NUREG/IA-0204

International

Agreement Report

OLKILUOTO 2 RELAP5/MOD3.2.1.2

Analysis of the Reactor Scram on

June 13, 1997

Prepared by:

Heikki Sj6vall

Teollisuuden Voima Oy

FIN-27160 OLKILUOTO

FINLAND

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

April 2007

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-0204

International

( I Agreement Report

OLKILUOTO 2 - RELAP5/MOD3.2.1.2

Analysis of the Reactor Scram on

June 13, 1997

Prcparcd by:

Heikki Sj6vall

Teollisuuden Voima Oy

FIN-27160 OLKILUOTO

FINLAND

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

April 2007

Prepared as part of

The Agreement on Research Participation and Technical Exchange

Undcr the International Code Assessment and Maintenance Program (CAMP)

Published by

U.S. Nuclear Regulatory Commission

ABSTRACT

The reactor scram, which occurred at Olkiluoto 2 June 13 in 1997 has been analysed with the

RELAP5/MOD3.2.1.2 computer program. The RELAP5 model of the Olkiluoto reactors is based on the

model prepared for analysing PSA level 1 success criteria. The purpose of the analysis was to

investigate how well RELAP5 is able to calculate the behaviour of process parameters after scram and

validate the RELAP5 model of the Olkiluoto reactors.

The main plant features connected with the event are briefly presented. Modelling aspects are

discussed regarding the RELAP5 analysis.

The RELAP5 result agrees with the measured reactor water level at the beginning of the transient

when the level drops due to the collapse of steam bubbles. Later the level rise due to supply of feed

water and the warming up of the reactor water inventory is slower in the RELAP5 analysis than

according to the measurements.

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requirements of the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et seq.).

Public Protection Notification

The NRC may not conduct or sponsor, and a person is not required to respond to, a request for

information or an information collection requirement unless the requesting document displays a

currently valid OMB control number.

iii

FOREWORD

This work was generated by international partners under the CAMP program and constitutes the

contributing organization's contribution to international thermal hydraulic code development. The

contents of this report should not be viewed as an official NRC endorsement of the products or

services of other persons or entities that may be referred to herein. Nor should this report be

viewed as binding the NRC in its rulemaking, licensing or adjudicatory process.

V

CONTENTS

Page

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

FO REW ORD .................................................................................................................... v

CO NTENTS ........................................................................................................................................ vii

LIST OF FIGURES ............................................................................................................................... ix

LIST OF TABLES .................................................................................................................................. ix

EXECUTIVE SUM MARY ..................................................................................................................... xi

1. INTRODUCTIO N ................................................................................................................... 1

2. OLKILUOTO 2 REACTOR SCRAM ON JUNE 13, 1997 .................................................. 3

3. O LKILUOTO I AND 2 RELAP5 M ODEL .......................................................................... 5

3.1 General ..................................................................................................................... 5

3.2 Steady State ........................................................................................................ 6

3.3 Transient ................................................................................................................... 6

4. INPUT DATA ........................................................................................................................ 7

4.1 Nodal m odel ........................................................................................................ 7

4.2 Reactor core and nuclear fuel .............................................................................. 8

4.3 Heat structures ...................................................................................................... 8

4.4 Steam separators .................................................................................................. 9

4.5 M ain circulation pum ps ........................................................................................ 9

4.6 M ain steam lines .................................................................................................. 10

4.7 Reactor level m easurem ent ............................................................................... 10

4.7.1 M ethod of M easurem ent ........................................................................ 10

4.7.2 Density com pensation .......................................................................... 11

4.7.3 Scope of Instrum entation ...................................................................... 12

4.7.4 Normal System Operating Conditions .................................................. 13

4.7.5 Abnorm al plant operating conditions .................................................. 14

4.7.6 Water level indication in the Olkiluoto RELAPS Model ........................ 16

4.8 Reactor Pressure M easurem ent ........................................................................ 16

4.8.1 Norm al system operating conditions .................................................... 16

4.8.2 Abnorm al system operating conditions ................................................ 16

4.8.3 Reactor Pressure indication in the Olkiluoto RELAPS Model .............. 17

4.9 Feed W ater .............................................................................................................. 17

5. RESULTS ........................................................................................................................... 19

5.1 Steady State ...................................................................................................... 19

5.2 Transient ............................................................................................................... 19

6. DIFFERENCES BETWEEN CALCULATIONS AND MEASUREMENTS ........................ 25

7. CONCLUSIONS ................................................................................................................. 27

8. REFERENCES ................................................................................................................... 29

vii

LIST OF FIGURES

Page

Figure 1. Olkiluoto 1 and 2 reactor pressure vessel. RELAP5 model nodalization ......................... 7

Figure 2. Olkiluoto 2 reactor scram June 13, 1997. Main circulation flow ......................................... 10

Figure 3. Olkiluoto 2 reactor scram June 13, 1997. Reactor fine level .............................................. 11

Figure 4. Olkiluoto 2 reactor scram June 13, 1997. Reactor pressure .............................................. 17

Figure 5. Olkiluoto 2 reactor scram June 13, 1997. Feed water flow ............................................... 18

Figure 6. Olkiluoto 2 reactor scram June 13, 1997. Feed water temperature ................................... 18

Figure 7. Olkiluoto 2 reactor scram June 13, 1997

Void fraction in the core high power region fuel bundles ................................................................. 20

Figure 8. Olkiluoto 2 reactor scram June 13, 1997.

Void fraction in the core medium power region fuel bundles .......................................................... 21

Figure 9. Olkiluoto 2 reactor scram June 13, 1997.

Void fraction in the core low power region fuel bundles .................................................................... 21

Figure 10. Olkiluoto 2 reactor scram June 13, 1997.

Void fraction in the upper plenum and in the steam separator standpipes ........................................ 22

Figure 11. Olkiluoto 2 reactor scram June 13, 1997.

Void fraction between the steam separator standpipes and in the

down comer at the level of the shroud head ..................................................................................... 22

Figure 12. Olkiluoto 2 reactor scram June 13, 1997.

Calculated mass of water in the reactor pressure vessel ................................................................. 23

LIST OF TABLES

Table 1. Olkiluoto 2 steady state prior to reactor scram on June 13, 1997 ......................................... 6

Table 2. OL1 and 0L2 reactor water level signals at the time of the 0L2 scram June 13, 1997 ..... 12

Table 3. OL1 and 0L2 current reactor water level signals ............................................................... 12

Table 4. Olkiluoto 2 steady state prior to reactor scram on June 13, 1997.

