NUREG/IA-0248

International

Agreement Report

Post-Test Analysis of Hot Leg 2x25%

Break at PSB-VVER Facility using

TRACE V5.0 Code

Prepared by:

P. Heralecky, M. Blaha

TES Ltd

Prazska 597

674 01 Trebic, Czech Republic

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
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NUREG/IA-0248

International

Agreement Report

Post-Test Analysis of Hot Leg 2x25%

Break at PSB-VVER Facility using

TRACE V5.0 Code

Prepared by:

P. Heralecky, M. Blaha

TES Ltd

Prazska 597

674 01 Trebic, Czech Republic

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 best estimate thermo-hydraulic computer code TRACE V5.0 has been assessed using Hot

Leg 2x25% break experiment at the large-scale test facility PSB WER. The PSB-WER facility

is a 1:3000 volume scaled model of WER 1000 NPP located in Electrogorsk, Russia. An

extensive TRACE input deck of PSB-WER facility was developed including all important

components of the PSB-WER facility: reactor, 4 separated loops, pressurizer, break unit, main

circulation pumps, steam generators, break section and important parts of secondary circuit.

The TRACE (TRAC/RELAP Advanced Computational Engine) is the latest in a series of

advanced, best-estimate reactor systems codes developed by the U.S. Nuclear Regulatory

Commission in frame of CAMP (Code Application and Maintenance Program). The TRACE

code is a component-oriented reactor systems analysis code designed to analyze light water

reactor transients up to the point of significant fuel damage. The original validation of the

TRACE code was mainly based on experiments performed on experimental facilities of typical

PWR design. There are some different features of WER design comparing to PWR. Therefore

the validation of the thermo-hydraulic codes for VVER types of reactors is required by

regulators. The purpose of performed analysis is to extend the validation of the TRACE code

focused on VVER type of NPP's. This work was performed in the frame of R&D project

sponsored by The Ministry of Industry and Trade of Czech Republic.

iii

CONTENTS

Pane

ABSTRACT ..................................................................................................................... iii

ACKNOW LEDG EM ENT ................................................................................................. vii

ABBREVIATIO NS ........................................................................................................... ix

1. Introduction .............................................................................................................. 1-1

2. Facility and test descriptions ................................................................................... 2-1

2.1 PSB-W ER Facility ......................................................................................................... 2-1

2.2 Experiment HL-2x25-2 ................................................................................................... 2-4

2.2.1 Facility configuration ........................................................................................... 2-4

2.2.2 Initial Conditions ................................................................................................. 2-4

2.2.3 Boundary Conditions (test scenario) .................................................................. 2-5

3. The TRACE Code ................................................................................................... 3-1

3.1 The TRACE code assessment ....................................................................................... 3-1

3.2 W ER typical features related to TRACE code assessment .......................................... 3-1

4. Input deck description ............................................................................................. 4-1

5. Results .................................................................................................................... 5-1

5.1 Steady-state calculation ................................................................................................. 5-1

5.2 Transient calculation ...................................................................................................... 5-2

5.3 Nodalisation study - comparison of 1D x 3D model results .......................................... 5-19

6. Run Statistics .......................................................................................................... 6-1

7. Conclusions ............................................................................................................ 7-1

8. References ............................................................................................................. 8-1

V
Appendices

APPENDIX A: Input Deck Nodalisation Schemes ............................................................ A-1

APPENDIX B: Measurement Localisation at PSB-W ER Facility .................................... B-1

Figures

Page

1. General View of PSB-W ER Facility ..................................................................................... 2-1

2. Principal Scheme of PSB-W ER Large Break Unit ............................................................... 2-3

3. Primary Pressure ................................................................................................................... 5-6

4. Primary Pressure (detail) ....................................................................................................... 5-6

5. Fuel Cladding Temperature (Top of the Core) ...................................................................... 5-7

6. Pressurizer Level ................................................................................................................... 5-7

7. HPIS + LPIS Flow (Boundary Condition) .............................................................................. 5-8

8. MCP Rotor Speed (Boundary Condition) .............................................................................. 5-8

9. Hydroaccumulators Levels .................................................................................................... 5-9

10. Fuel Rod Simulator Power (Boundary Condition) ................................................................ 5-9

11. Secondary Side Pressures ................................................................................................. 5-10

12. Steam Generators Levels .................................................................................................. 5-10

13. Broken Loop 3 Temperatures (Inlet of SG3) ..................................................................... 5-11

14. Broken Loop 3 Temperatures (Loop Seal) ......................................................................... 5-11

15. Intact Loop 4 Tem peratures (Inlet of SG3) ......................................................................... 5-12

16. Intact Loop 4 Temperatures (Loop Seal) ........................................................................... 5-12

17. Pressure Differences DP01-DP04 (Downcomer) ............................................................... 5-13

18. Pressure Differences DP05-DPO8 (Lower Plenum + Lower Part of FRS) ......................... 5-14

19. Pressure Differences DP09-DP12 (FRS + Lower Part of Upper Plenum) ......................... 5-15

20. Pressure Differences DP13-DP16 ( Upper Part of Upper Plenum ) .................................... 5-16

21. Break Flow ......................................................................................................................... 5-17

22. Primary Circuit Flow Balance ............................................................................................. 5-17

23. Reactor Collapsed Level .................................................................................................... 5-18

24. Temperature at the Break Donor Cell (From the UP Side) ................................................ 5-20

Tables

Page

1. Main Operational Characteristics of PSB-W ER Comparing to W ER-1000 ........................ 2-2

2. Test Facility Configuration in HL-2x25-02 Test ..................................................................... 2-4

3. Main Operational Characteristics of PSB-W ER Facility ....................................................... 2-4

4. Main Events During HL-2x25-2 Test ..................................................................................... 2-5

5. TRACE Components Statistic ............................................................................................... 4-2

6. Initial Conditions (TRACE calculation vs. experiment comparison) ...................................... 5-1

7. Chronology of main events (TRACE calculation vs. experiment comparison) ...................... 5-2

8. Run statistics ......................................................................................................................... 6-1

vi
ACKNOWLEDGEMENT

The project FI-IM5/150 was funded by The Czech Ministry of Industry and Trade in frame of

R&D activities.

vii

ABBREVIATIONS

ACC Hydro-accumulator

CL Cold Leg

CS Core Simulator

DC Downcomer

ECCS Emergency Core Cooling System

EREC Electrogorsk Research and Engineering Institute

FRS Fuel Rod Simulator

HL Hot Leg

HPIS High Pressure Injection System

IE Initiating Event

ITF Integral Effect Test Facility

LOCA Loss of Coolant Accident

LP Lower Plenum

LPIS Low Pressure Injection System

MCP Main Coolant Pump

MCP Main Coolant Pump

NEA Nuclear Energy Agency

NPP Nuclear Power Plant

NRC Nuclear Regulatory Commission

OECD Organization for Economic. Co-operation and Development

PKL Experimental Facility in Finland

PRZ Pressurizer

PWR Pressurized Water Reactor

R&D Research and Development

RPV Reactor Pressure Vessel

RPV Reactor Pressure Vessel

SCRAM Emergency Shutdown of a Nuclear Reactor

SETF Separate Effect Test Facility

SG Steam Generator

TC Thermo Couple

TH Thermal Hydraulic

UP Upper Plenum

WER Russian Pressurized Water Type Reactor

ix

1. INTRODUCTION

The assessment of PWR safety codes is mainly performed on the basis of experimental data

coming from scaled-down integral or separate test facilities. The TRACE code validation

process was mainly based on the data from experimental facilities or real NPP's of Western

PWR type as well. There is significant number of VVER type of reactors operating all over the

world and many other under construction or under preparation as well. VVER reactors are in

many aspects similar to Western PWR's. Therefore a lot of experimental data measured on

PWR's or PWR based test facilities are valuable also for VVER research. On the other hand,

the VVER design has several specific features such as larger volumes of primary coolant,

horizontal steam generators, different ECCS injection points and so on. Therefore the validation

of the thermo-hydraulic codes for VVER types of reactors is often required by national

regulators. The purpose of performed analysis is to extend the validation of the TRACE code

focused on VVER type of NPP's. The best estimate thermo-hydraulic computer code TRACE

V5.0 has been assessed using Hot Leg 2x25% break experiment at the large-scale test facility

PSB VVER. The PSB-VVER facility is a 1:3000 volume scaled model of VVER 1000 NPP

located in Electrogorsk, Russia. In order to perform code validation an extensive TRACE input

deck of PSB-WER facilty was developed including all important components of the PSB-VVER

facility: reactor, four separated loops, pressurizer, break unit, main circulation pumps, steam

generators, break section and important parts of secondary circuit.

