核动力工程

Pressurized Thermal Shock Analysis for Reactor Pressure Vessels

ZHENG Bin, ZANG Fenggang, SUN Yingxue

2012, 33(1): 1-3,13.

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According to RCC-M rule and NRC 10CFR50.61,the PTS analysis was performed for the reactor pressure vessel beltline with assumed defect.The end of life reference temperature TR of nuclear power plant was calculated and assessed with PTS screening criteria.The stress intensity factors were calcu-lated under PTS transients and assessed with RCC-M Rule.

A Modifying Method of Recorded Earthquake Ground Motion for Matching Design Response Spectrum and Peak Ground Displacement in Nuclear Power Plants

SHENG Tao, SHI Weixing, XIE Yitong, YUAN Jun

2012, 33(1): 4-8.

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A modification method of recorded earthquake ground motion was proposed for matching the design pseudo velocity response spectrum and target peak ground displacement in nuclear power plants.The method applies the second cardinal B-spline wavelet to decompose the recorded real records into the finite wavelet components.Then the wavelet components which mostly affect the peak ground displacement and peak ground acceleration can be searched out.By improving the wavelet-based modifying method and wave-let synthesis method,the earthquake ground motions have been modified to match the target response spec-trum and target peak values of displacement and acceleration.The results of the numerical examples demon-strate that the proposed algorithms boast high precision.

Sensitivity Analysis of Seismic Behavior of Nuclear Power Plant Electrical Building

HE Jia, WANG Haitao

2012, 33(1): 9-13.

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The generation algorithm of the artificial ground motion is analyzed,the dynamic analysis model and method for the nuclear power plant seismic behavior under soil-structure interaction are explored,and the effect of variations of both the soil property and the load on the building responses is studied via comparison of the floor response spectra.The results show that compared with the load,the soil property con-tributes more to the seismic behavior of nuclear power plants,and that the difference of seismic behavior arising from the parameter variations can not be fully enveloped by the broadened floor response spectra cal-culated with one specific soil property.Therefore,it is necessary to envelop the floor response spectra with various soil properties.

Experimental Study on Added Mass of Sump Strainer in Nuclear Power Plants

LIU Xiajie, WANG Dezhong, ZHANG Chengang, XIANG Wenyuan, GU Weiguo

2012, 33(1): 14-18.

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This paper investigates the experimental method to measure the added mass of the contain-ment sump strainer in nuclear power plants using the inertia moment measurement.Four types of structures,i.e.,solid cylinder,thin-walled cylinder,double-walled cylinder and sump strainer,are tested in the experi-ment.The added mass of the solid cylinder tested in the experiment is compared with the theoretical value,and it is proved that the method in this study is with high accuracy in testing the added mass.The results ex-hibit that the added mass of thin-walled cylinder increases 1.9 times compared with the solid cylinder,and small difference appears in the values of added mass between single-walled and double-walled cylinders.The added mass of sump strainer decreases 70% compared with the thin-walled cylinder because of the opening in the cylinder surface.However,the added mass of sump strainer is similar with the mass in magnitude,so that the added mass can not be neglected when designing the structure

Numerical Model of Post-DNB Transition Boiling Heat Transfer

LI Hongbo, CHEN Bingde, XIONG Wanyu

2012, 33(1): 19-24,50.

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In this paper a physical model for the transition boiling heat transfer is proposed.The corre-sponding mathematical descriptions are given in detail and the heat transfer characteristics of post-DNB tran-sition boiling is analyzed.The numerical model of post-DNB transition boiling heat transfer is obtained as the empirical value of the coefficient is determined by the experimental data.The numerical model is compared with the experimental data of different parameters and other numerical models,and the statistical deviations are calculated.The calculating results of the numerical model in this paper show good agreement with the experimental data,and the numerical model in this paper is with good applicability compared with other nu-merical models.

Preliminary Study on Heating Rate Calculation for PWR Internals by Three-Dimensional S_N Method

YANG Shouhai, CHEN Yixue, WANG Weijin, JIN Zhongmin, LU Daogang

2012, 33(1): 25-28.

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Based on the American H.B.Robinson-2 pressure vessel benchmark,the neutron spectrum in surveillance capsule and reaction rates for some species have been obtain by three-dimensional（3-D） discrete ordinates（SN） method code（TORT）.The results show that,the neutron spectrum in surveillance capsule makes good agreement with the benchmark,and the C/M of reaction rates for some species is 1.04±0.04 by TORT code.Then,the heat release rate distribution over reactor components has been calculated by TORT code and MCNP program for Ningde Nuclear Power Plant in Fujian China,and compared with each other.Finally,the application of TORT code in the calculation of the heating rate for PWR internals is presented and discussed.