Comparison between measurements and the RELAP5 calculation ................................................. 19

ix

EXECUTIVE SUMMARY

The reactor scram, which occurred at OL2 June 13 in 1997 has been analysed with the

RELAP5/MOD3.2.1.2 computer program. The RELAP5 model of the Olkiluoto reactors is based on the

model prepared for analysing PSA level 1 success criteria. The purpose of the analysis was to

investigate how well RELAP5 is able to calculate the behaviour of process parameters after scram and

validate the RELAP5 model of the Olkiluoto reactors.

The plant unit OL2 operated on June 13, 1997 at the old 105 % power (2272) during the test operation

for the power up rating. The periodic test of main steam line radioactivity level measurements was

being performed. The tests actuated the scram signal in each of the four subs A, B, C and D, one

channel at the time. The reactor scram operates with 2/4 logic, which means that a scram signal in two

of the four channels will lead to reactor scram. After the test of the channel D the scram signal (SS

chain) was not restored. The channel B was tested after the channel D, which led to actuation of

scram signal also in the channel B. Because the scram signal was actuated in two channels the

reactor was scrammed automatically.

Regarding the behaviour of the process parameters after the scram the most important observation

was that the reactor water level decreased below the scram level L2 (+3.1 m above top of active fuel).

This led to the stop of the pumps supplying water to the reactor water clean-up system. However,

because the reactor was scrammed before the L2 signal no other consequences followed. The

stopped pumps started again when the water level was again above L2.

The reason for the large reactor water level drop was the void collapse of the reactor after the scram.

The original steam separators and the shroud cover were replaced in the refuelling outages in 1997

with new ones in order to accommodate the uprated power level of 2500 MWth. The capacity of the

old steam separators would not have achieved sufficiently low moisture content of the steam. The

shroud cover height was increased by 0.4 m and the shroud cover volume was thus increased by 6.3

mi3 . This increase of the reactor two-phase mixture volume further increased, in addition to the larger

core void content due to the power uprating, the reactor level drop compared to the original design

values. The lowest reactor water level was +2.91 m above the top of active fuel at 24 seconds from

the scram signal. The lower fine level measurement connection is at the level +2.5 m above the top of

active fuel and the upper at the level +6.2 m.

The reactor scram, has been analysed with the RELAP5/MOD3.2.1.2 computer program. The

RELAP5 model of the Olkiluoto reactors is based on the model prepared for analysing PSA level I

success criteria. The purpose of the analysis was to investigate how well RELAP5 is able to calculate

the behaviour of process parameters after scram and validate the RELAP5 model of the Olkiluoto

reactors.

xi
Several changes were required in the Olkiluoto 1 and 2 models, which was prepared before the

scram, in order to reach good agreement between the calculation and the measurement data. The

changes concerned nodal volumes, node junction flow areas and pressure loss coefficients and heat

structures. Default values were retained for RELAP5 thermal hydraulic parameters.

The RELAP5 result agrees with the measured level at the beginning of the transient when the level

drops due to the collapse of steam bubbles. According to the analysis the fine level dropped from the

initial value +3.86 m above top of active fuel to the minimum value +2.91 m in 23 s. According to the

measurements the same minimum level was reached at 24 s after the scram. The level rise from 24 s

to 120 s due to supply of feed water and the warming up of the reactor water inventory is slower in the

RELAP5 analysis than according to the measurements. The reason can be that RELAP5 calculates

the amount of steam condensing in water is larger than in reality. The amount of steam in water is thus

smaller than the real void. The stronger steam condensation is also reflected by the observation that

the calculated water level drops faster than the measured level. RELAP5 also calculates at the

beginning of the transient a lower reactor pressure than what is measured. At this stage there is due to

feed water flow a large amount of sub cooled water in the vessel to condense steam.

xii
1. INTRODUCTION

On the west coast of Finland, in Eurajoki, Teollisuuden Voima Oy (TVO) owns and operates two 840

MWe boiling water reactors. The reactors Olkiluoto 1 (OL1) and Olkiluoto 2 (OL2) were supplied on a

turnkey basis by the Swedish company AB ASEA-ATOM (currently part of Westinghouse Electric

Company). The OL1 unit was first connected to the national grid in September 1978 and the OL2 unit

in February 1980.

The reactor core has 500 fuel bundles. The units have been uprated twice since the commissioning.

The thermal power of each reactor was increased from 2000 MWth to 2160 MWth in 1984 and to 2500

MWth in 1998. The corresponding nominal values of the net electrical output were 660 MWe,

respectively.

Four-fold redundancy is used throughout for systems having safety functions. The systems, e.g. safety

systems and safety related instrumentation, are divided into four sub-systems, designated A, B, C and

D which are subject to specific separation rules as regards the physical location of mechanical and

electrical components (including cabling), power supplies and signal transmission.

The reactor scram, which occurred at OL2 June 13 in 1997 has been analysed with the

RELAP5/MOD3.2.1.2 computer program. The RELAP5 model of the Olkiluoto reactors is based on the

model prepared for analysing PSA level 1 success criteria. The purpose of the analysis was to

investigate how well RELAP5 is able to calculate the behaviour of process parameters after scram and

validate the RELAP5 model of the Olkiluoto reactors.

RELAP5/MOD3.2.1.2. computer program simulates the thermal hydraulics of light water reactors. The

code is based on best-estimate methods to solve inhomogeneous and unsteady two-phase flow in one

dimension. The code development is financed by the United States Nuclear Regulatory Commission.

Several changes were required in the Olkiluoto 1 and 2 model, which was prepared before the scram,

in order to reach good agreement between the calculation and the measurement data. The changes

concerned nodal volumes, node junction flow areas and pressure loss coefficients and heat structures.

Default values were retained for RELAP5 thermal hydraulic parameters. TVO uses RELAP5 for the

analyses of PSA level success criteria and other thermal hydraulic analyses.

I

2. OLKILUOTO 2 REACTOR SCRAM ON JUNE 13, 1997

The plant unit OL2 operated on June 13, 1997 at the old 105 % power (2272 MWth) during the test

operation for the power uprating. The periodic test of main steam line radioactivity level measurements

was being performed. The tests actuated the scram signal in each of the four subs A, B, C and D, one

channel at the time. The reactor scram operates with 2/4 logic, which means that a scram signal in two

of the four channels will lead to reactor scram. After the test of the channel D the scram signal (SS

chain) was not restored. The channel B was tested after the channel D, which led to actuation of

scram signal also in the channel B. Because the scram signal was actuated in two channels the

reactor was scrammed automatically.

