NUREG/IA-0247

International

Agreement Report

RELAP5 Simulation of

Darlington Nuclear Generating Station

Loss of Flow Event

Prepared by:

D. Naundorf1

J. Yin'

A. Petruzzi 2

A. Kovtonyuk 2

'AMEC NSS Limited

700 University Avenue

Toronto, Ontario, Canada

2 Nuclear Research Group San Piero A Grado

University of Pisa

Pisa, Italy

A. Calvo, NRC Project Manager

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

February 2011

Prepared as part of

The Agreement on Research Participation and Technical Exchange

Under the International Code Assessment and Maintenance Program (CAMP)

Published by

U.S. Nuclear Regulatory Commission
AVAILABILITY OF REFERENCE MATERIALS

IN NRC PUBLICATIONS

NRC Reference Material

As of November 1999, you may electronically access

NUREG-series publications and other NRC records at

NRC's Public Electronic Reading Room at

htto://,xww.nrc.gov/readinq-rrn. html. Publicly released

records include, to name a few, NUREG-series

publications; Federal Register notices; applicant,

licensee, and vendor documents and correspondence;

NRC correspondence and internal memoranda;

bulletins and information notices; inspection and

investigative reports; licensee event reports; and

Commission papers and their attachments.

NRC publications in the NUREG series, NRC

regulations, and Title 10, Energy, in the Code of

Federal Regulations may also be purchased from one

of these two sources.

1. The Superintendent of Documents

U.S. Government Printing Office

Mail Stop SSOP

Washington, DC 20402-0001

Internet: bookstore.gpo.gov

Telephone: 202-512-1800

Fax: 202-512-2250

2. The National Technical Information Service

Springfield, VA 22161-0002

www.ntis.gov

1-800-553-6847 or, locally, 703-605-6000

A single copy of each NRC draft report for comment is

available free, to the extent of supply, upon written

request as follows:

Address: U.S. Nuclear Regulatory Commission

Office of Administration

Publications Branch

Washington, DC 20555-0001

E-mail: DISTRIBUTION. RESOURCE@NRC.GOV

Facsimile: 301-415-2289

Some publications in the NUREG series that are

posted at NRC's Web site address

..h....tt:/wvw.r.c.gov!readir.!-rm _doc-collectio.• ss /nurea.

are updated periodically and may differ from the last

printed version. Although references to material found

on a Web site bear the date the material was

accessed, the material available on the date cited may

subsequently be removed from the site.

Non-NRC Reference Material

Documents available from public and special technical

libraries include all open literature items, such as

books, journal articles, and transactions, Federal

Register notices, Federal and State legislation, and

congressional reports. Such documents as theses,

dissertations, foreign reports and translations, and

non-NRC conference proceedings may be purchased

from their sponsoring organization.

Copies of industry codes and standards used in a

substantive manner in the NRC regulatory process are

maintained atThe NRC Technical Library

Two White Flint North

11545 Rockville Pike

Rockville, MD 20852-2738

These standards are available in the library for

reference use by the public. Codes and standards are

usually copyrighted and may be purchased from the

originating organization or, if they are American

National Standards, fromAmerican National Standards Institute

11 West 4 2 nd Street

New York, NY 10036-8002

www.ansi.org

212-642-4900

Legally binding regulatory requirements are stated only

in laws; NRC regulations; licenses, including technical

specifications; or orders, not in

NUREG-series publications. The views expressed in

contractor-prepared publications in this series are not

necessarily those of the NRC.

The NUREG series comprises (1) technical and

administrative reports and books prepared by the staff

(NUREG-XXXX) or agency contractors

(NUREG/CR-XXXX), (2) proceedings of conferences

(NUREG/CP-XXXX), (3) reports resulting from

international agreements (NUREG/IA-XXXX), (4)

brochures (NUREG/BR-XXXX), and (5) compilations of

legal decisions and orders of the Commission and

Atomic and Safety Licensing Boards and of Directors'

decisions under Section 2.206 of NRC's regulations

(NUREG-0750).

DISCLAIMER: This report was prepared under an

international cooperative agreement for the exchange

of technical information. Neither the U.S. Government

nor any agency thereof, nor any employee, makes any

warranty, expressed or implied, or assumes any legal

liability or responsibility for any third party's use, or the

results of such use, of any information, apparatus,

product or process disclosed in this publication, or

represents that its use by such third party would not

infrinae privately owned ricahts.
NUREG/IA-0247

International

Agreement Report

RELAP5 Simulation of

Darlington Nuclear Generating Station

Loss of Flow Event

Prepared by:

D. Naundorf'

J. Yin1

A. Petruzzi 2

A. Kovtonyuk 2

'AMEC NSS Limited

700 University Avenue

Toronto, Ontario, Canada

2

Nuclear Research Group San Piero A Grado

University of Pisa

Pisa, Italy

A. Calvo, NRC Project Manager

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

February 2011

Prepared as part of

The Agreement on Research Participation and Technical Exchange

Under the International Code Assessment and Maintenance Program (CAMP)

Published by

U.S. Nuclear Regulatory Commission

ABSTRACT

The Darlington NGS consists of four 940 MWe CANDU reactors. A detailed RELAP5 model of a

Darlington NGS reactor has been created for use with RELAP5/MOD3.3. The model includes the

primary heat transport system, the steam generator feedwater and main steam supply systems, the

pressure and inventory control system, and the emergency coolant injection system. The modeling

of Darlington required the creation of new subroutines for RELAP5 to describe the functionality of

the process control systems and emergency control systems with sufficient accuracy. These

controller subroutines are standalone modules that are compiled and linked together with the

RELAP5 source to facilitate integration with future revisions to the code. The November 25, 1993,

loss-of-flow event at Darlington was simulated using this RELAP5 model. Excellent agreement was

obtained between the RELAP5 predictions and station measurements.

iii

CONTENTS

Page

A B S T R A C T ........................................................................................................................... iii

C O N T E N T S ........................................................................................................................... v

EXECUTIVE SUMMARY ................................................................................................. ix

ACKNOW LEDGMENT ...................................................................................................... xi

ABBREVIATIONS ................................................................................................................ xiii

1. IN T R O D U C T IO N ................................................................................................................... 1

2. DARLINGTON PLANT OVERVIEW ................................................................................ 3

2.1 Primary Side Systems .............................................................................................. 4

2.2 Secondary Side Systems ......................................................................................... 4

2.3 Reactor Regulating System ........................................................................................ 4

2.4 Special Safety Systems ............................................................................................ 5

3. DESCRIPTION OF LOSS OF FLOW EVENT .................................................................. 7

4. DARLINGTON CONTROLLER MODELING W ITH RELAP5 ............................................ 9

4.1 New Controller Input Data .......................................................................................... 9

4.2 Valve Stroking and Delay Times .............................................................................. 11

4.3 Testing of New Controllers ....................................................................................... 11

5. THE RELAP5 MODEL OF DARLINGTON ..................................................................... 19

5.1 RELAP5 Nodalization of Heat Transport System ..................................................... 19

5.2 RELAP5 Nodalization of the Secondary Side .......................................................... 22

5.3 RELAP5 Nodalization of the Pressure and Inventory Control System ..................... 25

5.4 RELAP5 Nodalization of the Emergency Coolant Injection System ......................... 26

6. RELAP5 SIMULATION OF LOSS OF FLOW EVENT ..................................................... 29

6.1 Modeling Assumptions Specific to Loss-of-Flow Assessment .................................. 29

6.2 RELAP5 Simulation Results and Comparison with Station Data .............................. 30

6.3 Sensitivity to Header Nodalization .......................................................................... 40

6.4 SUMMARY OF FINDINGS ..................................................................................... 41

7 . S U M M A R Y .......................................................................................................................... 4 3

v
Figures

Page

1. Simplified Schematic of the Heat Transport System ............................................................ 3

2. Integrated Testing - Governor Valve Position ................................................................... 12

3. Integrated Testing - Bleed Condenser Reflux Valve Position ............................................ 13

4. Integrated Testing - Bleed Valve Position ....................................................................... 13

5. Integrated Testing - Pressurizer Steam Bleed Valve Position ......................................... 14

6. Integrated Testing - Direct Feed Valve Position .............................................................. 14

7. Integrated Testing - Bleed Condenser Level Control Valve Position ................................ 15

8. Integrated Testing - Bleed Condenser Spray Valve Position ........................................... 15

9. Integrated Testing - Reactivity Worth of Light Water Zones .............................................. 16

10. Integrated Testing - Linear Power Error .......................................................................... 16