Numerical Simulation of Heat Transfer and Mixing Characteristics of Supercritical Pressure Water in Horizontal Tubes in Large Specific Heat Region

LEI Xianliang, LI Huixiong, YU Shuiqing

2012, 33(1): 29-33,38.

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Numerical simulation of the flow and heat transfer of supercritical pressure water in horizon-tal smooth tubes has been carried out.The diameter of the tube was Φ32 mm,and the length of the tube was 8000mm.Variation of heat transfer coefficients along the tube was obtained at pressures of 26MPa and 34MPa,respectively,and the evolution characteristics of the secondary flows in the tube were analyzed.The local parameters,such as the inner-wall temperature and the specific heat were adopted in the present study to reasonably explain the variation of the secondary flows of the supercritical pressure fluid in heated tubes.Acomparison of the natural convection to that of the force convection was made through dimensionless num-bers Gr/Re2 and Gr/Re2.7,and the characteristics of flow and heat transfer in horizontal tubes was well ex-plained.

Critical Heat Flux Tests in Rectangular Channel with Asymmetrical Heat Flux

XIE Shijie, XIONG Wanyu, WANG Jun

2012, 33(1): 34-38.

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Tests were performed in narrow rectangular gap channels to research the effect of asymmetri-cal heat flux on critical heat flux（CHF） using Freon-12 as the coolant.By testing four heated ratios in the research,we studied the effect of mass flux,inlet temperature and outlet quality on CHF.In addition,we studied the asymmetrical heat effect and found that the critical heat flux increases as the asymmetric heated ratio increases at low outlet quality,the effect is weakened when outlet quality increases,and it is the oppo-site that CHF changes with heated ratios at high outlet quality.

Experimental Study on Single-Phase Flow and Heat Transfer in A Narrow Rectangular Duct

MA Jian, HUANG Yanping, LIU Xiaozhong

2012, 33(1): 39-45.

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With deioned water as the test fluid,the experiments involved the single-phase isothermal flow and convective heat transfer with the boundary condition of constant heat flux were conducted in an upright narrow rectangular duct.The results show that the single-phase flow and heat transfer in the test duct did not deviate compared to the case in large-sized pipes,and can be well predicted by the conventional theoretical values and experimental correlations.

Visual Investigations of Temperature on Outer Wall of Narrow Rectangular Channel with Longitudinal Vortex

MA Jian, HUANG Yanping, HUANG Jun

2012, 33(1): 46-50.

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Four pairs of rectangular block as longitudinal vortex generators（LVG） are mounted periodi-cally in a narrow rectangular channel.The infrared thermal imaging method is used to measure the tempera-ture on the heating plate in this channel.The results show that LV which has certain effective distance and relative intensity induces quasi-periodic process of temperature descending and ascending alternately by means of reconstructing temperature field,and improves the whole convective heat transfer in the channel.The conclusions can be used to explain the heat transfer enhancement of LV for interrelated experimental in-vestigations

Study on Dynamic Lifting Characteristics of Control Rod Drive Mechanism

SHEN Xiaoyao

2012, 33(1): 51-55.

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Based on the equations of the electric circuit and the magnetic circuit and analysis of the dynamic lifting process for the control rod drive mechanism（CRDM）,coupled magnetic-electric-mechanical equations both for the static status and the dynamic status are derived.The analytical method is utilized to obtain the current and the time when the lift starts.The numerical simulation method of dynamic analysis recommended by ASME Code is utilized to simulate the dynamic lifting process of CRDM,and the dynamic features of the system with different design gaps are studied.Conclusions are drawn as:（1） the lifting-start time increases with the design gap,and the time for the lifting process is longer with larger gaps;（2） the lifting velocity increases with time;（3） the lifting acceleration increases with time,and with smaller gaps,the impact acceleration is larger.

Combination Test for PLC Embedded Software in Reactor Control and Protection System

SUN Kai, GONG Jianjun, KANG Jia, CHEN Peijun, DENG Jiangming, FENG Liang, LI Yuanwen, ZHU Xiaoyong, PU Yuruo

2012, 33(1): 56-59,77.

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The manual static test method and the dynamic simulation test method are combined to test the embedded software in the reactor control and protection system,which is called combination test for short.There are two steps for the combination test.Firstly,the manual static test method is used to verify the safety and rationality of PLC（programmable logic controller）,and then the dynamic simulation test method is used to verify its functional effectiveness.This combination test can be conducted with PC and the development and simulation tools for the software to be tested.The effectiveness of the combination test is verified by the hardware test method in the actual PLC operation environment.The result shows that the combination test is effective and is more efficient than the hardware test method.

