STEAM LINE BREAK ANALYSIS OF STEAM GENERATOR IN NUCLEAR

POWER PLANT USING RELAP5

Chang-Hoon Ha, Jong-In Kim and Moo-Yong Kim

DOOSAN Heavy Industries & Construction Co. LTD, 555 Guygok-Dong, Changwon, Gyeongnam,

641-792, Korea

1. Introduction

The steam generator in nuclear power plant has a function not only to supply continuously the high

quality steam for the turbine to produce electricity but also to sustain the pressure boundary to the high

primary pressure. Therefore, the component designer of steam generator should give consideration on the

structural integrity as well as the high performance for the design life. The steam generator goes through

numerous thermal hydraulic transients, such as pressure, flow rate and temperature and also the accident

postulated.

This study is to simulate the accident, especially in the steam line break (SLB) using RELAP5/MOD3

(herein called RELAP5) thermal hydraulic computer code. A RELAP5 (reference [1]) model was

developed to simulate the SLB of steam generator and to determine hydraulic loads across the internals.

Therefore, various thermal hydraulic parameters, such as pressure, flow rate, velocity and void fraction,

were obtained to evaluate the integrity of steam generator internals during the SLB accident.

CEFLASH-4B (reference [2]) computer code has been used to perform SLB analysis of steam generator

before. But preparing the input list of CEFLASH-4B is very difficult because the model consisted of

many nodes and flow paths. In addition, the initial condition of nodes and flow paths must be computed

using other computer program. As a result of this investigation a new analysis method which will be

conveniently used in thermal and hydraulic simulation of SLB with a good reliability, is developed in

this paper. The SLB analysis was performed using the RELAP5. The RELAP5 is the third major variant

of the RELAP5 advanced thermal-hydraulic code that was developed by the Idaho National Engineering

Laboratory (INEL) in 1979. Initially, this code was used principally by INEL analysts for understanding

Loss-of-Fluid Test and Semiscale Experimental Behaviour. Since then, the code has become widely

accepted throughout the world for analysing commercial and experimental reactor systems together with

their related scaled subsystems.

A RELAP5 model for the Replacement Steam Generator (RSG) of Sequoyah unit 1 (SQH#1) was

developed to analyse the SLB and system model preserved most of major features of the RSG including

component elevations, component volumes, flow rates and flow lengths.

2. Analytical Model and Condition

The RELAP5 equation set gives a two-fluid system simulation using a non-equilibrium, non-

homogeneous, six-equation representation. Constitutive models represent the interphase drag, the various

flow regimes in vertical and horizontal flow, wall friction, and interphase mass transfer. The code also

has the capability to simulate energy transfer to and from the fluid by the presence of slabs of material

adjacent to the fluid. Control systems and component models permit simulations of equipment

controllers, balance-of-plant equipment, and lumped-node representations of various processes. The

process of building a RELAP5 model is analogous to physically building the system that is being

modelled. Just as the physical system is composed of pipes, valves, pumps, and other components, so is

the RELAP5 model. Since the hydrodynamic simulation used within RELAP5 code is based on a one-

dimensional model of the transient flow for a steam-water non-condensable mixture, the numerical

solution scheme that is used results in a simple system representation using control volumes that are

connected by juncture.

The RELAP5 building blocks can be divided into four fundamental groups; thermal-hydraulic, heat

structures, trips, and control variables. The thermal-hydraulic group is composed of components

designed to simulate fluid passages and fluid handling equipment. Heat structures are designed to

simulate material mass and the interaction between the material mass and the fluid passage. Trips are

designed to simulate the signals that initiate equipment actions of various sorts (e.g., turning on a pump

at a desired time or causing a valve to open at one pressure but close at another). Finally, control systems

are designed to give the code additional modelling capability by allowing equipment control systems.