1-1

2. FACILITY AND TEST DESCRIPTIONS

Detail information about the PSB-VVER test facility's systems and elements is given in Ref 4.

Only a brief description of the PSB-VVER facility is given here. The hardware configuration for

HL-2x25-02 test is reported below.

2.1 PSB-WER Facility

The PSB-VVER test facility is a large-scale integral facility, whose structure is similar to that of

the primary circuit of VVER-1000 nuclear power plant. The facility consists of four loops, each of

them connected to the reactor model. Each loop includes a circulation pump, a steam generator

model, and cold and hot legs. Volumetric - power scale of the facility is 1:300. The hot and cold

leg elevations including the loop seals of the reference plant have been reproduced in the

PSB-VVER. Figure 1 shows an isometric picture of the test facility, its main characteristics are

given in Table 1.

Figure 1: General View of PSB-WER Facility

2-1
Primary and secondary circuits of the PSB-VVER facility are operated under nominal pressure

of reactor prototype. The reactor model consists of four elements: external downcomer, core

simulator, core by-pass, and upper plenum. The facilityalso has a pressurizer (PRZ) and

ECCS.

ECCS of the PSB-VVER test facility includes, as the reactor VVER-1 000, three subsystems: a

passive and two active systems - high pressure injection system (HPIS) and low pressure

injection system (LPIS). The passive system consists of four accumulators, which are

connected in pairs to inlet and outlet chambers of the reactor. Cooling water of active ECCS can

be supplied to three loops, both to cold and hot legs.

The PSB-VVER core model consists of 168 full-height indirectly electrically heated fuel rod

simulators with uniform power distribution. The rod simulator pitch (12.75 mm) and diameter

(9.1 mm) are identical to those of the reference reactor. The fuel rod simulators are arranged on

a triangular grid. The rod bundle cross section has the shape of regular hexagon with "wrench"

size of 168 mm. The core model represents the central part of the reference fuel rod assembly.

The PSB rod simulator bundle has 15 spacer grids with prototypic geometry.

Heat is removed from PSB-WER primary circuit by four steam generators (SG's). The

PSB-WER SG is a vertical vessel with two vertical headers inside. A bundle of horizontal spiral

heat-exchanging tubes of full size is mounted between the two headers. The PSB-VVER SG is

designed in such way that the reference tube bundle elevations and tube lengths to be

conserved, as well as the flow area, heat transfer surface and secondary fluid volume to be

matched the scale factor. On the secondary side, the feed water system and the main steam

lines are simulated. The turbine and the condenser are not modeled.

PSB-VVER pressurization system includes a pressurizer, surge lines, spray lines, and a relief

valve. By means of surge and spray lines the pressurizer can be connected to "broken" loop

(loop #4) or to one of the intact loops (loop #2) of the facility. The PRZ vessel height, the bottom

elevation and location of nominal level correspond to the reference ones. An electric heater with

a power of up to 80 kW is built in the lower part of the pressurizer vessel.

PSB main circulation pumps are used to provide forced circulation in primary circuit. The

circulation pumps are variable-speed ones of vertical centrifugal single-stage type and can

operate under two-phase fluid conditions.

Table 1: Main Operational Characteristics of PSB-WER Comparing to WER-1000

Parameter Units WER-1000 PSB-WER

Coolant, . ; - water water

Number of circulation loops - t_4

Primary circuit

Pressure MPa 15.7 *15.7

Coolant temperature (hot/cold leg) deg 290/320 290/320

Coolant flowrate m 1/h 82485 < 280

Core power MW 3000 10

Secondary circuit

Steam generator pressure MPa 6.3 6.3

Feed water temperature deg 220 < 270

2-2
I Thermal power of one SG I MW 1 750 1 2.5

To research thermal-hydraulics of large break accidents, there is special unit to reproduce

guillotine break of hot leg in PSB-VVER. Principal scheme of this unit is given in Figure 2.

There are installed two blowdown lines YE10 and YE20 to remove discharging coolant from

each side of the pipe. Fast acting valve YA03SO8, located between these two blowdown

lines, closes cross section of hot leg during the breaks of special rupture discs.

.TJ03SOI YB03WO0

TJ03SO2

YEOIPOJ YA03SO8 YEO2POI

6 010X16 0B1

YCO1B04 ") YEOISOI - YE02SOI to atmosphere

ZN_ • 4x2 • 0 14.x2 -

Tluottle j• •Throttle

Rupture discs Rupturediscs TP06SO7

YEIOP01 YE 0PO

rPo6Sa6 TPO6SO5

Y YE20SO2 TP06PO •-

V,59,6

•0159,6

YE30SOI •

Figure 2: Principal Scheme of PSB-WER Large Break Unit

To simulate fast break, each blowdown line contains special system of two rupture discs. To

produce the break, the valve YE20SO2 is opened to remove air from the cavity between

discs. As a result, after instantaneous rupture of two discs, coolant begins to discharge in

blowdown line. Throttles, determining a size of the break, are installed in the blowdown lines.

Valves YE01S01, YE02SO1 cut off blowdown lines from primary circuit for period of

preparation of experiment. Before beginning of experiment they are fully opened.

2-3
2.2 Experiment HL-2x25-2

The test HL-2x25-02 ,,Hot Leg Large Break 2x25%" was performed in the PSB-WER test facility

in Electrogorsk Research and Engineering Center (EREC) in Russia. The thermal-hydraulic

processes related to hot leg large break 2x25% were researched.

2.2.1 Facility configuration

The information on the test facility hardware and configuration of the system specific for

HL-2x25-02 test are given in the Table 2.

Table 2: Test Facility Configuration in HL-2x25-02 Test

Equipment Status

Pressurizer Connected to the loop #2

Core by-pass 2 throttles with 2 orifices of diameter 7 mm were installed at inlet

and outlet of core by-pass

HPIS One channel was connected to the cold leg of the loop #1

LPIS two channels connected to the cold and hot legs of the loop #3

ACC's ACC's #1 and 3 were connected to UP. ACC #2 and 4 was

connected to downcomer

SG's All SG's were connected through steam lines

Feed water heater In use. SG levels under steady-state were maintained by supply

of feed water

Large break unit The break unit was located at the hot leg of the loop #3 between

UP and SG. Leak channel in each blowdown line is a horizontal

throttle of 25 mm, L = 250 mm.

2.2.2 Initial Conditions

The main initial conditions of HL-2x25-2 test are given in the Table 3. The HL-2x25-2 test has

been performed under reduced initial core power corresponding to approximately 15% of

nominal power.

Table 3: Main Operational Characteristics of PSB-WER Facility

Parameter Units Value

Primary circuit

Pressure in upper plenum MPa 15.78

Coolant temperature (DC inlet / UP outlet) deg 290 / 317

Primary loops flow rates kg/s 2.32+2.37

Core power kW 1520

Core by-pass power kW 16

Coolant level in PRZ (gauge YP01L02) m 6.72

Secondary circuit

Pressure in SG's J MPa 7.84 / 7.84 / 7.84 /7.84

Level in SG's (gauges YB01+04L01) m 1.67 / 1.67/1.67 /1.67

ECCS

Pressure in ACC's (ACC 2 disconnected) MPa 5.88 / - / 5.88 / 5.88

2-4
I Level in ACC's (gauges TH01+04P01) m 4.84 - 4.90 / 4.90

2.2.3 Boundary Conditions (test scenario)

Detail information about the HL-2x25-2 test boundary conditions is given in Ref 5. The main

events of HL-2x25-2 test are described in the Table 4.

