verification during their development, minimizing the installation, commissioning and maintenance times. Testing of these safety systems is also computer assisted to reduce the time required for testing and reduce the possibility of spurious trips. Enhanced safety system monitoring and equipment monitoring systems are used for early detection of faults or deterioration of equipment. This can in turn allow more targeted maintenance activities and improved safety. Critical safety parameter monitoring is also included in the design. Fission chambers are used instead of ion chambers for measuring neutron power. Fission chambers can extend the bottom of the power measurement range by three decades, allowing outages and restart of the reactor to be performed without the installation of startup instrumentation. Instrumentation is included to cater for Beyond Design Basis Accidents (BDBA). Additional battery banks are also included in the design to cater for Station Black-Out (SBO) and extended SBO. Connection provisions for mobile power generators have also been included in the design. The safety systems are controlled by modern equipment qualified to international standards for safety systems. The instrumentation required for these systems is also assessed for the severe conditions to which they might be exposed. Lessons learned from Fukushima on protection from external events are included in the design.
Enhancement of Plant Availability by Outage Shortening in European Advanced Boiling Water Reactor (EU-ABWR)
Authors: Kamei Kazuhiro , Kataoka Kazuyoshi , Imasaki Kazuto , Saito Noboru
Published: February 2014

European Advanced Boiling Water Reactor (EU-ABWR) is developed by Toshiba. EU-ABWR accommodates an armored reactor building against Airplane Crash, severe accident mitigation systems, the N+2 principle in safety systems, the diversity principle and a large output of 1600 MWe. These features enable EU-ABWR’s design objectives and principles to be consistent with the requirements in the Finnish utility and the safety requirements of Finnish YVL guide.

By adopting Scandinavian outage processes, the Plant Availability is aimed to be greater than 95%. ABWRs have an excellent design potential to acheive short outage duration (e.g., shortening of maintenance and inspection duration by applying Fine Motion Control Rod Drive and Reactor Internal Pump). In addition, the EU-ABWR applies following key design improvements to reduce a refueling outage duration; a) Direct Reactor Pressure Vessel (RPV) Head Spray System, b) Self-standing Control Rods and c) Water shielding reactor pool. In this paper, coolability of RPV due to application of the Direct RPV Head Spray System is also verified with numerical evaluations by Computation Fluid Dynamics (CFD) analysis.
Study on Modifications of CPR1000 Nuclear Plants for Design Extension Conditions
Authors: Tang Hui , Gong Ai-cheng , Zhu Feng
Published: February 2014

Traditional design of CPR1000 nuclear plants has a lot of studies on engineering safety facilities which are important for Design Basis Conditions (DBC). Based on more recent studies on safety, such as Defense in Depth (DID) concept, Probabilistic Safety Analysis (PSA), and especially after lessons from Fukushima Nuclear Accident, more and more attentions are paid to Design Extension Conditions (DEC). This paper has studied the abilities and requirements of CPR1000 Nuclear Plants for DEC accidents mitigation. Some important modifications on mechanical systems are introduced.

DEC accidents are separated into DEC-A, and DEC-B. DEC-A accidents are caused by multiple failures. A main reason for a multiple failure is Common Cause Failure (CCF), such as CCF on Emergency Feed Water System (EFWS) pumps, Emergency Core Cooling System (ECCS) pumps, Emergency Diesel Generators or Cooling chain. Diverse design should be considered to deal with CCF. Several modifications are applied, such as enhancing EFWS turbine pumps, fast-backup between ECCS and Containment Spray System (CSS), mobile water makeup and mobile injection for primary and secondary side, etc. For DEC-B, which means severe accidents, several dedicated systems are added to CPR1000 design, such as hydrogen monitor and control, containment venting filter and In-Vessel Retention systems. Thanks to all the modifications, CPR1000 can sufficiently improve its defense for DEC accidents, which will help CPR1000 to reach a new safety level.
Improvement of SAMPSON Code for BWR Severe Accident Analysis
Authors: Liu Maolong , Ishiwatari Yuki , Okamoto Koji
Published: February 2014

The SAMPSON code has been developed in the IMPACT project in Japan to investigate severe accident phenomena for light water reactors. It integrates various analysis modules into a single code. The authors improved the fuel rod heat-up module of SAMPSON code by modeling the oxidation reaction of various core structures, including Zircaloy, stainless steel and B4C. And the creep failures of the Zircaloy fuel cladding and stainless steel monitoring guide tubes of the source range monitor (SRM) in the reactor core was also modeled for severe accident analysis.
Nuclear Data Statistical Treatment
Authors: Alrwashdeh Mohammad , Yu Jian Kai , Abdalla Aniseh , Wang Kan
Published: February 2014

Several methods have been developed in order to evaluate the best fit for nuclear data parameters, these methods relies on sequence logical steps should be followed to get accurate and reliable results, it subdivided into: (1) the physical model have been used (2) data types (3) statistical methods (4) problems. This paper will discuss the statistical methods used to evaluate the best fit for the nuclear data.

