Stress Analysis for Reactor Coolant Pump Nozzle of Nuclear Reactor Pressure Vessel

doi:10.4236/jamp.2013.16013

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Journal of Applied Mathematics and Physics, 2013, 1, 62-64

Published Online November 2013 (http://www.scirp.org/journal/jamp)

http://dx.doi.org/10.4236/jamp.2013.16013

Open Access JAMP

Stress Analysis for Reactor Coolant Pump Nozzle of

Nuclear Reactor Pressure Vessel

Lijing Wen, Chao Guo*, Tieping Li, Chunming Zhang

Nuclear and Radiation Safety Center, MEP, Beijing, China

Email: [email protected], *[email protected]

Received September 2013

ABSTRACT

Integrated reactor structura l design makes the pressure vesse l itself and loads more complicated, so stress concentration

makes strength failure easier at reactor coolant pump nozzle. The general purpose finite element program ANSYS/

WORKBENCH was used for 3D stress and fatigue analysis and the results of the evaluation are based on RCC-M crite-

ria. The integrated reactor structural design is evaluated to demonstrate with applicable criteria and ANSYS/WORK-

BENCH has better operability than ANSYS APDL on stress analysis of reactor pressure vessel.

Keywords: Numerical Simulation; Reactor Pressure Vessel; Stress Analysis

1. Introduction

Integrated reactor is new generation pressurized water

reactor developed by China [1]. Reactor structure is

highly integrated, for example, once through steam ge-

nerator placed in a pressure vessel; reactor coolant pump

is placed on the pressure vessel nozzles; press ure vessel

fixed and supported the main pump, control rod drive

mechanism and the top of the h eap struc ture. This struc-

tural design can simplify a loop system and eliminate

LBLOCA, but makes the pressure vessel itself and load s

more complicated. So the stress analysis is necessary to

ensure the effectiveness and safety of the design and the

integrity of th e reactor pressure vessel.

Most domestic researchers [2-5] using traditional

ANSYS APDL for stress analysis and evaluation of reac-

tor pressure vessel calculation. With the development of

computing software, ANSYS, Inc. has developed AN-

SYS/WORKBENCH software and pre-processing, cal-

culation and post-processing become more convenient

[6]. ANSYS/WORKBENCH is used in this paper.

2. Calculation Model

According to the structure features of the reactor pressure

vessel, the reactor pressure vessel cylinder and outlet

nozzle is → are simplified:

1) According to RCC-M, surfacing layer is not exposed

to any structural strength in criteria O, criteria C and cri-

teria D. So surfacing layer is ignored in numerical model.

2) Bolt holes on reactor coolant pump nozzle and the

effect of in-line casing and annular flow distribution

plate are ignored.

3) To reduce edge effects influence on the reactor

coolant pump nozzle, pressure vessel cylinder must be

long enough in calculation model. According to RCC-M,

the length in pressure vessel cylinder axial direction from

the center line of the reactor coolant pump nozzle is at a

length of

2.5 Rt

at least, whe r e R = (R1 + R2)/2. R1

and R2 are the inside and outside radius of the pressure

vessel cylinder, and t is the thickness of the pressure

vessel cylinder.

Three-dimensional finite element model of pressure

vessel cylinder and outlet nozzle is built by ANSYS/

WORKBENCH program as shown in the Figure 1. 1 /4

model of pressure vessel cylinder and complete model of

pressure vessel nozzle are built by hexahedral meshes

with the size of 40 mm.

Reactor pressure vessel design temperature is 343˚C.

Pressure vessels and outlet nozzle materials are 16MND5

forgings. According to RCC-M Parameter Manual, 16

MND5 material parameters at 350˚C are shown in Table 1.

3. Evaluation Conditions, Loads and

Boundary Conditions

According to RCC-M, stress strength must conform to

criteria O in the first operating condition when pressure

vessel nozzle is analyzed. Self-weight, internal pressure

and earthquake load are considered and each load is

composed of six components. Force and moment are

*Corresponding a uthor.

L. J. WEN ET AL.

Open Access JAMP

Figure 1. Reac tor pressure vessel simulation mode.

Table 1. Material parameters of 16MND5.

kg·m−3 E

GPa k

W·m−1·K−1 α

˚C −1 Sm

MPa Sy

MPa Su

MPa

7800 180 38.7 1.34 × 10−5 184 345 552

transmitted to the cylinder and nozzle ends through ex-

tension of pressure vessel cylinder.

Fixed boundary condition A is set in the bottom of the

vessel; symmetry boundary condition B is set on the sec-

tion of symmet r i c a l structur e of the reactor pressure ves-

sel; hydrostatic pressure C and F are set on the upper

section of reactor pressure vessel and reactor coolant

pump nozzle; the mechanical loads D and E caused by

self-weight and e ar thquake loa d are set on the upper sec-

tion of the reactor coolant pump nozzle; the internal

pressure G is set on the inner surface of all parts as

shown in the Figure 2.

