核动力工程

Studyon Severe Core Damage Progression Induced by Anticipated Transients without Scram and Its Mitigation in Qinshan NPP Unit1

CHE Ji-yao, CAO Xue-wu

2005, 26(3): 209-213,218.

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Abstract:
By using a simulator based on SCDAP/RELAP5/MOD3.1 computer code, the core damage progression and theeffectiveness of preventive and mitigation measures are analyzed with the anticipated transients without scram （ATWS） sequences caused by the loss of main feed water （LOFW）, the loss of off-site power （LOOP） and uncontrolled rod cluster control assembly （RCCA） bank withdrawal, combined with the failure of reactor control and protection system in Qinshan Nuclear Power Plant （NPP） unit1.

Optimization Calculation for In-Core Burnable Absorber Fuel Loading for Pressurized Water Reactor

YANG Bo, WU Hong-chun

2005, 26(3): 214-218.

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In this paper, genetic algorithms（GA） and tabu search（TS） algorithm are applied to optimize the burnable absorber fuel loading problem for nuclear power plant reactor. The tenth-cycle of Daya-Bay Nuclear Power Station is taken as the example, and three general kinds of burnable absorber, i.e., boron, Gd₂O₃ and IFBA, are optimized using GA separately. Calculation results demonstrate that GA is effective for optimizing the burnable absorber loading and the IFBA works the best for PWR. Finally a hybrid optimization method that combined with GA and TS is used. The initial optimized results of GA are taken as the initial point of TS searching. This method saves much calculation time.

Perturbation Calculation of Neutron Transport Equation with Interface Current Method

ZHANG Ying, CHEN Wei, XIE Zhong-sheng, CHEN Li-xin

2005, 26(3): 219-223.

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The perturbation calculation for neutron integral transport equation in hexagonal geometry is studied in this paper. Based on the calculation for interface current method of neutron integral transport equation in hexagonal geometry and the calculation for corresponding mathematical adjoint equation, the perturbation theory is used to calculate the reactivity variety in reactor hexagonal assembly when the parameters of cell in assembly occurred minimal change. The calculation result is satisfactory. It proves that the perturbation calculation method developed in this paper is correct.

Reflector Parameters Calculation Code Using the Transport Equivalent Method

WU Hong-chun, LIU Hai-feng

2005, 26(3): 224-227,232.

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In this paper, a calculation code （HBDC） for the parameters of core reflector is encoded based on the one-dimensional transport equivalent method and the discrete ordinates method ANISN. This code is applied to calculate the Zion-1 benchmark problem and iron-water reflector problem for verification. Numerical results demonstrate that this method is flexible for both the baffle reflector and iron-water reflector, and it can achieve high accuracy.

Investigation on Frictional Pressure Drop of Steam-Water Two-Phase Flow in an Internally Ribbed Tube

LI Yong-xing, YIN Fei, CHEN Ting-kuan, LI Hui-xiong

2005, 26(3): 228-232.

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Within the range of pressures from 9 to 22MPa, mass velocities from G 600 to 1200kg/（m²·s）, and heat fluxes from x 0 to 1.0, experiments had performed to investigate the frictional pressure drop of the steam-water two-phase flow in a six-head internally ribbed tube with the outer diameter of 38.1mm and the thickness of 7.5mm. The test section was thermally insulated as horizontal direction. Based on the experimental results, it was found that pressure had a noticeable effect on the frictional pressure drop of the mental results, it was found that pressure had a noticeable effect on the frictional pressure drop of the steam-water two-phase flow, and the frictional pressure drop factor of the steam-water two-phase flow decreased with an increase in pressure. The frictional pressure drop factor of the steam-water two-phase flow tends to one near the critical pressure. As steam quality increased, the frictional pressure drop factor of the steam-water two-phase flow first increased, and then it had a decreasing tendency. With an increase in mass velocity, the frictional pressure drop factor of the steam-water two-phase flow decreased. Correlations of the frictional pressure drop factor of the steam-water two-phase flow had been provided.