Table 4: Main Events During HL-2x25-2 Test

Event Time [s]

IE - Break Opening 0

Pressure in UP (YCO1P16) < 13.73 MPa -+ conditions for SCRAM 0.4

- initiation of MCP coastdown 0.4

- FW termination to all SG's 0.4

- initiation of core power and core bypass power reduction 3

- steamline separation from all SG's 18

HPIS activation 27

ACC's activation 35/-/35/35

LPIS activation 78

ACC's empty 151+172

Break termination (closing of blowdown lines) 335

HPIS termination 473

End of test (FRS power switched off) 1500

The experiment is started with opening of the valve YE20SO2 in the air drainage line from the

cavity between rupture discs. As a result, air pressure in the cavity drastically drops, and rupture

discs are ruptured in two blowdown lines practically simultaneously. Coolant discharging from

primary circuit is started. Simultaneously the valve YA03S08 is closed to prevent coolant flow in

hot leg of the loop #3 between blowdown lines. Interval of time to close the valve YA03S08 is

less 0.1 s. Interval of time to rupture discs is 0.001 - 0.01 s.

When the UP pressure Pup = 13.73 MPa, the facility control system starts to simulate operation

of NPP automatics in accordance with the SCRAM signal (station blackout takes place

simultaneously), which provides for the following actions to be performed at the test facility:

,, Power reduction on core simulator and BP is started

* Stop steam removal from steam generators

,, Feed water temperature is decreased from 220 to 150°C during 30 s

,, The procedure of MCP's coast down is started

,, Close feed water supply

* Power of PRZ heater is switched off.

There was rather nonstandard coastdown of MCP's realized. MCP's rotational speed was

temporarily stabilized at revolutions corresponding to 24% of nominal value for 173 second,

then standard coastdown continued.

After achievement of two conditions: 1) UP pressure decreases to the value Pup = 10.79 MPa

and 2) later the moment 22 s after the experiment start (interval of time to start operation of

diesel-generator), activation of high pressure injection system is simulated. Cooling water of

2-5
HPIS is supplied to cold leg of the loop #1. HPIS mass flow rate 0.12 kg/s is provided.

Temperature of cooling water is 280C.

HPIS operation is continued, while total volume of HPIS water supplied is achieved 0.052m3,

later HPIS water supply is terminated (simulation of emptying of HPIS water tank).

After upper plenum pressure decreasing below 5.88 MPa, hydro-accumulators start to operate.

After upper plenum pressure decreasing below 2.16 MPa, activation of low pressure injection

system is simulated. Cooling water of LPIS is supplied to hot and cold legs of the loop #3, mass

flow rate is 0.35 kg/s per each line. LPIS operation continues till the end of the experiment.

After upper plenum pressure decreasing below 0.4 MPa, the valves YE01-02S01 are closed,

and coolant discharging from primary circuit is terminated (simulation of containment pressure

achievement).

Feed water flow rate provides SG's collapsed levels at nominal value during the experiment.

The experiment is terminated by switching off CS and BP powers and by closing of LPIS.

2-6
3. THE TRACE CODE

The TRACE (TRAC/RELAP Advanced Computational Engine) code is a component-oriented

reactor systems analysis code designed to analyze light water reactor transients up to the point

of significant fuel damage. The TRACE code is a finite-volume two-fluid compressible flow code

with one, two and three dimensional flow geometry. The TRACE code can model heat

structures and control systems that interact with component models and the fluid solution. The

TRACE code can be directly coupled with the 3 D reactor kinetics code PARCS or with the other

TRACE jobs and other codes (CFD, CONTAIN ...) through its exterior communications interface

(ECI). TRACE uses what is commonly known as a 6-equation model for two-phase flow (mass

equation, equation of motion and energy equation for each phase). Additional equations can be

solved for non condensable gas, dissolved boron, control systems and reactor power. There are

five additional closure relationships for field equations: equations of state, wall drag, interfacial

drag, wall heat transfer and interfacial heat transfer. These constitutive models are semi

empirical equations. There are two numerical methods available in TRACE: semi-implicit

method and the stability enhancing two-step (SETS) method. All TRACE calculations presented

in frame of this report were performed using TRACE V5.0 code.

3.1 The TRACE code assessment

Confidence in the computational tools (codes) and establishment of their validity for a given

application depends on assessment. TRACE, like other two-fluid codes, is composed of

numerous models and correlations. When applied to full scale nuclear power plant conditions,

many of these models and correlations can be applied outside of their original database. By

assessing the code against thermal-hydraulic tests, it is possible to show that the code and its

constituent model packages can be extended to conditions beyond those for which many of the

individual correlations were originally intended (Ref 16). The assessment process however, can

also indicate potential deficiencies in the code. There are following four sources of data for code

assessment (Ref 17):

0 "Fundamental" experiments

0 Separate effect test facilities (SETF)

0 Integral test facilities (ITF)

* Real plant data

3.2 VVER typical features related to TRACE code assessment

The TRACE code validation process is mainly based on the data from experimental facilities or

real NPP's of Western PWR type. WER reactors are in many aspects similar to Western

PWR's. Therefore a lot of experimental data measured on PWR's or PWR test facilities is

valuable also for VVER research. On the other hand, the VVER design has several specific

features. From the hardware point of view the main differences between WER-1 000 and PWR

are the following (Ref 19):

* Horizontal steam generators with 2 headers

* Lower plenum internal structures

* Fuel assemblies with hexagonal fuel rod arrangements

3-1
* ECCS injection points

* Secondary side water volume of the steam generators is larger

* Operational conditions and set points of actuation of ECCS

* •Working conditions of secondary side of steam generators and set points for the

operation of feedwater and steam line

There are 50 operating of WER type (Ref 18). It is a meaningful number compare to 216

operating PWR reactors (Ref 18). Therefore corresponding attention should be given to code

validation for VVER type of reactors.

3-2

0
4. INPUT DECK DESCRIPTION

An extensive TRACE input deck of PSB-WER facility was developed including all important

components of the PSB-VVER facility: reactor, 4 separated loops, pressurizer, break unit, main

circulation pumps, steam generators, break section and important parts of secondary circuit.

The input deck was designed on the basis of PSB-VVER facility documentation (Ref 4, 20).

Nodalization diagrams of the PSB-VVER facility Input Deck are presented in Figure A-1 (reactor

+ primary circuit) and Figure A-2 (secondary circuit) in the Appendix A. Nodalization scheme of

the primary circuit includes all four loops equipped with MCP's and SG's.

Two models of the reactor pressure vessel (RPV) with internal structures were developed. Both

models of RPV are divided into 3 parts. The first part represents Downcomer (DC) + Lower

Plenum (LP), the second part represent Fuel Rod Simulator (FRS) + Upper Plenum (UP) and

the last part represents core by-pass from DC to UP. Both models employs VESSEL component

to model DC + LP (includes 26 axial layers, 1 azimuthal theta sector and 2 radial rings) and 3

PIPE components to model core by-pass. The Fuel Rod Simulator optionally employs one

component VESSEL (includes 32 axial layers, 4 azimuthal theta sector and 6 radial rings) or 2

PIPE components, where one represents main part of FRS and the other represents the hot

nozzles region.

Notice Results of the model which employs VESSEL component to model FRS is labeled "3D"

and the other model which employs 2 component PIPE is labeled "ID".

The pressurizer is modeled as PIPE component equipped with the heaters. It can be connected

to loop No 4 or to loop No 2 of the facility (as original facility design). Main circulation pumps are

modeled as PUMP components using original characteristics from Ref 20. HPIS and LPIS

systems are modeled as easy boundary conditions using FILL components. The large break

unit is simulated using BREAK component on the hot leg No 3.

Steam generator is modeled using multi-tube approach. The primary side of SG Input Deck

consists of 5 axial layers of heat exchanging tubes and two headers (the original facility SG's

consists of 34 tubes). Each axial layer is divided into 15 segments. The SG secondary part is

modeled as original three-channel complex with 10 axial layers (5 of them in the area of

exchanging tubes).

The feedwater system, the steam lines connected to all SG's and the common steam header

are also modeled. The BREAK component simulates the release of secondary steam from

steam header.

Components Statistic for both TRACE models - see the next Table 5.

