THERMAL DISTRIBUTION ANALYSIS IN PRESSURE ...

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THERMAL DISTRIBUTION ANALYSIS IN PRESSURE VESSEL WALL OF PWR

E. Saragi1, R. Himawan2,P.W. Kedoh3. 1,2Centre for Nuclear Reactor Technology and Safety, National Nuclear Energy Agency of Indonesia

(BATAN), Puspiptek Area, Building 80, Serpong,15310, Indonesia. 3Udayana of University, Bali

Email: [email protected]

ABSTRAK ANALISIS DISTRIBUSI THERMAL PADA DINDING BEJANA TEKAN PWR. Bejana tekan reaktor merupakan komponen sangat penting yang dikategorikan kedalam standar keselamatan kelas 1 dalam reaktor tipe Pressurized Water Reactor (PWR). Bejana tekan reaktor sering mendapatkan beban panas, radiasi, tekanan dan kemungkinan korosi. Keandalan dalam bagian ini sangat tergantung pada beban yang diterima seperti beban temperatur yang berubah -ubah. Untuk mengantisipasi agar bejana tekan dapat tetap berfungsi dengan baik maka dilakukan analisis struktur. Tujuan dari penelitian ini adalah mengevaluasi pengaruh temperatur terhadap desain struktur bejana tekan reaktor AP1000 menggunakan analisis termal transien. Analisis termal transien dilakukan pada dinding bejana tekan reaktor PWR. Analisis termal transien menggunakan metode komputasi pemodelan berbasis elemen hingga. Komputasi pemodelan menggunakan data bejana tekan reaktor AP1000. Komputasi pemodelan yang dilakukan dengan menvariasikan temperatur inlet dan outlet. Temperatur inlet dan outlet digunakan sebagai beban. Dinding bejana tekan dimodelkan secara 2-D menggunakan elemen axisimetri. Hasil analisis adalah distribusi temperatur. Hasil yang diperoleh terjadi penurunan temperatur dari 343 oC menjadi 340 oC pada waktu (t) 998.7 detik dan 427 oC menjadi 419 oC pada waktu 999.9 detik. Penurunan temperatur ini tidak akan menyebabkan kerusakan pada dinding bejana tekan AP1000. Penyelesaian masalah menggunakan code MSC-PATRAN berbasis elemen hingga. Kata Kunci: Dinding Bejana Tekan, PWR, Transien Termal, Code MSC-PATRAN.

ABSTRACT THERMAL DISTRIBUTION ANALYSIS IN PRESSURE VESSEL WALL OF PWR.The reactor pressure vessel is a very important component of which is categorized into Class 1 safety standard of the reactor type Pressurized Water Reactor (PWR). The reactor pressure vessel gets a load of heat, radiation, pressure and the possibility of corrosion and chemical. Reliability in this section is highly dependent on the load it receives such as changing temperature loads. To assure that the pressure vessel can remain functioning properly then do the structural analysis. The purpose of this study was to evaluate the effect of temperature on the structural design of the AP1000 reactor pressure vessel which used transient thermal analysis.Transient thermal analysis was done on the walls of the reactor pressure vessel PWR. Transient thermal analysis used computational modeling. Computational modeling used AP1000 reactor data. Inlet and outlet temperature is used as load. Computational modeling was done by varying the inlet and outlet temperature. The wall was modeled using 2-D by axisymmetric elements. The results of the analysis are the temperature distributions. The temperature drop are occurred from 343 °C to 340 °C at 998.7 seconds and from 427 oC to 419 oC at 999.9 seconds. This temperature drop will not cause damage in pressure vessel wall of PWR.The solving problem used MSC-PATRAN code which is based on finite element method Keywords: Wall Pressure Vessel, PWR, ThermalTransient, MSC-PATRAN Code

Thermal Distribution Analysis In Pressure Vessel ... ISSN: 2355-7524

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INTRODUCTION The reactor pressure vessel is a very important component of which is categorized

