Investigation on Modeling Thermal-Hydraulic System ...

Proceedings of the 2014 22nd International Conference on Nuclear Engineering ICONE22 July 7-11, 2014, Prague, Czech Republic

ICONE22-31096

Investigation on Modeling Thermal-hydraulic System of CPR1000 NPP Based on RELAP5 Cong Wang, Danmei Xie*, Peng Zhang, Xinggang Yu and Xiuqun Hou School of Power and Mechanical Engineering, Wuhan University , Wuhan, China Email: [email protected]

CPR1000 NPP adopts the improved Chinese pressurized water reactor technology and a representative demonstration project is Ling’ao phase II NPP with 1000MW capacity, which is designed and developed based on the construction and operation experience of Daya Bay and Ling’ao phase I NPPs. However, compared with the Daya Bay and Ling’ao phase I NPPs, CPR1000 NPP has many improvements at its design of conventional island thermodynamic system. Many of the changes and improvements are backfitted on the earlier unit following satisfactory development for the new unit. For example, the improvement substitutes half-speed turbine and electric feedwater pumps for full-speed turbine and turbine-driven feedwater pumps, etc[1,2]. Hence, it is necessary to build the complete thermal-hydraulic model with CPR1000 NPP features.

ABSTRACT Based on the best-estimate program RELAP5/MOD4.0, a fullscope thermal-hydraulic model with reference to CPR1000 nuclear power plant is established in this paper, which includes the thermal-hydraulic systems of conventional island as well as the primary nuclear island which has already been researched in traditional safety analysis. Therefore, this paper mainly details the numerical model of the turbine and other parts of the conventional island thermodynamic system. A comparison between the calculated results in steady-state and the actual data of reactor demonstrates a fine consistency, thus verifying the accuracy and reliability of the model. In addition, the steam parameter changes are numerically simulated during the steam turbine’s off-design operating condition such as back pressure variation and the variation trends are the same as the actual situation of nuclear power plants.

2. The RELAP5 model The RELAP5/Mod4.0 code contains one dimensional, nonhomogeneous and non-equilibrium model for two phase flow and a point model for reactor kinetics and uses finite difference method to solve the partial differential equations of mass, momentum and energy. In order to reduce complexity of PWR power plant thermal-hydraulic system modeling, the modular approach in software engineering is applied to the simulation, that is, decompose the complex system into several modules which function independently and can be separately designed and validated. As with this paper, each major component of nuclear and conventional island such as the reactor vessel, generator, pressurizer, coolant pump, turbine and secondary feedwater system is initialized respectively before incorporating the detailed conventional island thermal-hydraulic model into the full-scope NPP model. The nodalization of primary loop is shown in Figure 1.

1. Introduction Traditional safety analysis of nuclear power plant(NPP) mainly concentrates on establishing detailed models for the systems of primary loop while the system models of the secondary loop are always simplified and turbine components are always expressed with boundary conditions. With the development of nuclear power plant safety analysis methodology, the scope of its research gradually expanded to the entire nuclear power system which includes conventional island. Therefore, the establishment of a full-scope NPP thermal-hydraulic system model, especially a complete model of conventional island, is of great significance for safety analysis of NPP under various conditions.

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from 68520m3/h to 70116m3/h[3] and the main differences and improvements lie in secondary system, the Relap5 model presented below has a great detail in the development of the thermal-hydraulic systems of conventional island. During the full-scope thermal-hydraulic modeling process, the thermalhydraulic building blocks including pipe, junction, turbine, valve, pump, separator, time dependent volume along with heat structures, trips and control variables are employed in the model.

Steam generator

Pressurizer

3.

CPR1000 conventional island model

The conventional island system of CPR1000 NPP is composed by three steam generator secondaries, a turbo-generator, a two-stage moisture separator reheater, two condensers, two condensate pumps, four low pressure(LP heaters, a deaerator, two HP heaters and three electric feedwater pumps. The calculation models are described respectively as below.

Pump

3.1 Calculation model of turbine system The turbo-generator unit of CPR1000 adopts half-speed Arabelle 1000 type turbine, which is co-produced by France ALSTOM and Dongfang Steam Turbine Factory. This turbine is a quadrupleflow condensing steam turbine which consists of a one-flow integrated HP and IP cylinder, two double-flow LP cylinders and a moisture separator reheater (MSR). The flow chart of turbine system is shown in Figure 2.