4-1
Table 5: TRACE Components Statistic

TRACE Component 3D 1D Notes

Model Model

VESSEL 2 1 DC+LP (1 D, 3D); FRS+UP (3D)

PIPE 133 135 FRS+UP (1D model)

HSTR 170 143 -

POWER 2 2 FRS simulator + By-pass heating

VALVE 24 24 -

PUMP 4 4 MCP's

BREAK 3 3 Large Break unit + release of secondary

steam

FILL 7 7 HPIS, LPIS, Feedwater

Whole No of Components 345 319 -

4-2
5. RESULTS

5.1 Steady-state calculation

In order to achieve stable initial conditions of the HL-2x25-2 test, the steady state was

calculated for 1600 s. The following controllers were used for the first 1000 s:

* Pressurizer pressure controller

* Main steam header pressure controller

* Steam generators level controllers

The other controlled parameters (fuel rod simulator power, core bypass power, feedwater

temperature) were described as boundary conditions. It was necessary to decrease initial

pressure in SG's within 0.5 MPa compared with measured parameters in order to achieve good

results with primary temperatures. During the steady-state calculation, all SG's levels were

maintained by PI-controller, which was replaced by an original P-controller for the last 120 s to

get corresponding behavior of model before beginning of the experiment. All SG' levels

controllers were switched off for the last 600 s of the steady state to check the steady-state

calculation stability. Steady state calculation took approximately 100 min, main calculated and

measured parameters are compared in the next table.

Table 6: Initial Conditions (TRACE calculation vs. experiment comparison)

Parameters Units HL-2x25-2 TRACE

II 3D I 1D

Primar circuit

Upper plenum pressure (YCO1P16) MPa 15.79 ± 0.02 15.771 15.770

Pressure drop at FRS (YCO1P07-P10) kPa 29.39 27.25 28.59

Coolant temperature at FRS inlet (YC01T06) 0C 290.6± 0.14 288.9 288.8

Coolant temp. at FRS outlet (YA1TO4b) °C 317.8 ± 0.16 319.6 322.2

Loop-1 flow rate (YAO1FO1) kg/s 2.31 ± 0.03 2.275 2.138

Loop-2 flow rate (YA02FO1) kg/s 2.37 ± 0.01 2.280 2.138

Loop-3 flow rate (YA03FO1) kg/s 2.31 ± 0.01 2.284 2.140

Loop-4 flow rate (YA04F01) kg/s 2.19 + 0.01 2.285 2.1"38

FRS power (YCO1NO1) kW 1533.0 ± 14.4 1533.0 1533.0

Core by-pass power (YCO1N02) kW 16.1 ± 0.3 16.1 16.1

Collapsed level in PRZ (YP01L02) m 6.97 ±0.02 7.125 6.835

Secondary circuit

Pressure in SG-1 (YBO1P01) MPa 7.839 ± 0.06 .7.319 7.326

Pressure in SG-2 (YB02POI) MPa 7.755 + 0.07 7.322 7.320

Pressure in SG-3 (YB03PO1) MPa 7.805 0.06 7.310 7.308

Pressure in SG-4 (YB04PO1) MPa 7.734 ± 0.07 7.326 7.323

Collapsed level in SG-1 (YBO1LO1)

Collapsed level in SG-2 (YB02LO1)

Collapsed level in SG-3 (YBO3L01)

m

m

m

1.66 ± 0.01

1.66 ± 0.02

1.67 ± 0.01

1.693

1.684

1.684

1.689

1.685

1.685

5-1
Collapsed level in SG-4 (YB04LO1) I m 1 1.60 ± 0.03 1 1.668 1 1.670

Hydroaccumulators_

ACCU-1 pressure (TH01 P01) MPa 5.924 ± 0.003 5.946 5.946

ACCU-2 pressure (TH02PO1) out of service MPa 5.874 ± 0.004 5.887 5.887

ACCU-3 pressure (TH03PO1) MPa 5.873 ± 0.005 5.889 5.889

ACCU-4 pressure (TH04P01) MPa 5.890 ± 0.004 5.910 5.910

ACCU-1 collapsed level (TH01P01) m 4.84 ± 0.002 4.845 4.845

ACCU-2 collapsed level (TH02PO1) m 4.82 ± 0.002 4.822 4.822

AGCU-3 collapsed level (TH03P01) m 4.91 ± 0.001 4.910 4.910

ACCU-4 collapsed level (TH04PO1) m 4.90 ± 0.001 4.896 4.896

5.2 Transient calculation

The post-test calculation of HL-2x25-2 experimental test at PSB-VVER facility started at the

time 0 s with initiating event - double ended break 25% at the hot leg of the Loop No 3. A

comparison of calculated and experimental times of the occurrence of main events is presented

in the next Table 7.

Table 7: Chronology of main events (TRACE calculation vs. experiment comparison)

Time fsl

PSB TRACE

Event HL-2x25-2 3D 1D

Initiating Event - Leakage opening 0 0 0

Pressure in UP (YC01P16) < 13.73 MPa 0.4 0.34 0.31

= Reactor SCRAM signal

- start of MCP coastdown 0 0.34 0.31

- start of FRS and BP power reduction 3 3.64 3.61

- signal for stopping feed water supply (RL01-04S01) 0.4 0.34 0.31

- signal for stopping steam discharge (RA06SO1) 0 0.34 0.31

PRZ emptying (according to YP01L02 / complete 11 / - 13/28 12/27

emptying)

Start of HPIS injection into Loop 1 22 28 28

Pressure in UP (YCO1P16) < pressure in ACCU's

- start of ACCU-1 operation 34 31 32

- start of ACCU-2 operation 34 32 33

- start of ACCU-3 operation 35 32 34

- start of ACCU-4 operation 36 32 34

Start of LPIS injection into Loop 3 65 71 67

Levels in ACCU's (TH01-04L01) < 0.1 m

- ACCU's 1-4 separated 129-134 138 - 158 144 - 166

Break termination (closing YE01S01 a YE02SO1) 335 335 335

- containment pressure achievement

HPIS injection termination 473 470 470

End of the experiment - FRS power switched off 1500 1500 1500

5-2
The graphs of main measured and calculated parameters are presented bellow. The

parameters labeled "3D" came from model which employs VESSEL component to model FRS.

The.parameters labeled "ID" came from model which employs 2 PIPE components to model

FRS.

The following description in this chapter is provided for 3D model results. Comparison of results

from 3D vs. 1D model has been done in the next Chapter 5.3.

Calculation started at 0 s by simultaneously opening both blowdown lines and closing valve

YA03S08 between both blowdown lines.

Due to high initial break flow (see Figure 21), primary parameters were rapidly decreasing - see

primary pressure at Figure 4, pressurizer level at Figure 6 and cladding temperature at Figure 5.

At 0.05 s PRZ electric power was switched off.

At 0.34 s UP pressure decreased below 13.73 MPa (Figure 4) and power to core simulator and

by-bass started to decrease along to the specified function (Figure 10). The initial rate of PZR

pressure decrease was lower than one in FRS, which was caused by presence critical flow rate

in the surge line, see Figure 4. PRZ pressure became equal to primary pressure at 28 s (after

PZR emptying). Collapsed level in PZR was decreased to 1.885 m (in accordance to the gauge

YP01L02).

Detailed analysis of the first period calculation is presented below. There can be identified 3

stages.

First stage of calculation (0 - 1.1 s) is characterized by fast depressurization, primary pressure

was very quickly decreased from 15.7 MPa to 10 MPa as a result of a high initial break flow

(the initial rate of primary pressure decrease was approximately 5.2 MPa/s.) The first stage of

calculation was in very good agreement with experimental data, see Figures 4, 6 to compare

primary pressure and collapsed water level in pressurizer - most important parameters, which

describes dynamic behavior of the whole primary circuit.

During the second stage of the calculation (1.1 - 28 s ) the rate of primary pressure decrease

become lower (approximately 150 kPa/s). Primary pressure became lower than secondary one

at 16 s (Figures 3, 11). Thus, reverse of heat exchange in SG's took place. Rate of primary

pressure decrease after the reverse of heat exchange in SG's was decreased to 90 kPa/s.

During the second stage of calculation the value of primary pressure was slightly lower than

during experiment, but pressure progression coincide with experimental data very good.

