bluuebunny/crossref_metadata_2025 huggingface dataset:
Study on Neutronic Characteristics and Neutronic-Thermohydraulic Coupling Behavior of SCWR With Thorium-Based Fuel
Authors: Liu Shichang , Cai Jiejin
Published: February 2014

A novel thorium-based fuel assembly for SCWR has been introduced in this paper, taking advantage of the neutron spectrum of SCWR. Single channel thermohydraulic model was integrated with calculation codes “dragon” and “donjon”, developing a computation platform for neutronics assessment and thermohydraulic analysis of the proposed fuel assembly.

Results show that the thorium-based fuel assembly meets design rules of SCWR. Meanwhile, it has considerable advantages on both efficient fuel utilization and lower minor actinide generation.
Liquid Lead Bismuth Embrittlement Susceptbility of Optifer-IX FM Steels
Authors: Gao W. , Zhao Y. , Yang F. , Zhang T.
Published: February 2014

This paper describes the results of uniaxial tensile tests on unirradiated and irradiated ferritic and martensitic (FM) steel Optifer-IX, in Argon or liquid Pb-Bi eutectic (LBE) environment at strain rate of 1 × 10−5s−1 during the temperature range from 150∼400°C. The results were compared with those have been obtained from T91 and F82H FM steels. It suggested that the normal Optifer-IX exhibited non LBE induced embrittlement effect; while besides irradiation induced embrittlement effect, the irradiated ones may endure the additional embrittlement effect induced by LBE. The fracture surfaces showed ductile mode for unirradited ones, mixed mode type for the irradiated ones respectively.
The Challenges and Opportunities in Developing Nuclear Engineering Education in Indonesia After Fukushima Accident
Authors: Putero Susetyo Hario , Rosita Widya , Sihana Fnu , Santosa Haryono Budi , Muharini Anung
Published: February 2014

Indonesia plans to build the first nuclear power plant (NPP) to solve the country’s energy problems. It needs full skilled human resources in the field of nuclear technology. Department of Engineering Physics, Gadjah Mada University is the only university in Indonesia offering a nuclear engineering curriculum which is established since 1977. The main obstacle of nuclear engineering education development is the unclear government policy. The government postpones its plan several times since the first nuclear research reactor was established in Bandung (1972) due to the presence of anti-nuclear groups, especially after Fukushima accident. This paper would like to describe our experiences in order to response the challenge mentioned above. As an education institution, Gadjah Mada University has a mission to develop science and technology for leveraging the nation competitiveness. According to that goal, the nuclear engineering development has to be independent to the presence of NPP in Indonesia, since many NPP is establishing in the world. As addition, according to the Nuclear Energy Regulatory Agency (BAPETEN), there are 2,825 institutions, including medical institutions, which are currently utilizing radioactive materials in Indonesia. As addition, Indonesia now is playing important role in the radio-pharmacy production, particularly in Asia. So, there are several opportunities as basis for developing the new nuclear engineering curriculum. In the new nuclear engineering curriculum at Gadjah Mada University, student is offered two concentrations named Nuclear Energy System and Medical Physics Technology. The first one is oriented to support the nuclear reactor system development including G4 NPP, research reactor and radioisotopes production reactors. They study how to plan, to operate and to develop from cradle to grove of a nuclear reactor system that means from front-end to back-end. In the other hand, the second one is developed to graduate competence human resources who could apply their knowledge in medical field. They would be a partner of radio-oncologist in hospitals. Besides the hard skills, student’s character has to be educated in order to create their competitive excellences based on safety culture, global and national vision. We believe that the above mentioned hard and soft skills will assist student to survive in today’s global competition, especially in nuclear technology utilization.
The Treatment Technology of Radioactive Sodium in the Frame of Fast Reactors Decommissioning
Authors: Wang Guozhi , Du Haiou , Xie Chun , Wang Rongdong , Liu Chen
Published: February 2014

A number of sodium fast reactors around the world have been shut down permanently, most of which use sodium as coolant in liquid form. Owing to the extremely high chemical reactivity of sodium, radioactive sodium waste arising during decommissioning reactors requires unique treatment technologies. The paper introduces various technologies about treatment of radioactive sodium, as follows: drained bulk radioactive sodium treatment, polluted radioactive sodium treatment. By comparing the discussed technologies in the following aspect of advantages, disadvantages, limitations, industry experiences, hazards, and final waste products, the paper concludes with recommendations about the preferential treatment technology of radioactive sodium.
The Human Reliability Analysis in Fire PSA Using Spar-H Method
Authors: Wang Yao
Published: February 2014

According to existing research results, fire risk makes a significant contribution to the total risk of a nuclear power plant (NPP). So fire probabilistic safety analysis (PSA) for NPPs is becoming more and more important in recent years. How to perform human reliability analysis (HRA) which is an essential part of PSA is therefore being paid more and more attention in fire PSA. This paper describes the characteristics and special considerations of HRA in fire PSA, and demonstrates in fire PSA how to use SPAR-H method which is so-called an advanced second-generation HRA method and is being widely used in PSA for Chinese NPPs. The study results can be a reference for other HRA analysts to use SPAR-H method in fire PSA models or other PSA models in Chinese NPPs or the world-wide nuclear industry.
Experimental Studies of Thermal Hydraulics of a HLMC Flow Around Heat Exchange Surfaces
Authors: Iarmonov Mikhail , Novozhilova Olga , Bokov Pavel , Beznosov A. V.
Published: February 2014