2.1 Analytical Model

The SQH#1 RSG is of the vertical, inverted U-tube, feed ring type and shown on Fig.1. The pressure

vessel incorporates a primary head, a main shell containing the heat transfer surface, and a enlarged

steam drum at the top. The heat transfer surface, which also provides the interface between the primary

and secondary systems, consists of a bundle of 4983, alloy 690, u-tubes hydraulically expanded into and

welded to the tubesheet. The tubesheet forms the pressure boundary between the primary and the

secondary sides. The primary fluid enters the primary head through one inlet nozzle, passes through the

tubes and leaves via the two outlet nozzles in the primary head. The primary head contains a divider plate

that separates incoming and outgoing primary fluid. The tube bundle is situated within a shroud which

defines the secondary side two-phase circulation flow paths. The steam is separated from the fluid in the

steam drum by separators and dryer. The remaining liquid drops to the can-deck and flows down the

downcomer, outside of the shroud, and enters the top of tubesheet.

Figure 1 The replacement steam generator of Sequoyah Unit 1

In developing the RELAP5 model, the principal modelling objective was to preserve all secondary-side

volumes and elevations. As a consequence, many of the flow areas within the unit are representative of

the average flow area and were obtained by dividing the flow volume in each section of the steam

generator by the flow length. On the primary-side, only the fluid volume contained within the tubes was

Steam Outlet Nozzle

Steam Dryer

Moisture Separator

Feedwater Nozzle &

Feed Ring

Shroud

U-Tube & Tube Bundle

Tubesheet

Primary Head

Downcomer Region Tube Support

191192

170-11176-1

180-J2

180

180-J1

171-J1

170-12174

174

178

174-J2

174-J1

171

171-J3

182

183

186189

190

184

185

176-1

174187

188

1

3

4

5

6

7

8

9

10 11

12

13

14

15

16

17

18

19

20

176-1 170-11

177107 109

106 110

2

3

4

5

6

7

8

9

10 21

2

3

4

5

6

7

8

9

10

preserved. The primary inlet and outlet plenum dimensions and volumes were not preserved since these

regions are considered to be non-active and do not affect results.

The secondary-side volumes are calculated as the total secondary-side volume excluded all major

internals. The downcomer volume includes all of the annulus outside of the shroud from the top of lower

shell to the tubesheet.

Referring to the RELAP5 nodalization diagram shown in figure 2, there are three pipe components in the

RELAP5 model. The components numbered 108, 170 and 176 are the three pipe components.

Component 108 represents primary fluid volume inside of tubes. The first and last part of component 108

represent fluid volume inside of tube within tubesheet at the hot side and cold side respectively.

(Upper part of SG) (Lower part of SG)

Figure 2 The schematic diagram of RELAP5 analytical model

The first 8 part of Component 170 represents secondary fluid volume between each tube support plates.

The 9th part of Component 170 represents U-bend region of tube, and the 10th and 11th part of

Component 170 represents riser, and The 12th part of Component 170 represents fluid volume inside

primary separator. The first part of Component 176 represent downcomer transition region and the other

parts of Component 176 represent lower downcomer. The separating region from top of separator to

dryer was modeled separator component. The separator component provides a simplified representation

of the actual process. In general, this component accepts the inlet flow, performs an idealized prescribed

separation of the liquid and vapor phases. When in the separating mode, it passes the vapor out the

separator outlet junction and passes the liquid out the liquid return junction. Single volume 174

represents the volume of fluid contained within the drum liquid region. The normal water level was input

as a void fraction of this volume. Single volume 178 represents the volume of fluid within drum and

outside of separator. Branch 180 represents the flow path from the separator outlet to dryer inlet. Single

volume 182 and 184 represent the volume above the dryer. Single junction 185 represents the steam

outlet and trip valve 191 represents the break line of steam to the containment building which is modeled

as single volume 192. Time dependant juncture 187 represents the feed water control valve to control the

water level. Cylindrical heat structures were used to model the heat transfer path between the primary

and secondary side.