11. Integrated Testing - Calculated Thermal Power .............................................................. 17

12. Integrated Testing - Power Error from Fast Program ....................................................... 17

13. Integrated Testing - Log Rate Power .............................................................................. 18

14. RELAP5 Nodalization of Darlington Heat Transport System ........................................... 20

15. Darlington Fuel Channel Assembly ................................................................................. 21

16. RELAP5 Nodalization of Darlington Secondary Side ....................................................... 23

17. RELAP5 Nodalization of Darlington SG Shell Side .......................................................... 24

18. RELAP5 Nodalization of Darlington Pressure & Inventory Control System ..................... 25

19. RELAP5 Nodalization of Darlington Emergency Coolant Injection System ..................... 27

20. Darlington LOF Event - NE RIH Pressure ....................................................................... 31

21. Darlington LOF Event - NE RIH Temperature ................................................................ 32

22. Darlington LOF Event - NE ROH Pressure ..................................................................... 33

23. Darlington LOF Event - Comparison of Pressurizer and ROH Conditions ...................... 34

24. Darlington LOF Event - NE ROH Temperature .............................................................. 34

25. Darlington LOF Event - Pressurizer Pressure ................................................................ 35

26. Darlington LOF Event - Pressurizer Level ........................................................................ 36

27. Darlington LOF Event - NE SG Level ............................................................................... 36

28. Darlington LOF Event - NE SG Shell Side Pressure ....................................................... 37

29. Darlington LOF Event - North HT Bleed Valve Flow ....................................................... 38

30. Darlington LOF Event - South HT Bleed Valve Flow ...................................................... 38

31. Darlington LOF Event-Steam Pressure in Bleed Condenser ........................................ 39

32. Darlington LOF Event - Bleed Condenser Level .............................................................. 39

33. Sensitivity of NW ROH Pressure to Header Nodalization ................................................ 40

34. Sensitivity of NW ROH Outlet Mass Flow Rate to Header Nodalization .......................... 41

vi
Tables

Page

1. Darlington Loss of Flow Event Sequence ............................................................................ 7

2. Darlington Control Functions Implemented in RELAP5 .................................................... 10

3. Summary of RELAP5 Components in HT System ........................................................... 21

4. Summary of RELAP5 Components in Secondary Side ..................................................... 24

5. Summary of RELAP5 Components in Pressure and Inventory Control System ................ 26

6. Summary of RELAP5 Components in Emergency Coolant Injection System ................... 27

vii

EXECUTIVE SUMMARY

The Canadian nuclear industry is currently considering replacing its thermal-hydraulics codes, and

the USNRC codes TRACE and RELAP5, which are made available via the Code Applications and

Maintenance Program (CAMP), are potential candidates. Accordingly, AMEC NSS is assessing

these codes for their applicability to CANDU safety analysis. This assessment includes

demonstrating that the codes have the functionality required to model features unique to CANDU

reactors and their validation against data specific to CANDU reactors.

An important step towards demonstrating the applicability of the CAMP codes to CANDU reactors is

showing their ability to predict known station transients. This report presents the results of such a

comparison between a RELAP5 simulation and plant measurements for a loss-of-flow event at the

Darlington Nuclear Generating Station.

The report first provides an overview of the CANDU process and safety systems relevant to loss-offlow events at the Darlington Nuclear Generating Station (Section 2). A description of the

November 25, 1993, loss-of-flow event at Darlington Unit 4 is then presented (Section 3). To

facilitate modeling this event with the RELAP5 code, Darlington-specific control functionality was

added to the RELAP5 code, and a Darlington RELAP5 input data set was created. These

Darlington-specific code modifications are described in Section 4, together with testing that was

performed with the modified code. Test results demonstrated excellent agreement between

RELAP5 code predictions of benchmark controller responses. The Darlington RELAP5 input data

set is then described in Section 5.

Results from the RELAP5 simulation of the loss-of-flow event are presented in Section 6, showing

excellent agreement between predictions and measurements for the parameters of interest.

Section 6 also includes results of a sensitivity simulation, which demonstrates that the header

nodalization has no significant impact on the loss-of-flow simulation results.

ix

ACKNOWLEDGMENT

The authors acknowledge the substantial contributions of several individuals. At the San Piero a

Grado Nuclear Research Group (GRNSPG) of the University of Pisa, Italy: M. Cherubini, W.

Giannotti, M. Kovtonyuk, and especially Prof. F. D'Auria. At AMEC NSS, Toronto, Canada: B.

Mclaughlin and 0. Orlov.

The authors also wish to thank Mr. M. O'Neill of Ontario Power Generation for granting access to

the Darlington NGS plant data and permitting simulation of this event. The review of this report by

Mr. W.K. Liauw, Ontario Power Generation, is also gratefully acknowledged.

xi
c
ABBREVIATIONS

ASDV Atmospheric Steam Discharge Valve

CAMP Code Applications and Maintenance Program

CANDU Canadian Deuterium Uranium

CSDV Condenser Steam Discharge Valve

ECI Emergency Coolant Injection

FP Full Power

GRNSPG Gruppo di Ricerca Nucleare San Piero a Grado

HTHP Heat Transport High Pressure

HTLF Heat Transport Low Flow

HTS Heat Transport System

IWST Injection Water Storage Tank

LOF Loss of Flow

LRV Liquid Relief Valve

MCA Mechanical Control Absorber

NGS Nuclear Generating Station

PAWCS Post Accident Water Cooling System

P&IC Pressure and Inventory Control

PVM Parallel Virtual Machine

RIH Reactor Inlet Header

ROH Reactor Outlet Header

RRS Reactor Regulating System

SDS1 Shutdown System 1

SDS2 Shutdown System 2

SG Steam Generator

tc Time constant

TUF Two Unequal Fluids

xiii

1. INTRODUCTION

The Canadian nuclear industry is currently considering replacing its thermal-hydraulics codes, and

the USNRC codes TRACE and RELAP5, which are made available via the Code Applications and

Maintenance Program (CAMP), are potential candidates. Accordingly, AMEC NSS is assessing

these codes for their applicability to CANDU safety analysis. This assessment includes

demonstrating that the codes have the functionality required to model features unique to CANDU

reactors and their validation against data specific to CANDU reactors.

An important step towards demonstrating the applicability of the CAMP codes to CANDU reactors is

to show their ability to predict known station transients. This report presents the results of such a

comparison between a RELAP5 simulation and plant measurements for a loss-of-flow event at the

Darlington Nuclear Generating Station.

Simulation of this loss-of-flow event required creation of a RELAP5 model of a Darlington reactor,

as well as the incorporation of custom controller routines in the code. The latter was required

because the full functionality of the process control systems could not be captured fully with the

existing version of the code.

This report includes a description of Darlington NGS, the loss-of-flow event upon which the

RELAP5 simulation is based, the plant model and the resulting comparison of the code predictions

with actual plant measurements.

1

2. DARLINGTON PLANT OVERVIEW

The Darlington Nuclear Generating Station consists of four CANDU reactors. A simplified schematic

of the main process system for one CANDU reactor is shown in Figure 1. The process systems can

be divided into primary side and secondary side systems. The primary side covers the nuclear side

of the plant in which heavy water is circulated through the reactor core, transporting heat to the

steam generators. The secondary side covers the light water, conventional side of the plant used to

generate steam to supply the turbine.

Figure 1 Simplified Schematic of the Heat Transport System

3
2.1 Primary Side Systems

The main system on the primary side of the plant is the Heat Transport System (HTS). The HTS

circulates heavy water through 480 horizontal channels in the reactor core, each containing a

pressure tube with thirteen fuel bundles. Coolant circulation is arranged in two loops, each

supplying one half of the channels in the reactor core; Figure 1 is a simplified schematic of one

such loop. Each loop passes through the core twice and therefore has two Reactor Inlet Headers

(RIHs) and two Reactor Outlet Headers (ROHs), with the RIH in a given core pass supplying

coolant to 120 channels. Circulation is provided by four heat transport pumps, two in each loop.

