ML20128Q107
| ML20128Q107 | |
| Person / Time | |
|---|---|
| Site: | South Texas  |
| Issue date: | 07/31/1977 |
| From: | Calhoun D, Miller D WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20128P958 | List: |
| References | |
| WCAP-9135, NUDOCS 8507260586 | |
| Download: ML20128Q107 (12) | |
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STRUCTURAL ANALYSIS OF REACTOR COOLANT LOOP FOR THE SOUTH TEXAS PROJECT
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UNITS NO.1 AND 2 VOLUME 1 ANALYSIS OF THE REACTOR COOLANT LOOP PIPING D. V. Calhoun July 1977 I
APPROVED:
W 3-D. F. Miller, Manager Reactor Coolant Loop Analysis Work Performed Under Shop Order No. TGX 145 WESTINGHOUSE ELECTRIC CORPORATION Nuclear Energy Systems P. O. Box 355 Pittsburgh, Pennsylvania 15230 8507260586 850715 PDR ADOCK 05000498 A
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TABLE OF CONTENTS S
Section Title Page 1
INTRODUCTION 11
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2 CRITERIA 2-1 2 1.
Introduction 2-1 2 2.
Primary Stresses 21 2 3.
Fatigue 22 3
LOADING 31 3-1.
Design Loads 31 3-2.
Weight 31 3-3.
Earthquake 3-1 3-4.
Loss of Coolant Accident (LOCA) 32 3-5.
Operatirg and Transient Loads 34 3-6.
Thermal Expansion 3-4 3-7.
Pressure 3-4 3 8.
Transients 3-7 3 9.
Loading Classification 37 3-10.
Design Conditions 37 3-11.
Design Temperature 3-7 3-12.
Design Pressure 3-7 3 13. Design Mechanical Loads 3-7 3-14.
Normal Conditions 3-8 3-15.
Upset Conditions 38 3 16. Emergency Conditions 3-8 3 17. Faulted Conditions 3-8 3-18. Test Conditions 38 3-19.
Load Combinations 39 i
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TABLE OF CONTENTS (cont)
Section Title Page 4
ANALYTICAL METHODS AND MODELS 41 4 1.
Modeling Considerations 41 4 2.
Static Analysis 41 4 3.
Dynamic Analysis 45 4 4.
Seismic Analysis 45 4 5.
Hydraulic Analysis 48 46.
Loss of Coolant Accident Analysis 4-13 4 7.
Fatigue Analysis 4-15 4-8.
Thermal Transients 4-15 4-9.
Load Set Generation 4-18 4 10. Flexibility Factors and Stress indices 4 19 5
RESULTS OF STRESS EVALUATION 51 5 1.
Primary Stress Evaluation 51 3
5 2.
Design Conditions 51 5 3.
Emergency Conditions 5-2 5-4.
Faulted Conditions 52 5-5.
Fatigue Evaluation 5-4 5-6.
Seismic Loadings 54 5-7.
Transient Thermal Loadings 5-5 58.
Thermal Expansion Loadings 5-5 5 9.
Loading Combinations 5-5 5-10. Stress Intensities and Usage Results 56 6
CONCLUSIONS 6-1 I
7 PIPING AND EQUIPMENT DRAWINGS 7-1 6
71.
Piping Drawings 7-1 i
72.
Equipment Drawings 71 t
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LIST OF ILLUSTRATIONS Figure Title Page 11 Reactor Coolant System, Flow Diagram 12 1
12 Simplified Diagram of the NSSS 14 1-3 Reactor Coolant Loop Coordinate System 16 31 Location of Postulated Breaks 35 4-1 Reactor Coolant Loop Model 42 42 Steam Generator Lumped - Mass Model 47 43 Reactor Coolant Pump Lumped - Mass Model 49 44 Reactor Pressure Vessel Lumped - Mass Model 4 10 45 THRUST RCL Model Showing Hydraulic Force Locations 4 11 46 Time History Dynamic Solution for LOCA Loading 4-14 4-7 Through Wall Thermal Gradients 4 17 i
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LIST OF TABLES Table Title Page 31 Postulated Break Locations for the LOCA Analysis of t.
the Primary Coolant Loop 36 32 Load Combinations and Stress Evaluation 3 10 4-1 Description of Node Points on the Reactor Coolant Loop Model 4-3 5-1 Stress Analysis Summary 53 i'
i V
SECTION 1 INTRODUCTION '
The South Texas Project nuclear power plant reactor coolant loops for Units 1 and 2 are sub-lected to a very detailed structural and mechanical evaluation to ensure that the public's health I
and safety are protected. This detailed evaluation compares the results obtained from piping system analyses with the acceptance criteria, ASME Boiler and Pressure Vessel Code Section ill I
Nuclear Power Plant Components N (hereafter referred to as the code) for the conditions stated in the South Texas Project Piping Design Specification ASME Ill Code Class 1 ANS Safety Class l l25 (hereaf ter referred to as the design specification). This report concerns itself with the structural evaluation of the reactor coolant loop and primary equipment supports sys-tem under all design loading conditions. In particular, volume 1 contains the piping stress eval-uation and system analysis description, volume 2 the primary equipment support evaluation, and volume 3 the branch nozzles evaluation and the pipe evaluation in the close vicinity of the nozzles.