Dynamic Analysis of Diesel Generator System after Loss of Offsite Power Event Occurring in Nuclear Power Plants

LI Zhe

2012, 33(1): 60-65.

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Based on the GO methodology and Markov method,the dynamic analysis of emergency die-sel generator system for protecting the nuclear power plant from Station Blackout,which is caused by Loss of Offsite Power event,is made with duration of 24 hours.In addition,the accurate reliability calculation problem is solved for the repairable system with dependant maintenance relation,and the logic relation of emergency response system is fully simulated by creating the "Backup Operator" of the GO methodology.By combining the two reliability analysis methods,which is used suitably for the emergency response system of diesel generators with dependant maintenance relation,the application range for the two methods is expanded,and the effect of station blackout event on the safety operation of nuclear power plants can be obtained more accurately.

Research on Cognitive Reliability Model for Main Control Room Considering Human Factors in Nuclear Power Plants

JIANG Jianjun, ZHANG Li, WANG Yiqun, PENG Yuyuan, ZHOU Cheng, ZHANG Kun

2012, 33(1): 66-73.

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Facing the shortcomings of the traditional cognitive factors and cognitive model,this paper presents a Bayesian networks cognitive reliability model by taking the main control room as a reference back-ground and human factors as the key points.The model mainly analyzes the cognitive reliability affected by the human factors,and for the cognitive node and influence factors corresponding to cognitive node,a series of methods and function formulas to compute the node cognitive reliability is proposed.The model and cor-responding methods can be applied to the evaluation of cognitive process for the nuclear power plant opera-tors and have a certain significance for the prevention of safety accidents in nuclear power plants.

Microstructural Analysis of Irradiated UO₂ Dispersed Fuel during High Temperature Failure

WU Xiaoyong, WANG Fei, WEN Bang

2012, 33(1): 74-77.

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The blister annealing has been performed on UO₂ dispersed fuel plate under various tempera-ture conditions.The microstructure of UO₂ particles changing with the rise of temperature is studied by using the microscope.Fission gas releases at high temperature and the bubbles burst the coat to generate cracks and result in the local fracture and fall off of UO₂ particles.A lot of holes occur in the reaction layer outside of UO₂ coat because of the diffusion reaction intensified by the high temperature.

Thermal Strain of Fuel Rods during Transient of Operation Transformation

HAO Yalei, YANG Xingtuan, JIANG Shengyao, LIU Zhiyong

2012, 33(1): 78-82.

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The analytical model was created using the RELAP5 code,and the rate of coolant temperature alteration and thermal strain of fuel rods during the transient from forced circulation to natural circulation were calculated and analyzed.The results show that,RELAP5 code is a proper tool to analyze the thermal strain of fuel rods during the transient from force circulation to natural circulation;the dramatic changes of thermal strain have effects on the mechanical properties of fuel rods;and the thermal stress of fuel rods dur-ing other transient can also be analyzed using RELAP5 code.

Corrosion Properties of Sealing Surface Material for RPV under Abnormal Working Conditions

LIU Jinhua, WEN Yan, ZHANG Xuemei, HUO Songmin, GONG Bin, HE Yanchun

2012, 33(1): 83-87.

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Based on the corrosion issue of sealing surface material for RPV in some nuclear projects,the corrosion properties of sealing surface material for RPV under abnormal working conditions were investigated.The corrosion behavior of 308L stainless steel were studied by using autoclave in different contents of Cl-solutions,and these samples were observed and analyzed by means of the metalloscope and Scanning electron microscope（SEM）.Results show that no pitting,crevice and stress corrosion occurred,when the content of Cl-was lower than 1 mg/L at the temperatures of 270℃ and the pressure of 5.5 MPa.However,with the increase of the content of Cl-,the susceptibility to pitting,crevice and stress corrosion of 308L was enhanced remarkably.

Investigation of Oxide Scale on 304NG Exposed in Supercritical Water

YIN Kaiju, QIU Shaoyu, TANG Rui, ZHANG Qiang, ZHANG Lefu

2012, 33(1): 88-92.

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The corrosion behavior of an austenitic stainless steel 304NG exposed to supercritical water（SCW） at 500℃,550℃,600℃ and 650℃/25 MPa is investigated by means of scanning electron micro-scope/energy dispersive X-ray spectroscopy and X-ray diffraction.The surface morphology is composed of oxide islands and relative smooth base oxide.Oxide islands is a dual-layered oxide scale,which is mainly composed of an outer magnetite（Fe₃O₄） layer and an inner magnetite/spinel-mixed（Fe₃O₄/FeCr₂O₄） layer,base oxide is a single-layered magnetite/spinel-mixed（Fe₃O₄/FeCr₂O₄） oxide scale.The oxide scale has a tendency to spall,which increases with the increasing of exposure temperature.

