research article full scope modeling and analysis on the...

Research ArticleFull Scope Modeling and Analysis on the Secondary Circuit ofChinese Large-Capacity Advanced PWR Based on RELAP5 Code

Dao-gang Lu,1,2 Fan Zhang,1,2 Dan-ting Sui,1,2 Xue-zhang Xi,3 and Lei-bo Yu3

1School of Nuclear Science and Engineering, North China Electric Power University, Beijing 102206, China2Beijing Key Laboratory of Passive Nuclear Safety Technology, North China Electric Power University, Beijing 102206, China3State Nuclear Electric Power Planning Design & Research Institute, Beijing 100095, China

Correspondence should be addressed to Fan Zhang; [email protected]

Received 2 March 2015; Revised 3 June 2015; Accepted 4 June 2015

Academic Editor: Iztok Tiselj

Copyright © 2015 Dao-gang Lu et al. This is an open access article distributed under the Creative Commons Attribution License,which permits unrestricted use, distribution, and reproduction in any medium, provided the original work is properly cited.

Chinese large-capacity advanced PWR under construction in China is a new and indispensable reactor type in the developingprocess of NPP fields. At the same time of NPP construction, accident sequences prediction and operators training are in progress.Since there are some possible events such as feedwater pumps trip in secondary circuit may lead to severe accident in NPP, trainingsimulators and engineering simulators of CI are necessary. And, with an increasing proportion of nuclear power in China, NPPwill participate in regulating peak load in power network, which requires accuracy calculation and control of secondary circuit. Inorder to achieve real-time and full scope simulation in the power change transient and accident scenarios, RELAP5/MOD 3.4 codehas been adopted to model the secondary circuit for its advantage of high calculation accuracy. This paper describes the model ofsteady state and turbine load transient from 100% to 40% of secondary circuit using RELAP5 and provides a reasonable equivalentmethod to solve the calculation divergence problem caused by dramatic two-phase condition change while guaranteeing the heattransfer efficiency. The validation of the parameters shows that all the errors between the calculation values and design values arereasonable and acceptable.

1. Introduction

Speeding up the development of nuclear power is an inev-itable way to solve the energy crisis and environmental prob-lems caused by pollution of fossil energy. Since the large-capacity advanced PWR is a new and indispensable reactortype in the developing process of nuclear power plant fields,the enthusiasm of research on it increases gradually. Chineselarge-scale advanced PWR which is under construction inChina is the research subject in this paper. On the onehand, with an increasing proportion of nuclear power in theelectric power in China, nuclear power plant (NPP) will par-ticipate in regulating peak load in power network by loadand frequency adjustment; on the other hand, there are somepossible events such as steam turbine failure and feedwaterpumps or other pumps trip in secondary circuit which maylead to severe accident inNPP. Taking these two requirementsinto account, training simulators and engineering simulatorsof conventional island (CI) are necessary.

The proportion of transient analysis codes for conven-tional island of nuclear power plant system is small. SASSYSdeveloped by Argonne National Laboratory is a general pur-pose one-dimensional thermal-hydraulic liquid-metal reac-tor (LMR) systems analysis code [1]. It is validated using thedata of Experimental Breeder Reactor number 2 (EBR-II).SASSYS has some typical models for transient analysis, suchas core thermal hydraulic model and heat transfer betweenfuel component models. The adding of Balance-of-Plantmodel to SASSYSmakes it capable of achieving transient sim-ulation for thermal hydraulic system in conventional island.Some accident scenarios have been studied using SASSYSsuch as closure of the turbine admission valve. Another codecalled NATDEMO developed by Argonne National Labora-tory is special for EBR-II, including primary, feedwater, andsteam systems [2]. Another code named DSNP provides afull scope simulation of power plant based on componentor system model in libraries, which are continually updatedwith new, improved, and verified modules [3]. Verification of

Hindawi Publishing CorporationScience and Technology of Nuclear InstallationsVolume 2015, Article ID 913274, 7 pageshttp://dx.doi.org/10.1155/2015/913274

2 Science and Technology of Nuclear Installations

SGSG

HP

MSR

MSR

IP LP LPLP

Condenser A Condenser B Condenser C

Secondreheat

First reheat

7A

7B6B6A

4A

4B

3A 3B

1A 2A 1B 2B 1C 2C

6A

6B

7A

7B

Main water and steam trainExtraction

DRT

DRT

1#2#3#4#Sealing heater

Regulatingvalve

Dea

erat

or

Boosterpumps

6#7#

A CBCondensate pumps

1A Feedwater heaterDrain water

Low pressure feedwater heaters

Feedwaterpumps

High pressurefeedwater heaters

Figure 1: Flow sheet of feedwater and steam system of Chinese large-scale advanced PWR.