Temperature and velocity fields in high-temperature lead coolant flows in a circular clearance for controlled oxygen impurity content in a flow were experimentally studied at the Nizhny Novgorod State Technical University by R.E. Alekseev (NNSTU). Temperature and velocity fields were simultaneously studied in “cold” and “hot” parts of the circuit in the following operating conditions: the lead temperature is t = 400–550 °C, the thermodynamic activity of oxygen is a = 10−5–100; the Peclet number is Pe = 500–7000, the coolant flow velocity is w = 0.1–1.5 m/s, and the average heat flux is q = 50–160 kW/m2. It has been found that the oxygen impurity content and characteristics of protective oxide coatings affect temperature and velocity fields in round and circular channels. This is due to the fact that oxygen in a coolant and oxide coatings on the surfaces limiting a liquid metal flow influence characteristics of the wall boundary region. The heat transfer process that occurs when HLMC transversely flows around heat exchange pipes is investigated now at the NNSTU. The experimental facility is a combination of two high-temperature liquid-metal stands, i.e., FT-2 with the lead coolant and FT-1 with the lead-bismuth coolant combined with an experimental section. The temperature of a heat-exchange surface is measured by thermocouples of diameter 1 mm mounted in walls of heat-exchange pipes. Velocity and temperature fields in a high-temperature HLMC flow are measured by special sensors placed in the flow cross section between rows of heat-exchange pipes. Heat transfer characteristics and temperature and velocity fields in a high-temperature lead coolant flow are studied in the following operating conditions: the lead temperature is t = 450–500 °C, the thermodynamic activity of oxygen is a = 10−5–100, and the coolant flow rate through the experimental setup is Q = 3–6 m3/h, which corresponds to coolant flow velocities of V = 0.4–0.8 m/s. Integrated experimental studies of characteristics of the heat transfer that occurs when the lead coolant transversely or obliquely flows around pipes have been carried out for the first time and the dependences Nu = f(Pe) for controlled content of thermodynamically active oxygen impurity and sediments of impurities have been obtained. It is assumed that the obtained experimental data on distribution of velocity and temperature fields in a HLMC flow will permit to study heat transfer processes and to use them for developing program codes for engineering calculations of heat exchange surfaces (steam generators) with a HLMC flow around them.
Reliability Analysis for Power Supply System in a Typical Reprocessing Facility Based on GO Methodology
Authors: Wang Renze , Feng Zongyang , Li Pengbo
Published: February 2014

Since the nuclear energy industry in China has developed rapidly, the demand for spent fuel reprocessing is becoming more and more distinct. In a reprocessing facility, power supply system (PSS) is important for operation-stability and safety-guarantee of the whole facility. Therefore, the reliability of the PSS deserves in-depth investigation and analysis.

Fault tree (FT) methodology is the most accustomed methodology for system-reliability analysis and it has been standardized and applied widely. GO methodology is another effective methodology for system-reliability analysis but it has not been applied widely.

In this work, GO methodology was applied to analyze the reliability of the PSS in a typical reprocessing facility. First, for modeling expediently, based on the fact that tie breakers are set in the system, tie breaker operator was defined by myself. Then, GO methodology modeling and quantitative analysis were performed, minimal cut sets (MCSs) and average unavailability of the system were obtained. Finally, parallel analysis between GO methodology and fault tree methodology was also performed. The results from the two methodologies are completely coincident.

The results of this work show the following two points:

1. Setup of tie breakers in the PSS of the reprocessing facility is rational and necessary;

2. For reliability analysis of the PSS of the reprocessing facility, parallel with fault tree methodology, GO methodology has two distinct advantages: Its modeling is much easier and its chart is much more succinct.
Fangjiashan Nuclear Power Plant Unit1 Fuel Handling and Storage System (PMC) V&V Overview
Authors: Ma Gang
Published: February 2014

The product of Fangjiashan Nuclear Power Plant (NPP) Unit1 Fuel Handling and Storage (PMC) System is the first Chinese PMC system, with its own intellectual property rights. During the development and design lifecycle for the product, China Nuclear Control System Engineering Co., Ltd (CNCS) was responsible for Verification and Validation (V&V) processes of the PMC system, taking a significant role for ensuring the reliability and safety of the software in PMC system. The PMC V&V project is also the first ever independent V&V project of the China nuclear industry.

V&V is a technical discipline of systems engineering. The purpose of V&V is to help the development team build quality into the system during the life cycle. The V&V processes determine whether the product developed satisfies its intended use and user’s needs. V&V provides an objective assessment of products and processes throughout the life cycle. This assessment demonstrates whether the requirements are correct, complete, accurate, consistent, and testable.

CNCS V&V PMC project team performed PMC V&V activities in compliance with IEEE Std 1012-1998, which is endorsed by NRC Regulatory Guide 1.168-2004. After critical analyses by V&V team, the Software Integrity Level (SIL) of PMC is SIL3, so associated V&V activities must conform to the minimum tasks list in IEEE Std 1012-1998 Table 1. After 2 years of hard working, Fangjiashan NPP Unit1 PMC system V&V activities have been completed. The whole activities and methodology was passed the final evaluation by the customer and was highly compliment by the international V&V expert committee at year of 2012.