Effect of Nb and Cu on α/β Phase Transition Temperature and Corrosion Behavior of New Type Zirconium Alloys

LUO Wei, WANG Jun, XIONG Ji, YING Shihao, YANG Xiaoxue, ZHANG Taiping

2012, 33(1): 93-97.

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The microstructure and the performance of corrosion resistance of Zr-Nb-Cu alloys have been investigated in the superheated steam of 500℃/10.3MPa.The results showed that the alloy Zr-1.0Nb-0.05Cu possessed the best corrosion resistance,which was far superior to that of Zr-4 and N18.The second phase par-ticles which contain Nb,Fe and Cr formed in Zr-Nb-Cu alloys,which was a factor that affected the corrosion resistance of Zr alloys.During heating up of Zr-Nb-Cu alloys in the DSC experiment,the corrosion induced hydride dissolved and hydrogen solutionizing in the matrix.When the temperature reached to α/β phase tran-sition temperature（about 550℃）,β phase transformation occurred.The addition of Nb element can decrease the temperature of hydrogen-induced β phase transformation,and the increasing of Cu content can reduce the hydrogen absorbing amount,and improve the corrosion resistance of the alloy significantly.

Experimental Research on Simulated Radioactive Wastewater Treatment by Ultra-Filtration

KONG Jingsong, WANG Xiaowei

2012, 33(1): 98-100.

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In order to reveal the performance of the ultra-filtration adopted in the radioactive wastewater treatment,experiments were carried out on a pilot UF equipment using the wastewater containing the simulated nuclides.Results showed that the nuclides could be rejected by UF efficiently and the rejection ratio is influenced by pH,with different effects on the rejection of Fe³⁺,Co²⁺,Mn²⁺ and Ni²⁺.Theoretical calculation was performed to explain the phenomenon and rejection mechanism was interpreted.

Research on Rejection Performance of Reverse Osmosis to Cobalt in Simulated Radioactive Wastewater

KONG Jingsong, TIAN Yanjie

2012, 33(1): 101-103,142.

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In order to reveal the rejection performance of the reverse osmosis applied in the radioactive wastewater treatment,treatment experiments were carried out on a pilot reverse osmosis equipment using wastewater containing cobalt nuclide.Results showed that the rejection ratio of reverse osmosis to cobalt was almost not affected by the operation pressure and the ratio of reclaiming,and had no direct relation with the salt rejection ratio.The ratio of cobalt rejection reached 90% in the experiment condition and could meet the requirement on the disposal of radioactive wastewater produced by nuclear powered installations.

Optimization Problem of Radiation Protection for Site Work based on Genetic Algorithm

ZHOU Xiuling, CHEN Baowei, AN Yongfeng, GUO Ping

2012, 33(1): 104-106,123.

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In this paper,Genetic Algorithm is applied for optimizing the temporary shielding plan of pipes when a valve needs to be repaired in the field of radiation protection.Firstly,the mathematical model of the radiation dose for shielding of two pipes is established,and then the model is converted to a constrained combinational optimization problem.The Genetic Algorithm is used to solve this optimization problem.The experimental results show that the Genetic Algorithm can give an accurate complete radiation protection scheme.

Study on Refueling Pattern and Its Application in Tianwan Nuclear Power Plant

YAN Jing, WANG Wei

2012, 33(1): 107-111,133.

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Refueling for the reactor core is a top priority for a nuclear power plant unit in the course of yearly refueling overhaul.Taking the fuel loading diagram for the 3rd cycle of No.1 and No.2 unit of Tianwan NPP as an example,this paper introduces the operation process for three refueling methods,i.e.,the all-in out,the shifting within core and chainre fueling.Time efficiency for these three refueling methods is estimated,and the advantages and applicable scenarios for these three refueling methods are analyzed and compared.The actual refueling operation for Tianwan NPP verified that the latter two refueling methods are feasible,and can greatly improve the efficiency of refueling.

Analysis of Boron Concentration Deviation and ¹⁰B Abundance Evolution in Primary Loop of Pressurized Nuclear Plants

TAN Shijie, LIAN Zhikun, LI Zhijun, ZHOU Xiaoling, HU Ruping, GAO Jinghui

2012, 33(1): 112-116.