Regarding the behaviour of the process parameters after the scram the most important observation

was that the reactor water level decreased below the scram level L2 (+3.1 m above top of active fuel).

This led to the stop of the pumps supplying water to the reactor water clean-up system. However,

because the reactor was scrammed before the L2 signal no other consequences followed. The

stopped pumps started again when the water level was again above L2.

The reason for the large reactor water level drop was the void collapse of the reactor after the scram.

The original steam separators and the shroud cover were replaced in the refuelling outages in 1997

with new ones in order to accommodate the uprated power level of 2500 MWth. The capacity of the

old steam separators would not have achieved sufficiently low moisture content of the steam. The

shroud cover height was increased by 0.4 m and the shroud cover volume was thus increased by 6.3

mi3. This increase of the reactor two-phase mixture volume further increased, in addition to the larger

core void content due to the power uprating, the reactor level drop compared to the original design

values. The lowest reactor water level was +2.91 m above the top of active fuel at 24 seconds from

the scram signal. The lower fine level measurement connection is at the level +2.5 m above the top of

active fuel and the upper at the level +6.2 m.

The behaviour of the reactor water level and reactor pressure are presented in the report with the

corresponding RELAP5 results in the figures 3 and 4. Main circulation flow, feed water flow and feed

water temperature given as input for the RELAP5 calculation are presented in the figures 2, 5 and 6.

Reactor water level, reactor pressure, main circulation flow and feed water flow are obtained from OL2

process computer print out of the reactor scram event. The feed water temperature is obtained from

OL2 measurement computer print out.

3

3. OLKILUOTO 1 AND 2 RELAP5 MODEL

3.1 General

The model includes the reactor pressure vessel and the main steam lines to the main steam isolation

valves. Other reactor process system components are modelled as boundary conditions utilizing the

time dependent volumes and junctions of RELAP5.

The essential parts of the reactor protection systems are modelled by trip and control variables. The

following reactor protection system signals are included:

-reactor scram SS signals (SSl to SS15)

-containment isolation I signals

-automatic primary system depressurization TB signals

-low reactor water level (L3 +2.0 m above top of active fuel) Xl signal

-low reactor pressure (L3 1.2 MPa) X5 signal.

The containment is not included in the RELAP5 Olkiluoto input file. That is why the following reactor

protection systems are not modelled:

-reactor building room supervision Y signals (Y1 to Y16)

-main steam line supervision A signals (Al to A23)

-feed water line supervision M signals (M1 to M8 and MT9)

-safety system room supervision HA, HB, HC, HD signals

-high radioactivity level in the reactor hall X2 signal

-high condensation pool temperature X3 signal

-low pressure difference between diesel backed normal operation second cooling system and the

lower drywell space.

Neither reactor kinetics included in the RELAP5 Olkiluoto input file. That is why the following reactor

protection systems are not modelled:

-reactor screw stop V signals (Y1 to Y16)

-control rod withdrawal locking at power less than 8 % (SI to S5)

-control rod withdrawal locking at power equivalent to or greater than 8 % (El to E4)

-reactor scram B signals during refuelling outage

The control rod drive system provides two independent ways for insertion of the control rods, one

electro-mechanical system for normal operation with fine motion and one hydraulic system for fast

scram insertion.

5
3.2 Steady state

The operation point at the time of the reactor scram in Olkiluoto 2 June 13, 1997 is presented in the

table 1. The data is obtained from process computer.

Table 1. Olkiluoto 2 steady state prior to reactor scram on June 13, 1997.

Reactor

thermal power

(MW)

2272

Pressure (MPa)

Temperature

Flow rate (kg/s)

Reactor fine

Steam dome

pressure

Feed water

Feed water

Main steam line

Downcomer

7.0

180

1131

1158

7850

3.86

(m above top of

active fuel)

3.3 Transient

The scram signals SS1 (manual scram) is assumed to be released at the beginning of the transient

calculation at 0 sec into the transient. Manual scram is used because the signals, which actually

released the scram, are not modelled in the Olkiluoto RELAP5 model. The hydraulic scram system

starts to insert control rods into the core at 0.2 sec from the scram signal. The reactor fission power is

zero at 4.2 sec, when the control rods are fully in the reactor core. After that the core thermal power is

assumed to follow the ANS 5.1-1979 standard decay heat curve.

The main circulation pump run-down starts 0.2 sec from the scram signal.

The feed water flow and temperature are set according to the data obtained from the process and

measurement computers. The pressure regulator will control the reactor pressure at 7.0 MP.

6
4. INPUT DATA

4.1 Nodal model

The RELAP5/MOD3.2.1.2 nodal model of Olkiluoto reactors is presented in the figure 1. The nodes

are drawn according to the scale regarding both height and flow area. Due to drawing technical

reasons flow area of some nodes, however, is not according to scale.

RELAP5/MOD3.2.1.2. OLKILUOTO 1 AND 2 REACTOR

PRESSURE VESSEL NODALIZATION.

02

a 8

93

-02

10

-- -

2 .2

line-31- - -- 1 -a Ma312 syte-Fee --- -e 

313 M-i - p--- -i~lto -u 

s5r2 syte - -323- -o 

-327- - - - - -uiir -ee -ae s se

O--iluoo -. 1

.0 .4

System abbre.iations

311.: Main Steam hres

312. Feed water system

3 1 3- Main cicu~ation pumps

323-: Core spray system

327- Auxihary Feed water system

Figure 1. Olkiluoto 1 and 2 reactor pressure vessel. RELAP5 model nodalization

7
4.2 Reactor core and nuclear fuel

The data of reactor core nodal volumes, flow areas and heat structures are according to ATRIUM 10-

9Q nuclear fuel.

The radial and axial power distributions of the model are based on POLCA4 calculations. POLCA4

was the incore fuel management code used by TVO at the time of the transient. Based on the radial

power distribution the 500 fuel bundles of the reactor core are divided in four regions, which describe

the fuel bundle groups of different power levels:

-low power region, radial power coefficient p,=0. 3 7 5 , 88 fuel

-bundles

-medium power region, radial power coefficient p,=0.998, 260 bundles

-high power region, radial power coefficient p,=1.364, 151 fuel bundles

-hot channel, radial power coefficient p,=1.514, 1 fuel bundle

In the axial direction the core is divided in 25 nodes.