into Class 1. The reactor pressure vessels are the most important pressure boundary components of the nuclear structures, systems and components because their function is to contain the nuclear core under elevated pressures and temperatures. Additional RPV functions are to provide structural support for the reactor vessel internals and the core. One of the problems in the safety system of a nuclear power plant is a pressure vessel failure. Pressure vessel must be able to withstand thermal load [1-4], especially regarding the materials used on reactor pressure vessel. In 2010, Roziq Himawan has conducted research on Transient thermal response of pressurize vessel wall under spraying process [5]. Zhang et al. [6] derived an analytical solution for determining the stress distribution of a multi layered composite pressure vessel subjected to an internal fluid pressure and thermal load. In a pressurized water reactor, main construction is the primary cooling system (reactor systems) and a secondary cooling system (steam system). The second function of the system is confining radioactive materials from spreading out of the reactor. In the primary system, water is not allowed to boil by giving a high enough pressure. High -pressure cooling water, and High Temperature ( Pressure 173.95 kg/cm2 and a temperature of 343°C ) of the primary system is passed to steam generator with primary circulation pump. M. Chen et.al [7] has also performed structural integrity analysis of RPVs pressurized thermal shocks (normal and accidental conditions). To investigate the effects of thermal stresses, under normal operating conditions of the plant, on the corner surface cracks at the set-in nozzle-cylinder intersection, a typical RPV [8] of the Westinghouse pressurized water reactor (PWR) is selected for the fracture mechanics analysis. The reliability of the reactor pressure vessel materials is depend on the reactor pressure vessel. Material pressure vessel must be able to withstand the load which is received as thermal and pressure loads or other loads. The material of the PWR pressure vessel uses SA508 or SA533B [7, 9-10].

To assure that the material used in the reactor pressure vessel remains in good condition there should be analysis based on the thermal load. The purpose of this study is to evaluate the transient thermal analysis on the walls of the reactor pressure vessel. Thermal analysis on the reactor pressure vessel that will be evaluated influence the operating temperature of the reactor pressure vessel structure design. Thermal analysis which is performed in this study is based on time dependent temperature load. Modelings done on the walls of the pressure vessel are not considering connection and nozzle on the reactor pressure vessel. Pressure vessel wall must be able to withstand the heat that flows in it. This modeling used MSC-Patran [11]. Sample of data used AP 1000 reactor data. From the simulation it will be obtained the temperature drop (the amount of thermal).This temperature drop will be used to determine the integrity of the reactor pressure vessel THEORY THE GOVERNING EQUATION OF TEMPERATURE DISTRIBUTION

The Solution of the temperature distribution in pressure vessel wall of PWR used the finite element method. The governing equation for cylinder transient thermal in an isotropic is given by [12- 16];

Qz

Tk

zr

Trk

rr

Tc

1

(1)

Initial condition: T(r,z,0)= To. f(r).f(z) for r ≤ 0; z ≤ L Boundary condition : These are specified zero temperatures at r = 0 and r = L T(0,z, τ)=0 for 0 ≤ z ≤ L,

T(r,z,τ )= 0 for 0 ≤ z ≤ L, and no-flow boundary condition z = 0 and z = L

LrforLrz

T

Lrforrz

T

00),,(

00),0,(

Prosiding Seminar Nasional Teknologi Energi Nuklir 2016 ISSN: 2355-7524

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Where k is the thermal conductivity of the materials, τ is used to represent time, ρ is mass density, c is specific heat, r is radius of pressure vessel, Q is the heat generation rate per unit volume and T is temperature. METHODOLOGY In this study, computational modeling is done to determine temperature distribution on pressure vessel of PWR using transient thermal analysis. Calculation of transient temperature distribution is done using Equation 1. Completion of equation 1 using code MSC – Patran. Geometry of reactor pressure vessel is complex as shown in Figure 1. Therefore, the analysis of pressure vessel will be modeled on the walls of the pressure vessel only. The model were not include connection and nozzle on the reactor pressure vessel. Modeling of pressure vessels was made in 2-dimensions using axisymmetric element. The size of the pressure vessel geometry used inside and outside diameter and loading is inlet and outlet temperature, which are axisymmetric. Pressurize vessel made of carbon steel type SA- 533B class1 which has thermal properties[7,9-10]. In the analysis, the material vessel assumed to be isotropic and homogeneous.

Figure 1. Pressure vessel of AP1000 reactor [1]

Modelling was done using MSC-Patran code which is based on transient thermal analysis. The processes are carried out as follows [11];

Pre-processing a. Create geometry of AP1000 reactor pressure vessel. Geometry modeling is

done with a 1/2 circle of pressure vessel wall. b. Perform meshing (dividing the field into smaller elements called

discretization/ meshing). Meshing using axisymmetric elements. Axisymmetric elements are classified as "2D solid elements" in PATRAN. First mesh the surface with Tria6 elements.

c. Define material and element properties. Material specification used Table 2 as the heat transfer and thermal expansion coefficient.

d. Create a transient load case. The model is subjected to load of the temperature. Loadings condition used Table 3 .