4. Results and Evaluation

The calculated stress intensity contours shown in Figure

3. The maximum stress is in the junction of the cylinder

nozzle. According to RCC-M, stress is divided into pri-

mary membrane stress, secondary membrane stress and

bending stress to be evaluated.

A typical evaluation section usually is the structural

discontinuous section where stress intensity is high due

to mechanical loads. Stress is analyzed by stress lineari-

zation in ANSYS/WORKBENCH. Firstly, the high stress

intensity area is searched on the stress intensity contours,

Figure 2. Load s an d bou ndary condi t i on s .

Figure 3. Stress intensity contours.

and then the two nodes which can run through the thick-

ness of pressure vessel cylinder and nozzle are selected

in the structure discontinuous area. Result data can be

mapped to the path formed by connecting two nodes, and

then the numerical calculation results can be analyzed by

the selected path. Based on the above principles, twelve

paths are set in the structure continuous areas, and stress

concentration areas and structural discontinuous areas are

shown in Figur e 4.

Table 2 sho ws the results of stress analysis on evalua-

tion section. The bending stress of the whole disconti-

nuous area on the path 2, 3, 5, 8, 10 and 11 are secondary

stress, and they are not evaluated in criteria O.

According to RCC-M, primary stress must conform to

criteria O in the first operating condition. The criteria O

requires that Pm ≤ Sm, PL ≤ 1.5Sm, and Pm (PL) + Pb ≤

1.5Sm. Pm is general primary membrane stress, PL is pri-

mary local membrane stress, Pb is primary bending stress

L. J. WEN ET AL.

Open Access JAMP

Figure 4. Evaluation section and paths.

Table 2. Stress linearization results.

Path Pm PL Pm + Pb

1 149.5 - 167.0

2 - 168.5 *

3 - 199.5 *

4 34.3 - 47.6

5 - 33.3 *

6 40.3 - 49.8

7 38.6 - 49.5

8 - 8.1 *

9 20.7 - 38.3

10 - 199.7 *

11 - 167.2 *

12 127.4 - 137.3

and Sm is Allowable Stress.

Path 1, 4, 6, 7, 9 and 12 are in the structure continuous

areas. The membrane stre ss there is general primary

membrane stress and the maximum “Pm(149.5 MPa)” is

lower than “Sm(184 MPa)”. The bending stres s there is

primary bending stress and the maximum “Pm + Pb

(167.0 MP a)” is lowe r than “1.5Sm (276 MPa)”.

Path 2, 3, 5, 8, 10 and 11 are in the structural discon-

tinuous areas. The membrane stress th ere is primary local

membrane stress and the maximum “PL(199.7 MPa)” is

lower than “1.5Sm (276 MPa) ”. The bending stress there

is secondary stress and it is not evaluated in criteria O.

As can be seen from the above, the stru ctu re designed

can satisfy the intensity requirement in the criteria O of

RCC-M.

6. Conclusions

1) The str ess of reactor coolant pump nozzle of nuclear

reactor pressure vessel is analyzed and evaluated by

ANSYS/WORKBENCH. The results show that the

structure designed c an satisfy the intensity requirement in

RCC-M.

2) ANSYS/WORKBENCH is more convenient than

the traditional ANSYS APDL, whi ch can be widely used

in the geometry modeling, loading and post processing.

REFERENCES

[1] The French Nuclear Island Equipment Design and Con-

struction Rules of Association, “PWR Nuclear Island

Mechanical Equipment Design and Construction Rules

(RCC-M)”, Paris, 2000.

[2] J. H. Yu, “Analysis and Evaluation for the Reactor Ves-

sels of CEFR,” M.S. Thesis, China Institute of Atomic

Energy, Beijing, 2004.

[3] Z. Qin, “Research and Design of China Integral Reactor

Plant Innovative Containment,” Nuclear Power Engineer-

ing, Vol. 27, No. 6, 2006, pp. 91-93.

[4] J.-W. Wang, Z.-H. Zhang, G.-F. Wu and L.-M. Fan,

“Reactor Vessel Stress Analysis for CPR1000 Nuclear

Power Plant,” Atomic Energy Science and Technology,

Vol. 42, No. Supp1, 2008, pp. 459-499.

[5] W. Yang, L.-G. Zheng and Y. Yang, “Stress and Fatigue

Analysis for Outlet Nozzle of Nuclear Reactor Pressure

Vessel,” Atomic Energy Science and Technology, Vol. 42,

No. Supp1, 2008, pp. 505-508.

[6] X.-Y. Cheng, X.-W. Feng, G.-Z. Li, B.-Q. Zhu and B.

Zhang, “Stress Analysis of Pressure Vessel Nozzle Based

on ANSYS/WORKBENCH,” Petro & Chemical Equip-

ment, Vol. 14, No. 2, 2011, pp. 5-8.