Heat transported from the reactor core is rejected to the steam generators. Pressure on the primary

side of the plant is maintained by the Pressure and Inventory Control (P&IC) system. A key

component of the P&IC system is the pressurizer.

During normal operation, pressure in the HTS is controlled by adjusting the steam pressure in the

pressurizer. ROH pressure measurements that exceed the pressure setpoint cause the pressurizer

steam bleed valves to open. The valves discharge steam to the shell side of the bleed condenser,

reducing the pressurizer steam pressure and therefore the HTS pressure. Conversely, low ROH

pressure measurements cause the steam bleed valves to close, and the pressurizer heater power

to increase, increasing the steam pressure in the pressurizer.

Inventory in the HTS is controlled by maintaining the pressurizer liquid level at the power dependant

setpoint. A liquid level higher than the setpoint drives the bleed valves open and feed valves closed.

A liquid level lower than the setpoint drives the bleed valves closed and the feed valves open.

Overpressure protection is provided by Liquid Relief Valves (LRVs). The LRVs connect the HTS to

the shell side of the bleed condenser, and open if the HTS pressure exceeds the LRV opening high

pressure setpoint.

2.2 Secondary Side Systems

On the secondary side of the plant, light water is supplied to the shell side of each of the four steam

generators by the Steam Generator Feedwater system. The resulting steam is transported to the

turbine through the piping of the Main Steam Supply system.

Steam generator pressure control is achieved through the Atmospheric Steam Discharge Valves

(ASDVs) and the Condenser Steam Discharge Valves (CSDVs). The ASDVs can relieve up to 10%

of the full power steam flow to the atmosphere in the event of an increase in steam generator

pressure above the setpoint. The CSDVs can relieve up to 70% of the full power steam flow. The

CSDVs bypass the turbine and discharge directly to the condenser and are opened automatically

following a turbine trip.

2.3 Reactor Regulating System

The reactor regulating system controls the neutron power in the reactor core by manipulating either

.the light water zone controllers or the mechanical control absorbers.

Fourteen light water zone controllers are located in vertical columns throughout the reactor core.

Neutron flux is regulated by adjusting the water level in compartments within these controllers. An

increase in zone controller water level results in a higher neutron absorption rate (reduction of

reactivity) and the lowering of reactor power. A decrease in zone controller water level results in a

4
reduction in neutron absorption and an increase in reactor power.

Four Mechanical Control Absorbers (MCAs) are used for coarse control of reactor power. The four

controller rods are grouped into two banks. They are used if the power error or the light water zone

controller levels exceed predefined values. The MCAs are also inserted if a reactor stepback is

initiated.

Stepbacks (a step change reduction in reactor power) and setbacks (a ramped reduction in reactor

power) are initiated if predefined conditions are met. An example of a condition that results in a

reactor stepback is a turbine trip. An example of a condition that results in a reactor setback is a

high pressurizer level.

In the absence of a stepback, the rate at which the MCAs are driven into the core depends on the

size of the power error. In the case of a stepback, the MCAs are dropped into the core under

gravity.

2.4 Special Safety Systems

Special safety systems relevant to the loss of flow event discussed in this report are Shutdown

System 1 (SDS1) and Shutdown System 2 (SDS2). Activation of SDS1 results in the insertion of

negative reactivity into the reactor core through dropping of shutoff rods. Activation of SDS2 results

in liquid poison injection to the moderator that surrounds the reactor core. Each shutdown system

has multiple trip parameters that can initiate a reactor trip. Two trip parameters on both SDS1 and

SDS2, which protect against loss of flow events, are the Heat Transport Low Flow (HTLF) and the

Heat Transport High Pressure (HTHP) trips.

The SDS1 HTLF trip is based on flow measurements from three flow instrumented channels in each

of the four core passes. Low flow measurements from any two channels in a given core pass result

in an HTLF trip signal. The SDS2 HTLF trip logic is identical to the SDS1 trip, but it utilizes

independent flow measurement channels and instrumentation.

The SDS1 and SDS2 HTHP trips are activated when the pressure measurements on any ROH

exceed the HTHP trip setpoint.

5

3. DESCRIPTION OF LOSS OF FLOW EVENT

On November 25, 1993, a loss-of-flow event occurred at Darlington Unit 4. The event was initiated

by a switchyard transformer explosion, resulting in the loss of Class IV power to Unit 4 and a

consequential turbine trip. The Reactor Regulating System responded by initiating a reactor

stepback to 60 %FP. The HT pumps tripped because of the loss of Class IV power, significantly

reducing the coolant flow through the channels. The resulting increase in heat transport system

pressure was sufficient to activate opening of the LRVs, which provided pressure relief via

discharge to the shell side of the bleed condenser. The bleed condenser high pressure alarm was

activated shortly thereafter.

As the HT pumps continued to run down, the measured flow in the SDS channels dropped below

20.5 kg/s, the HTLF setpoint for both SDS1 and SDS2 at reactor powers greater than 70% FP. A

SDS1 trip was initiated at this time. Due to instrument sampling time, the SDS2 low flow trip was

initiated shortly afterwards. A summary of the event sequence is shown in Table 1.

Time Relative to

Event Time Loss of Class IV Power

(s)

Loss of Class IV Power 15:32:31.0 0.0

Turbine Trip* 15:32:31.0 0.0

Stepback on Turbine trip* 15:32:31.5 0.5

CSDVs open 15:32:31.8 0.8

ASDVs open 15:32:32.0 1.0

SDS1 low flow trip 15:32:33.9 2.9

Liquid Relief Valves open 15:32:34.5 3.5

SDS2 low flow trip 15:32:42.6 11.6

* Timings shown are estimates.

Table I Darlington Loss of Flow Event Sequence

In CANDU loss-of-flow safety analysis, the focus of the full system simulations is on system

response up to the time of reactor trip and shortly thereafter. This facilitates the assessment of fuel

cooling up to reactor trip time, the capacity of the interim heat sink at the time of reactor trip, and

heat transport system overpressure protection.

In the longer term, flow patterns in the horizontal channels can become difficult to predict in the

absence of forced circulation. Flows in the channels are driven by coolant density differences

7
around the circuit, and it becomes difficult to preclude flow reversal or temporary flow stagnation in

some channels. Longer term fuel cooling is conservatively assessed assuming temporary flow

stagnations in some channels.

Assessment of fuel cooling is beyond the scope of the present report, which instead focuses on the

following global system parameters:

Core flows. Flow rundown is a direct consequence of the loss of Class IV power, and indication

of low flow resulted in a reactor trip for this event. As will be discussed later, RELAP5 single

channel models of the flow instrumented safety channels are not available in the current

Darlington NGS nodalization, and flow rundown in the safety channels is therefore not directly

assessed. Instead, the header conditions that would be used to drive single channel analysis

are qualitatively assessed.1

Reactor outlet header pressures and the pressurizer level. Coolant swell caused by the flow

reduction is reflected in the pressurization transient. The heat transport high pressure trip

(based on ROH pressure measurements) provides effective back-up trip coverage fora number

of loss-of-flow events.

Steam generatorlevel. The levels provide an indication of the heat transfer between the primary

and secondary sides of the plant. The levels are also reflective of the available steam generator

inventory (i.e., interim heat sink) following the loss of Class IV power, which is a key parameter

required to mitigate the consequences of total loss of Class IV power events.

'The development of a 480 single-channel model is planned in the next phase of the program for validating

RELAP5 against data specific to CANDU reactors, in order to assess directly the flow rundown in the safety

channels.

8
4. DARLINGTON CONTROLLER MODELING WITH RELAP5

Simulation of the Darlington loss of flow event requires modeling of Darlington process control

systems. Implementation of these controllers using built-in RELAP5 basic control components in the

input data was investigated and found to be incapable of capturing all aspects of controller logic. In

particular, it was found that conditional evaluation is not a function of RELAP5, nested loops (or

simple loops) are extremely complex and difficult to interpret, and a maximum of 10,000 control

components are permitted.

Instead, the Darlington control systems were implemented by developing new controller subroutines

and compiling these with the existing RELAP5 source code. The coding was performed in a

modular fashion to facilitate integration with future releases of RELAP5. To accommodate the new

routines, some additional changes to the existing code were required to allow the new controller

data to be read and the control variables initialized. The version of RELAP5 to which the Darlington controllers have been added is RELAP5/MOD3.3 Patch 03.