This paper discusses how to establish V&V organization and infrastructure, and then how to perform V&V activities in high quality, and the specific new and best practice methods, approach and tools for executing V&V tasks. Because the PMC V&V project is the first ever independent V&V project of the China nuclear industry, the methods, approach and tools which was executed in the whole lifecycle and depicted in this paper could help to guide and be referred by the future V&V activities and projects.
The Study of Cavitation Characteristics of a Heavy Liquid-Metal Coolant
Authors: Bokov P. A. , Beznosov A. V. , Lvov A. V. , Iarmonov M. V. , Bokova T. A.
Published: February 2014

Adequate design solution and maintenance of circuits with fast reactors cooled by lead and lead-bismuth coolants require taking into account the peculiarities of hydrodynamics of these coolant flows. The design pressure of saturated vapors of lead and its alloys at temperatures of 400–550 °C is 10 −18−10 −10 atm, which is significantly less than that of sodium or water. Processes of traditional cavitation cannot occur in a flow of heavy liquid-metal coolants because of their specific character. The main circulation pumps of reactor circuits are one of their basic elements. In fact, the flow-type parts of these pumps and other vane pumps operating in lead and its alloys cannot be calculated by traditional methods in terms of cavitation characteristics; appropriate calculation formulas are not currently available. To study cavitation processes in a heavy liquid-metal coolant flow, the authors have carried out the experiments aimed at:

- determining the conditions of disconnection of liquid lead and lead-bismuth eutectic column;

- determining the cavitation characteristics of a centrifugal pump transferring lead at a temperature of 500 °C;

- studying the characteristics of ejector (Venturi nozzle) in a liquid metal;

- studying the cavitation erosion effect of the lead coolant on impeller vanes of an axial-flow pump in a limited volume in the FT-4-A stand at a lead flow rate of up to 1200 t/h;

- studying the cavitation characteristics of an axial-flow pump in the FT-4 stand at shaft speeds of 13.34–25 Hz.

These studies are performed with the lead coolant at temperatures of 450°−550°C, oxygen in lead from 10−4−10−5 to 100, flow rates from 20 to 1800 t/h, which corresponds to velocities of the lead coolant flow from 1.0 to 26 m/s.

The experiments have shown that as distinct from water, traditional cavitation processes in a heavy liquid-metal coolant (HLMC) flow are not recorded. The probable cavitation mechanism is gas cavitation. The allowance for the specific character of hydrodynamics of HLMC flows is necessary for adequate design engineering and maintenance of some elements of the reactor circuit.
The Research on Core Melting Process: Oxidation
Authors: Wang Jun , Tian Wenxi , Lu Jianan , Ma Yingying , Su Guanghui , Qiu Suizheng
Published: February 2014

Beyond-design basis accidents in the AP1000 may result in reactor core melting and are therefore termed core melt accidents. The aim of this work is to develop a code to calculate and analyze the oxidation of a single fuel rod with total failures of engineered safeguard systems under a certain beyond-design basis accident such as a gigantic earthquake which can result in station blackout and then total loss of coolant flow. Using the code, the responses of the most dangerous fuel rod in the AP1000 were calculated under the accident. A discussion involving fuel pellets melting, cladding rupture and oxidation, and hydrogen production then was carried out, focused on DNBR during coolant pump coastdown, the cladding intactness under different flow rates in natural circulation, and the delay effect on cladding rupture due to cladding oxidation. By the analysis of calculated results, several suggestions on guaranteeing the security of fuel rods were provided.
Reactivity Hold-Down Technique for a Soluble Boron Free PWR Using TRISO Particle Fuel
Authors: Dai Xiang , Cao Xinrong
Published: February 2014

TRISO coated particle, developed for HTGR initially, has advantages of nuclear proliferation-resistance and fuel integrity against the release of fission products. In this paper, a 350MWt small sized PWR core design utilizing TRISO fuel concept is presented. TRISO particles are dispersed in graphite matrix to form the fuel compact, and then the fuel compact is clad by Zircaloy-4 cladding to form a fuel rod. The graphite matrix increases thermal conductivity of fuel compact, so that the fuel average temperature would be well below conventional PWRs’.

In order to simplify reactor design, operation and maintenance, soluble boron free concept while operation is introduced. The emphasis of the study is put on the reactivity hold-down technique for the 350MWt PWR core. Excess reactivity is suppressed through a combination of Pu-240 adding with Gd2O3 loading. Pu-240 is added into UO2 fuel kernel of some assemblies, and Gd2O3 rods are loaded in other assemblies. The non-fissile plutonium isotope Pu-240 has a considerably high thermal neutron capture cross section compared to U-238, so that the Pu-240 added fuel can greatly suppress excess reactivity over burnup. Besides, reactor core life would be extended by adding proper amount of Pu-240 for its converting into Pu-241 which is a fissile isotope. Combining Pu-240 adding with Gd2O3 loading, the designed core reaches an average core burnup of approximately 58GWD/t, as well as a core life of nearly 6EFPY.
Matrix Method of Characteristics Based on Modular Ray Tracing
Authors: Wu Wen-bin , Wang Kan , Li Qing
Published: February 2014