Third stage (28 - 160 s) is characterized by emergency water supply. At 28 s cooling water was

started to be supplied to primary circuit from HPIS, mass flow rate was 0.12 kg/s (the same like

during PSB test), see Figure 7. HPIS operation was terminated at 470 s, after total volume of

HPIS water supplied had achieved 0.052 M 3 . At 32 s primary pressure decreased below a value

of pressure in ACC's and cooling water from ACC's started to be supplied to DC and UP, see

ACC levels in Figure 9. Termination of water supply from ACC's occurred between 138 and

158 s, after level in a particular ACC decreased below 0.1 m (in order to prevent

non-condensable gas from penetrating into primary circuit). At 65 s started LPIS and continued

in operation until the end of the calculation, mass flow rate was 0.35 kg/s to the cold leg and to

the hot leg as well. Due to HPIS and ACC's water supply the rate of primary pressure decrease

5-3
increased up to 167 kPa/s at 44 s. After termination ACC's operation the primary pressure was

fairly stable and the rate of primary pressure decrease went up to very low value until the

moment of blowdown line closing at 335 s. Calculated time of beginning acting particular

emergency water supply from HPIS, LPIS and ACC's was comparable to experimental data,

see Table 7 (the longest time difference equaled to 6 s). Calculated stable primary pressure in

the end of ACC's discharge, after 115 s, was slightly higher than during experiment. As a result

of higher calculated primary pressure, ACC's discharge time was slightly longer then during

experiment, see Figure 9.

Figures 13 - 16 present a comparison of calculated and experimental transients of coolant

temperatures at inlet of SGx (YAxxT04) and at LOOPx seal (YAxxT26) in broken Loop No 3 and

intact Loop No 4 (behavior of temperatures in other intact loops No 1 and 2 is analogous).

Figures show that calculated temperatures in intact loop No4 behaved in a similar manner

whereas broken loop No3, where an essential difference was observed due to presence of the

break here. All calculated temperatures were in quite a good accordance with experimental

ones.

Figure 5 presents cladding temperature at the top of the rod bundle. After 0.34 s when UP

pressure decreased below 13.73 MPa, power to core simulator started to decrease along to the

specified function so cladding temperature was rapidly decreasing and was near saturation

temperature until the location of cladding temperature was filled with cooling water. The

calculated and experimental transient was in good agreement during the period when the break

was opened, but after break had been terminated (at 335 s - see Figure 33) and core simulator

begun to be flooded, calculated transient was slightly faster than experimental one. The reason

of this behavior is discussed in more detail in the next paragraphs.

The whole calculated transient can be divided into two main stages. The first stage from 0 to

335 s is characterized by blowdown of primary circuit due to coolant water discharge from

break. The second stage from 335 s, when the break was terminated, till end of calculation

where primary circuit was intensively filled with water as a result of LPIS operation. Water

inventory and its distribution in reactor (DC, FRS and UP) is described as collapsed water in

FRS and UP, see Figure 23, and as differential pressures over the reactor, see Figures 17 - 20.

The differential pressure measurements location in the reactor are given in Figure B-1 in

Appendix B.

During an early time period of blowdown stage 0 to 28 s, when no cooling water was supplied,

collapsed water decreased dramatically (see Figure 23) and upper part of DC and UP was

emptied - see differential pressures at Figure 15, 16 and Figure 21 - 29 (lower absolute value of

pressure drop means lower liquid volume between gauges of differential pressure

measurement). After cooling water started to be supplied to primary circuit from HPIS at 28 s,

the rate of the reactor collapsed level decrease got lower. After ACC'S discharge begun at 32 s

the total mass flow supplied from HPIS and ACC's was higher than the break coolant discharge,

see. Figure 22. As a result of higher cooling water supply than the break discharge, level in

reactor started slightly increasing. In theend of ACC's coolant water discharge and mainly after

its termination, coolant water supply from HPIS and LPIS was not able to compensate the break

discharge and the reactor collapsed level started again to decrease until termination of break

discharge, see Figures 21, 23. All described behavior of collapsed level in the reactor during

ACC's discharge are clearly visible on pressure differences in FRS and UP, see Figures 17 - 19.

5-4
Calculated reactor water inventory and its distribution over the reactor were in a fairly good

accordance with experimental ones, but calculated water distribution slightly differs from

experimental one. In case of calculation, more liquid was kept in lower plenum and lower part of

FRS than in case of the experiment. Calculated absolute value of pressure differences had

higher value than experimental ones - see Figures 17 - 20. As a result such a liquid distribution

the calculated break mass flow was lower than during the experiment. Lower break discharge

means higher water inventory in the reactor, see Figure 23.

At 335 s, the same time like in experiment, break was terminated and valve YA03SO8 between

both blowdown lines was opened to make possible water flow through loop No3. After closing of

blowdown lines, filling of primary circuit was realized intensively, see reactor collapsed level at

Figure 23. Calculated progress of filling of primary circuit was faster than calculated - the

reason (higher amount of liquid in lower plenum) was described in previous paragraphs. Filling

of particular regions of the reactor is described by rising of absolute value of pressure

differences over the reactor. If value of pressure difference is nearly 0, it means emptied region

- only vapor on it. Contrariwise, if value of pressure difference is near nominal value and is

stable, it means region is filed - only liquid on it, see Figures 17 - 20.

Pressurizer filling started at 1235 s which is good coincidence with experiment where it

happened at 1300 s.

5-5
PSB-VVER test facility

Hot Leg Large Break 2x25%

Blowdown own' phase ase Reflood phase ! ___' ___,, I ___ .... . ... I • J...... I14 - 32 t L . ............

--- +27 s HPIS activation --. -4-- . --

+35s-ACCs activation

) ! ; . +78 s--LPlS activation ) " ) " •

' 4 ACI empyI I'

-,+335 s- Break terminated 7

-11100 f 4 .) '-'I ia I :0i) IM 4) (luo 7MK MW -MX) 1 IW 1M II t 1 l 3M f14 14M11 Ilf

Time W~

Mup3I'V.- %MOPI

Figure 3: Primary Pressure

PSB-VVER test facility

Hot Leg Large Break 2x25%

-8 -1R - 'lr - COIP16 TR,(UE LIP prWir., V403131

06 -- - . J.. .... .... ....... - .. . .... . . -- . .... 1 ------ - .. .. .. - - - ..... ..... . .. . .... . .... ....

-, S - ........... .7.. . . ------------

0

-10 O in 20 30 40 5O 60 70 80 90 100

Time Ws)

Figure 4: Primary Pressure (detail)

5-6
PSB-VVER test facility

Hot Leg Large Break 2x25%

-- i -I-h--,-.I-. - -.

- +27 s HPIS activation .---

+35 s ACCs activation

)..... ..... .. | .. ;.. ............ .... ......... - - .... . ... r-!...... . . . ..

.17!

I +7s •is activaton '• Top of the core flooded

+76Bea terminatedivtio

1 ...7 ~~. ... .. ... .. .... ... .... ... . .. . .

....... .... .

i7: +33 a ',-Break terminated:

Blowdown phase Reflood phase !

-I(0 (1 ICn ':(n 3(06 4W,(X0 l00 (X) 700 () 9IX, ( IO f 110 1200 13(10 14W1 1.500

'li im (S)

Figure 5: Fuel Cladding Temperature (Top of the Core)

PSB-VVER test facility

[lot l.eg Large Break 2x25%

14 . , .O h* )(o

- ( , iit,

TRACE shows step-wise

.. flooding of the loop region

(comparing to the experment) -

., . . .1D model shows slower.

flooding of the loop region

4 -. • (comparing to the 3D model) I

: , Hot log elevation :

Cold leg elevation .. -- ..0. -- . -.-. •- .- -. . -7 -- .... • - - Zz" -/;J =; • -.

-"I ,I 0 10 '1 X I 3W 400 511.) 7 0 W K) IMO 114X) 12M 13M 14M 13M

llrnK (.1

Figure 6: Pressurizer Level

5-7
PSB-VVER test facility

Hot Leg Large Break 2x25%

030 - : --------- ,-. 0 PBIIPLJTJOIF02

- -1 PSD LPIS TJ O3FI . - PIB P• R TJ03F02

ii - f~ýlM HPIS CL LOON PD..-34D3 ID

TRiACE OPIS ML LOOP3 V-.-3411 ID ~~IRACE [P1 Ci. LDP3 (b..- 3.2) 3D TRACE 1PIS CL LOON 0-34) ID

0.40 " '" " '- TIACE LPIS CL LOOP] f -- 342) 3

... . _

0.10 ___ ___

000 -. -,, - _ ,D33 :- -100 0 100 200 300 400 500 600 700 800 900 1000 1100 1200 1300 1400 1500

mc• IS)

Figure 7: HPIS + LPIS Flow (Boundary Condition)

PSB-VVER test facility

Hot Leg Large Break 2x25%

I ,; . ____ 0-0 PSNI MPinua-.OD 0 30101 V . .. ............. ...... : ... P Mu iCPPI W.U L I

- M si MCP3 rolaDoD W1. YDO2RII

" .. . . " A PSEMCP4 mt OOVdL.YD04IRII

TRACE MCPI rt I o~m- 11

IID.3C0 MCPS . M . A. 1D03.0ll 

-- TRACE MCPI .l. I. IDD.3- 211I 3D 4 5 :.. ._ . .... ... o .. ..... .- _ _ - ...... ..... _.._.. . . M C E , M . ,, 1 3D

TRACE MCPI m. W1. (omeLgn-6) ID

TF1D.CE MCP1 tOE WI(3033-161)1 I

40 TRACE MCP4 mot. W.1. (Ol=Mtt- 2511 ID

.____ ...-.. - - - TRACF P-WN .1.IM.Ilg.-I r)

. . . . . ....... .i . . .. . .. . '. . . . . . . : . . . . . ..• .. . . . . . . . . i .. i .