4.1 New Controller Input Data

The new Darlington controller routines cover the control functions shown in

The list of modeled functions covers all controller responses required for simulation of the

Darlington November 23, 1995, loss-of-flow event. Additional controller functions that are not

required for the loss-of-flow simulation, e.g., emergency coolant injection, were also created for

future analysis of other accident scenarios.

Steam Generator Pressure Control System, comprising the following subroutines:

Live steam limiter

Steam generator pressure control

Condenser pressure o Governor valve

Re-heater steam flow

Turbine pressure

o Turbine trip

Normal mode pressure and inventory control, comprising the following subroutines:

o Bleed condenser level control

o Bleed condenser pressure spray control

o Normal mode inventory control

Normal mode pressure control

Steam Generator Feedwater Control, comprising the following subroutines:

o Feedwater pump recirculating valves o Reheater return flow

* First and second order filter subroutine for simulation of the measurement lags

* Linear interpolation subroutine

9
* Steam Generator Level Control Emergency Coolant Injection, comprising the following

subroutines:

High pressure bypass

High pressure pump startup

o High pressure recirculation valves

o Low pressure recirculation valves

o Low pressure pump start

o Steam generator cool down

Steam generator emergency cooling

* Reactor Regulating System, comprising the following subroutines: o Thermal power calculation

o Demand power routine - slow program

o Demand power routine - fast program

o Light water zone controller

Mechanical control absorber rods

o Shut-off rods withdrawal

Reactor Power Stepback, comprising the following subroutines: o Reactor SDS1 or SDS2 trip

D Turbine trip

o HT pump trip

o High HT pressure trip

[ High zone flux trip

a High power log rate

o Low pressurizer level trip o High HT pump suction pressure trip

o Low steam drum level trip

High reactor power error trip

Absorber Rods Insertion

Reactor Shut Down Systems, comprising the following subroutines:

o High HT pressure trip o Low HT pressure trip

Low HT flow trip

Low pressurizer level trip

Low steam drum level trip

o High reactor power trip

High reactor power log rate trip

Low boiler feed line pressure trip

Shut-off rods and poison injection

Table 2 Darlington Control Functions Implemented in RELAP5

Control variables in RELAP5 typically act as functions, i.e., following execution of the control action,

the control variable is assigned a value. In contrast, the CANDU control components have been

created as subroutines, thus allowing multiple outputs. The number of input/output and constant

cards depends on the requirements of the controller. The general input structure of the CANDU

control system components is such that the controller inputs are entered first, followed by outputs,

and then constants.

10
A summary of the types of input and output parameters is shown below:

Input variables

" Input parameters originating from the thermal hydraulics code

o Historical input parameters containing the state of a control variable from the previous

program execution o Input parameters originating from other control systems o Constants that define the behavioral characteristics of the control system

Output variables

o Parameters returned to the thermal hydraulics code for control

o Terms required for the next iteration of the controller, or variable demands internal to

controllers

All data passed to or from a controller is included in the controller input file. This allows each

controller to be activated in the code individually, provided that input from other controllers is

mimicked by user supplied input.

Since the CANDU control system components produce multiple outputs, output variables are

required to which the outputs are written. The CONSTANT control component can be used for this

purpose. However, to prevent confusion regarding variable "constants", a CANDUOUT component

has been created. The CANDUOUT component is identical to a CONSTANT component except in

name. The component by itself has no function except to return one of the multiple output

arguments from a CANDU control system component. The type of output is determined by its

connection to the CANDU control system component.

4.2 Valve Stroking and Delay Times

An additional feature that was created to complement the controller routines is a valve stroking and

delay time control component. The output from the Darlington controllers was filtered through the

valve stroking and delay time control component prior to assigning a valve position to the servo-type

control valves.

The valve stroking and delay time control component operates as a typical RELAP5 control

component in that it is a function with a single output. It is identified by the new control component

type VLVPOS.

4.3 Testing of New Controllers

The new controller routines have been tested in three stages, i.e., unit testing, system testing, and

integrated testing.

Unit testing of the controllers exercised individual logic sequences of the controllers. To facilitate

this, drivers were used to execute single time steps based on user supplied test data. The controller

outputs were then compared to expected controller outputs, which were calculated analytically in

EXCEL using the controller's logic and equations. Agreement was obtained between all code

predictions and analytucal calculations.

System testing of the controllers exercised complete control systems using drivers with user

11
supplied test data. The controller outputs were then compared to expected controller outputs, which

were also calculated analytically using the controller's logic and equations. Agreement was

obtained between all code predictions and analytical calculations.

Integrated testing of the controllers exercised the control systems after linking the routines with the

RELAP5 executable. Controllers were tested by comparing the RELAP5 controller responses to a

benchmark. The selected benchmark was the controller response predicted by the TUF code to a

generic single pump trip event with a turbine trip credited and reactor shutdown not credited. The

same generic transient input data was supplied to both the TUF and RELAP5 controllers, although

it is likely that the controller sampling times differed marginally.

The results of the integrated testing are summarized in Figure 2 through Figure 13. Note that the

only intent of these figures is to demonstrate that the new Darlington controllers yield the same

response as the benchmark following integration into the RELAP5 code. As shown in these figures,

there is excellent agreement between the RELAP5 and benchmark controller responses.

1.0

0.9

0.8

" 0.7

20.6

•03

0.2

o 0.4-

to 0.3

0.2-

0.1

0.0

RELAP5 controller

* Benchmark

0 5 10 15

Time (s)

20 25 30

Figure 2 Integrated Testing - Governor Valve Position

12
1.0

0.9

'00.8-

0 0.7

0

C.

w 0.6

0.5

0.4

q 0.3

0.2

0.1 1

0.0

RELAP5 controller

F Benchmark

0 5 10 15 20 25 30

Time (s)

Figure 3 Integrated Testing - Bleed Condenser Reflux Valve Position

1.0

0.9

0.8

0.7

c0.6

R 0.5

> 0.4

" 0.3

0.2

0.1

0.0

0 10 15 20 25

Time (s)

30

Figure 4 Integrated Testing - Bleed Valve Position

13
1.0

0.9

0.8

ro0.7

:• 0.6

0.5

0.4

E

0.3

0.2

0.1

0.0

RELAP5 contrllner

M Benchmark

0 5 10 15

Time (s)

20 25 30

Figure 5 Integrated Testing - Pressurizer Steam Bleed Valve Position

1.0

0.9

0.8

0.7

: 0.5

• 0.4

._• 0.3-

0

0.2

0.1

0.0

- RELAP5 controller 7 * Benchmark

~-uu~m.-u:u----------- ---------- u*u*

0 5 10 15

Time (s)

20 25 30

Figure 6 Integrated Testing - Direct Feed Valve Position

14
1.0

0.9

2 0.8 0 - RELAP5 controller

0

K0.7 a Benchmark

m 0.6

r 0.5

. 0.4

S0.3

0.2

0.1

0.0

0 5 10 15 20 25 30

Time (s)

Figure 7 Integrated Testing - Bleed Condenser Level Control Valve Position

1.0

0.9

0 RELAP5 controller

0.8 Benchmark

. 0.7

0.6

0.5

0.4

0.3

- 0.2 -

0.1 -

0.0 1

0 5 10 15 20 25 30

Time (s)

Figure 8 Integrated Testing - Bleed Condenser Spray Valve Position

15
0.0

,E

-0.5

0

B -1.0 -RELAP5 controller

OL N Benchmark

0

-2-1.5

-2.5

Time (s)

Figure 9 Integrated Testing - Reactivity Worth of Light Water Zones

1.6

1.4 0

1.2

• 0.8

0 0.6 RELAP5 controllers

C/ N Benchmark

0.4

0.2

0.0

0 5 10 15 20 25 30

Time (s)

Figure 10 Integrated Testing - Linear Power Error

16
1.0 1.0 .inmeýn ~ u u

0.8

m

0.6

0.4

0.2

- RELAP5 controller

a Benchmark

0.0

0 5 10 15 20 25

Time (s)

Figure 11 Integrated Testing - Calculated Thermal Power

30

0.18

0.16

0.14

•- 0.12

E

0 0.10

0.