DSNP has been done by comparing the simulation programscalculation with EBR-II experiment data. The modules areused to simulate the dynamic response of reactor systems’transient conditions and BOP system perturbations.

Sharma et al. simulate BOP system of Prototype FastBreeder Reactor (PFBR) based on RELAP5/MOD 3.4 code,and some of the important events possible in the steam watersystem have been analyzed using this code to verify the effect-iveness of the procedure adopted, such as one CEP trip andstandby not starting and one condenser cooling water pump(CCWP) trip [4]. Yang et al. developed a dual RELAP5-3D-based engineering simulator for Lungmen advanced boilingwater reactor (ABWR), which has two separate RELAP5-3D modules for reactor system and BOP system individuallysynchronized on a platform [5].

Since the accuracy, authenticity, and reliability of eachpart’s model in system related to the effect of the whole plantsimulation directly, RELAP5/MOD 3.4 code is used as themodel tool in this study [6]. This paper describes the model

of steady state and turbine load transient from 100% to 40%ofCI. This model provides an easy interface with nuclear islandsystem and a basic simulation for future accident analysis.

2. Secondary Circuit Model Description

Chinese large-scale advanced PWR is a two-circuit reactor[7]. As shown in Figure 1, thermal hydraulic system in con-ventional island consists of steam generator (SG) secondaryside, high pressure turbine (HPT), two moisture separatorreheaters (MSRs), intermediate pressure turbine (IPT), threelow pressure turbines (LPTs), three condensers, three con-densate extraction pumps (CEP), ten low pressure heatersin four stages (three number 1 low pressure heaters (LPH1),three LPH2, two LPH3, and two LPH4), deaerator, three feed-water pumps (FP) with dedicated booster pump (BP), fourhigh pressure heaters in two stages (two number 6 high pres-sure heaters (HPH6) and two HPH7), and correspondingvalves and pipes.

Science and Technology of Nuclear Installations 3

026028

050

030032

034 036 031038

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MSR A

IPT

LPT A

Condenser B

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1 6 7

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LPT B

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Condenser C Condenser A

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t

b

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b

086b

snglvol snglvol

t t

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T: TurbineB: Branch

P: Pipe

t t

t t

t t

b b

Separator Separator

093sv

091tv

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342 343t 344

364 366 368

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360 362

380 382374 376

394 396 400 402 404 406 408

413

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416b

b

b

Figure 2: Nodal map of thermal-hydraulic system in conventional island (part I).

TheRELAP5model of secondary circuit of Chinese large-scale advancedPWRhas been built based on the practical sys-tem layout of NPP design shown in Figure 1. Figures 2-3 showthe nodal map of thermal-hydraulic system in CI consistingof turbines, MSRs, condenser, condensate water pumps,LPH, deaerator, feedwater pumps, booster pumps, HPH, andcorresponding valves and pipes. In traditional simulation

of CI, parallel running equipment is simplified as one withequivalent function. Different from it, the amount of equip-ment is coincided with actual layout of NPP in this model,which is convenient for simulating disconnection opera-tion situation of certain equipment.

The turbine is the equipment where the internal energyof high temperature and pressure steam transforms into


452

470

460

494B

496

516

536

500P

520P

540P

498

518

538

900P 902898

830

850

840

860

880

870742B

CWP

Low pressure feedwater heater

No. 3 Deaerator

Feedwater pump

No. 7

478482P486490 488 480492524

544

522

542

504 502899 901

906P 908904 905 907

912P 914910 911 913

Regulating

valve

576

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600P

620P

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918P 920916

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584 582917 919

924P 926922 923 925

930P 932928 929 931

634B 630B632

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946P 948942

654 652

644 642943 947

936P 938934 935 937

666

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952P 954950

684 682

674 672951 953

958P 960956 957 959

944 945

940 941

730P

726B

688

700 708716

From MSR

From 6#

From HPTB

702

710

720

686

689

776

786

780P

790P

778

788794 792

784 782 638

986P 988984 985 987982 983

994P 996992 993 995990 991

834

844

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890

503 501

523 521

543 541

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Sealing heater

01

10

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743741

Standby

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447P449450B 445

882744 883

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664665

No. 6

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750P

760P

748

758

968P 970962

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752963 969

978P 980972 973 979

964 965

974 975

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976 977

770

772

961 971

961 971

754P

No. 4

No. 1No. 2

High pressure feedwater heaterbooster pump

SG inlet

From condenser

Figure 3: Nodal map of thermal-hydraulic system in conventional island (part II).