Solving the model Model may be considered to be ready for analysis after creating mesh, applying material properties, boundary conditions and loadings. Transient thermal analysis was carried out for 0,50 s until 10000 s.

Post-processing

Thermal Distribution Analysis In Pressure Vessel ... ISSN: 2355-7524

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After completing the analysis work, results of the model were investigated using post-processor. The analysis included the time-dependent evaluation of temperature on the vessel walls.

RESULTS AND DISCUSSION

The modeling is done by simplifying the geometry of the pressure vessel as shown in Figure 2. Input data used pressure vessel of AP1000 reactor data as shown in Table 1 and Table 2. Load temperature is used as transient load on the pressure vessel wall which used data on Table 3. Normal operation is representated by inlet and outlet temperature of 343 oC and 300oC, respectively. Inlet temperature of 427oC was chosen because it the maximum allowable inlet temperature. Allowable operating temperature limits on SA533B1 material is from 343 oC up to 427oC[17-18]. And the AP 1000 reactor vessel design pressure and temperature is 17.1 MPa and 343 oC, respectively [19]. Temperature variations were used to determine the integrity of the pressure vessel. The temperature-dependent thermal and mechanical properties of RPV are applied in the analysis. Transient thermal analysis is solved based on equation 1.

Figure 2. Modeled geometry

Table 1.Data of pressure vessel AP1000 [7, 9-10].

No. (Approximate values) 1. Design pressure (psig/Kgcm-2) 2485/173,95 2. Design temperature (°F/oC) 650 / 343,33 3. Overall height of vessel and closure head, bottom head

outside diameter to top of control rod mechanism (ft-in) 45-9

4. Number of reactor closure head studs (in) 45 5. Diameter of reactor closure head/studs (in) 7 6. Outside diameter of closure head flange (in.) 198 7. Inside diameter of flange (in.) 148.81 8. Outside diameter at shell (in.) 176 9. Inside diameter at shell (in.) 159

10. Inlet nozzle inside diameter (in.) 22 11. Outlet nozzle inside diameter (in.) 31 12. Clad thickness, nominal (in.) 0.22 13. Lower head thickness, minimum (in.) 6 14. Vessel beltline thickness, minimum (in.) 8 15. Closure head thickness (in.) 6.25

Prosiding Seminar Nasional Teknologi Energi Nuklir 2016 ISSN: 2355-7524

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Table 2. Specifications material SA 533B Class 1[1, 9, 13, 14-15,18]

Temperature(oC) Modulus of

elasticity (Gpa) Poisson

ratio Heat conductivity

(W/m oC)

Thermal expansion

(1/oC) ~350 193 0,3 38,7 15,1E-06

Table 3. Loads data [13,17]

No. Inlet Temperature (oC)

Outlet Temperature (oC)

1. 343 300 2. 343 250 3. 343 200 4. 427 300 5. 427 250 6. 427 200

Discretizations of the modelwere performed by dividing the field into small elements as shown in Figure 2. The results obtained in the form of transient temperature distribution as shown in Figure 3.

Figure 3a.Transient temperature distribution for temperature of 343 oC to temperature of 300 oC

Figure 3b.Transient temperature distribution for temperature of 343oC to temperature of 250 oC

Thermal Distribution Analysis In Pressure Vessel ... ISSN: 2355-7524

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Figure 3c.Transient temperature distribution for temperature of 343oC to temperature of 200 oC

Figure 3d. Transient temperature distribution for temperature of 427oC to temperature of 300 oC

Figure 3e.Transient temperature distribution for temperature of 427oC to temperature of 250 oC.

Figure 3f. Transient temperature distribution for temperature of 427oC to temperature of 200 oC

Figures 3show temperature distribution on RPV’s wall for different inlet and outlet temperature. For instance, Figure 3a, which represent a normal condition,show that temperature distribution will be decreased from inlet temperature 343 °C to 340 °C at 998.702 seconds. Figure 3f shows the maximum temperature drop up to 15 oC from 427 oC to 412°C at 999.9201 seconds.This temperature decrease will not damage the material of AP 1000 reactor vessel.. CONCLUSION. Thermal distribution analysis in pressure vessel wall of PWR was already performed. Simulation results show that the maximum occurring difference of temperature distribution is 15 oC. This maximum temperature drop occured with inlet and outlet temperature of 427°C and 200°C, respectively. This temperature drop will not cause damage on pressure vessel of