20

, ..... . .

10 .. ... . ..... .t

,., ,_ I _

-100 -50 0 50 100 150 200 250 300 350 400

TIme Ws)

Figure 8: MCP Rotor Speed (Boundary Condition)

5-8
PSB-VVER test facility

Hot Leg Large Break 2x25%

0-0 PB"AI~dTHOLDI -- TRAMEIA., IJvd3DO)JD -- TRACE IIAI I.d W~300) ID

2.5

60 80 100 120

Time WS

200

Figure 9: Hydroaccumulators Levels

PSB-VVER test facility

Hot Leg Large Break 2x25%

18000- S o WI)

TRACEnp-h- 9503D

t~o .. • .- i'.:l : . I ___ ._, ,___•_

bOO - 1" •}; i -- : i .! : .: .- -:. .. i , ., . 4600 .1 L

1400

'-0 0 .. .. ..

Soo0 - - , . : I . . .' • , ! ....... . ]-. : ;. ': .. . .

-100 0 100 200 300 400 500 600 700 800 900 1000 1100 1200 1300 1400 1500

Time is)

Figure 10: Fuel Rod Simulator Power (Boundary Condition)

5-9
PSB-VVER test facility

Hot Leg Large Break 2x25%

0-0 PS: SSG IDI OIl

1I3PSI II,!fve 2 YUOZLOI

PSI SO11 1"3Is YB033LI

PSO 1C4 evel YEN I0

]RACE SGZ IM] CC602) 311

TRA1CE 563 levl L c02)l 3D

7 TRIACE I :=l (c11 M),f 3D

- TRACE SID WCe 1,512) ID

ThAýCE I2Levl :6651170) ILL

- TCElSG4 k.d 1,75. DI

TImC (S)

Figure 11: Secondary Side Pressures

PSB-VVER test facility

Hot Leg Large Break 2x25%

MR- SC. p W2 I.. .l.

ME SGI preSM Ys3131P01

-- L- PSIIIC

4

M -I0SIPBI

TISC 5 1 pess~re II13-602AOS 3D1

[RIIACE SGI w -4501131 I M1 DII2A!

--4 . . ... ..... .... ... '.• .. . • .i . k .- , . . , . .. . . ; . . . . • - : - • . •

2 . . .. . "" : .. .... .-...... .. . . .. 7 . . .t .- ! . i • . , i •. ] . .. .. .

-100 0 100 200 300 400 500 600 700 800 900 1000 1100 1200 1300 1400 1500

Time (s)

Figure 12: Steam Generators Levels

5-10
PSB-VVER test facility

Hot Leg Large Break 2x25%

0-0 PSOCnIfI teoenflYAMNO 7 TtRAE -1 -M. WhOIGU 3D TDACE -oImItOi.eMAMM~ID

200

150

Time Ws)

Figure 13: Broken Loop 3 Temperatures (Inlet of SG3)

PSB-VVER test facility

Hot Leg Large Break 2x25%

350. ..- - - - . - P SBM1,f*•,A.. ,N ,T2,

-~~ -.- ThA( roL.C ,p.( b$li ID

300 4-----

E~ IS

E 50

10

100 0 100 200 300 400 500 600 700 800 900 1000 1100 1200 1300 1400 1500

Time (s)

Figure 14: Broken Loop 3 Temperatures (Loop Seal)

5-11
PSB-VVER test facility

Hot Leg Large Break 2x25%

0-0 PSI -&.1tg--nR~YA04TR4

-- TRACE WOin~ ;MWI, (61123) 30 __TRACE C0ol~w: enWR (aI,83)11)

E

9-

ime (s)

Figure 15: Intact Loop 4 Temperatures (Inlet of SG3)

PSB-VVER test facility

Hot Leg Large Break 2x25%

-TRACE-U.bn SB.M (~I,82Sl3D

TRAC. r .m=,,: ,. (611151 .L.

-200

E150 F-T

100

)0 ' . . 1 . 1 , 1 .1

-100 0 100 200 300 400 500 600 700 800 900 1000 1100 1200 1300 1400 1500

Tnme (s)

Figure 16: Intact Loop 4 Temperatures (Loop Seal)

5-12
9

CD

0

0c

0l

0m

UI U

... . .........

.... .....

0

C

0

0

Si

e in 0 LM 0 0 LI 0

1 1 C1 Il

(uddDp doap -)xnssUd (tdjM) do.p ainssald

0

0

0

0

N

0

0

0

0

0

cc

0

0

Co

0

0

0

0

N

0

"- .1. • .. ' ': : ...... ' " .... •.. ..

-•- .•i -.•" . ..•. i • .. •; I •. '.. . ..:. .:.:: ....: .. . .I t ,...; ....... ....• .

2!J

.! .. ..... ... i : ... ..; i. .,

.. .. ......... .. 1 " " ' ' .... ... . ' " . .. .. . .

.. .... .... .. ....... •.. .... . ...... . . .

.. •. . .. ... i t..

.... ...i... ;.. , ......+. ; .i l .i ...... . .... ...... .. • ... : ; . .... i... •.. . . .. .. . , .. .

... ..:...±..... •...... . . , .. ! . , . • . , , .. .. .

. . . . .. ... ..... . i.. .

...... . . I. . .

(Wdl) doip O.InssO-d

0c

0

0

0o

CD

...... i.

0 ..... 0......

I - -

(edl[) doaip z)fl9ss;Jd

0

Figure 17: Pressure Differences DP01-DP04 (Downcomer)

5-13
oeB

o -

II

ii~~~i. .. .. ..!.• .....• 

.... ...... .... i .. i -

0

0

0N

0

0

0

0

0

0

0

0

0

0

0

0

0

0

0

0

0

cJ

(L2dj) doap ajflss3Jd (v(Tr) dojIp a.mssa id

00

"'00

uww

) U U

.

.. ..... ...

........ . . .....

....... .

.. .. ...... ........... ......... ....

.... .. ... .. ......... .........

ts.

..... 

...... 

... .. ..... 

...... ............

... .... .. . .....................

0

0

0

0

0

0

0

0

0

C

0

0

0

0

0

0

00

0

0

IN

0

0

0

0

I

(uPr) doip aiussaid

0 Ln Ln

(uept) doap amssajId

LM

Figure 18: Pressure Differences DP05-DPO8 (Lower Plenum + Lower Part of FRS)

5-14
... ......

4."

..... .... .....

K 

0. 

.... ... ... ... .. . .. .. .. .. .... .-. .

. ... ... . ..... .

.. ....... ... ........q .......F .

0

0

r4- 0

0,

N

0

0%

I

0

0

0

0

N

0

0

0

-4

0

0

QO

0

0

0

0

0

0

N

0

(u,"¶) doxp Uflssa.id

0

0

'I.

0

0

N

0

0

0

0

0

0

0

0

0

0

0

N

0

cJ

0

0

0

0

N

0

0

0

0

0

cc

0

0

0

0

0

0

N

0

0 Lfn 0 in

I

(v,)I) doap ainss~id

Un 0 kA~ Lfn

(LeP1) dojxp ainssa.41

Figure 19: Pressure Differences DP09-DP12 (FRS + Lower Part of Upper Plenum)

5-15
.. . .. . . . .

II.

Z.

..........

...........

0

0

0

0

N

0

0

0

0

0~-~

o

.Coo

ww~

0

0

0

0

N

0

0

0

0

0

eo~

I

(ed~¶) doip aiflssa3Jd

00u

.... ... .. ... ..