The Matrix Method of Characteristics (MOC) is an alternative to traditional MOC, in which repeatedly characteristics sweeping is needed. In the Matrix MOC, a linear algebraic equation system, which is represented by coefficient-matrix, can be formed by sweeping only once, and then solving the linear system takes the place of repeatedly characteristics sweeping. There are many linear solvers to be use, like direct LU decomposition and widely used Krylov subspace iterative methods. In the past, 2D Matrix MOC was implemented based on the long characteristics in which tremendous memory is consumed to hold the geometrical information. Then, 2D Matrix MOC based on modular ray tracing was implemented to overcome the difficulty. Modular ray tracing just holds the geometrical information of typical modules like cells or assemblies, consequently needs less memory. In this study, Filippone-type modular ray tracing in which different azimuths own the same number of characteristics lines was used, and AutoCAD VBA programing produced the geometrical information. Also, the Matrix MOC based on modular ray tracing was implemented to solve half, fourth and eighth reactor core in which additional modules, e.g. half-module, fourth-module and eighth-module were involved and need special considerations. Numerical results demonstrate that Matrix MOC based on modular ray tracing can obtain good efficiency and accuracy.
GEANT4 Simulations Response of Cylindrical Silicon-on-Insulator Microdosimeter Array Detector for Neutrons Detection
Authors: Lei Ming , Liu Shu-Huan , He Chao-Hui , Xiong Cen
Published: February 2014

In the experiment about the radiation protection and radiation biology, microdosimeter is often used as a tool for tissue equivalent microdosimetry measurement. The characteristics of neutron energy response for the cylindrical SOI (Silicon-on-Insulator) microdosimeter arrays detector designed with different technology structures are simulated and compared with GEANT4 toolkit. In this work, the distributions of neutron linear energy spectra which influenced by influence angles of incident particles, types and depth of conversion materials are studied. The characteristics of neutron energy response with different kinds of conversion materials of the SOI microdosimeter arrays are mainly analyzed. Meanwhile, we also calculate dose distributions of linear energy spectra in a kind of novel cylindrical Silicon-on-Insulator microdosimeter arrays detector (with sensitive sizes to 1um and 2um) due to 252Cf neutron source. Comparisons between simulated data with the SOI microdosimeter measurement and previous reported experimental data with TEPC (tissue-equivalent proportional counter) show a rather good agreement. Differences partly due to different kinds of conversion materials are analyzed. The results show that LDPE (low-density polyethylene) is a better kind of tissue equivalent material for neutron microdosimetry, but the differences in the detection of higher linear energy range 10–100KeV/um between LDPE and other tissue-equivalent materials may be caused by the lack of microelements for LDPE material and the sensitivity to the angle of incidence in the linear energy range 10–100KeV/um also should not be neglected for the distorted spectrum caused by setting angular deviation of more than 45 degree on the detector surface. The main reason of the influence on the linear energy spectra from the larger angular deviation incidence is that the design of cylindrical shell structure of sensitive volume could increase the recorded length of the tracks of secondary particles of larger angular incidence neutron included in the sensitive volume cell. The results of calculations and analyses could provide some important information for the optimization to the SOI microdosimeter measurement.
The Study of Qinshan III PHWR Thermosyphoning Capability
Authors: Sheng Xin
Published: February 2014

Thermosyphoning capability, also known as the natural circulation is an important nuclear safety indicator in NPP. In the absence of any forced circulation of the NPP primary system, the cooling of the fuel to remove the decay heat can be provided by thermosyphoning to the steam generator (SG).

Based on the theory of natural circulation and simulated test, this paper will further identify thermosyphoning capability of Qinshan III PHWR (CANDU6). Analysis shows that, the ultimate reactor heat sinks are SGs and related safety valves in case of loss main pumps. Thermosyphoning capability was confirmed as effective and acceptable.

Furthermore, loss of Power accident happened during Qinshan III Unit#1 normal operating condition. We will investigate the impact of reactor core cooling and validate thermosyphoning as the capability of Qinshan III PHWR heat sink on accident.

Finally, this paper will give some suggestions to NPP operator’s actions issues with reactor core temperature difference, reactor inlet headers sub-cooling margin, reactor outlet header temperatures, and others. Meanwhile, paper also discuss on the improvement of CANDU future design.
Creep Strength of Chromium-Containing Conventional and ODS Steels in Oxygen-Controlled Pb at 650°C
Authors: Yurechko Mariya , Schroer Carsten , Wedemeyer Olaf , Skrypnik Aleksandr , Konys Jürgen
Published: February 2014