2.2 Condition Analyzed

One power level, 100% power, was analyzed in this study and the steam generator operating parameters

for the 100% power level is taken from reference [3] and [4]. The normal water level is 500 inches above

the tubesheet. The operating parameters for 100% power level are shown below.

Table 1 Operation parameters for 100% power level

Primary Side Condition Secondary Side Condition

� Pressure : 15.513 MPa

� Inlet Temperature : 320.06 oC

� Outlet Temperature : 286.67 oC

� Flow Rate : 4.646E+03 kg/sec

� Heat Load : 866.904 kW

� Saturation Pressure : 6.17 MPa

� Steam Flow Rate : 477.532 kg/sec

� Feed Water Temperature : 224.33 oC

� Circulation Ratio : 4.97

� Water Level : 12.7 m

3. Method of Solution

The thermal hydraulic information presented in this paper was obtained using the RELAP5 which

calculate the behaviour of a reactor coolant system during transient and can be used for simulation of a

wide variety of hydraulic and thermal transients in both nuclear and non-nuclear systems involving

mixtures of steam, water, non-condensable, and solute.

The result of RELAP5 largely depends on user's approach made in the application of the code to the

specific system. Therefore the results of RELAP5 to the SQH#1 RSG model are deliberately reviewed as

described below.

The elevation variation check through the steam generator circulation loop is properly implemented by

reducing the height of last part of downcomer component, which make the re-entrance elevation of first

part of tube bundle to its realistic value.

Component pressures are also checked through the secondary system in order to prevent unrealistic

pressure drop due to Courant limit error which happens in the component with very small fluid volume

relative to other components. Heat structure input like tube thickness, tube inside diameter and overall

tube length within a fluid volume is also reversely checked by comparing the resulting transferred heat

with desired values.

Separator model is checked by reviewing whether it give proper liquid fall back path condition and

vapour outlet path condition. As a result, high quality steam condition is obtained in vapour outlet path

and subcooled water condition is obtained in the downcomer region which is downstream of liquid fall

back path.

The analysis results of RELAP5 are compared with those of CEFLASH-4B (reference 6). CEFLASH-4B

is a one-dimensional, two-phase, thermal-hydraulic code which calculates the time-dependant behaviour

of the fluid state resulting from a steam line rupture or an operational transient. The solution of

CEFLASH-4B proceeds by numerically integrating the momentum equation applied to the flow paths,

which is represented as juncture component of RELAP5, while maintaining mass and energy balances in

the nodes which are represented as volume and pipe in RELAP5. The conditions applied on each

analytical computer code, RELAP5 and CEFLASH-4B, are listed in Table 2.

Table 2 Conditions applied on analysis computer code

Condition RELAP5 CEFLASH-4B

Model Structure Based on Volume & Juncture Based on Node & Flow Path

Two-Phase Non Homogeneous Homogeneous

Critical & Choked Flow Henry-Fauske Flow

Moody Critical Flow

Modified Henry-Fauske/Moody

Flow

Opening time 0.0 sec 0.001 sec (1 ms)

Discharge Coefficient 1.0, 1.2 1.0

The three cases of applied condition for RELAP5 are investigated for various results and are shown on

Table 3.

Table 3 Cases of applied condition for RELAP5

Case I.D Flow Model Discharge Coefficient(DC)

DC 1.0 Henry-Fauske Flow 1.0

DC 1.2 Henry-Fauske Flow 1.2

MDC 1.0 Moody Critical Flow 1.0

4. Analytical Results

4.1. The Results of System Initialization

Before the SLB analysis, the initial condition of 100% power level is set up on RELAP5 model. When

the analytical system of RELAP5 model becomes a steady state, the trip valve opens to simulate the

steam pipe rupture. Table 4 shows the results of hydraulic parameters on the initial condition.