The core by-pass includes the volume between the fuel bundles, the volume between the core

boundary fuel bundles and core shroud and the water channel volume of the fuel bundles. In the axial

direction the by-pass channel is divided in 25 nodes.

The RELAP5 point neutron kinetics model is not used for the reactor power calculation, but the steady

state is calculated with a constant core thermal power of 2272 MWth. After the reactor scram the core

power generation is assumed to follow the decay heat curve according to the ANS 5.1-1979 - standard. Two percent of the reactor power is assumed to be transferred directly to the coolant.

4.3 Heat structures

The reactor pressure walls are modelled by a heat structure of carbon steel, the thickness of which is

139 mm in the skirt, 187 mm in the bottom and 85 mm, 138 mm and 371 mm in the top head. The

vessel has a 5 mm stainless steel cladding except in the bottom, which has a 5 mm Inconel cladding.

The cladding is modelled as one layer and the rest of the wall is divided in five or seven layers. The

thickness of one calculation layer is thus 17.0 to 52.2 mm. The default RELAP5 material properties of

the materials are used for thermal conductivity and specific heat. The reactor vessel outer surface is

assumed thermally insulated and the heat losses to the containment are not modelled. In reality the

thermal losses at reactor operating temperature of 286 °C are ca. 700 kW.

The thickness of the reactor internal heat structures varied from 0.75 mm (fuel cladding) to 45 mm

(reactor shroud and shroud head). The same material properties are used both for fuel cladding

(Zircaloy-2) and fuel channels (Zircaloy-4). The material properties are presented in the table 3.

8
4.4 Steam separators

OL1 and OL2 have 174 GENE AS2B steam separators by General Electric. They are modelled with

the mechanistic RELAP5 steam separator model for General Electric separators.

4.5 Main circulation pumps

The total core coolant flow in the reactor pressure vessel is calculated by summating the flow

measurements in eight strategically located fuel assemblies.

The distribution of the core coolant flow throughout the core is determined by calibrated orifice plates

in each fuel assembly seat. The selected seats are provided with pressure tappings for measuring the

pressure drop across the orifice plate. The tappings are located immediately above the plate and

approx. 500 mm under the fuel assembly support plate. The small-bore tubes from the pressure

tappings are run in pairs through guide tubes to penetrations in the bottom of the pressure vessel and

thence to the flow measurement differential pressure transmitters.

The instrumented orifice plates are calibrated individually during charging of the reactor using a

calibration assembly incorporating a reference flow meter of the turbine type. The calibration assembly

is designed to simulate an ordinary fuel assembly hydraulically, to ensure that the flow pattern will be

the same as during normal operation of the reactor.

The eight flow measurement channels are sub-divided into two channels in each sub-group A, B, C

and D.

The flow measurement signals from the transmitters are summated to form combined flow signals in

four independent averaging units, one in each sub-group.

The flow meters have an inherent temperature dependency since the density is included as a factor in

the calibration constant. The maximum error due to normal variations in coolant temperature is

estimated as ±1%. The error is negative at low coolant temperatures (i.e. low main circulation pump

flow).

The reactor pressure vessel core coolant flow is measured continuously and is compared with the

neutron flux density in order to ensure that the thermal operating margins are maintained. The analog

signals from the four flow averaging units may be read in the control room, while these and the eight

individual flow signals from the fuel assemblies are also supplied to the station computer.

The core coolant temperature is measured continuously, displayed and recorded in the control room,

and supplied to the station computer, as is the pressure rise signal from the main circulation pumps.

Together with the measurement of core coolant flow and the main circulation pump speeds, the latter

measurements are used to establish the pump operating points, and to check that the correction for

total core coolant flow is correctly adjusted.

The six main circulation pumps are modelled by RELAP5 pump component. For the steady state

calculation a control function was used to regulate the main circulation flow to 7850 kg/s, which flow

was measured at the time of the reactor scram. In the transient calculation the controller was removed

and the pump was steered down by changing directly the pump rotation speed. The pump run-down

was performed as a function of time so that the calculated main circulation flow agreed as well as

possible with the measured value (figure 2).

9
OL2 - REACTOR SCRAM JUNE 13,1997. RELAP5 MOD3.2.1.2

Main circulation flow

0

IL

0 10 20 30 40 50 60 70 80 90 100 110 120

Time from scram signal (a)

Figure 2. Olkiluoto 2 reactor scram June 13, 1997. Main circulation flow.

4.6 Main steam lines

The main steam lines are modelled to the main isolation valves (node 311). For the steady state

calculation a control function was used to regulate the steam flow out the reactor pressure vessel so

that a constant steam dome pressure of 7.0 MPa was attained. The controller was used also in the

transient calculation. The pressure controller simulates the actual reactor pressure regulator in a very

simplified way. However, the function is satisfactory for the analysis.

4.7 Reactor level measurement

4.7.1 Method of measurement

The level instrumentation is designed to measure the water level in the reactor pressure vessel

downcomer. The indirect method, by which the pressure head is measured at a fixed point below the

lowest level which must be indicated, is used for this purpose.

Since in this case the level is being measured in a closed vessel under pressure, with condensable

vapor above the free surface of the liquid, the method used must employ differential pressure

measurement. A differential pressure measuring device is arranged with the negative pressure side

connected to an instrument tapping on the vessel at the lower limit of the desired range and the

positive pressure side connected through a reference vessel to a tapping on the steam chamber of the

vessel at or above the upper limit of the range. A specific water level is maintained in the uninsulated

reference vessel by means of condensing steam, the excess water being returned to the reactor

pressure vessel through an overflow.

10
The dp-cell is installed below the level of the negative pressure tapping. The connections to the

tapping and the reference vessel are of small-bore piping sloping continuously upwards towards the

vessel to allow the escape of gas.

OL2 - REACTOR SCRAM JUNE 13, 1997. RELAP5 MOD3.2.1.2

Reactor fine level

.... Calculated fine level

4.10 - Measured One level

3.390 -

3.70- .

o 350,

*•

3.30 . .

3.10-*

2 .9 0 I l I I I I I

0 10 20 30 40 50 60 70 80 90 100 110 120

Time from scram signal (s)

Figure 3. Olkiluoto 2 reactor scram June 13, 1997. Reactor fine level.