E 0.08

0.06

n 0.04

0.02

0.00

0 5 10 15 20 25

Time (s)

Figure 12 Integrated Testing - Power Error from Fast Program

30

17
0.015

0.010L

0.005

--•0.0001

5 10 15 20

-0.005 - RELAP!

0 Benchr

-0.010

Time (s)

Figure 13 Integrated Testing - Log Rate Power

18
5. THE RELAP5 MODEL OF DARLINGTON

The Darlington RELAP5 model has been created primarily from plant reference data. The model

comprises the heat transport system, the steam generator feedwater and steam supply systems,

the pressure and inventory control system, and the emergency coolant injection system.

For each of these systems, standalone models were created and tested separately in null transients

with appropriate boundary conditions supplied to represent interfacing systems. This permitted

assessment of the response of each system without feedback from the interfacing systems. The

systems were also combined in the overall Darlington RELAP5 input model and tested with a null

transient.

5.1 RELAP5 Nodalization of Heat Transport System

The RELAP5 nodalization of the Darlington heat transport system (HTS) is shown in Figure 14. The

model consists of the tube side of all four steam generators, the four HTS pumps, the four RIH and

four ROH, feeders, channels, end-fittings and associated connecting piping.

Modeling of the channels in each core pass was done by parallel averaging of the channel flow

paths in a given core pass. Therefore, each core pass is represented by one or more aggregate

average channel. For three of the four core passes, one aggregate channel is used per pass

representing the 120 channels comprising each pass. For the core pass originating from the

Northwest (NW) RIH, the channels have been subdivided into six groups, each containing 20

channels. The core pass originating from the NW RIH therefore consists of six aggregate channels.

The RELAP5 model includes a detailed nodalization of the fuel channel end-fittings. Each channel

in the core has two end-fittings, connecting the individual feeders from the headers to the channel

(Figure 15). The end-fittings are designed to allow on-power fueling of the reactor.

In the reference RELAP5 nodalization, the heat transport system RIHs and ROHs have been

divided into six and seven nodes respectively. In typical Darlington safety analysis, single nodes

are used to represent reactor headers. The effect of modeling the headers by six nodes rather than

one node was assessed through a sensitivity study to be discussed later in this report.

Darlington uses 37-element fuel bundles. Each bundle consists of a center fuel element surrounded

by three concentric rings of fuel elements referred to as the inner ring, intermediate ring, and outer

ring. Each Darlington fuel string is therefore represented by four heat structures for the center pin,

inner, intermediate and outer rings, respectively. Each heat structure is divided into axial mesh

points to allow power to be distributed along the length of the fuel string. An additional heat

structure with multiple mesh points is included to simulate the presence of the pressure tube,

calandria tube and the gap between the pressure and calandria tubes. The external surface of the

calandria tube is in contact with the moderator fluid, which has a bulk temperature fixed at 700C.

Heat structures are also associated with the feeders, headers, other connective piping above the

headers, and steam generator U-tubes. The number of heat structures and other hydraulic

components in the RELAP5 model of the heat transport system is summarized in Table 3.

19
PU7 P 48NW P1 7)ý 1

I 5- 755 PEER C N0 (8 79

6)NE 714 7 7.88 NW

R,,757 TBED E 7, 77 7. 3 -ý N 8.82717437M KM RM IE( ) •HROM

1__1_E"

LIN E (1il )0 1 720

7723 72 8 724 "3

77. V• 7" 12

764 1Z 7O 31 e3

WEST 80EAST

TUBE SO, 8W SESGT0E

7 1

Figure 14RELAP5 Noalizatio ofarinto HetTrnpotSytm

. 1_ 12

.0w 029 .

,°, : ' o';• m ° I 824 US° 13 932 - ,827

WE ,ST .6 

." so 

V., 

.5B 87ý EAST,,,

SOUT "C

Figur 14 R L P Bo aiaioEfD rigo ea rnp r yt m

,IH "SS7 . TO BEO 2S
CHANNEL CLOSURE 8 SHIELDING SLEEVE

2 FEEDER COUPLING I Jk-LET.qEIELDPLUG

3 LINER TUBE II OUTBOARD END FITTING SLEEVE

4 END FITTING BODY BEARING AND JOURNAL

5 CHANNEL ANNULUS BELLOWS 1t SPACER SLEEVE

6 FUELLING MACHINE-SID0 1UBESHEET 12 SPLIT (INBOARDI END FITTING SLEEVE

7 ENO SHIELD SHIELDING BALLS acARING AND JOURNAL

CHANNEI. " BED' SIDE

CHANNEL FLOATiNG' SIDE

13 TUBE SPACER 19 LAICH ASSEMBLY

14 PRESSURE TUBE 20 OUTLET SHIELD PLUG

I5 FUEL BUNDLE 21 LINER TUBE 16 CALANDRIATUR[, 22 SHIELDING SLEEVE

IT CALANoRIA-SIOC "UBESIHEET 23 POSITIONING ASSEMBLY YOKES

lB ENDSHIFLD|AI'IICE "uISE 24 POSITIONING ASSEMBLY STUD

Figure 15 Darlington Fuel Channel Assembly.

Parameter Value

No. of hydrodynamic components 250

No. of volumes 1674

No. of junctions 1721

No. of heat structures 2470

No. of mesh points 31467

Table 3 Summary of RELAP5 Components in HT System

21
5.2 RELAP5 Nodalization of the Secondary Side

The secondary side RELAP5 model covers the following main components:

" the steam generator feed water system from the deaerator up the shell side of the steam

generators;

" the shell side of the steam generators; and

" the steam supply system from the shell side of the steam generators up to the turbine.

The RELAP5 nodalization of the secondary side is shown in Figure 16.

The starting point, i.e., the deaerator, and the ending point, i.e., the turbine, are both modeled as

boundary conditions through time-dependent volume components. The main steam generator feed

water pump component in the RELAP5 nodalization represents three parallel pumps. The pumps

are modeled using the time-dependent junction component in RELAP5, where the flow rate-head is

specified instead of the "time-dependent variable" relationship.

The High Pressure (HP) Heaters immediately downstream of the feedwater pumps are modeled

using heat structures with a user defined power source to increase the feedwater temperature to

approximately 1770C prior to entering the steam generators.

Feedwater flow to each steam generator (SG) is regulated by level control valves. Three such

valves are shown in the RELAP5 nodalization for each SG, i.e., two large valves and one small

valve. Under normal operation, only one large valve is used to regulate feedwater flow.

Operation of the small valves would only occur at low reactor powers.

The shell side of the NE steam generator is shown in Figure 17. Appropriate heat structures have

been added to the hydraulic part of the SG nodalization. In particular, heat transfer can occur

through:

" the shroud between the downcomer and the preheater, and between the downcomer and

the upper and lower boiling regions;

" the divider plate between the preheater and the lower boiler region; and

" the tubesheet between the preheater and the SG outlet bowl, and between the lower boiling

region and the SG inlet bowl.

Heat losses to the environment from the external vessel of the SG are set to zero by representing

the vessel as an insulated heat structure.

The number of heat structures and other hydraulic components in the RELAP5 model of the

heat transport system is summarized in Table 4.

22
Figure 16 RELAP5 Nodalization of Darlington Secondary Side

23
- ~

I C)fJ

191 i ý3 5

RE/tEA4 MR

LIRINS

-- S61CS

131l

Figure 17 RELAP5 Nodalization of Darlington SG Shell Side

Parameter Value

No. of hydrodynamic components 250

No. of volumes 1475

No. of junctions 1503

No. of heat structures 372

No. of mesh points 4986

Table 4 Summary of RELAP5 Components in Secondary Side

24
5.3 RELAP5 Nodalization of the Pressure and Inventory Control System

Pressure and inventory control of the heat transport system is performed through adjustments of the

pressure and level in the pressurizer. Of secondary importance to heat transport system conditions

is the bleed condenser. Hot bleed flow from the heat transport system is discharged to the shell

side of the bleed condenser. The hot bleed flow flashes as it passes through the bleed valves,

choking the flow. Due to the choked flow, variation of the conditions in the shell side of the bleed

condenser does not have a significant effect on heat transport system conditions during

overpressure transients such as occur for loss of flow events.