Two martensitic conventional steels, T91 (modified 9Cr1Mo) and P92 (NF616), and two ODS steels with 12 and 14 mass.% chromium content were subjected to creep-rupture tests in stagnant oxygen-controlled Pb with co = 10−6 mass.% at 650°C. The 9Cr conventional steels were tested in liquid metal at 75–200 MPa, while the stress for the tests of ODS steels were chosen in the range of 190–400 MPa. Reference tests were carried out in stagnant air under similar loading. Both ODS steels show obvious change in secondary creep rate and failure appearance at 330–350 MPa, while the martensitic steel P92 tested in Pb at 75 MPa (tR = 13,090h) shows a significant drop in creep strength accompanying with a change from ductile to brittle fracture and a considerably reduced necking. Effects of microstructure evolution during the exposure at the elevated temperature and risk of LME due to a direct contact to liquid metal are discussed.
CFD Simulation of 5×5 Rod Bundles With Split-Type Spacers
Authors: Podila K. , Bailey J. , Rao Y. F. , Krause M.
Published: February 2014

Atomic Energy of Canada Limited (AECL) has initiated a program to develop Computational Fluid Dynamics (CFD) capability for simulating single- and two-phase flows in rod-bundles. In the current work, a 5×5 rod assembly with a split-type spacer grid is simulated with ANSYS Fluent 14 using unsteady simulations with a fully conformal hybrid mesh (wall y+∼30). This work represents results of AECL’s recent participation in the OECD/NEA organized CFD benchmarking exercise on the MATiS-H experiment performed at the Korean Atomic Energy Research Institute (KAERI). The sensitivity to turbulence models is tested using the standard k-ε and the Reynolds stress model (RSM). Reasonable agreement is achieved between the calculated and experimental velocity values in the region close to the spacer grid, whereas turbulence intensity values are underpredicted compared to the experiments.
Elementary Analysis for the Decommissioning Schedule of High Temperature Reactors
Authors: Xiong Yin , Li Shengqiang , Hao Yalei , Jiang Shengyao
Published: February 2014

From the safety consideration, the decommissioning of High Temperature Reactors should be successfully achieved or do not have a direct impact on the advanced nature of the high-temperature reactor technology and public acceptability. Therefore, based on the decommissioning experience at home and abroad, the relevant research of decommissioning in the national major topics was referenced. Combined with the technical features of High Temperature Reactors, adopted decommissioning tactics and implementation plan suitable for the High Temperature Reactors was presented. It hopes to enhance the safety features for the whole lifetime of High Temperature Reactors.
Elementary Analysis of the Quality Assurance for Probabilistic Safety Analysis
Authors: Xiong Yin , Li Shengqiang , Hao Yalei , Jiang Shengyao
Published: February 2014

The quality of Probabilistic Safety Assessments (PSA), frequently named as probabilistic risk assessment (PRA), is one of the core issues in advanced nuclear power plant applications. Although numerous guidance and regulations have been provided to enhance the PSA quality, the inherent uncertainty of current PSA and project management process still make it to be puzzled. Based on the implementation of current PSA development status and limitation, a new integral model accounting for interrelated economic, health, security, quality, environmental and economic matters as well as quality attributes is proposed in this article. We aim to provide a new idea for other investigators and hope it can be help to further enhance the quality of PSA in industrial applications.
A Study of Rules of Material Groupings in ASME IX QW-420 by Decision Tree Method
Authors: Guo Lifeng , Zhang Yaochun , Shi Libo , Dong An
Published: February 2014

Based on ASME IX QW-420, each base metal is assigned a P-No., which is a critical material parameter and used extensively in the ASME Codes, including welding procedure qualification, qualification of welder and welding operator, preheat and post weld heat treatment., etc. However, unlike EN and ISO standards, ASME Code doesn’t explicitly specify the rules of material grouping, which brings inconvenience in some circumstances. In this study, the 1872 base materials in ASME IX QW-420 (2007 Edition) are investigated from the point of view of material chemical composition, mechanical properties and weldability, with the aid of decision tree method. Through the study, a complete set of rules is given to classify a certain material into a P-No., which provides an aspect to understand ASME material grouping and facilitates use of the Code, and more importantly, helps to implement the principle of ASME Code in local standards.
Numerical Prediction of the Outcomes of Binary-Droplet Collision in Steam-Water Separator
Authors: Zhang Huang , Bo HanLiang
Published: February 2014

A theoretical model is carried out to describe the regimes of binary-droplet collision in steam-water separator (SWS). Different from previous researches which seldom concerned the interaction between droplets moving in the SWS, this study indicates that droplet collision, especially the binary-droplet collision, is a key factor to influence the efficiency of SWS. Four regimes of binary-droplet collision including bounce, coalescence, reflexive separation and stretching separation are considered. A brief summary is showed how this theoretical model is established according to previous papers. Then we use this model to study the probability of the four regimes and the states of droplets (e. g., number, size and velocity) after binary-droplet collision. The results perform that, at low symmetric Webber number (Wes), coalescence is the dominate regime. While, with the increasing Wes, the probability of stretch separation goes up quickly. Otherwise, coalescence would not occur frequently at this time. The probability of reflexive separation always stays at a low level. In addition, bounce would not happen whatever the Wes is. From the calculation results of post-collision drops, binary-droplet combine as one drop in coalescence. In reflexive or stretch separation, the large drop and the small one would remain their velocities but their radii might decrease, for satellite droplets may be generated.
PSA Conservatism Analysis and Quantification for the LOFW Accident
Authors: Kong Tony X. , Deng Chunrui , Vrbanic Ivan , Basic Ivica
Published: February 2014