4.7.2 Density compensation

The major advantage afforded by the method of measurement described above is that no instruments

require to be installed in or on the reactor pressure vessel, thus avoiding the considerable

environmental problems which this would entail. However, there is a corresponding disadvantage in

that the results of measurement are dependent on the density of the medium. The influence of

variations in density may be considerable. The differential pressure corresponding to a specified water

column is approximately 30% lower at the normal operating conditions of 7 MPa and 2860C than at

room temperature

The results of measurement are also affected, albeit to a lesser extent, by the steam density and by

the density of the reference water column i.e. the water present in the instrument line from the

reference vessel to the level of the negative pressure tapping. Below this level, the heads due to the

water columns in the water lines to the differential pressure transmitter cancel each other. Thus, the

location of the transmitter is of minor importance.

The level measuring units are designed to provide the most accurate possible indication of the level of

the boundary surface between the steam and water phases (the latter of which may consist of a

mixture of water and small steam bubbles or voids) during normal operation, start-up and shut-down.

Thus, the units are provided with compensators, which automatically correct the differential pressure

signals for variations in density.

I1
The density of the water is a function of the temperature, pressure and void fraction, while the steam

density is a function of the pressure and temperature. However, since the saturated steam in the

reactor is in thermal equilibrium with the water in the downcomer, the temperature is governed by the

pressure. Therefore, the reactor pressure and the temperature of the reference water column are used

as the controlling input parameters for the density compensator.

4.7.3 Scope of instrumentation

Each of the instrument sub-groups (A, B, C and D) includes a fine and a coarse measuring unit, both

of which are equipped with density compensators. The positive sides of both units are connected to

the same reference vessel (+6.4 m above the top of active fuel), while the negative sides are

connected at different levels to individual branches on the pressure vessel (fine level +2.5 m, coarse

level +0.2 m). Each sub-group is assigned its own reference vessel and instrument tappings. In

addition, each group is provided with a reference temperature sensor in the form of a resistance

surface thermometer mounted on the line to the reference vessel and supplying a control signal to the

density compensators in the subgroup through a transmitter.

The instrumentation also includes two core level measuring units (sub-groups C and D, positive side

+6.4 m)) with the negative sides connected to tappings on the moderator tank (-6.2 m).

In addition, the unit includes a single high level measuring unit. The individual reference vessel for this

unit is mounted below the pressure vessel flange (+8.2 m) while the negative pressure side is

connected to one of the fine level measurement tappings (-6.2 m). This unit is not provided with

density compensation.

The level signals at the time the analyzed OL1 scram were initiated at the levels presented in the table

2.

Table 2. OLI and OL2 reactor water level signals at the time of the OL2 scram June 13, 1997.

H +4.7mATAF

2 +4.2 m ATAF

H +3.9 m ATAF Normal level

1 +3.6 m ATAF

LO +3.1 m ATAF

LI +2.0 m ATAF

L2 +0.7 m ATAF

L3

L4

After the OL2 scram the reactor water level was increased due to the large level drop and the current

level signals are presented in the table 3.

rable 3. OL1 and OL2 current reactor water level signals.

H +5.0miATAF

2 +4.3 m ATAF

H +4.2 m ATAF Normal level

1 +3.6 m ATAF

LO +3.1 m ATAF

LI +2.0 m ATAF

L2 +0.7 m ATAF

L3

L4

12
4.7.4 Normal system operating conditions

The signals from the differential pressure transmitters used for level measurement are supplied to

density compensators, which are also supplied with signals from the coarse pressure measuring unit

and the reference temperature sensor in the corresponding sub-group. The outputs from the density

compensators are in the form of analog signals proportional to the water level in the reactor pressure

vessel downcomer.

These signals are fed to limit switches which produce limit value signals and also to one or more

isolation amplifiers. The control instrumentation (i.e. controllers, indicating instruments, recorders and

station computer inputs) is connected to these amplifiers. The provision of sufficient level information

is ensured by this means, while galvanic isolation between the sub-groups is also maintained.

The foregoing does not apply to the high level measuring unit which is not equipped with a density

compensator. The transmitter for this measuring unit is calibrated for correct level indication when the

reactor is cold (30-600C). The level signal is transmitted to an indicating instrument in the control room.

A number of systematic error sources are present under normal operating conditions. These are due

partly to the flow of steam and water through the reactor vessel and partly to the inflow of feed water.

The pressure sensed at the level measurement tappings and transmitted to the differential pressure

transmitters is the static pressure of the medium, steam or water. However, the application of

Bernoulli's equation demonstrates that the level measurement method requires that the total pressure,

i.e. the sum of the static and dynamic pressures, be measured. Thus, a measuring error equivalent to

the difference between the dynamic pressure components at the positive and negative tappings, is

present. This error is greatest in the case of the fine level measuring units, since the water at the

negative tappings of these is more or less static. The error is also dependent on the reactor power

since the flow velocity is proportional to the quantity of steam generated. At full steam flow, the

maximum error is estimated as +0.05 m, which is negligible with respect to the safety margins of the

level limits.

A measuring system error may also be caused by voids or undercooling in the downcomer, since the

density compensators are designed to operate with void-free water at saturation temperature.

The level reading supplied by the level measurement units represents in reality the equivalent level of

void-free water above the negative instrument tappings. The true level of the boundary surface

between the water and steam phases may differ from the indicated level. In the downcomer, the

deviation is determined mainly by the void fraction while the drop in steam pressure in the steam

dryers and the radial pressure drop due to the main circulation flow between the steam separator

branches, are additional factors effecting the deviation inside the steam dryer casing. The latter factor

results in convexity of the boundary surface among the steam separators. The steam dryer pressure

drop is about 600 Pa (0.09 m water height at saturation temperature). It is estimated that the convexity

water level gradient is about 0.16 m.

At normal full power operation, the sum of these effects results in the level among the central steam

separator groups being slightly higher (about 0.07 m) than the indicated level. This fact leads to a

systematic negative indicating error. At half power, the error is of the order of 0.10-0.14 m.

Under normal station operating conditions, the fine level measuring units supply the feed water

controller with information regarding the water level in the reactor pressure vessel. The feed water

controller regulates the speed of the feed pumps (and thereby, the supply of feed water to the reactor

pressure vessel) so that the water level in the downcomer is maintained at the desired value within

permissible limits.

13
In the event of operating upsets, the alarm system will alert the operator if the indicated level should

exceed the H1 limit setting or fall below the Li setting.

The level indicated by the coarse and fine level measuring units is displayed on indicating instruments

and recorders in the control room, and is also supplied to the station computer. Under normal

operating conditions, the only function of the coarse unit is to provide back-up indication for the fine

unit.