Figure 18 RELAP5 Nodalization of Darlington Pressure & Inventory Control System

The RELAP5 nodalization of the pressure and inventory control system is shown in Figure 18. The

pressurizer is shown on the right side of the figure. The bleed condenser shell and tube sides are

shown on the left and middle portion of the diagram, respectively. Beginning from the piping exiting

the shell side of the bleed condenser, other components in the model are the bleed cooler,

purification system, D20 storage tank, D20 feed pump(s), and the associated control and check

valves.

25
The bleed valves are relevant to the prediction of conditions in the heat transport system during a

loss-of-flow induced overpressure transient. Modeling of the choked flow through these valves has

been performed based on the manufacturer supplied valve recovery coefficient.

The RELAP5 code does not support the recovery coefficient method of predicting choked flow rates

in control valves. Rather the default criterion for determining choking is the Henry-Fauske

correlation, which did not capture choking in the bleed valves during preliminary simulations. As a

consequence, the flow rates through the bleed valves were originally overpredicted. Therefore, the

valve recovery coefficient method was implemented in the input data to adjust the RELAP5

predicted flows. A series of control variables, single volumes and time dependent junctions was

used for this purpose.

The number of heat structures and other hydraulic components in the RELAP5 model of the

pressure and inventory control system is summarized in Table 5.

Parameter Value

No. of hydrodynamic components 150

No. of volumes 358

No. of junctions 362

No. of heat structures 40

No. of mesh points 200

Table 5 Summary of RELAP5 Components in Pressure and Inventory Control System

5.4 RELAP5 Nodalization of the Emergency Coolant Injection System

Loss-of-flow assessment at Darlington does not require the use of the Emergency Coolant Injection

(ECI) System. As such, the RELAP5 model implemented for ECI is intended for application in future

analysis and will not be discussed in detail in this report. The RELAP5 nodalization of the

emergency coolant injection system is shown in Figure 19.

26
Figure 19 RELAP5 Nodalization of Darlington Emergency Coolant Injection System

In the event of ECI initiation, the coolant water supply is initially drawn from the Injection Water

Storage Tank (IWST) shown on the left hand side of the figure. The flow passes through three low

pressure pumps and two high pressure pumps prior to injection into each of the headers in the

affected reactor.

Following depletion of the IWST inventory, ECI operation transitions to a recovery mode in which

the injection water is drawn from sumps in the reactor vault. This mode of operation is not currently

modeled and would require the representation of the Post Accident Water Cooling System

(PAWCS).

The ECI model does not contain any heat structures. Other hydraulic components are summarized

in Table 6.

Parameter Value

No. of hydrodynamic components 98

No. of volumes 444

No. of junctions 452

Table 6 Summary of RELAP5 Components in Emergency Coolant Injection System

27

6. RELAP5 SIMULATION OF LOSS OF FLOW EVENT

The loss-of-flow event described in Section 3 has been simulated using the RELAP5 Darlington

controllers (Section 4) and Darlington input model (Section 5). Additional assumptions specific to

the loss of flow simulation, and the simulation results, are presented below.

6.1 Modeling Assumptions Specific to Loss-of-Flow Assessment

The assumptions applied in this assessment can be divided into two groups, i.e., those required to

define the initial operating state, and those required to define the equipment credits following the

initiating event.

The initial operating state has been defined based on the station data. To achieve this operating

state, the following assumptions were applied to the holding transient:

" The reference 100% reactor thermal power to the coolant is 2651 MW. The indicated reactor

power prior to the event was 99.8% full power.

" There is no flux tilt across the reactor core, i.e., the flux is distributed symmetrically.

" The steam generator pressure control is in hold mode with an average steam generator

pressure of 5.05 MPa(g).

" The reactor outlet header pressure control is in normal mode with a setpoint of 9820 kPa(g).

" The initial bleed valve position has been adjusted to yield an average total bleed flow from the

heat transport system of 12.2 kg/s.

" The temperature of the feedwater in the deaerator is 1530C.

" All four heat transport pumps are operating.

" Three steam generator feedwater pumps are operating.

" One D20 feed pump is operating.

" All process systems are functioning normally.

Following the initiating event, the plant response depends largely on which pieces of equipment

continue to operate following the loss of Class IV power. Class IV power is one of four classes of

power in the Power Distribution System in CANDU Nuclear Plants, each with varying degrees of

reliability and continuity. Class IV power is supplied from the operating unit or the bulk electrical

grid. Of relevance to the present assessment, Class IV power supplies both the main heat transport

pumps and the steam generator feedwater pumps. Under normal operation, Class IV power also

supplies Class III power, which powers some equipment relevant to the loss-of-flow event

simulation. Loads on Class Ill power can be interrupted for several minutes without affecting plant

safety. From an examination of plant data, it appears that Class III power was restored within a few

seconds of the initiating event.

29
The above considerations were used to inform assumptions regarding post-event equipment

credits. Assumptions for simulation of the loss-of-flow event are as follows:

" A turbine trip occurs at time zero of the event.

" All four heat transport pumps trip at time zero of the event.

" All three steam generator feed water pumps trip at time zero of the event.

" The auxiliary steam generator feed pumps are started on Class III power 2 seconds after the

initiating event.

" The operating D20 feed pump is tripped at time zero of the event due to the temporary

interruption in Class Ill power. Based on the bleed condenser spray flow trend in the station

data, pump restart did not occur until about 85 s into the event. Since the period of interest in

the present simulation is 0 s to 80 s, the D20 feed pumps remain unavailable for the duration of

the simulation.

" A reactor stepback occurred at 0.5 s due to the turbine trip. Previous work related to this event

specified a window of 0 to 1 s within which the stepback could have occurred.

" A reactor trip occurred at 2.9 s due to low flow in the flow instrumented safety channels. Models

of individual flow instrumented safety channels have not been created. As such, the reactor trip

time based on station data has been imposed on the RELAP5 model.

" The ASDVs are available and function normally during the event.

" The CSDVs are available and function normally during the event until the condenser vacuum is

lost.

Based on station data, the condenser vacuum is lost causing the CSDVs to close at about

13.5 s. The condenser is not modeled in the RELAP5 input data. However, based on

considerations of the CSDV control logic, the valve response was approximated by initiating

forced closure of the CSDVs at 13.5 s and stroking the valves closed over a 2 second period.

The bleed condenser level is controlled by the bleed condenser level control valve. Under

normal operation, this valve is controlled only on the bleed condenser level error. However, a

temperature override exists to close this valve in the event of high temperature readings

downstream of the bleed cooler. Based on the station data, the temperature override was not

activated during the loss-of-flow event. As such, the temperature override was masked in the

RELAP5 simulation.

6.2 RELAP5 Simulation Results and Comparison with Station Data

The RELAP5 predictions of thermal-hydraulic parameters for the loss-of-flow event are shown in the

figures below. Where available, station data is also shown for comparison. In all figures, event

initiation is time zero.

30
Due to the loss of all four heat transport pumps, the loss-of-flow event affects each of the four heat

transport system core passes equally. Similarly, the response of each of the four steam generators

is essentially the same. As such, the discussion that follows focuses on the thermal-hydraulic

parameters associated with one core pass, that originating from the North East (NE) RIH and the

associated downstream ROH. The response of the steam generator upstream of the NE RIH will

also be presented, i.e., the NE SG.

Figure 20 shows the response of the NE RIH pressure to the heat transport pump trips.

Immediately following the heat transport pump trip at time zero, the header, which is downstream of

the heat transport pumps, drops in pressure. However, the initial pressure drop is quickly

terminated due to the mismatch between the power to the coolant and the heat transported out of

the core by the coolant. Therefore, about 1 s into the transient, the header pressure trend

temporarily reverses, increasing in pressure. At 2.9 s into the transient, the reactor is tripped on

heat transport low flow and the neutron power drops sharply. The drop in thermal power to the

coolant lags the drop in neutron power, but within the first 5 s the reduction is sufficient to

depressurize again the NE RIH due to coolant shrinkage.

-12

11.5

10.511

wi 10 Z

9.5

9

8.5

-20 -10 0 10 20 30 40 50 60 70 80

Time (s)

Figure 20 Darlington LOF Event - NE RIH Pressure

The corresponding temperature in the NE RIH is shown in Figure 21. The time constants

associated with temperature measurement instrumentation can vary widely, depending on the

instrument type and location. Quantification of instrument specific time constants for use in the

temperature comparisons were not performed as part of this assessment. Instead, RELAP5

temperature predictions have been filtered using two different time constants (tc) to demonstrate

the sensitivity of the measured parameter to the instrumentation time constant.