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The ¹⁰B abundance evolution under two conditions,i.e.,with and without boronizing,is calculated and analyzed with theoretic derivation method,and the evolution pattern of ¹⁰B abundance for one cycle in PWR is provided.The comparison of the calculated and measured ¹⁰B abundance shows that the abundance equation considering the boronizing is accurate.With this,the theoretical boron concentration provided by the fuel management software can be corrected and validated.According to the equation and analysis method,the boron concentration deviation problem could be well understood or even solved.

Multi-State Reliability for Pump Group in Nuclear Power System Based on UGF and Semi-Markov Process

SHANG Yanlong, CAI Qi, ZHAO Xinwen, CHEN Ling

2012, 33(1): 117-123.

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In this paper,multi-state reliability value of pump group in nuclear power system is obtained by the combination method of the universal generating function（UGF） and Semi-Markov process.UGF arithmetic model of multi-state system reliability is studied,and the performance state probability expression of multi-state component is derived using semi-Markov theory.A quantificational model is defined to express the performance rate of the system and component.Different availability results by multi-state and binary state analysis method are compared under the condition whether the performance rate can satisfy the de-manded value,and the mean value of system instantaneous output performance is also obtained.It shows that this combination method is an effective and feasible one which can quantify the effect of the partial failure on the system reliability,and the result of multi-state system reliability by this method deduces the modesty of the reliability value obtained by binary reliability analysis method.

Experimental Study on Effects of Double Pumps Switching on Water Supply Flow Rate

WANG Xin, HAN Weishi

2012, 33(1): 124-128.

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Flow characteristics in the process of switching one centrifugal pump to the other was investi-gated experimentally using a closed loop with two centrifugal pumps and two check valves.Characteristics of the check valves responding and the flow rate changing during the process of switching was studied by ex-perimental data analysis.The results show that in the switching process with high and low original flow rate,the restoring time is 26s and 21s respectively;the lowest flow rates are 59.4% and 87.2% out of that in normal water supply,and the average deficit of feed water is 20.8% and 7.5% respectively.Compared to double-pump switching with low flow rate,a longer transition time,more intense flow fluctuations and increased water loss are observed with high flow rate,which has significantly effects on the stability of water supply.

Neutronics Calculation of Fusion-Fission Hybrid Breeding Thorium Fuel Assembly

MA Xubo, CHEN Yixue, QUAN Guoping, WANG Yue, HAN Jingru, LU Daogang

2012, 33(1): 129-133.

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A preliminary comparative study of the physical properties among 17×17 fuel assembly in PWRs for prototype between uranium assembly and hybrid breeding thorium-based assembly has been investigated respectively using the DRAGON software.The parameters such as fuel temperature coefficient,moderator temperature coefficient and that variation as a function of operation period have been investigated.Results show that the neutron properties of uranium-based assembly and hybrid breeding thorium-based assembly are similitude,but MA mass of hybrid breeding thorium-based assembly is evidently less than those of the uranium assembly.

Economic Evaluation of Fast Reactor Fuel Cycling

HU Ping, ZHAO Fuyu, YAN Zhou, LI Chong

2012, 33(1): 134-137.

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Economic calculation and analysis of two kinds of nuclear fuel cycle are conducted by check off method,based on the nuclear fuel cycling process and model for fast reactor power plant,and comparison is carried out for the economy of fast reactor fuel cycle and PWR once-through fuel cycle.Calculated based on the current price level,the economy of PWR one-through fuel cycle is better than that of the fast reactor fuel cycle.However,in the long term considering the rising of the natural uranium’s price and the development of the post treatment technology for nuclear fuels,the cost of the fast reactor fuel cycle is expected to match or lower than that of the PWR once-through fuel cycle.

Study on Graphical Modeling and Simulation of Secondary Loop in HTR-PM

DONG Liyu, ZHOU Zhiwei, ZHOU Yangping, SUI Zhe, ZHOU Shuyong, LI Fu

2012, 33(1): 138-142.

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In this paper,the development of a compact engineering simulator of the HTR-PM by em-bedding THERMIX code into the vPower simulation environment is introduced firstly.Then the structure of the simulation system of secondary loop is presented in detail,based on the vPower simulation platform.Steady state and transient conditions of the HTR-PM have been studied by the help of this simulation system.The results show that the errors of the steady-state value and design value are within 2%.At the rated operating conditions,the typical transient response curve of the main valve down-step and feedwater flow reducing have rational trend,and the indicators and the final stable value of the dynamic response meet the requirements.The condenser circulating water disrupts for the closing of the circulating water pump causes low condenser vacuum,and the regulating valve quickly closes,the main steam flow decreases and the turbine trips.

2012 Vol. 33, No. 1