Moi

0

0

~l.

0

0

0

. ........... .... .. ....... .. ....

...........

...... ..... ......... ............ .

.... ........

........

7i-

.. ........

. . . . . . .. .. .

0

N

0

'. 0en0 4 0

I I I

(L~dpj) doxp a-MSSad

0

Cý 00

C'D

0

0=

N=

* 00

.~00

*0>.>.

-uww

ill

"'"/....

cJ

.4-

0

C

0

0

N

0

...... ........ .. .

.. .................

... ...... .. ..... ... .......

...... .........

TT.

0

0

C

0

N

0

0

0

0

0

0

0

CO

C

0

'I,

C

0

N

0

(e,pj) doap aissa.id

tM LI 0 Int 0 inC

IN I

(vcP3) doip 2Jlssaj~T

Figure 20: Pressure Differences DP13-DP16 ( Upper Part of Upper Plenum)

5-16
PSB-VVER test facility

Hot Leg Large Break 2x25%

):,(~~~ - - !iIo. ~ o•tI

VI

L ...... ........ ......i .. ..... .. ..! . . ...' .. ......... ....

Differebt ilowrate from UP side to break

I0 co.mpa~ing 10 (blue line) to 30 model (red line)

... Break termination

-0 Al 0 is1 210I6 2 •11 ur) 3Y-0 400

lime• is)

Figure 21: Break Flow

PSB-VVER test facility

Hot Leg Large Break 2x25% 40 ".' . , = . . r . , . .:. . . . -t . -- IMC• iot=1 b.,•akfno,,,re 3O

4 • , . • i . . . , , 1- TRAC(HMS S TOllba T ri ota l lIPI, flownlfeID k 

+A now.pf a i .oWD IDI.fo flow m3D

37, .. . • 4. .. .... 1 

oZ. O .. . -1 .. . . - . $ o:fov.I

30 .. .'i i. s.-I.. . . . .. . .. . .

50 20 50 3030 0

O F .. i • -:... . : , ' • i : : -: :

Time W .. T+

Figure 22: Primary Circuit Flow Balance

5-17
PSB-VVER test facility

Hot Leg Large Break 2x25%

I .... " ... l ....A.. _ _ , . .. . i_ _0.4.. . .. ... .. . .. .4_.. I .... ar,,•, d141

C-l le ge at nsaSM Md. 

1!L I - __i__.___ -•:[ -_i_ ._ i,

-'.0 ( 0 (0 • I tl .(41

Ho le lvto

-I (K) 0 I ~ 10() MI) 4(X) ;00'l CAWl 7(9)i R 0 IMI91 IWOl I10(X U010 13M(• 1400 ISM•(

Figure 23: Reactor Collapsed Level

5-18
5.3 Nodalisation study - comparison of 1D x 3D model results

The progress of transient using 1D model was very similar to transient with 3D model, but

noticeable improvement of water inventory and liquid distribution over the reactor was observed,

see reactor collapsed levels at Figure 23 and pressure differences at Figures 17 - 20. That

improvement was caused by higher break mass flow with 1D model - see Figure 21. Higher

break mass flow had induced lower amount of liquid phase in the core (FRS) which led to better

results.

The amount of break discharge was influenced by fluid temperature at break donor cell. In case

"ID" model, the temperature at the break donor cell (from the UP side) was significantly lower

(see Figure 24) so the break mass flow was higher as well. That temperature difference at the

break was caused by different flow pattern of both models. In case of "3D" model cold water

injected from ACC's to 21-st level of VESSEL component flowed through central rings of the

VESSEL component downwards to the core region, where liquid was heated by hotter fluids

coming from core. At the bottom of UP the heated liquid flow gradually changed its direction

from central to outer rings and from downwards to upwards to the hot nozzle region where the

broken LOOP3 was. Whereas in case of "1 D" model of FRS such flow pattern was not

observed, in fact the 1D nodalisation of FRS excludes this behavior, and cold water injected

from ACC's to 21-st level of PIPE component flowed to the next lower volume and then directly

to hot nozzles region, so there was nearly no heating period.

Comparison of both calculated transients shows better agreement of calculation that employs

"ID" model of FRS. The comparison leads us to a conclusion that in case PSB-WER test

facility, which reactor vessel diameter is relatively small (180 mm), 3D flow pattern is not

significant. Thus simpler FRS modeled with 2 PIPE components gives a reasonably good

results for analysed LOCA transient. But it has to be mentioned that significantly different

situation might be in case a real reactor vessel of VVER 1000, which inner diameter is

approximately 4 m and 3D flow pattern could be very important.

5-19
PSB-VVER test facility

Hot Leg Large Break 2x25%

Is MA bKg~d T ~lb.2W313ft n- t..Mn "lK'C %03) ID

TS IUA. b"wld ! ,l-21VJA(1.11 k ID)

7., 100 12.5 150 175 219)'- 2.2, 250 277, 300 32- 31.0

Time (N

Figure 24: Temperature at the Break Donor Cell (From the UP Side)

5-20
6. RUN STATISTICS

The transients were calculated on calculation server with Intel Xeon 5440 processor 2.83 GHz

under GNU/Linux Debian 5.0 Lenny x64. The run statistics is shown in Table 8. The calculations

run substantially slower than real time.

Table 8: Run statistics

1D Model 3D Model

Number of components 319 345

Number of time steps 56626 130 832

Transient time 1 500s 1 500s

CPU time 8548s 31 219s

CPU time / Transient time 5.7 20.8

6-1

7. CONCLUSIONS

The R&D program FI-IM5/150 "Validation of the TRACE Code for WER Reactors" was

introduced . In frame of this program was the PSB-VVER facility evaluated as the most suitable

and qualified for the TRACE validation concentrated on WER-1 000 specific design and the

data from test HL-2x25-2 (Hot Leg Large Break LOCA 25%) was chosen as proper basis for

validation of the TRACE code, especially for reflood phase of LOCA.

Two TRACE 5.0 input decks of PSB-WER experimental facility were introduced in this paper.

The VESSEL component was applied to model fuel rod simulator (FRS) and upper plenum (UP)

in the first Input Deck (labeled 3D). Two PIPE components were applied to model FRS and UP

in the second Input Deck (labeled 1 D). The reason for variant modelling of the PSB-WER

reactor vessel was the aspiration to discover the best approach before modelling the real

WER-1000 plant.

Comparative post-test calculations of HL-2x25-2 test were performed using both (1 D and 3D)

Input Decks. Relatively good results was achieved comparing primary pressure and cladding

temperatures at the top of the core during first - blowdown phase (for both models).

Agreement between measured and calculated data was not satisfactory during second - reflood

phase of the test. TRACE 5.0 calculations predicted faster reflood of the core (for both models).

As the main reason of such behavior was identified different distribution of liquid phase over the

primary circuit (especially loop region) comparing to the HL-2x25-2 test data.

Significant deviations between results from 1 D and 3D model were identified analysing partial

parameters. 1D Model (using PIPE component to model FRS and UP) gave better results

comparing to using 3D model with VESSEL component for FRS and UP. The key reason of

such behavior was investigated. We have realised, that different flowpaths of liquid phase from

ACC's and LPIS injection into UP to break leads to significant difference of the temperature of

the liquid phase discharging through break. Thus the integral break flow is higher using 1D

model.

The comparison leads us to a conclusion that in case PSB-VVER test facility, which reactor

vessel diameter is relatively small (180 mm), 3D flow pattern is not significant. Thus simpler

reactor model gives a reasonably better results for analysed LOCA transient. But it has to be

mentioned that significantly different situation might be in case a real reactor vessel of VVER

1000, which inner diameter is approximately 4 m and 3D flow pattern could be very important.

Although a lot of work had been done in frame of this project, further investigation is necessary

to asses the TRACE code application for WER-1 000 design. We are ready to recalculate

HL-2x25-2 test after issuing proposed TRACE Code upgrades (improved spacer grid and level

tracking models). Moreover, we are finishing similar work using darta from UP-1 1-07 test (Upper

Plenum 11% Small Break LOCA) from PSB-WER experimental facility in these days. We are

ready to submit results in frame of CAMP the same way as HL-2x25-2 test results.