Experimental Study on High Pressure Steam-Liquid Two-Phase Flow Pressure Drop Instability in Parallel Tube

ZHANG Ji-hui, CHEN Ting-kuan

2005, 26(3): 233-236.

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Steam-liquid two-phase flow pressure drop instability in vertical parallel tubes was studied experimentally in the high-pressure steam-water test loop. The effects of system pressure, mass velocity, inlet sub-cooling, exit throttle and compressible volume on the instability were determined. The threshold of pressure drop instability was obtained. The dimensionless correlations for the prediction of the instability in the vertical parallel tubes were given. The result may be used for the design of the large boiler and steam generators.

Numerical Verification of Field Synergy Principle in Parallel Channels

TIAN Wen-xi, YU Fang-wei, QIU Sui-zheng, JIA Dou-nan, SU Guang-hui, WANG Gui-fang

2005, 26(3): 237-241.

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Numerical simulation of turbulent flow with re-circulation was conducted by SIMPLE algorithm with two-equation k-ε model. Extension of computational region method and wall function method were quoted to regulate the whole computational region geometrically, given the inlet Reynold number keeps 10000, by changing the height of the solid obstacle. The result showed that the overall wall heat flux and heat exchange Nu increased with the decreased of the angle between the velocity vector and the temperature gradient. It verified that the field synergy principle based on 2-D boundary laminar flow may be applied to complex turbulent flow even with re-circulation.

Numerical Simulation of Interface Movement in Gas-Liquid Two-Phase Flows with Level Set Method

LI Hui-xiong, DENG Sheng, ZHAO Jian-fu, CHEN Ting-kuan, WANG Fei

2005, 26(3): 242-248,267.

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Numerical simulation of gas-liquid two-phase flow and heat transfer has been an attractive work for a quite long time, but still remains as a knotty difficulty due to the inherent complexities of the gas-liquid two-phase flow resulted from the existence of moving interfaces with topology changes. This paper reports the effort and the latest advances that have been made by the author and his colaborators, with special emphasis on the methods for computing solutions to the advection equation of the Level set function, which is utilized to capture the moving interfaces in gas-liquid two-phase flows. Three different schemes, i.e. the simple finite difference scheme, the Superbee-TVD scheme and the 5-order WENO scheme in combination with the Runge-Kutta method are respectively applied to solve the advection equation of the Level Set. A numerical procedure based on the well-verified SIMPLER method is employed to numerically calculate the momentum equations of the two-phase flow. The above–mentioned three schemes are employed to simulate the movement of four typical interfaces under 5 typical flowing conditions. Analysis of the numerical results shows that the 5-order WENO scheme and the Superbee-TVD scheme are much better than the simple finite difference scheme, and the 5-order WENO scheme is the best to compute solutions to the advection equation of the Level Set. The 5-order WENO scheme will be employed as the main scheme to get solutions to the advection equations of the Level Set when gas-liquid two-phase flows are numerically studied in the future.

Effect of Heat Treatment on the Corrosion Resistance for New Zirconium-Based Alloy

LIU Wen-qing, LI Qiang, ZHOU Bang-xin, YAN Qing-song, YAO Mei-yi

2005, 26(3): 249-253,287.

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After being treated in different ways, N18 zirconium alloy specimens are exposed in 0.01mol · L^-1 LiOH aqueous solution at 350℃, 16.8MPa. The microstructures of these specimens are carried out by high-resolution transmission electron microscopy （HRTEM）. The specimen treated by 800℃/C. R/ 500℃ has best corrosion resistance among all the specimens. TEM examination shows, in addition to Zr（Fe,Cr）₂, there exist Zr-Nb-Fe SPPs, which containing much more Nb element, in the specimen. The existence of Zr-Nb-Fe SPPs is helpful to the reduction of niobium content in αZr solid solution and results in an improvement in the corrosion resistance of N18 zirconium alloy.

Study on Aging Embrittlement of 17-4PH Martensite Stainless Steel at 350℃

WANG Jun, ZOU Hong, WU Xiao-yong, QIU Shao-yu, SHEN Bao-luo

2005, 26(3): 254-258.