Table 4 Results of Initialization of 100% Power Condition for RELAP5 SLB Model

Operating Parameters From RELAP5 From Thermal Hydraulic

Performance Analysis Ratio

Steam Flow Rate (kg/sec) 476.5 477.5 0.998

Recirculation Flow Rate (kg/sec) 2359.5 2375.1 0.993

Heat Transfer Rate (kW) 8.55 E+05 8.67 E+05 0.986

Water Level (m) 12.7 12.7 1.00

4.2. Condition Analyzed

The method described in section 3 is used to obtain pressure loads on the steam generator internals which

include the steam dome, the separator deck, U-bend region and tube supports. Results are obtained for a

2.0 second blowdown simulation at 100% power. Peak pressure loads on steam generator internals are

presented Table 5.

Table 4 Results of Peak Pressure Loads (kPa)

RELAP5 CEFLASH-4B Location

DC 1.0 DC 1.2 MDC 1.0 MDC 1.0

Steam Dome 25.5 32.9 27.6 18.2

Separator Deck 66.4 84.7 68.7 79.1

U-Bend Region 20.5 23.8 21.0 25.3

Tube Bundle 98.7 106.0 100.0 97.6

Upper ATSG 14.3 16.3 14.6 19.8

Middle ATSG 14.3 16.1 14.5 18.4

Lower ATSG 11.2 11.7 11.4 19.4

Figure 3 and 4 present the pressure variation and the flow rate as a function of time for 100% power

case, respectively.

Pressure Variation at Steam Dome Pressure Variation at Separator

Figure 3 Pressure Variation at Upper SG

Flow Rate at Upper Tube Bundle Flow Rate at Lower Tube Bundle

Figure 4 Flow Rate Variation at Tube Bundle

5. Conclusion

The SLB analysis of SQH #1 RSG was successfully performed using RELAP5. As a result of this

investigation, it was found that the RELAP5 hydrodynamic model was well constructed because

RELAP5 hydrodynamic component represented actual hydrodynamic system component. Furthermore,

no additional calculation to get initial condition of RELAP5 hydraulic model is needed because the

initial condition can be obtained by steady state solution of the code itself.

The RELAP5 model developed for this analysis also can be used to perform the various hydraulic

performance analyses of steam generator. These hydraulic performance analyses include finding the

relationship of water level to moisture carry under, producing the thermal hydraulic responses for

pressure, temperature and flow rate inside of steam generator during the plant operation, and predicting

water level variation during power-up transient.

6. References

[1] Nuclear Safety Analysis Division, “RELAP5/MOD3.3 Code Manual, Volume I through VII”,

Information Systems Laboratories, Inc., Rockville, Maryland, Idaho Falls, Idaho, Prepared for the

Division of Systems Research Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory

Commission Washington, DC 20555, NUREG/CR-5535/Rev P3-Vol I through VII.

[2] Michonski, M., “Software Verification and Validation Report, CEFLASH-4B, Version f4b.1.1”,

VV-FE-0178-0, Transient thermal-hydraulic code used for the licensing analysis of blowdown

loads during a large break loss-of-coolant accident, ABB Combustion Engineering Nuclear

Operations, Windsor, Connecticut, October 1994.

[3] Basol, M., “Design Specification for Replacement Steam Generator for TVA Sequoyah Nuclear

Power Plant Unit 1”, Specification No. S1RSG-CD-C201, Rev.01, CE Nuclear Power LLC,

Chattanooga, TN, January 15, 2001.

[4] J.C. Lowry, “Thermal and Hydraulic design Data Report”, Prepared for Tennessee Valley

Authority by Westinghouse Electric Company, June 11, 2002.

[5] J.C. Lowry, “Thermal and Hydraulic Analysis Report”, Prepared for Tennessee Valley Authority

by Westinghouse Electric Company, August 27, 2002.

[6] J.R.Schwall, “Hydraulic Loads for Steam Line Break”, Westinghouse Electric Company, 2002.