mechanical energy [8]. In RELAP5, the modified energy,momentum, and continuity equations have been used tocalculate the stages of turbine. An effective factor should beset based on the simple consideration of momentum andenergy to represent the ideal inner energy conversion process.In order to simulate the steam extraction in turbine here,the several stages are modeled by typical nuzzle-shape areainstead of lumped-parameter model. The number of stagescorresponds with the extraction train. A reasonable simplifi-cation is used in the model of extraction trains from turbineto heaters and deaerator; that is, instead of real connectionwith heater shell side, time dependent junction and timedependent volume are used to simulate the boundary ofextraction. As shown in Figure 2, there are one HPT, one IPT,and three LPTs modeled precisely. HPT is modeled by threestages (026, 032, and 038), IPT is modeled by two stages (220and 226), and each LPT is modeled by four stages: 300, 306,310, and 313 for LPT A, 315, 321, 325, and 328 for LPT B, 330,336, 340, and 343 for LPT C.

To prevent corrosion, MSRs pull moisture from theexhaust of turbines and reheat [9]. According to the operatingprinciple of the model of each MSR which consists of threeparts, two multiple steam separators are modeled using sepa-rator component, namely, 102 and 161; the first level reheatersare modeled by two pipes with a heat structure, whose

shell sides are extraction steam boundaries, and second levelreheaters’ shell sides are live steam boundaries.

As shown in Figure 3, there are three condensate pumpsin total, two of which are running normally while the thirdone is standby. When one of the running pumps trips, thestandby pump comes into working. Condensate pumps aremodeled by pump components, namely, 452, 460, and 470(standby). There are three feedwater pumps (830, 840, and850, resp.) and three booster pumps (860, 870, and 880,resp.). For simulating these nine pumps exactly, not onlyare geometrical parameters and operation-related parametersset, but also characteristic curves (head and torque curves)of each pump are modeled by homologous curve data. Thereare eight operation octants for each pump, all data are setto completely describe the single-phase pump operation, anddata of two normal regimes are given particular concern.

Feedwater heaters are used to preheat water deliveredto steam generators. Preheating the feedwater reduces theirreversibility involved in steam generation and thereforeimproves the thermodynamic efficiency of the system [10].There are six levels of feedwater heaters in Chinese large-capacity advanced PWR.The temperature of feedwater ariseswhen it flows through every level of heaters and reaches tothe saturation temperature eventually. All of the heaters aremodeled by two heat transfer pipes with a heat structure


Table 1: Validation of steady-state parameters.

Parameter Calculation value Nominal value ErrorTurbine inlet pressure 5.95MPa 5.95MPa 0.0Deaerator pressure 1.027MPa 1.028MPa 0.0Condenser outlet temperature 300.3 K 300.55 K 0.08%LPH outlet temperature 433.07 K 432.45 K 0.14%Deaerator outlet temperature 453.33 K 454.25 K 0.20%HPH outlet temperature 495.46K 498.85 K 0.68%Turbine inlet mass flow 2249.62 K 2249.62 K 0.0Condensate pump mass flow 663.7 kg/s per pump 663.7 kg/s per pump 0.0Feedwater pump mass flow 747.1 kg/s per pump 749.54 kg/s per pump 0.33%Condensate pump speed 1480 rpm 1480 rpm 0.0Booster pump speed 1496 rpm 1496 rpm 0.0Feedwater pump speed 5122 rpm 5122 rpm 0.0

Table 2: Time sheet for power change in transient analysis.

Time (s) 0 300 400 700 800 1100 1200 1500 1600 1900 2000 2300 2400 2700Power 100% 100% 90% 90% 80% 80% 70% 70% 60% 60% 50% 50% 40% 40%

using the material of stainless steel. Extraction steam fromturbines in shell side of heater is modeled by time dependentjunctions and time dependent volumes, the parameters ofwhich are controlled the same as the turbine extractionusing controller component in RELAP. Isolate valves aremodeled in each subcircuit to provide the ability of simulatingthe disconnection operation situation (isolate valves are notpresented in the nod map for concision). Shell sides offeedwater heaters contain two-phase flow and heat transfer,especially in high pressure feedwater heater. Various sourcesof HPH’s inlet boundary in shell side, including extractionsteam from turbine and water forMSR and next level heaters’drain water, contribute to rapid and dramatic change of two-phase flow, leading to calculation divergence. To solve thisproblem, an equivalentmethod is used inHPH’s heat transfermodel. All inlet boundary conditions of HPH’s shell sides areset to saturated water; that is, void fraction is zero, and aninternal heat source with heat power equaling the enthalpydifference between the actual condition of shell sides’ heatsource and the saturated water boundary we set is addedto the water sides. This method guarantees the heat transfereffect in HPH.