The indication of core level is also displayed in the control room, although this does not fulfill a function

under normal operating conditions. This indication is relatively inaccurate due to the wide measuring

range and the low situation of the negative instrument tappings. With the reactor on load, undercooling

of the water in the downcomer gives a systematic deviation of the order of +0.25-0.30 m in relation to

the fine measurement.

Analog signals from the pressure measuring channels are used under normal operating conditions for

indicating and recording purposes in the control room, and for supplying measured value signals to the

station pressure control system. The analog signals are also supplied to the station computer. An

alarm is initiated if the pressure falls below LI or exceeds H1 in the event of an operating disturbance.

The output signals from the two pressure vessel temperature measuring points may be recorded on a

recording instrument. Pressure and temperature relationships will be unbalanced during start-up and

shut-down. Since the operation of the density compensators is based on the assumptions that the

reactor water is at saturation temperature and that the water columns are of equal temperature, the

indication error in the level measurement channels may be somewhat greater than during normal

operation.

During cold shut-down (e.g. during refueling) the level is measured up to the level of the pressure

vessel flange by means of the high level measuring unit.

The core coolant measuring units are calibrated for correct indication at normal reactor water

temperature. The flow indicated during the heating-up period will be too low. At reactor temperatures

of 500C and 1800C, the indication errors will be approx. 15% and approx. 8% respectively.

The temperatures of two selected points on the reactor pressure vessel cover and cover flange are

recorded during start-up and shut-down.

The rate of temperature rise of the coolant flow, which is a measure of the power, is measured and

displayed during nuclear heating using the computer. The prevailing rate of temperature rise is

displayed on the screen with a time lay of approx. 1 minute.

4.7.5 Abnormal plant operating conditions

Level limit signals to other systems are supplied through the logic circuits of the reactor protection.

Level H2 initiates reactor scram high reactor water level condition SS5, level L2 reactor scram low

reactor water level condition SS4, level L3 condition Xl, and level L4 containment isolation condition

12 and automatic reactor pressurization conditionTB1. An alarm is initiated at level H1 and reactor

scram is initiated at H2 in the event of an abnormal rise in reactor pressure vessel level due, for

example, to a fault in the feed water control system. The H2 setting is specified to ensure that reactor

scram takes place under all operating conditions before the moisture content of the outlet steam

becomes excessive.

14
In analogous manner, an alarm is initiated at LI and reactor scram occurs at L2 as the level falls. In

addition, partial operation of the auxiliary feed water system is initiated by L2. System 327 is

subsequently subject to on-off control by Li and HI. Level L2 is specified to ensure that reactor scram

occurs without fail before steam can flow through the steam separator water outlets. Li also reduces

the speeds of the main circulation pumps, thus reducing reactor power.

If the level should continue to fall, due for example to loss of water caused by pipe failure, the coarse

measurement level signals L3 and finally L4 will be initiated. Containment I isolation together with low

pressure X5 initiates water injection by the core spray system. Under these conditions, it must be

expected that the water in the downcomer may be heavily undercooled, which will result in a

systematic positive error indication by the coarse measurement units. However, at L4, the error will not

exceed 0.1 m even at the maximum conceivable degree of undercooling (100 °C). This value is

negligible.

Conditions in the reactor pressure vessel may be monitored with the aid of the core level measuring

units if extremely low levels are reached.

Rapid reductions in pressure in the reactor pressure vessel may be caused by conditions such as a

major pipe failure or certain types of malfunction in the turbine system. This will result in the

generation of voids in those sections of the reactor pressure vessel in which the water is at or close to

saturation temperature (i.e. flashing will occur). The voids will give rise to an increase in the volume of

the water phase, resulting in a rapid surge in water level in the downcomer. It is essential that the

containment isolation valves in the steam lines be closed before the water level reaches the steam

outlets, since the valves may otherwise be damaged by water hammer.

The sequence of events in circumstances of this type are dependent both on the nature of the incident

and on the operating status of the reactor when the incident occurs. In most cases, closure of the

containment isolation valves will be initiated from other systems such as isolation monitoring system,

although there are other cases in which the impulse must be supplied by level limit signal H2 or the

limit switches after the high pass filters connected to the coarse pressure measuring units.

In this instance, one of the problems is the fact that voids are also generated in the downcomer and

the negative error in the level indication thereby increases during the incident. If the total void fraction

in the downcomer exceeds a certain limit, the downcomer level will reach the level reference vessel

tappings before the indicated level has reached H2. Since the signal obviously cannot increase

following this, H2 will never be initiated. However, there is a counter-effect in the form of a pressure

increase at the negative tapping due to the flow of water in the downcomer which is interpreted as a

level increase by the level measuring system. Due to this effect, the indicated level is increased by

approx 0.2 m at a surge rate of 0.5 m/s.

The above factors are considered in specifying level setting H2, to ensure that H2 is initiated under all

relevant operating conditions, in sufficient time before the water level reaches the steam outlets. Both

the closing times of the containment isolation valves and the time delay in the level instrumentation

are taken into account. Certain improbable turbine failures may cause so rapid pressure reductions in

the reactor pressure vessel that there is a risk that these cannot be handled by the level

instrumentation. To obtain closure of the main steam line isolation valves at these failures, rapid

pressure reduction is detected by high pass filtering of the signals from the coarse pressure measuring

units. If the pressure in the reactor pressure vessel is reduced faster than can be tolerated during a

certain minimum of time, signals are obtained from limit switches for main steam line A isolation.

Coarse level measurement is usually unimportant in cases in which flashing occurs. However, there

are cases in which operation of the low level trips may conceivably occur, such as failure of a feed

water line during hot shut-down. Steam will then be generated in the whole reactor pressure vessel,

even below the feed water distributors and the coarse level measuring units will show a negative error

15
over entire range. This, however, is an advantage under such operating conditions since the critical

parameter is the total amount of water in the reactor pressure vessel and the indicated level is a better

measure of this parameter than the actual level of the water surface.

4.7.6 Water level indication in the Olkiluoto RELAP5 model

The Olkiluoto RELAP5 model includes fine, coarse and core level measurements. The level is

indicated as the collapsed water level height between the positive and negative side of the

measurement tappings because the level reading supplied by the level measurement units represents

in reality the equivalent level of void-free water above the negative instrument tappings,

-fine level (negative side +2.5 m, positive side +6.4 m)

-coarse level (negative side +0.2 m, positive side +6.4 m)

-core level (negative side -6.2 m, positive side +6.4 m).