The temperature in the NE RIH is dictated in large part by the temperature on the secondary side of

31
the upstream steam generator. For this event, there is an increase in the upstream steam generator

pressure and therefore temperature. The increase in SG temperature is reflected in the increasing

NE RIH temperature.

RELAP5

- RELAP5, tc = 8.7 s

310 - RELAP5, tc = 17.4s

3 Station

U 300

290 LUw

280

E

EJ 270

260

-20 -10 0 10 20 30 40 50 60 70 80

Time (s)

Figure 21 Darlington LOF Event - NE RIH Temperature

The ROH pressure downstream of the NE RIH is shown in Figure 22. For historical reasons, this

outlet header is labeled the NE ROH. The immediate drop in pressure observed in the RIH does not

occur in the ROH, due in part to the saturated coolant conditions in the header. Instead, the

pressure trend in the ROH, which is directly downstream of the reactor core, is dominated by the

mismatch between heat input to the core and heat transport from the core. The credited HTS low

flow reactor trip at 2.9 s promptly terminates the neutron power, but because of the thermal lag, the

ROH pressure continues to rise, reaching the Liquid Relief Valve opening setpoint at 3 s. Opening

of the LRVs mitigates the HTS overpressure transient. The LRVs close again at approximately 6 s.

Within the first 5 seconds of the transient, the combination of coolant shrinkage due to a reduction

in power to the coolant, and opening of the LRVs, reverses the pressure trend and the pressure

begins to decrease.

32
o. 10.5

zO 01

9.5

9

8.5

-20 -10 0 10 20 30 40 50 60 70 80

Time (s)

Figure 22 Darlington LOF Event - NE ROH Pressure

The drop in ROH pressure is limited by the rate at which the pressurizer depressurizes. The

pressurizer is initially filled with saturated D20 and steam at equilibrium conditions and is connected

directly to the NE ROH. As the pressure drops below the initial equilibrium condition, the saturated

liquid in the pressurizer vaporizes, slowing the depressurization transient. The pressure trends for

the NE ROH and pressurizer are shown in Figure 23. Also shown in this figure is the void in the NE

ROH. The presence of void in the HT system also slows the depressurization transient as shown

between 20 and 25 s in the transient. After approximately 25 s, the void in the HT has collapsed,

and the depressurization continues, driven by the vapor pressure in the pressurizer.

Qualitatively, the overall agreement between station measurements and predictions of conditions in

the RIH and ROH is good. This provides confidence that single channel analysis of flow rundown in

the safety channels would also be in good agreement with station data if assessed.

The temperature in the NE ROH is shown in Figure 24. Three RELAP5 temperature trends are

shown, i.e., the direct RELAP5 temperature prediction, and two filtered trends which demonstrate

the effect time constants associated with instrumentation can have on the measured value. The

initial temperature increase shown in the RELAP5 prediction is caused by the mismatch between

the power to coolant and the heat transport from the core discussed previously.

33
a'

a.

I

2

w

2

a.

1

0.9

0.8

0.7

0.6

0.5

0.4

0.3

0.2

0.1

0

-20 -10 0 10 20 30 40 50 60 70 80

Time (s)

Figure 23 Darlington LOF Event - Comparison of Pressurizer and ROH Conditions

RELAP5

- " RELAPS, tc= 8.7 s

4RELAP5, Ic = 17.4

UStation

300 '

0

IM

290

LUZ

2• 280

I2E

-20 -10 0 10 20 30 40 50 60 70 80

Time (s)

Figure 24 Darlington LOF Event - NE ROH Temperature

34
Comparisons of the RELAP5 predictions and station measurements for the pressurizer pressure

and level are shown in Figure 25 and Figure 26. The RELAP5 pressure is underpredicted compared

to station measurements in the later portion of the transient. This difference is also reflected in the

ROH pressure trend presented previously in Figure 22. Quantification of measurement uncertainty

was not part of this assessment and as such a quantitative assessment of the agreement between

predictions and measurements was not performed. However, qualitatively, the overall agreement

between the predictions and measurements is good. This provides confidence that the RELAP5

code is capable of capturing pressure transients of importance to CANDU safety analysis.

As noted, the inventory in the steam generators at the time of reactor trip is of interest in CANDU

safety analysis for events in which SG feedwater is lost or interrupted. The available inventory

allows the SGs to be used as an interim heat sink until a longer term heat sink, or alternate means

of SG inventory make-up, are established. Although loss-of-flow events are not limiting with respect

to SG inventory at time of trip, comparison of SG level predictions with measurements does provide

a preliminary means of assessing the SG model. The NE SG level transient is shown in Figure 27

and is in excellent agreement with station data. The corresponding pressure in the NE SG is shown

in Figure 28. The variations in SG pressure have a second-order effect on the HTS conditions, in

that they change the shell-side temperature of the SG by approximately 10C for every 0.1 MPa

change in pressure. Heat transfer to the SGs is therefore temporarily reduced marginally with the

increase in SG pressure. The shape of the predicted SG pressure trend within the first 13.5 s is

driven in large part by the assumed response of the condenser. During this time period, steam

discharge from the SG is primarily to the condenser. The condenser was not modeled in RELAP5,

rather a condenser pressure response was assumed as described in Section 6.1. The SG pressure

response does not have a significant impact on the HT system response.

RELAPS

U Station

11.5

11

zo10.5

U,,

10,

9.5

-20 -10 0 10 20 30 40 50 60

Time (s)

Figure 25 Darlington LOF Event - Pressurizer Pressure

70 80

35
N 1

5.5

5

4.5

-20 -10 0 10 20

Figure 26 Darlington LI

15

14

V 13

'U

12

11

10

-20 -10 0 10 20

Figure 27 Darlington

30 40 50 60 70 80

Time (s)

OF Event - Pressurizer Level

30 40 50 60 70 80

Time (s)

LOF Event - NE SG Level

36
.5.2

5

CL 4.8

U,

4.6

4.4

4.2-

-20 -10 0 10 20 30 40 50 60 70 80

lime (s)

Figure 28 Darlington LOF Event - NE SG Shell Side Pressure

The response of the pressure and inventory control system is generally of secondary importance in

this loss-of-flow event. One exception to this is the response of the HTS bleed valves. These valves

connect the suction of the HTS pumps to the shell side of the bleed condenser. The shell side of the

bleed condenser nominally operates at 1.6 MPa(g). By contrast, the pressure at the HTS pump

suction is approximately 10 MPa(g) under nominal operating conditions. Therefore, the pressure

drop across the bleed valves is large and the hot HTS coolant flashes, choking the flow as it

discharges to the bleed condenser. Since the flow is choked, conditions in the bleed condenser

shell and downstream pressure and inventory control system piping have little impact on the

response of the HT system. However, the flow through the bleed lines does impact the HT system

response. The bleed flows from the north and south HTS loops are shown in Figure 29 and Figure

30. The RELAP5 calculated bleed flows overpredict the measured bleed flows. Reasons for the

overprediction include: use of the same controller for both bleed valves (CV3 and CV4); and the

symmetry of the Darlington NGS model, in contrast with the asymmetry of station data for the

measured bleed flow. This overprediction contributes to the underprediction of pressures observed

in earlier figures (Figure 22 and Figure 25).

Conditions in the shell side of the bleed condenser are shown in Figure 31 and Figure 32.

Consistent with the overprediction of bleed flow, the bleed condenser level and pressure are

overpredicted. Because of the bleed valve choking, these conditions have no significant effect on

the HTS conditions.

Overall, there is excellent agreement between RELAP5 predictions of system thermal-hydraulic

conditions and station measurements for the November 25, 1993, loss-of-flow event at Darlington.

37
= -20 -10 0 10 20 30 40 50 60 70

Time (s)

Figure 29 Darlington LOF Event - North HT Bleed Valve Flow

R..APo

UStation

80

-20 .10 0 10 20 30 40 50 60 70

Time (s)

Figure 30 Darlington LOF Event - South HT Bleed Valve Flow

80

38
4 MI I0.

3.5

3

2.5 mm

2

41 1.5

C.