Dissolved gases in feedwater may lead to severe corrosiondamage in thermal-hydraulic system; thus, deaerator is usedto avoid it in all power plants [11]. The spray-type deaeratoradopted in this system consists of a horizontal cylindricalvessel, which serves as both the deaeration section and thefeedwater storage tank. The pressure of deaerator is requiredto be maintained by extraction steam from turbine to ensurethe feedwater pressure is enough to meet the need of netpositive suction head at pump inlet. Main body of deaeratoris modeled by a combination of a ten-volume pipe (730)and a branch (726). Time dependent volumes represent thepreheating source fromMSR, number 6HPH, andHPT (700,708, and 716, resp.), which will mix up with condensed waterin branch and then flow to pipe 730. Time dependent volumes690 and 743 are set to keep pressure of deaerator, and time

dependent junction 741 is set to govern the water level byproportional-integral controller.The dramatic change in voidfraction of two-phase mixture flow at top of deaerator leadsto preheating inadequacy at the exit. This is solved by using aheat structure with internal heat source adding to five formervolumes of pipe 730. The valves of thermal-hydraulic systemin conventional island are all modeled according to theirfunction and operation logic.

Boundary data for various operation conditions has beenset in input data deck, more specifically the turbine loadfor 100%, 90%, 80%, 70%, 60%, 50%, and 40%. In additionto achievement of required steady state for 100% powercondition, transient simulation of turbine load change from100% full power to 40% low power step by step has also beenmodeled.

3. Validation and Discussion

Steady states for turbine load from 100% to 40% individuallyare performed by running 5000 s without perturbation. Allcalculation values are compared with nominal design values,and the errors between them are less than 2%. Since keyparameters’ errors are not more than 1%, the various powerconditions are modeled correctly and values of 100% powersteady state are tabulated in Table 1.

Transient analysis is achieved by performing changepower by time as shown in Table 2, running 2700 s in total.

The validation of transient calculation has been carriedout, and all the calculation values of steady state of turbineload for 90%, 80%, 70%, 60%, 50%, and 40% are comparedwith the design values; the errors between them are less than10%, and the major parameters’ errors are less than 5%. Someof the typical parameters’ values and trends of transient runare illustrated in Figures 4–6. Figure 4 shows transient changeof mass flow for turbine inlet, booster pumps inlet, and singleSG inlet, and the parameters’ values decline stage by stage


0 2700240021001800150012009006003000

500

1000

1500

2000

2500

Mas

s flow

(kg/

s)

Time (s)

Turbine inletBooster pumps inlet

Single SG inlet

Figure 4: Transient change of mass flow for turbine inlet, boosterpump inlet, and single SG inlet.

0 300 600 900 1200 1500 1800 2100 2400 2700

400

420

440

460

480

500

520

540

560

Turbine inletDeaerator outlet

LPH outlet

Tem

pera

ture

(K)

Time (s)

Figure 5: Transient change of temperature for turbine inlet, LPHoutlet, and deaerator outlet.

as power decreases. Since the pump component is sensitiveto the working point, the pressure surge caused by densitydifference and fluid accumulation in deaerator leads somesmall fluctuation for booster pumps inlet mass flow. Due tothe tolerance of this model and main feedwater regulatingvalve, the fluctuation does not pass to the SG inlet mass flow.The mass flow at booster pump inlet is twice as much asit at single steam generator inlet, which is reasonable. Themass flow at turbine inlet is a little less than flow at boosterpump inlet (equaling SG inlet total mass flow).The difference

0 300 600 900 1200 1500 1800 2100 2400 27000

2

4

6

8

10

SG inletFeedwater pump outlet

Deaerator

Pres

sure

(MPa

)

Time (s)

Figure 6: Transient change of pressure for SG inlet, deaerator, andfeedwater pump outlet.

between them is exactly live steam extraction before steamfrom SG to HPT, which is delivered to MSR for reheating.