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The transformation of microstructure and hardness with the extension of aging time on the 17-4PH Martensitestainless steel at 350℃ is studied in this paper, and the change of dynamic fracture toughness and fractographys of the stainless steel for various holding time at this temperature are also studied by instrumental impact test and scanning electron microscope. The results indicate that the crack initiation energy （E_i）, crack propagation energy （E_p）, absorbed-in-fracture energy （E_t） and dynamic fracture toughness （K_1d ）of this type of alloy Charpy v-notch sample is decreased with the continuation of time at 350℃, it means that the toughness of the alloy is degraded, and the hardness of the steel is ascended when aging time is expanded and reaches the maximum at 9000h. The fractography of this steel changes from dimple fracture into cleavage fracture and inter-granular rapture.

Study on Brazability of Low-melting-point Ag-Al-Mn-Si Brazing Filler

YANG Jing, QIU Shao-yu, ZHU Jin-xia, WANG Fei, LIU Xiao-rong

2005, 26(3): 259-263.

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Low-melting-point Ag-Al-Mn-Si brazing filler were produced. Refer to concerned standards and test procedures, spreadability and clearance fillabitity for the filler metal on brazed titanium alloy to stainless steel, and mechanical properties of brazed joints have been tested. Test results indicated that linear temperature for liquid phase was under 850℃, the spreadable area was 22.5cm² on titanium alloy and 13.2cm² on stainless steel, the slow-flow length was bigger than 80mm, the tensile intension of joints was 242MPa and shear intension 154MPa, and rupture place was in intermetallics layers next to stainless steel. The micro-examination of brazed joints showed that Al in brazing filler would prevent dissolution and diffusion of Ti; Mn, Si in brazing filler would induce crackle formation next to stainless steel. Titanium alloy and stainless steel can be brazed using Ag-10Al-1Mn-0.5 Si brazing filler.

Effects of γ-ray on the Fireproof Materials

WEI Zhao-rong, ZHU Shi-fu, ZHAO Bei-jun, ZOU Hong, YANG Wen-bin, WEI Yong-lin

2005, 26(3): 264-267.

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In order to guarantee the fire safety of nuclear power plants, we have studied the radio resistance of two kinds of fireproof materials—the cable fireproof coating and organic fireproof putty, which are widely used in nuclear power stations. The results indicate that the γ-ray over 1000 kGy can seriously deteriorates the bending-resistance performance, the wet-resistance and heat-resistance, the cold-resistance and heat-resistance performances of the cable fireproof coating. The γ-ray over 500 kGy not only possess great effects on the corrosion performance of the organic fireproof putty but decreases the fireproof capability greatly.

Experimental Study of Heat Pipeof Nanometer Particles

LIU Jun-hong, GU Jian-ming, LIU Hui, DONG Dong-fu

2005, 26(3): 268-271.

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The application of magnetic fluid with nanometer magnetic particles in heat pipe was studied, and the effect of particle concentration on the heat transfer characteristic of the heat pipe was investigated. The experiment showed that the heat transfer ability of the heat pipe with magnetic fluid receded compared with its carrier liquid within the experimental range, and there was an optimum concentration, in which the heat transfer ability of the heat pipe was best.

PID Parameters Setting on Level Control System of Steam Generator in Qinshan Phase II NPP

DONG Hua-ping, ZHANG Jian-min

2005, 26(3): 272-276.

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Based on SimPort simulation platform of nuclear power plant, a simulation model on the level control system of steam generator in the unit one of Qinshan Phase II NPP was established. Using this model, transient simulation experiments and researches with different conditions were conducted and the setpoints of the level control system were obtained. To level controller, K_P, T_I and T_D are 4.25, 425s and 10s respectively. To flow controller, K_P is 1.0 and T_I is 13s. These setpoints are in agreement with the actual values and can be referred by engineering technician.

Analysis of Vibration of Exhaust Valve Pipeline in Nuclear Power Plant

TAN Ping, WANG Bin

2005, 26(3): 277-279,296.