The conservatism was introduced in the probabilistic safety analysis (PSA) while successes criteria for the event tree were determined. The conservatism has strong relation with the uncertainty but independent with the uncertainty, and the conservatism could significantly influence the CDF results, contribution of dominant sequence, and sensitivity analysis. For methodology, the conservatism in PSA was identified, conservative numbers were defined to quantify the conservatism. Addressing the loss of feed water (LOFW) accident at the PWR, the probability distributions for the function events of the event tree were defined, and the conservatism was quantified based on the PSA insights and sensitivity analysis. The methodology and the application of the conservative numbers were discussed and conclusions were drawn.
Progress in the Large-Scale Simulation of Convective Boiling Heat Transfer Under Turbulent Conditions Using TransAt
Authors: Caviezel D. , Lakehal D. , Gong Y.
Published: February 2014

A detailed numerical analysis of the turbulent convective flow along the heated rods of an idealized PWR sub-channel is investigated. The key predicted quantity is the length at which the rod surface temperature reaches nucleation temperature, approximated here as the saturation temperature. The turbulent effects are modeled using LES with very high grid resolution (∼6 million cells) allowing resolving the viscous-affected layer. In a second step, boiling nucleation is triggered artificially and vapor bubbles are allowed to interact with the turbulent flow. We’ll thoroughly discuss the results, including comparing the turbulence data basis with single-phase flow conditions, and the effect of boiling on the flow.
About Harmonization of Deterministic and Probabilistic Approaches to the NPPs Safety Substantiation
Authors: Ponomarenko Grigory L.
Published: February 2014

Severe accident on the Fukushima NPP lowered the public trust for nuclear power engineering and must lead to the revision of the NPPs safety substantiation in the trend of the more strict regulatory body’s requirements. This relates to the deterministic (DSA) and probabilistic (PSA) safety analysis. In the author’s opinion, the most effective increase in the safety should be reached due to convergence or harmonization of deterministic and probabilistic methods. In particular, at fulfillment of DBAs in a number of cases one should consciously move in the direction of more realistic analysis of the actually BDBAs emergencies (in the traditional understanding) with the imposition of more than one Initiating Event (IE) or more than one Single Failure. In turn, when the PSA is fulfilled one should move from the realistic analysis without taking into account uncertainties in the direction of their reasonably conservative calculation. This work presents the practical realization of harmonious interaction of methods DSA and PSA based on the example of the estimation of the minimally sufficient CRs CPS for WWERs. A new conceptual approach to the definition of a minimum required quantity of CRs as part of the Spatial Effects Methodology is presented here and in papers [15, 16]. This approach combines the neutron-physical, thermal-hydraulic and probabilistic aspects, using DBC’s Fuel Criteria generally accepted for CD realization. Analysis is performed with the up-to-date coupled code KORSAR/GP [1–3] on example of the mode MSLB, which has the most dependence on EP with respect to any other accident.

This paper presents an approach that allows:

- to assess the EP success criteria with multiple stuck CRs for the purposes of PSA. In other words — assess the maximum number of stuck CRs when CD does not yet occur;

- justify the low (P_EPfail ≤ 10−5, in accordance with Russian State Standard [4]) probability of EP failure on demand, for different amounts of CRs in a range from 49 till 121 pcs. Furthermore, this probability allows substantiate the low partial CD probability P_CDpartial < ∼10−10 for concrete scenarios with multiple stuck CRs as a negligible contribution in the total CD probability which should be P_CDtotal < 10−5;

- a joint optimization for positioning and quantities of CRs and power monitoring detectors (SPND) in central cell of FAs. Indeed, if the total number of CRs in WWER-1000 becomes more than 79–85 pcs, it creates some inconvenience, particularly the need to shift the power monitoring detectors from the central cell of fuel assembly into another noncentral cell.
Electronic Structure and Mechanical Properties of Zircaloy-2 and Zircaloy-4: A First Principle Study
Authors: Lu Chunhai , Chen Wenkai , Chen Min , Ni Shijun , Zhang Chengjiang
Published: February 2014

The local-density approximation (LDA) coupled with the virtual crystal approximation (VCA) method electronic structure is applied to evaluate elastic constants, bulk modulus, shear modulus, Young’s modulus and Poisson’s ratio mechanic properties of metal zirconium, Zircaloy-2 and Zircaloy-4. The results show that there is no obvious difference in band structure and total density of state (DOS) between metal zirconium and zirconium alloy. However, p and d electron partial density of state (PDOS) presents the slight difference between metal zirconium and zirconium alloy. Zircaloy-2 and Zircaloy-4 present better elastic mechanical properties than metal zirconium. The metal zirconium and zirconium alloy show the anisotropic mechanical properties.
Behavior of a Magnetically Suspended Rotor Dropped Into Auxiliary Bearings
Authors: Xiao Zhen , Yang Guojun , Shi Zhengang , Yu Suyuan
Published: February 2014