4.8 Reactor pressure measurement

The level reference vessels are connected to a number of pressure transmitters. One fine and one

coarse pressure measuring unit are included in each sub-group (A, B, C and D) and one extreme

coarse pressure measuring unit in sub B. By high pass filtering of the signals from the coarse pressure

measuring units fast pressure reduction in reactor pressure vessel is supervised.

4.8.1 Normal system operating conditions

The static head due to the water column between the level reference vessel and the pressure

transmitters is included in the transmitter calibrations. Thus, transmitter outputs are proportional to the

pressure in the steam space of the reactor pressure vessel. The signals are supplied to limit switches

which produce pressure limit signals and to isolation amplifiers. The latter supply analog signals to the

instrumentation and to other systems requiring information on pressure.

The signals from the coarse level measuring units are also supplied to the density compensators in

the same sub-group. Further, these signals are high pass filtered. The outputs from the filters are

supervised by limit switches.

The dynamic pressure component due to the steam flow is negligible in comparison with the reactor

pressure and is completely unimportant to this measurement, nor would there appear to be any other

systematic error sources of importance to the measurement.

4.8.2 Abnormal system operating conditions

The above discussion regarding possible malfunctions in the level measuring channels, and their

consequences, applies also to the other pressure and differential pressure measuring systems, with

obvious modifications.

Signals are supplied to the reactor pressure relief system at pressure limits H4, H3, H2, H1 and L2. In

addition, alarms are initiated at pressure limits H1 and L1.

Monitoring of extremely high reactor pressure has been added to recorder in main control room.

16
4.8.3 Reactor pressure indication in the Olkiluoto RELAP5 model

The reactor pressure in the Olkiluoto RELAP5 model is indicated as the steam dome pressure.

OL2 - REACTOR SCRAM JUNE 13, 1997. RELAP5 MOD3.2.1.2

Reactor pressure

7.20

I ... Calculated pressure

7.10 - Measured pressure

7.00

6.90 "

6.80

6.70

6.60-

6.50

6.40

0 10 20 30 40 50 60 70 80 90 100 110 120

Time from scram signal (s)

Figure 4. Olkiluoto 2 reactor scram June 13, 1997. Reactor pressure.

4.9 Feed water

In the Olkiluoto RELAP5 model feed water flow is regulated by a controller so that the steady state

reactor water level was the same as the measured value 3.86 m above the top of active fuel.

The controller was removed in the transient calculation. The feed water flow rate and temperature

were instead set as a function of time according to the measured values (figures 5 and 6).

17
OL2 - REACTOR SCRAM JUNE 13, 1997. RELAP5 MOD3.2.1.2

Feed water flow

1200

1100

1000

v. 700,

. 600

500

400

300

200

100

0

0 10 20 30 40 50 60 70 80 90 100 110 120

Time from scram signal (s)

Figure 5. Olkiluoto 2 reactor scram June 13, 1997. Feed water flow.

OL2 - REACTOR SCRAM JUNE 13, 1997. RELAP5 MOD3.2.1.2

Feed water temperature

182

180

178

0 176

E 174

172

'.=

170

168

166 I I I i

0 10 20 30 40 50 60 70 80 90 100 110 120

Time from scram signal (s)

Figure 6. Olkiluoto 2 reactor scram June 13, 1997. Feed water temperature.

18
5. RESULTS

5.1 Steady state

The operation point of the 0L2 reactor scram of June 13, 1997 is presented in table 4. The table gives

both the values of the main parameters calculated with RELAP5 and also the corresponding values

obtained from the process computer.

Table 4. Olkiluoto 2 steady state prior to reactor scram on June 13, 1997. Comparison between

measurements and the RELAP 5 calculation.

Process parameter Measurement Calculation

Reactor 2272 2272

Thermal power (MW)

Pressure (MPa) Steam dome 7.0 7.0

Pressure difference 0.147

Between core inlet and

outlet

Temperature (-C) Feed water 180 180

Core inlet 275.5

Core outlet 286.3

Flow rate (kgfs) Feed water 1131 1131

Main steam line 1158 7850

Downcomer 7850 6731

Core fuel 1119

Core by-pass

Reactor fine level 3.86 3.86

(m above top of active fuel) 178.2

Mass of water in reactor pressure

Vessel (103 kg)

5.2 Transient

Figure 3 presents the calculated fine level from 0 s to 120 s after the reactor scram together with the

measured fine level. The RELAP5 result agrees with the measured level at the beginning of the

transient when the level drops due to the collapse of steam bubbles. According to the analysis the fine

level dropped from the initial value +3.86 m above top of active fuel to the minimum value +2.91 m in

23 s. According to the measurements the same minimum level was reached at 24 s after the scram.

The level rise from 24 s to 120 s due to supply of feed water and the warming up of the reactor water

inventory is slower in the RELAP5 analysis than according to the measurements.

The figure 4 presents the calculated and measured pressure from 0 s to 120 s from the scram. The

calculated pressure is at 28 s 0.28 MPa lower than the measured pressure. At 120 s the calculated

pressure is 0.14 MPa higher than the measured pressure.

19
Figures 7 to 11 present the calculated void in the core low, medium and high power region fuel

bundles, in the upper plenum (volume between the core and shroud head), in the steam separator

stand pipes, in the volume between the steam separator standpipes and in the downcomer at the level

of the shroud head. It can be pointed out that in the steady state the void fraction is 0.15 in the

volume between the steam separator standpipes and 0.11 in the downcomer at the level of the shroud

head.

Figure 12 presents the amount of water in the reactor pressure vessel. In the steady state the mass of

water is 178 metric tons.

OL2 -REACTOR SCRAM JUNE 13,1997. RELAP5 MOD3.2.1.2

Void fraction in the core high power region fuel bundles

0.90

0.80

0.70

Z 0.60

> 0A0,

0.30

0.20

0.10

- -Mddle

F, ~ a' a'

A\ \ .-..

-. /

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

U-ua + '

0 10 20 30 40 50 60 70

Time from scram signal (s)

80 90 100 110 120

Figure 7. Olkiluoto 2 reactor scram June 13, 1997. Void fraction in the core high power region fuel

bundles.