The difficulty in finding a real and reasonable solution to the fitting of data can be made easier by choosing the right fitting method. Different methods will converge differently depending on several parameters like the correct choose of the fitting function and number of fitting parameters.

Here we will discuss the uses of two methods, first is Differential evaluation for nonlinear data to optimize a problem by iterative to improve the solution with regard to the quality.

The second one is nonlinear regression method, this method used a model function is not linear in the parameters, and to estimate the relationships among variables, by capturing the trend in the data by assigning a single function in order to minimize the sum of residuals (Or Chi square) with respect to a set of parameters a = {A1, A2,…..An}.
Important Severe Accident Research Issues After Accident at Fukushima Daiichi Nuclear Power Station
Authors: Sugimoto Jun
Published: February 2014

After the accident at Fukushima Daiichi Nuclear Power Station several investigation committees issued reports with lessons learned from the accident in Japan. Among those lessons, some recommendations have been made on severe accident research. Similar to the EURSAFE efforts under EU Program, review of specific severe accident research items was started before Fukushima accident in working group of Atomic Energy Society of Japan (AESJ) in terms of significance of consequences, uncertainties of phenomena and maturity of assessment methodology. Re-investigation has been started after the Fukushima accident in this working group. Additional effects of Fukushima accident, such as core degradation behaviors, sea water injection, containment failure/leakage and re-criticality have been covered. The review results are categorized in ten major fields; core degradation behavior, core melt coolability/retention in containment vessel, function of containment vessel, source term, hydrogen behavior, fuel-coolant interaction, molten core concrete interaction, direct containment heating, recriticality and instrumentation in severe accident conditions. In January 2012, Research Expert Committee on Evaluation of Severe Accident was established in AESJ in order to investigate severe accident related issues for future LWR development and to propose action plans for future severe accident research, in collaboration with this working group. Based on these activities and also author’s personal view, the present paper describes the perspective of important severe accident research issues after Fukushima accident. Specifically those are investigation of damaged core and components, advanced severe accident analysis capabilities and associated experimental investigations, development of reliable passive cooling system for core/containment, analysis of hydrogen behavior and investigation of hydrogen measures, enhancement of removal function of radioactive materials of containment venting, advanced instrumentation for the diagnosis of severe accident and assessment of advanced containment design which excludes long-term evacuation in any severe accident situations.
RELAP5 Code Study of ROSA/LSTF Experiment on a PWR Station Blackout (TMLB’) Transient
Authors: Takeda Takeshi , Nakamura Hideo
Published: February 2014

RELAP5 code post-test analysis was performed on one of abnormal transient tests conducted with the ROSA/LSTF simulating a PWR station blackout (SBO) transient with the TMLB’ scenario in 1995. The RELAP5 code predicted well the overall trend of the major phenomena observed in the LSTF test, and indicated remaining problems in the predictions of reverse flow U-tubes in steam generator (SG) during long-term single-phase liquid natural circulation. Sensitivity analyses were done further to clarify effectiveness of depressurization of and coolant injection into SG secondary-side as accident management measures to maintain core cooling, based on the LSTF post-test analysis. SG secondary-side depressurization was initiated by fully opening the safety valve in one of two SGs with the incipience of core uncovery. Coolant injection was done into the secondary-side of the same SG at low pressures considering availability of fire engines. The SG depressurization with the coolant injection was found to well contribute to maintain core cooling by the actuation of accumulator system during a PWR SBO (TMLB’) transient.
Fast Reactor Design Using the Advanced Reactor Modeling Interface
Authors: Cheatham Jesse , Truong Bao , Touran Nicholas , Latta Ryan , Reed Mark , Petroski Robert
Published: February 2014

The Advanced Reactor Modeling Interface (ARMI) code system has been developed at TerraPower to enable rapid and robust core design. ARMI is a modular modeling framework that loosely couples nuclear reactor simulations to provide high-fidelity system analysis in a highly automated fashion. Using a unified description of the reactor as input, a wide variety of independent modules run sequentially within ARMI. Some directly calculate results, while others write inputs for external simulation tools, execute them, and then process the results and update the state of the ARMI model. By using a standardized framework, a single design change, such as the modification of the fuel pin diameter, is seamlessly translated to every module involved in the full analysis; bypassing error-prone multi-analyst, multi-code approaches. Incorporating global flux and depletion solvers, subchannel thermal-hydraulics codes, pin-level power and flux reconstruction methods, detailed fuel cycle and history tracking systems, finite element-based fuel performance coupling, reactivity coefficient generation, SASSYS-1/SAS4A transient modeling, control rod worth routines, and multi-objective optimization engines, ARMI allows “one click” steady-state and transient assessments throughout the reactor lifetime by a single user. This capability allows a user to work on the full-system design iterations required for reactor performance optimizations that has traditionally required the close attention of a multi-disciplinary team. Through the ARMI framework, a single user can quickly explore a design concept and then consult the multi-disciplinary team for model validation and design improvements. This system is in full production use for reactor design at TerraPower, and some of its capabilities are demonstrated in this paper by looking at how design perturbations in fast reactor core assemblies affect steady-state performance at equilibrium as well as transient performance. Additionally, the pin-power profile is examined in the high flux gradient portion of the core to show the impact of the perturbations on pin peaking factors.
Phenomena Identification Ranking Table (PIRT) for the MAAP Enhancement Project
Authors: Sakai Norio , Horie Hideki , Yanagisawa Hiromasa , Fujii Tadashi , Mizokami Shinya , Okamoto Koji
Published: February 2014