Reactor vessel Fig.1 Nodalization of primary loop Since the primary component and system parameters of CPR1000 NPP are basically the same as those of Ling’ao Phase I NPP except that the designed mass flow of primary loop increases

Fig.2. Flow chart of turbine system A lumped-parameter turbine model is used in RELAP5 code where the stage group is then represented using modified energy, continuity, and momentum equations. An efficiency factor based upon simple momentum and energy considerations is used to represent the nonideal internal processes. Taking the HP cylinder model of CPR1000 NPP for example, the model is of four stages with the first stage an artificial turbine (i.e., constant efficiency, with efficiency=0) which is a condition for the RELAP code, so in actual there are three turbine stages (402,403,404) corresponding to three steam extractions[4,5]. Pressure of 61.1bar is maintained at the first artificial turbine, and pressure gradient is established between the following three stages. Moment of Inertia and the rotational speed of each stage is 138570.6 Kg.m2 and 157.08 radians/sec respectively. As suggested in the RELAP manuals the CCFL is on for the first

artificial turbine stage and for the remaining other stages it is off. Branches are added to the turbine system to extract a steam bleed for feedwater heating and moisture separation. Time dependent volume (TDV) is inbuilt to provide boundary conditions for the preliminary steady-state debug of turbine component. The calculation model of HP cylinder is shown in Figure 3. 401 402

403

404

Fig. 3. Nodalization of HP turbine

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the condenser through heat structures. Therefore, attention should be paid in the following aspects such as: the maintenance of the volume pressure, parameters setting between condensers and LP cylinders, the relatively high heat transfer coefficient and low steam resistance.

The momentum equation of turbine model is expressed as equation (1) with the nominal radius calculated as equation (2). 1 2 1 1−η V2 − V12 = (p2 − p1 ) (1) 2

2

R=

ρ V1

2ω(1−r)

(2)

Where 𝑉1 、 𝑉2 are velocity at the inlet and outlet of each turbine stage respectively; 𝑝1 、𝑝2 are pressure at the inlet and outlet respectively; 𝜌 is average density; 𝜂 is the actual efficiency; 𝜔 is the rotation speed；𝑟 is design reaction degree.

In the main steam system, the modeling difficulty mainly lies in the simulation of extraction system. The extract steam as the heat source of heaters and deaerator in the feedwater system connects the main steam system and feedwater system. The control of such parameters as temperature, pressure and mass flow rate, are of great importance. It is relatively difficult to control the flow accurately in modeling and its fluctuations always affect the quality of feedwater. Also, because of the sensitivity of the extraction pipe flow area, the accuracy of data is significant.

In order to correct the enthalpy error caused by drawbacks of old turbine model, the model in this paper adopts the following improvements: energy equation modified to include dissipation in turbine; momentum equation for turbine inlet junction changed from central difference to backward difference; moisture separator option added to turbine component and so on[6].

3.3 Calculation model of FW system

3.2 Calculation model of VVP system

The calculation model of feedwater system is shown in Fig.5. It is composed of low-pressure feedwater heater system, highpressure feedwater heater system, electric feedwater pumps (ABP&AHP&APP) and meanwhile coupled with drainage system and steam extraction system.

The calculation model of the main steam system is as shown in Fig.4. The major components include the model of steam header, the integrated HP and IP Cylinder, LP Cylinders, moisture separator reheater (MSR), condenser as well as the turbine bypass system and cooling water circulation system.

The HP heater #6 and #7 extract steam from the HP cylinder and drain water from the heater components of the MSR to heat the feedwater. The models of LP and HP heaters consist of three parts: feedwater side, steam side and heat structures. Each steam extraction of turbine cylinders is corresponding to the steam side of the LP and HP heaters.

The valve model includes the main steam valve(MSV) and control valve(CV) before HP cylinder and the regulating steam valve(RSV) and regulating control valve(RCV) before IP cylinder. The MSR model contains a separator component and two heater components. Phase transition from steam to water occurs inside

Fig.4. Calculation model of VVP system

Fig.5. Calculation model of FW system

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steady state analysis of nuclear island and especially conventional island with an important degree of detail. Main parameters of nuclear island and furthermore the pressure and extraction flow parameters of each turbine stage are calculated using this model with maximum error 1.09%, which demonstrates the accuracy and reliability of the model.

Calculating results and analysis

4.1 Calculation results of nuclear island Incorporate the detailed thermal-hydraulic model of conventional island and nuclear island together, and then main parameters are calculated at steady state shown in Table 1. A comparison between the calculated results and the actual data of reactor demonstrates a fine consistency, thus verifying the accuracy and reliability of the model.