Auxiliary bearings (ABs), as the back-up support when the Active Magnetic Bearing (AMB) system doesn’t work, play an important role on the assurance of AMB system’s safety. According to the structure design of ABs-rotor system on back-up helium circulator of 10MW high temperature gas-cooled reactor (HTR-10) equipped with AMB, this paper presents a simplified finite element model (FEM) in Ansys, applied in the transient response of the dropped rotor. The ABs load characteristics were determined based on Hertz contact theory. Furthermore, a special spring element named Combin40, which is coupled to a gap, was introduced to communicate information of displacement and force between auxiliary bearings which were simplified into nonlinear spring elements and rotor which was simplified into beam elements. Based on the developed FEM, the behavior of vertical and horizontal magnetically suspended rotor dropped into the ABs was analyzed respectively and the safety of the rotor and ABs was evaluated. The results provide an important reference to the ABs design and application in HTR-10 and High Temperature Reactor-Pebblebed Modules (HTR-PM).
Thermal-Hydraulic Conceptual Design of Water-Cooled Blanket of a Fusion-Fission Hybrid Reactor for Energy Production
Authors: Guo Haibing , Peng Xianjue , Shi Xueming , Ma Jimin
Published: February 2014

A Thermal-hydraulic conceptual design of the energy production blanket for a fusion-fission hybrid reactor is proposed in this paper. The blanket uses U-10Zr alloy as its fuel, in which the uranium abundance is natural. It uses water as coolant as well as moderator, so that the Pu-239 bred from neutron capture reaction of U-238 can be depleted by thermal neutrons, while the fast neutrons can cause U-238 fission. A properly designed blanket needs no separation of Pu-239 from spent fuel, and makes full use of U-238 to produce energy. Based on the preliminary neutronic conceptual design of the blanket, a detailed blanket model has been established, and the distribution of deposited energy in it is calculated by MCNP. In this conceptual model, the coolant flows in tubes, which go through from the bottom to the top of the blanket, and the fuel fills in the gap between tubes. A set of criteria have been developed for the thermal-hydraulic design, consequently the restrictions on tube size, wall thickness, pitches and power density are determined. The study on thermal-hydraulic characteristics of the blanket is carried out by the 3D computational fluid dynamic code FLUENT, in which the variation of transverse cross section due to the special geometry of Tokamak can be considered. The temperature and heat flux distribution at typical positions are analyzed, and the dependence of thermal-hydraulic performance on fuel-coolant volume ratio is discussed. It can conclude that the conceptual design proposed in this paper has satisfactory performances, all parameters satisfy the preset thermal-hydraulic criteria, and have large enough safety margin.
The Thermal Balance Calculation of AP1000 Turbine Roll-Up by Non-Nuclear Steam
Authors: Wang Shunda , Li Xiuli
Published: February 2014

The initial turbine non-nuclear steam roll-up is an important test during nuclear power plant startup. The purpose of the test is to discover and solve turbine problem before reactor achieve critical. The benefit of this test is milestone may be shorter and save time. Use AP1000 equipments design parameters to calculate the basement of this test, by the way, we can find a safe method to startup.

In this test, the heat is produced by reactor coolant pumps and PZR heaters. During roll-up, to avoid critical zones of turbine and reduce thermal stress, we hold on some speed points. At low speed period we can make up heat to make sure we can achieve the purpose of this test. So the calorimetric balance between primary side and second side is very important.

Non-nuclear steam has many benefits for nuclear power plant startup. First, the secondary loop completes build before the primary system, but as one of the most important equipments, turbine must be examined to make sure it can roll-up and operation without no disease, so before primary system equipments installation finished we can save lots of time to solve turbine problems; the second one, during this duration, all of equipments of primary system can perform their function in the first time, we also make sure system function can perform correctly and timely. So, as a big picture, the milestone of startup can be shorter by replace nuclear steam rolling-up the turbine.
Prospective Applications of Flow Drag and Noise Reduction Technology for Nuclear Reactors
Authors: Feng Beibei , Chen Darong , Yang Xingtuan , Jiang Shengyao
Published: February 2014

Flow drag and noise reduction technology is a multidisciplinary field currently relative to hydrodynamics, materials and physics. A comprehensive review of the literature on the principles, limitations and engineering applications of the technology was performed in this paper. Due to restriction of experimental conditions, it is rarely adopted to use to reduce flow resistance associated with noise for nuclear reactors. Assessment of effects of flow drag and noise reduction on circulation efficiency of cooling water and helium gas was done based on riblet structured surface, which would efficiently reduce the flow resistance and noise with high reduction rate and was available to realize for nuclear reactors. Results of this analysis showed that vortex induced in riblet grooves was the dominant factor resulting in flow drag and noise reduction on the riblet structured pipeline inner wall of cooling water and helium gas. Accordingly, higher circulation efficiency and heat transfer efficiency would eventually lead to better performance of nuclear reactors.
Lessons Learned From Automation of Nonconformance Reporting (NCR) Process at Los Alamos National Laboratory (LANL)
Authors: McFarland Shyla A. , Van Valkenburg Taunia S. , Sandquist Gary M. , Tepley Daniel J. , Petru Ernie F.
Published: February 2014