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Pipeline system for conveying pressurized steam often operates under time-varying conditions due to the valve operations. This may cause vibration problems as a result the pipeline system suffered vibration damage. In this paper, a finite element formulation for the exhaust dynamic equations that include the effect of all pipe supports, and hangers is introduced and applied to the dynamic analysis of the pipeline system used in a nuclear power plant. The vibration response of steam-conveying pipeline induced by valve exhaust has been studied. The model is validated with a fieldwork experimental pipeline system. The mechanical vibrations from steam exhaust valves can be eliminated by careful design of the valve plug and seat.

Establishment and Solution of Dynamic Simulation Model for Spray Desuperheater

NING De-liang, PANG Feng-ge, GAO Pu-zhen

2005, 26(3): 280-283,290.

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Based on the conservation of mass and conservation of energy and reasonable hypotheses for the dynamic thermal system, the dynamic simulation model of spray desuperheater was established using lumped parameters method. Using Simulink workbox in the software MATLAB, the differential equations of the mathematical model was solved directly, and exact results were obtained. It greatly simplified the solution process of equations.

Integrated Management of Operation and Maintenance Activities Based on Information Flow in Nuclear Power Plants

YE Zhi-qiang, XU Jia-shu

2005, 26(3): 284-287.

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The processes of operation, maintenance and management are the principal parts of the activities of one enterprise. Based on the Workflow Management Coalition workflows meta model, this paper has analyzed the attributes of the information flow in the information system, and introduced the application of information flow in the integrated management during the operation and maintenance of one nuclear power plant.

Project Management in Nuclear Equipment Manufacture

LIU Jian-cheng

2005, 26(3): 288-290.

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Project management for a nuclear power plant shall first consider the completion of the management organization. The organization of nuclear equipment quality assurance program and project management consists of 5 departments such as the nuclear power container department, the manufacture department and the quality assurance department. The general manager takes the overall responsibility for the quality of the nuclear press bearing equipment, and the vice general manager takes responsibility for the quality, technology and schedule related with the manufacture of the equipment, and organize the organization department for the audit. The director of the quality assurance department takes the responsibility for the establishment and completion of the quality assurance program, with enough rights authorized by the general manager, including the right not bounded by the cost and schedule, and confirms the implementation of the program by related departments and personnel. The manufacture schedule shall be prepared to ensure the implementation feasibility, process continuity and flexibility. The schedule shall be followed and monitored for the whole process, to check and feedback the implementation.

Statistics and Analysis of WANO Human Factor Events

ZHANG Li, ZHAO Ming

2005, 26(3): 291-296.

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940 WANO operating event reports from 1993 to 2002 were analyzed, among which 551 were found relative to human errors. Human errors are still one of the major causes of accidents in a nuclear power plant （NPP）. In the paper, the following conclusions are concluded: （1） human errors in test, maintenance and calibration leading to the potential failure of the system are the main causes of human factor incidents; （2） there is a comparatively high probability of human errors in the period of start-up and shutdown of a reactor which are liable to cause severe accidents; （3） the probability of human errors is not necessarily related to the type of a reactor ;（4） the main human errors include the error judgment on the symptoms and manipulation which most probably cause severe transient and the failure of safety system or improper shutdown; （5） lack of theoretical knowledge, poor manipulation, organization management failure, inappropriate procedures, carelessness and lack of check-up are the major root causes of human errors.

Study of Bumpless Transfer Based on AC800M Controller

FENG Jun-ting, ZHANG Liang-ju, DING Shu-ling

2005, 26(3): 297-300,304.

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This paper described a manual and automation transfer method for nuclear reactor power control system by digitalized technology design. Some key technologies of bumpless transfer based on standard digitalized controller are introduced. Bumpless transfer techniques are solved by software instead of analog automatic follow line.

Fire-Protection Zoning of Reactor Building for Daya Bay Nuclear Power Station

QIN Zhong

2005, 26(3): 301-304.

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Period safety review （PSR） has been conducted for Daya Bay nuclear power station according to the requirement of China Nuclear Regulatory Commission, and the fire hazard analysis is one important part of PSR. Based on the real status and characteristics of Daya Bay nuclear power station, this paper introduces thefire areas and zoning of reactor building, to provide the basic information for future safety review.

2005 Vol. 26, No. 3