In addition, the steam parameter changes are numerically simulated during the steam turbine’s off-design operating condition such as back pressure variation and the variation trends are the same as the actual situation of nuclear power plants.

4.2 Calculation results of turbine system Comparing the simulated pressure and extraction flow of each turbine stage with the plant data[8], the results are shown in Table2 and the maximum error is less than 1%. In addition, the steam parameter changes are numerically simulated during back pressure variation condition. In the further calculation when reducing the back pressure from 0.0056MPa to 0.0050MPa, the results show that the exhaust enthalpy and dryness change from 2330.98kJ/kg and 0.903 to 2315.05kJ/kg and 0.8985 respectively, as a result, the turbine efficiency is improved by 0.273%. Table 1 Main parameters of nuclear island parameters Plant Relap Nuclear power(MW) 2895.0 2895.0 Turbine power(MW) 1086.9 1098.7 Pressurizer pressure(MPa) 15.5 15.505 292.4 293.13 Cold leg temperature(℃) 327.6 328.97 Hot leg temperature(℃) Primary coolant flow(kg/s) 19476.7 19476.7 Steam flow form SG(kg/s) 1613.4 1617.32 steam pressure(MPa) 6.73 6.747 Feedwater flow to SG 1613.4 1617.32

When the backpressure of the half-speed turbine of CPR1000 NPP decreases from 0.0056MPa to 0.0050MPa, the turbine efficiency will increase by 0.273%, which verifies its advantage of better availability at different temperatures over the full-speed turbine of Ling’ao phase I NPP. Because the optimal backpressure of the Ling’ao phase I NPP turbine is 0.0075MPa, in the other domestic area where water temperature is lower, its efficiency would not be improved. REFERENCES [1] PU Jilong. The Formation of the Improved Chinese Pressurized Water Reactor Technology—CPR1000 [J]. China Engineering Science, 2008,10(3):54-57(In Chinese). [2] WU Jiakai. Thermal System Characteristic and Design Improvement for CPR1000 Nuclear Power Plant Conventional Island[J]. Northeast Electric Technology, 2012,12:39-41(In Chinese). [3] XIAO Min, HAO Sixiong，HAN Qinghao，LI Xianfeng, LIU Daohe. Primary Study on Core Concept Design and Safety Margin of CPR1000 [J]. Nuclear Power Engineering, 2005,26(6):11-18(In Chinese) [4] The RELAP5 Code Development Team. RELAP5/Mod3 Code Manual[R]. UREG/CR-5535-V2. Idaho National Engineering Laboratory, 1995. [5] Ashutosh Tiwari, Brij Mohan Sharma, Abhishek Srivastava, Rajesh Kumar, H.G. Lele, Prabhat Munshi. RELAP/SCDAP Code Analysis of PHWR LP Turbine[R]. Indian Institute of Technology Kanpur. [6] Davis C.B., Weaver W.L. Improvements to the RELAP5 Turbine Model[R]. 2003 IRUG Meeting INEEL, 2003. [7] C. Llopis , F. Revent´os, L. Batet, C. Pretel, I. Sol. Analysis of Low Load Transients for the Vandell’os-II NPP Application to Operation and Control Support[J]. Nuclear Engineering and Design 237 (2007) , 2014-2023. [8] Zhu Jizhou, Shan Jianqiang, Zhang Bin. Operation of pressurized water reactor nuclear power plant[M]. Automatic Energy Press ,2008.(In Chinese).

Error(%) 0.0 1.09 0.05 0.25 0.42 0.0 0.24 0.25 0.24

Table 2 Pressure and extraction flow of each turbine stage Pressure(bar) Extraction flow(kg/s) Plant Relap Plant Relap 1st bleed from HP 27.55 27.82 80.029 79.993 2nd bleed from HP 18.48 18.73 76.213 76.005 3rd bleed from HP 9.731 9.856 53.937 53.724 1st bleed from IP 5.721 5.813 44.075 44.196 2nd bleed from IP 3.129 3.151 68.052 68.293 1st bleed from LP 0.9611 0.9668 70.929 70.591 2nd bleed from LP 0.5421 0.5445 15.992 15.882 0.2415 0.2427 52.093 52.317 5.

Conclusions

A full-scope thermal-hydraulic model with reference to CPR1000 NPP is established in this paper, which is available in

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