The Los Alamos National Laboratory (LANL) Quality Performance and Assurance (QPA) Division was tasked with the opportunity to improve the NCR process and to support ASME NQA-1-2008 and ASME NQA-1A-2009; Quality Assurance Requirements for Nuclear Facility Applications [1] compliance. This process improvement plan involved input from NCR Coordinators, Engineering, Project Management, Facility Operations, Suspect Counterfeit Item Coordinators, and Waste Certifying Officials, among others. The focus of the project was to fully automate the NCR process through a phased systematic approach that would be useable to a wide audience. During the short and long-term automation process development for nuclear and non-nuclear facility nonconforming items, noteworthy benefits and lessons learned were identified and implemented.
Theoretical Analysis of a Natural Circulation System Under Rolling Condition Based on RELAP5/MOD3.3 Code
Authors: Lian Haibo , Zhu Hongye , Yang Xingtuan , Jiang Shengyao
Published: February 2014

Natural circulation is an important process for certain advanced reactor’s main loop and passive heat removal systems. However, in marine conditions the thermo-hydraulic characteristics of natural circulation will change because of the ship motions such as inclination, rolling and heaving, which introduces extra body forces in to the system. In this paper, we conducted theoretical studies on the natural circulation behaviors in a symmetrical two-circuit loop under rolling conditions. A RELAP5/MOD3.3 code is developed based on the basic control equations and empirical formulas. Based on this code the natural circulation behavior under a larger range of rolling angle and period is also investigated. It is found that with the increase of rolling angle and decrease of rolling period, the magnitude of flow fluctuation increases. The fluctuation period of mass flow rate do not always consists with the rolling period under the comprehensive actions of complex body forces. Under the rolling with large angle and short period, the fluctuation period of flow rate is only half the rolling period.
Experimental and Numerical Study on Natural Circulation Under Inclined Condition
Authors: Zhu Hongye , Yang Xingtuan , Lian Haibo , Jiang Shengyao
Published: February 2014

Natural circulation systems are broadly used in marine environments. When accidents happen, these systems may work under inclined condition. In this paper, we conducted a series of experimental study on the thermohydraulics behavior of natural circulation in a symmetric two-circuit loop under the inclined angle from 0∼45°. A CFD model is also set up and predicts the results well in comparing with the experiments. Both experimental and numerical analysis show that with the increase of inclined angle, the total circulation flow rate decreases. When the loop inclines about the axis perpendicular to the circulation, one circulation is depressed while the other is enhanced; accordingly the disparity between the branch circulations arises and increases with the increase of inclined angle. The flow pattern of the circulation under larger inclined angle (45∼90°) condition is also studied by CFD model. At large inclined angle the circulation is mainly happens in one circuit. Also based on this model, the influences of flow resistance distribution and loop configuration on natural circulation are predicted. The numerical results show that to design the loop with the configuration of big altitude difference and small width are favorable to confine the influence of inclination; however too small loop width will cause sever reduction of circulation ability in large angle inclination.
Selective Adsorption Properties and Stable Solidification of Cs by Insoluble Ferrocyanide Loaded Zeolites
Authors: Ikarashi Yuki , Masud Rana Syed , Nakai Tomonori , Mimura Hitoshi , Ishizaki Eiji , Matsukura Minoru , Hosoi Yoshinobu
Published: February 2014

In Fukushima NPP-1, large amounts of high-activity-level water (HALW) accumulated in the reactor, turbine building and the trench in the facility is treated by circulating injection cooling system. The development of highly functional adsorbents and stable solidification method contributes to the advancement of the decontamination system and environmental remediation. The present study deals with (1) preparation of insoluble ferrocyanide loaded zeolites, (2) selective uptake of Cs+ in seawater, (3) estimation of Cs immobilization ratio and stable solidification. Various kinds of Cs-selective composites loaded with insoluble ferrocyanides (CoFC, NiFC) into the zeolites (zeolite A (A51, A-51J), zeolite X (LSX), chabazite (modified chabazite) and natural mordenite (SA-5)) matrices have been prepared using successive impregnation/precipitation methods by Tohoku University. As for Cs+ adsorption, these composites had relatively large uptake (%) above 95%, distribution coefficients (Kd) above 103 cm3/g and excellent adsorption kinetics even in seawater. The immobilization ratio (%) of Cs for the CoFC saturated with Cs+ was estimated at different calcination temperatures up to 1,200°C in advance. The immobilization ratio was less than 0.1% above 1,000°C, indicating that the adsorbed Cs+ ions are completely volatilized and insoluble ferrocyanides had no immobilization ability for Cs. In contrast, the insoluble ferrocyanide-loaded zeolites had excellent Cs immobilization ability; in the case of insoluble ferrocyanide-loaded natural zeolites (NiFC-SA-5, CoFC-modified chabazite), the immobilization ratio was above 99% and 96% even after calcination at 1,000°C and 1,100°C, respectively, indicating that nearly all Cs ions are immobilized in the sintered solid form. On the other hand, the immobilization ratio for the insoluble ferrocyanide-loaded A and X zeolites (NiFC-A (A51, A51J), NiFC-X) tended to decrease with calcining temperature; for example, the immobilization ratio for NiFC-X at 1,000°C and 1,100°C was estimated to be 74.9% and 55.4%, respectively, and many spots concentrating Cs were observed on the surface. The difference in immobilization behavior between natural zeolites and synthetic ones is probably due to the phase transformation and surface morphology at higher temperature above 1,000°C. The stable solidification of insoluble ferrocyanides was thus accomplished by using the excellent Cs immobilization abilities of zeolite matrices (Cs trapping and self-sintering abilities).