A phenomena identification ranking table (PIRT) was constructed to identify important phenomena that should be considered in simulating the accident progression for the Fukushima Dai-ichi nuclear power station and predicting the molten core and debris distribution in the reactor system with the MAAP code. The accident scenario, starting from reactor scram to PCV failure, was divided into four time phases in which fuel rod degradation, molten core relocation, and reactor vessel failure were selected as the phase-changing events. Phenomena that would have occurred in the accident were identified in the plant system including the reactor vessel and the containment that consist of 16 sub-components, consistent with the MAAP nodalization for the BWR system. The importance of the identified phenomena was evaluated for each time phase in cooperation with the experts in the Atomic Energy Society of Japan. Eventually total of 1047 phenomena were identified, of which 386 were ranked as ‘highly’ important for the analysis code to evaluate the behavior of molten fuel rods and reactor internal materials. These important phenomena are compared with those considered in the current MAAP analysis models to make sure what should be enhanced to improve the analysis capability for the accident progression.
Preliminary Study of Graphite Dust Resuspension in Steam Generator of HTR-10
Authors: Zhang Tianqi , Peng Wei , Yang Qijia , Yu Suyuan
Published: February 2014

The resuspension rate of graphite dust in steam generator of high temperature gas-cooled reactors (HTRs) is calculated both in theoretical Rock’n’Roll model and imperial model derived from precious experiments. Dust diameter, friction velocity, as well as roughness of wall is investigated in detail.
The Development and Assessment of a New CHF Correlation for PWR Fuel Assemblies
Authors: Bai Ning , Liu Wei , Zhu Yuanbing , Shan Jianqiang , Zhang Bo , Gou Junli , Ren Zhihao , Li Jinggang
Published: February 2014

The CHF in PWR fuel assemblies is usually predicted by the local flow correlation approach based on subchannel analysis while the effects of spacer grids, cold walls, non-uniform heat flux, etc are investigated. By using the subchannel code ATHAS to calculate each set of bundle CHF data, the local thermal-hydraulic parameters at DNB occurrence point were obtained. In present study, the minimum DNBR point method was applied to develop a new CHF correlation for PWR fuel assemblies. The so-called “three-step method” and “magnitude analysis method” were used to determine the shape and the expression of each item, respectively and the least square method was applied to determine the coefficients of the correlation. Based on the large database of CHF tests, the CHF correlation named ACC correlation has been developed to calculate the risk of DNB. The analysis and assessment results indicate that the ACC correlation can fit the experimental data well with high prediction accuracy and correct parametric trends. Coupled with subchannel code ATHAS, this correlation can simulate the thermal-hydraulics performances of PWR fuel assemblies exactly.
Ultrasonic NDE for Irradiation-Induced Material Degradations
Authors: Isobe Yoshihiro , Etoh Junji , Sagisaka Mitsuyuki , Matsunaga Takashi , Freyer Paula , Garner Frank , Okita Taira
Published: February 2014

We have developed a prediction model of ultrasonic wave changes caused by microstructural changes in 304 stainless steel based on the theory of ultrasonic wave propagation and experimental results. The correlation between microstructure and ultrasonic waves was simulated in the model with microstructure components such as grain boundaries, voids, precipitates and dislocations taken into account. Material parameters used for the prediction model were determined from the literature, microscopic observations, and experimental results obtained from archive materials.

Calculations were performed assuming respective and integrated changes of irradiation-induced microstructures and spectrum changes of ultrasonic signals due to microstructural changes were evaluated. Calculations for archive material were also performed for validation, and the results were reasonably consistent with experimental data. It was found that by selecting appropriate indicators one can identify the nature of microstructures and quantify the microstructural changes due to irradiation.
The Application of Virtual Reality Technology to the Conical Roof Lifting for the AP1000 Nuclear Power Plant
Authors: Shen Yang , Shen Hongwei
Published: February 2014

It mainly discusses the application of virtual reality technology to the conical roof lifting for the AP1000 nuclear power plant. By choosing the sketch-up and AutoCAD soft-wares, this paper established the 3d models of the conical roof and the 2600T Lampson crane in the manner of parameter. Based on virtual construction theory, these 3 d models were used to implement the simulation tests for the whole lifting progress in virtual environment so as to obtain optimized construction scheme and to train crane operator. And also the optimum running track for the crane is provided by the analysis, and the most economical enlargement plan for the 2600T Lampson crane landing pad is also provided.