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Normal operation and safe shutdown of a nuclear plant depends upon the design adequacy and structural integrity of the reactor coolant loop. To demonstrate design adequacy and structural integrity of the reactor coolant loop, analyses are performed for loading under normal condi-tions, seismic disturbances, and postulated loss of coolant accident conditions. The results of these analyses are compared with the allowable stresses of the code in accordance with the de-sign specification. The results of this information are reported in volume 1 of the stress report.
A nuclear power plant, based on the closed cycle pressurized water reactor (PWR) concept, utilizes two separate fluid systems which interface at a heat exchanger. These systems are known as the primary and secondary systems. Heat is produced in the core by the fission process and is transferred from the primary coolant to the secondary system through a heat i
exchanger normally referred to as a steam generator. The primary coolant cycle is completed when the water is pumped back to the reactor vessel by the reactor coolant pump (figure 11).
In the Westinghouse pressurized water reactor (PWR) system, the primary system is designed l
for a pressure of 2485 psig and a reactor design temperature of 650 F. The secondary system
- 1. AsME Boiler and Pressure Vessel Code Section lil Nuclear Power Plant Components." ASME. New York,1974. up to Winter 1975 addenda.
- 2. Westinghouse Equipment Specification 953385 Rev 0. Piping Design Specification AsME Ill Code class 1. ANs safety Cises 1 - for Houston Lighting & Power Company. South Temos Prosect units 1 and 2." Salmey. K. R. (Westinghouse Proprietary).
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Figure 11. Reactor Coolant System, Flow Diagram
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operates at a lower temperature and pressure than the primary system to allow the transfer
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of heat and to promote the formation of high quality saturated steam.
The Westinghouse PWR consists of closed reactor coolant loops as shown in figure 12. Each reactor coolant loop contains one coolant pump and one steam generator. The reactor coolant pumps are Westinghouse vertical, single steam, mixed flow pumps of the shaft seal type, the t
steam generators are Westinghouse vertical U tube units. The steam generator consists of two basic parts: an evaporator section, and a moisture separation section. The evaporator section features a U tube bundle where heat from the reactor is transferred through tbc tube walls to convert pure secondary side feedwater into steam. The moisture separation section consists of
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a set of moisture separators which remove entrained water from the steam.
An electrically heated pressurizer connected to one reactor coolant loop maintains reactor coolan} system pressure during normal operation, limits pressure during normal operation, limits pressure variations during plant load transients, and keeps system pressure within design limits during abnormal conditions.
Auxiliary system components are provided to charge the reactor coolant system, add makeup water, purify reactor coolant water, provide chemicals for corrosion inhibition and reactor control, cool system components, remove decay heat when the reactor is shut down, and pro.
vide for emergency safety injection.
k The chemical and volume control system (CVCS) performs the following functions:
a Fills the reactor coolant system Provides a source of high pressure water for pressurizing the reactor coolant system a
when cold Maintains the water level in the pressurizer when the reactor coolant system is hot a
Reduces the concentration of corrosion and fission products in the reactor coolant a
Adjusts the boric acid concentration control e
Provides high pressure seal water for the reactor coolant pump seats a
The residual heat removal system (RHRS) transfers heat energy from both the core and the reactor coolant system during plant shutdown and refueling operations. The system is also used (in conjunction with the safety injection system) for emergency core cooling under the postu-1 lated pipe rupture accident conditions.
The primary function of the safety injection system (SIS) is to supply borated water to the reactor coolant system to limit fuel rod cladding temperature in the postulated but unlikely i
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event of a loss of coolant accident. A secondary SIS function is to provide a means for intro-ducing a spray of borated water into the containment as an additional dynamic heat sink.
The Emergency Boration System (EBS) protects the Reactor Coolant System against the effects of an incident causing a loss of fluid from the secondary side (steam side), identified as a
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steam break. The EBS provides negative reactivity in the form of concentrated boric acid to counteract the reactivity increase resulting from the steam break.
I The reactor coolant loop and primary equipment supports system are part of the nuclear steam supply system (NSSS). Figure 1-3 shows a simplified plan view of the NSSS.
This report presents the computation of stresses in the reactor coolant loop piping for all l
loading conditions stipulated in the design specification. In addition, loads on supports for the loading conditions stated above are generated for use in the support evaluation. (See volume 2 of this report.) Finally, this report compares the piping stresses with the acceptance criteria as stated in the code.
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