Prosiding Seminar Nasional Teknologi Energi Nuklir 2016 ISSN: 2355-7524

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PWR. Results of the time-dependent temperature of the RPV’s wall can be used in further study for the time-dependent stress distribution analysis. ACKNOWLEDGEMENT Author is grateful to the Centre for Nuclear Reactor Technology and Safety, National Nuclear Energy Agency of Indonesia for funding the research activity in BATAN's DIPA 2015/Number:SP DIPA 080.01.1.450310/2015 REFERENCES

1. USMAN TARIQ MURTAZA, M. JAVED HYDER, The effects of thermal stresses on the elliptical surface cracks in PWR reactor pressure vessel, Theoretical and Applied Fracture Mechanics 75, 124–136 (2015).

2. X. CHARLES,Stress Analysis of Pressure Vessel Due to Load and Temperature, Middle-East Journal of Scientific Research 20 (11), 1390-1395 (2014).

3. D. FERRENO, R. LACALLE, I. GORROCHATEGUI, F. GUTIERREZ-SOLANA, Analysis of dynamic conditions during thermal transient events for the structural assessment of a nuclear vessel, Engineering Failure Analysis 17 ,894-905 (2010).

4. INTERNATIONAL ATOMIC ENERGY AGENCY, Pressurized Thermal Shock in Nuclear Power Plants: Good Practices for Assessment Deterministic Evaluation for the Integrity of Reactor Pressure Vessel, IAEA-TECDOC-1627, IAEA, Vienna (2010).

5. Roziq Himawan,”Transient thermal response of pressurizer vessel wall under spraying process”, Prosiding Seminar Nasional Teknologi Energi Nuklir 2014, Pontianak, 19 Juni 2014.

6. Zhang Q, Wang Z W, Tang C Y, Hu D P, Liu p Q, Xia l Z, Analytical Solution of the Thermo-mechanical stresses in a Multi-layered Composite Pressure vessel considering the influence of the closed ends, Int J of Press vessels and Piping, 98 (2012), p102.

7. M. Chen, F. Lu, R. Wang, A. Ren, Structural integrity assessment of the reactor pressure vessel under the pressurized thermal shock loading, Nucl. Eng. Des. 272 (2014) 84–91.

8. ESS, Assessment and management of ageing of major nuclear power plant components important to safety: PWR pressure vessels, International Atomic Energy Agency, IAEA-TECDOC-1120, 1999.

9. ASME Boiler and Pressure Vessel Code, Section III, Division 1: Rules for construction of nuclear facility components, New York, 2010.

10. Barry H. Rabin, Ph.D, Characterization of A508/A533B Pressure Vessel Steel, VHTR R&D FY12 Technical Review Meeting, May 22-24, 2012, Salt Lake City, Utah.

11. PATRAN User’s Guide – MSC Software, 2012. 12. Xiao bing Feng, Thomas Lewis, Michael Neilan, Discontinuous Galerkin finite element

differential calculus and applications to numerical solutions of linear and non linear partial differential equations, Mathematics, Volume, June 2016, Pages 68–91.

13. P.F. Wang, Y. Liu, B.T. Jiang, J.S. Wan, F.Y. Zhao, Nodal dynamics modeling of AP1000 reactor for control system design and simulation, Annals of Nuclear Energy Volume 62, December 2013, Pages 208–223.

14. E. Saragi, Modeling of temperature distribution in TRISO fuel-based on finite element method, AIP Conference Proceedings. 1448, 270 (2012).

15. Mutahir Ahmed, Rafi Ullah Khan, Saeed Bad shah, Sakhi Jan, Finite Element Investigation of Geometry Effect on Pressure Vessel under Combined Structural and Thermal Loads, International Journal of Engineering and Advanced Technology (IJEAT) ISSN: 2249 – 8958, Volume-4 Issue-2, December 2014.

16. Olaf Kolditz,Hua Shao,Wenqing Wang,Sebastian Bauer, Thermo-Hydro-Mechanical-Chemical Processes in Fractured Porous Media:Modelling and Benchmarking, Springer International Publishing Switzerland, 2015.

17. M.Y.H. Bangash, Structures for Nuclear Facilities: Analysis, Design, and Construction, hand book, Springer 2011.

18. NN, Using SA508/533 for the HTGR Vessel material, Idaho National Laboratory, June 2012.

19. NN, AP 1000 European Design Control Document, Westinghouse, 2009.