7-1

8. REFERENCES

8-1
1. TRACE V5.0 THEORY MANUAL, Field Equations, Solution Methods, and Physical

Models; Division of Systems Analysis and Regulatory Effectiveness, Office of Nuclear

Regulatory Research, U. S. Nuclear Regulatory Commission, Washington, DC 20555-

0001; September 2008

2. Bajorek S. at al. TRACE V5.0 USER'S MANUAL, Volume 1: Input Specification; Division

of Systems Analysis and Regulatory Effectiveness; Office of Nuclear Regulatory

Research, U. S. Nuclear Regulatory Commission, Washington, DC 20555-0001; October

2008

3. Bajorek S. at al. TRACE V5.0 USER'S MANUAL, Volume 2: Modeling Guidelines; Office

of Nuclear Regulatory Research, U. S. Nuclear Regulatory Commission, Washington,

DC 20555-0001; October 2008

4. 0. I. Melikhov, V. I. Melikhov; Analysis of LOCA thermal-hydraulics of VVER 1000.

Description of PSB-WER Test Facility; Moscow Power Engineering Institute (Technical

University), NPP Department; Moscow; 2008

5. 0. I. Melikhov, V. I. Melikhov; Analysis of LOCA thermal-hydraulics of VVER 1000. Hot

Leg Large Break 2x25%; Moscow Power Engineering Institute (Technical University),

NPP Department; Moscow; 2008

6. P. Heraleckq,; Engineering Handbook modelu PSB-1.0; TES-Z-09-175; TES s.r.o. Tfebid;

prosinec 2009

7. P.D. Bayless, 0. Melikhov, V. Melikhov, Y. Parfenov, 0. Gavritenkova, I. Elkin, I.

Lipatov; RELAP5/MOD3.2 Assessment Using INSC SP-PSBV1; Seventh International

Information Exchange Forum On SAFETY ANALYSIS FOR NUCLEAR POWER

PLANTS OF WER AND RBMK TYPES; Slovakia, Pie.t'any; 28-30 October 2003

8. B. Neykov, M. Cherubini, F.D. Auria; Relap 5 Thermal Hydraulic analysis 11% upper

plenum leakage for PSB-WWER test facility and WWER NPP

9. A. Del Nevo, F.D. Auria, M. Mazini, M. Vylov, I.V. Elkin, A. Suslov; The Design of PSBWER Experiments Relevant to Accident Management; Journal of Power and Energy

Systems; Vo12, No. 1, 2008

10. V. Blinkov, 0.1. Melikhov; Effect of Steam Generator Nodalization on RELAP5 Simulation

of Loss-of-Feed-Water Transient in PSB-VVER Facility; Seventh International

Information Exchange Forum On SAFETY ANALYSIS FOR NUCLEAR POWER

PLANTS OF VVER AND RBMK TYPES; Slovakia, Pie.t'any; 27-31 October 2003

11. L. Sabotinov, P. Chevrier; Post-Test Analysis of 11% Break at PSB-WER Experimental

Facility using Cathare 2 Code; Journal of Power and Energy Systems; Vo12, No. 2, 2008

12. Thermal-Hydraulic Codes News, Volume 09,. 4NoLS. Nuclear Regulatory

Commission; December 2009

13. P. Heraleckq,; Srovnavaci v•,podet testu PSB-WER UP-11-07 v prostfedi k6du TRACE

5.0; TES-Z-09-179; TES s.r.o.; Tfebi,; prosinec 2009

14. Preliminary Safety Assessment Report,. Rev.2 - Input Data for Safety Analysis of NPP

Dukovany, Revision V5. 01-March-2007

15. Staudenmeier J. TRACE Code Modelling Capabilities. Proceedings of TRACE/SNAP

User Workshop. Potomac, Maryland, .27 - 29 March 2006

8-2
16. Bajorek S. at al. TRACE V5.0 Assessment Manual - Main Report. Division of Risk

Assessment and Special Projects, Office of Nuclear Regulatory Research, U. S. Nuclear

Regulatory Commission

17. Petruzzi A. and Auria F. Thermal-Hydraulic System Codes in Nuclear Reactor Safety

and Qualification Procedures. Science and Technology of Nuclear Installations Volume

2008, Article ID 460795

18. WNA Reactor Database. March 3, 2008

19. OECD Support Group on the VVER Thermal-Hydraulic Code Validation Matrix.

Validation Matrix for the Assessment of Thermal-Hydraulic Codes For WER LOCA and

Transients. NEAICSNI/R(2001)4. 01 June 2001, Paris, France

20. 0. I. Melikhov, V. I. Melikhov; Analysis of LOCA thermal-hydraulics of WER 1000.

Single-phase characteristics of the PSB-WER main circulation pumps; Moscow Power

Engineering Institute (Technical University), NPP Department; Moscow; 2008

8-3

APPENDIX A

INPUT DECK NODALISATION SCHEMES

B I

51

\ ::- -W|i f :i:- -,• ..-:.•tt -:-+-. V V vV t

5A0 I

A-1-
ii

,,, • .. . . -i -i J ... ........ .... ................ ! 1 ........ -.. 

.... ......... ....... .- . .

* I I

-- I

Figure A- 2 Nodalisation. Scheme of Primary Circuit of PSB-WVER Facility (11D Model)

A-2
II .

C4J

A-3

K ... , 1 ,~ ; I'rI :

Figure A-3 Nodalisation Scheme of Secondary Circuit of PSB-WER Facility (for both models)

A-3

APPENDIX B

MEASUREMENT LOCALISATION AT PSB-VVER FACILITY

yCOI - reaCtOl m'odel

-<990

Figure B-1 pSB-WER Reactor Model Measurements

B-i
I-.

0

0

0

Figure B-2 PSB-VVER Loop 1 and SG-1 Model Measurement

B-2
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-0248

(See instructions on the reverse)

2. TITLE AND SUBTITLE 3. DATE REPORT PUBLISHED

Post-Test Analysis of Hot Leg 2x25% Break at PSB-WER Facility using TRACE V5.0 MONTH YEAR

Code February 2011

4. FIN OR GRANT NUMBER

5. AUTHOR(S) 6. TYPE OF REPORT

Petr Heralecky, Martin Blaha Technical

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.)

TES Ltd

Prazska 597

674 01 Trebic, Czech Republic

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 or less)

The best estimate thermo-hydraulic computer code TRACE V5.0 has been assessed using Hot Leg 2x25% break

experiment at the large-scale test facility PSB VVER. The PSB-WER facility is a 1:3000 volume scaled model of WER

1000 NPP located in Electrogorsk, Russia. An extensive TRACE input deck of PSB-WER facility was developed

including all important components of the PSB-WER facility: reactor, 4 separated loops, pressurizer, break unit, main

circulation pumps, steam generators, break section and important parts of secondary circuit. The TRACE (TRAC/RELAP

Advanced Computational Engine) is the latest in a series of advanced, best-estimate reactor systems codes developed

by the U.S. Nuclear Regulatory Commission in frame of CAMP (Code Application and Maintenance Program). The

TRACE code is a component-oriented reactor systems analysis code designed to analyze light water reactor transients

up to the point of significant fuel damage. The original validation of the TRACE code was mainly based on experiments

performed on experimental facilities of typical PWR design. There are some different features of WER design comparing

to PWR. Therefore the validation of the thermo-hydraulic codes for WER types of reactors is required by regulators. The

purpose of performed analysis is to extend the validation of the TRACE code focused on WER type of NPPs. This work

was performed in the frame of R&D project sponsored by The Ministry of Industry and Trade of Czech Republic.

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

Electrogorsk, Russia- WER-1000 NPP unlimited

FI-IM5/150 project 14. SECURITY CLASSIFICATION

TRACE input deck (This Page)

TES Company unclassified

PWR safety codes (This Report)

Czech Ministry of Industry and Trade unclassified

PSB-WER facility 15. NUMBER OF PAGES

Hot Leg 2x25% break experiment

R&D project 16. PRICE

Code Application Maintenance Program (CAMP)

NRC FORM 335 (9-2004) PRINTED ON RECYCLED PAPER


Fderal Recycling Program

NUREG/IA-0248 Post-Test Analysis of Hot Leg 2x25% Break at

PSB-WER Facility using TRACE V5.0 Code

February 2011

UNITED STATES

NUCLEAR REGULATORY COMMISSION

WASHINGTON, DC 20555-0001

OFFICIAL BUSINESS