0.5 -

-20 -10 0 10 20 30 40 50 60 70

Time (s)

Figure 31 Darlington LOF Event - Steam Pressure in Bleed Condenser

80

~1

co

1.8 -

1.6

1.4

1.2

1

0.8

0.6

0A4

0.2

RELAPS

StatIon

-20 .10 0 10 20 30 40 50 60 70 80

Time (s)

Figure 32 Darlington LOF Event - Bleed Condenser Level

39
6.3 Sensitivity to Header Nodalization

Sensitivity analysis was performed to assess the effect of heat transport system header

nodalization on the heat transport system response. In the sensitivity case, the headers were

represented by single nodes instead of the multiple node representation in the reference case. The

boundary conditions for the reference and sensitivity cases were identical, with only the header

nodalization changed.

The sensitivity analysis was performed on an earlier version of the simulation than that presented in

Section 6.2. As such, the boundary conditions applied to both the reference and sensitivity case

were not identical to those applied in the analysis documented in Section 6.2. Modeling of the flows

exiting the shell side of the bleed condenser differed from the assessment in Section 6.2. Since the

shell side of the bleed condenser is essentially decoupled from the heat transport system due to

choking of the bleed valves, the difference has no significant impact on the heat transport system

response. All other boundary conditions were identical to those used for the analysis in Section 6.2.

Therefore, the predicted difference in heat transport system response between the reference and

sensitivity analysis results is representative of the expected difference that header nodalization

would have on the predictions presented in Section 6.2.

The results of the sensitivity and reference runs are shown in Figure 33 and Figure 34. In these

figures, event initiation time is 1500 s, run D1 is the reference simulation with the multiple node

headers, and run SA3 is the sensitivity simulation with the single node headers.

1. le+07

1.05e+07

0-.

1)-

le+07

9.5e+06

-p-882010000 -D1 (NW ROH 882)

..... -p-885010000- ...... DI (NW R0H 885) ... ..

0--0 p-881010000 - SA3 (NW RON)

! ....... .... ......... ........ .............. ............. ....... ..... ...i ............. i. ..........

i....... ...i....... i . ......... i . i ............. ... ......... • ..... .... ...... : ; .. .......... • .... ................. 9e+06

1500 1520 1540 1560

Time (s)

1580 1600

Figure 33 Sensitivity of NW ROH Pressure to Header Nodalization

40
2000 .

- mflowj-882020000 - D1 (pipel to NE boiler) -(............ Dl (pipe1 to NE boiler .

mflowj-881070000 - SA3 (pipe 1 to NE boiler)

1500.......... . 5nflowj-881080000 - SA3 (pipe2 to NE boiler)

•_•• ~ . . .. . ........ .. .. .. .. . . . .... ... .............. . . . .:

r~ 1000. ............. ..

Ca

c• 0 t .. . . ...... .. . ......... . ......... ... . .. ..... .............. . ... !. ... ........ ....... ................

0 '

L;",

1500 1520 1540 1560 1580 1600

Time (s)

Figure 34 Sensitivity of NW ROH Outlet Mass Flow Rate to Header Nodalization

As shown in Figure 33, header nodalization has no significant effect on the predicted header

pressure for the analyzed loss of flow event.

With respect to flow predictions, Figure 34 demonstrates that the results are also not significantly

affected by header nodalization. Marginal differences are noticeable between approximately 0 s to

1 s, 8 s to 9 s, and 15 s to 25 s into the transient. During these times, the reference nodalization

exhibits small asymmetries between flow predictions in the two parallel pipes connecting the header

to the steam generator inlet. Use of a single node representation produces flow predictions that are

averages of the asymmetric flows in the reference case. The differences are small and have no

significant impact on the overall transient simulation results.

Therefore the loss-of-flow results are not sensitive to header nodalization.

6.4 SUMMARY OF FINDINGS

Overall excellent agreement was obtained between the RELAP5 predictions and the station

measurements presented in Section 6.2. Of the three thermal-hydraulic parameters of interest to

the present assessment (Section 3), the header pressures and steam generator levels compared

well to station data. Assessment of core flows could not be performed due to the absence of single

channel RELAP5 models. However, the overall qualitative agreement between the station

measurements and the predicted system response provides confidence that core flows would

likewise compare favorably if assessed.

41

7. SUMMARY

A detailed RELAP5 system model of one of the operating units at the Darlington Nuclear

Generating Station has been developed based on reference plant data. The model includes the

heat transport system, feedwater system and steam supply system (comprising the secondary

side), pressure and inventory control system, and the emergency coolant injection system.

Controllers specific to Darlington have also been developed in modular fashion to allow these to be

easily compiled with the base RELAP5 code. The controllers relevant to a loss-of-flow assessment

have been tested against benchmark controller responses and found to be in excellent agreement.

The loss-of-flow event at Darlington on November 25, 1993, was simulated with the RELAP5

Darlington model. The simulation results are not sensitive to header nodalization. Good agreement

was obtained between station measurements and the RELAP5 predictions.

43

NRC FORM 335 U.S. NUCLEAR REGULATORY COMMISSION 1. REPORT NUMBER

(9-2004) (Assigned by NRC, Add Vol., Supp., Rev.,

NRCMD 3.7 and Addendum Numbers, If any.)

BIBLIOGRAPHIC DATA SHEET NUREG/IA-0247

(See instructions on the reverse)

2. TITLE AND SUBTITLE 3. DATE REPORT PUBLISHED

RELAP5 Simulation of Darlington Nuclear Generating Station Loss of Flow Event MONTH YEAR

Febrii-ar'V 2011

4. FIN OR GRANT NUMBER

5. AUTHOR(S) 6. TYPE OF REPORT

Douglas Naundorf (AMEC NSS) Alessandro Petruzzi (GRNSPG) Technical

Jianghui Yin (AMEC NSS) Andriy Kovtonyuk (GRNSPG)

7. PERIOD COVERED (Inclusive Dates)

8. PERFORMING ORGANIZATION - NAME AND ADDRESS (If NRC. provide Division, Office or Region, U.S. Nuclear Regulatory Commission, and mailing address; if contractor,

provide name and mailing address.)

AMEC NSS Limited Nuclear Research Group San Piero A Grado

700 University Avenue University of Pisa

Toronto, Ontario, Canada Pisa, Italy

9. SPONSORING ORGANIZATION -NAME AND ADDRESS (if NRC. type "Same as above; if contractor, provide NRC Division, Office or Region, U.S. Nuclear Regulatory Commission,

and mailing address.)

Division of Systems Analysis

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, D.C. 20555-0001

10. SUPPLEMENTARY NOTES

A. Calvo, NRC Project Manager

11. ABSTRACT (200 words orless)

The Darlington NGS consists of four 940 MWe CANDU reactors. A detailed RELAP5 model of a Darlington NGS reactor

has been created for use with RELAP5/MOD3.3. The model includes the primary heat transport system, the steam

generator feedwater and main steam supply systems, the pressure and inventory control system, and the emergency

coolant injection system.

The modeling of Darlington required the creation of new subroutines for RELAP5 to describe the functionality of the

process control systems and emergency control systems with sufficient accuracy. These controller subroutines are

standalone modules that are compiled and linked together with the RELAP5 source to facilitate integration with future

revisions to the code.

The November 25, 1993, loss-of-flow event at Darlington was simulated using this RELAP5 model. Excellent agreement

was obtained between the RELAP5 predictions and station measurements.

12. KEY WORDS/DESCRIPTORS (List words or phrases that will assist researchers In locating the report.) .13. AVAILABILITY STATEMENT

Darlington NGS unlimited

CANDU reactors 14. SECURITY CLASSIFICATION

RELAP5 (This Page)

San Piero a Grado Nuclear Research Group (GRNSPG) unclassified

University of Pisa, Italy (This Report)

AMEC NSS unclassified

Ontario Power Generation 15. NUMBER OF PAGES

Code Application Maintenance Program (CAMP)

thermal-hydraulics codes 16. PRICE

TRACE

NRC FORM 33S (9-2004) PRINTED ON RECYCLED PAPER
APrined Jonreycled,:

Federal Recycling Program

NUREGIIA-0247 RELAP5 Simulation of Darlington Nuclear Generating Station Loss of Flow Event February 2011

UNITED STATES

NUCLEAR REGULATORY COMMISSION

WASHINGTON, DC 20555-0001

OFFICIAL BUSINESS