The heat transfer power between first and secondary sideof SG is described by (1), where𝐾 is heat transfer coefficient,which is seen as a constant, and 𝐴 is heat transfer area.𝑇av is average temperature of coolant in primary loop, 𝑇

𝑠

is the saturate temperature of secondary side of SG. Theheat transfer power is proportional to difference between 𝑇avand 𝑇

𝑠. Concerning both thermodynamic cycle efficiency in

second loop and reactivity adjustment requirement caused bythe change of coolant volume, the design NPP adopted is thatreactor inlet temperature maintained constant while coolantaverage temperature declines within a certain acceptablerange as power reduces, and steam temperature increases aspower reduces. Consider

𝑃 = 𝐾𝐴 (𝑇av −𝑇𝑠) . (1)

Figure 5 shows transient change of temperature for tur-bine inlet, low pressure feedwater heaters outlet, and deaer-ator outlet. The temperature in turbine inlet increases grad-ually as load reduces, which is consistent with the design ofNPP. Since the turbine extraction to heaters decreases as thepower decreases, the water temperature at each level heatersand deaerator exit also declines.

Figure 6 shows transient change of pressure for deaerator,feedwater pump outlet, and SG inlet. The deaerator pressureis controlled the same as the extraction steam, which declinesas power drops. The head of pump is inversely proportionalto the flow, so the pressure at the outlet of pumps increases aspower decreases.

Since the errors of calculation results of steady state arein an acceptable range and turbine load reduction transienttrend is reasonable, the model of conventional island iscorrect.


4. Conclusions

A full scope modeling for secondary circuit in conventionalisland of Chinese large-scale advanced PWR has been devel-oped based on RELAP5. Steady state for turbine load from100% to 40% individually is achieved by this model, andthe validation of the parameters shows that all the errorsbetween the calculation values of steady state and transientand practical design values are less than 10%, and the majorparameters’ errors are less than 2%, which is acceptable.

This model is developed according to actual plant layoutwithout simplification in the amount of equipment, which isconvenient for simulating disconnection operation situationof some equipment. The equivalent internal heat sourcemethod is used both in HPFHs’ noncontact heat exchangemodel and deaerator’s mixing heat transfer model to solvethe calculation divergence problem caused by dramatic two-phase condition change while guaranteeing the heat transferefficiency. The transient model of turbine load from 100%to 40% provides an easy way to observe the parameters’change in different conditions. Further study will be focusedon optimization of the model, peak load simulation, andaccident scenario based on this model.

Conflict of Interests

The authors declare that there is no conflict of interestsregarding the publication of this paper.

References

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[2] H. P. Planchon, R. M. Singer, D. Mohr, E. E. Feldman, L. K.Chang, and P. R. Betten, “The experimental breeder reactor IIinherent shutdown and heat removal tests—results and anal-ysis,”Nuclear Engineering andDesign, vol. 91, no. 3, pp. 287–296,1986.

[3] H. A. Larson, E. M. Dean, J. F. Koenig, J. G. Gale, andW. K. Lehto, “Dynamic simulator for nuclear power plants(DSNP): development, verification and expansion of modules,”in Proceedings of the International Conference on Power PlantSimulation, CONF-851007-15, Cuernavaca, Mexico, November1984.

[4] P. Sharma, K. Natesan, P. Selvaraj, V. Balasubramaniyan, andP. Chellapandi, “Dynamic modeling of steam water system ofprototype fast breeder reactor using RELAP code,” Annals ofNuclear Energy, vol. 68, pp. 209–219, 2014.

[5] C.-Y. Yang, T. K. S. Liang, B. S. Pei, C. K. Shih, S. C. Chiang,and L. C. Wang, “Development and application of a dualRELAP5-3D-based engineering simulator for ABWR,” NuclearEngineering and Design, vol. 239, no. 10, pp. 1847–1856, 2009.

[6] Q. Zhang, R. Guo, C. Zhang, X. Chen, and B. Wang, “Radioac-tive airborne effluents and the environmental impact assess-ment of CAP1400 nuclear power plant under normal opera-tion,” Nuclear Engineering and Design, vol. 280, pp. 579–585,2014.

[7] Nuclear Safety Analysis Division, RELAPS/MOD3.3 CodeMan-ual Volume I: Code Structure, System models, and SolutionMethods, Information Systems Laboratories, 2001.

[8] M.Dulau andD. Bica, “Mathematicalmodelling and simulationof the behaviour of the steam turbine,” Procedia Technology, vol.12, pp. 723–729, 2014.

[9] T. C. Elliott, K. Chen, and R. Swanekamp, Standard Handbookof Powerplant Engineering, McGraw-Hill, New York, NY, USA,2nd edition, 1997.

[10] M. Alvarez-Fernandez, L. D. Portillo-Valdes, and C. Alonso-Tristan, “Thermal analysis of closed feedwater heaters in nuclearpower plants,” AppliedThermal Engineering, vol. 68, no. 1-2, pp.45–58, 2014.

[11] S. C. Stultz, Steam: Its Generation and Use, Babcock & Wilcox,41st edition, 2005.

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