20
OL2 - REACTOR SCRAM JUNE 13,1997. RELAP5 MOD3.2.1.2

Void fraction in the core medium power region fuel bundles

1.00

0.90

0.80

0.70

c 0.60

0.50

0.40

0.30

0.20

0.10

0.00

0 10 20 30 40 50 60 70 80 90 100 110 120

Time from scram signal (s)

Figure 8.

bundles.

Olkiluoto 2 reactor scram June 13, 1997. Void fraction in the core medium power region fuel

OL2 - REACTOR SCRAM JUNE 13, 1997. RELAP5 MOD3.2.1.2

Void fraction in the core low power region fuel bundles

1.00

0.90

0.80

0.70

0.60

.2

• 0.50

> 0.40

0.30

0.20

0.10

0.00

0 10 20 30 40 50 60 70

Time from scram signal (s)

80 90 100 110 120

Figure 9.01kiluoto 2 reactor scram June 13, 1997. Void fraction

bundles.

in the core low power region fuel

21
OL2 - REACTOR SCRAM JUNE 13, 1997. RELAP5 MOD3.2.1.2

Void fraction in the upper plenum and

in the steam separator stand pipes

0

to

0 10 20 30 40 50 60 70 80 90 100 110 120

Time from scram signal (s)

Figure 10. Olkiluoto 2 reactor scram June 13, 1997. Void fraction in the upper plenum and in the

steam separator standpipes.

OL2 - REACTOR SCRAM JUNE 13. 1997. RELAP5 MOD3.2.1.2

Void fraction between the steam separator standpipes and

In the downcomer at the level of the shroud head

I NrA

0.90

0.80

0.70

- 0.60 o=

00

0.50

0

> 0.40

...... Volume betwen the steam separator

standpipes

-Downcomer

0.30

0.20

0.10

0.00

0 10 20 30 40 50 60 70

Time from scram signal (s)

80 90 100 110 120

Figure 11. Olkiluoto 2 reactor scram June 13, 1997. Void fraction between the steam separator

standpipes and in the downcomer at the level of the shroud head

22
OL2 - REACTOR SCRAM JUNE 13, 1997. RELAP5 MOD3.2.1.2

Calculated mass of water in the reactor pressure vessel

a

0

.5

5~

.5

10 20 30 40 50 60 70 80 90 100 110 120

Time from scram signal (s)

Figure 12. Olkiluoto 2 reactor scram June

vessel.

13, 1997. Calculated mass of water in the reactor pressure

23

6. DIFFERENCES BETWEEN CALCULATIONS AND MEASUREMENTS

RELAP5 calculations agreed with the measurements at the beginning of the accident when the void in

the reactor pressure vessel decreased rapidly because of the reactor scram. After the reactor water

level minimum the void in the reactor core and in the water volumes above core started to increase

again. At this point the difference between the calculated and the measured level increased. The

reason is be that RELAP5 calculates the amount of steam condensing in water is larger than in reality.

The amount of steam in water is thus smaller than the real void. The stronger steam condensation is

also reflected by the observation that the calculated water level drops faster than the measured level.

RELAP5 also calculates at the beginning of the transient a lower reactor pressure than what is

measured. At this stage there is due to feed water flow a large amount of subcooled water in the

vessel to condense steam.

The calculated water temperature is at the end of the transient higher than in reality because of the

stronger steam condensation. Because the calculated pressure is essentially the same as the

saturated steam pressure corresponding to the water temperature of the volumes above the core, the

calculated pressure is higher than the measured value. The stronger steam condensation explains

also the slower water level rise, because the amount of steam in water is smaller. Towards the end of

the calculation the calculated and measured get closer.

At the end of the RELAP5 calculation the level increase is caused by the thermal expansion of water.

At the end of the calculation the void fraction is about 0.1 in the upper part of the high power fuel

bundles. The steam condenses almost completely in the upper plenum volume between the core and

the shroud head. In the upper part of the down comer and between the steam separators the void

fraction is about 0.04.

25

7. CONCLUSIONS

The reactor scram, which occurred at the Olkiluoto 2 BWR plant on June 13 in 1997 has been

analysed with the RELAP5/MOD3.2.1.2 computer program. The RELAP5 model of the Olkiluoto

reactors is based on the model prepared for analysing PSA level 1 success criteria. The purpose of

the analysis was to investigate how well RELAP5 is able to calculate the behaviour of process

parameters after scram and validate the RELAP5 model of the Olkiluoto reactors.

Several changes were required in the Olkiluoto 1 and 2 model, which was prepared before the scram,

in order to reach good agreement between the calculation and the measurement data. The changes

concerned nodal volumes, node junction flow areas and pressure loss coefficients and heat structures.

Default values were retained for RELAP5 thermal hydraulic parameters.

The RELAP5 result agrees with the measured level at the beginning of the transient when the level

drops due to the collapse of steam bubbles. According to the analysis the fine level dropped from the

initial value +3.86 m above top of active fuel to the minimum value +2.91 m in 23 s. According to the

measurements the same minimum level was reached at 24 s after the scram. The level rise from 24 s

to 120 s due to supply of feed water and the warming up of the reactor water inventory is slower in the

RELAP5 analysis than according to the measurements. The reason can be that RELAP5 calculates

the amount of steam condensing in water is larger than in reality. The amount of steam in water is thus

smaller than the real void. The stronger steam condensation is also reflected by the observation that

the calculated water level drops faster than the measured level. RELAP5 also calculates at the

beginning of the transient a lower reactor pressure than what is measured. At this stage there is due to

feed water flow a large amount of subcooled water in the vessel to condense steam.

27

8. REFERENCES

1. 2-KK-R6-2/97, OL2 - Reactor scram 13.6.1997

2. NUREG/CR-5535, INEL-95/0174, Vol.1, RELAP5/MOD3 Code Manual, Code Structure, System

Models, and Solution Methods

3. NUREG/CR-5535, INEL-95/0174, Vol.2, RELAP5/MOD3 Code Manual, User's Guide and Input

Requirements

4. NUREG/CR-5535, INEL-95/0174, Vol.3, RELAP5/MOD3 Code Manual, Models and Correlations

5. NUREG/CR-5535, INEL-95/0174, Vol.4, RELAP5/MOD3 Code Manual, User's Guidelines

6. ENE4/22/96, OLI/0L2 - UPSA - Level 1 PSA success criteria for 2500 MWth reactor power, S14-

08. Feed water system 312 pipe guillotine break, research report, ViT Energy, May 2, 1996

7. S21-TR-M-143/95, TVO 1/11 - Code input data to account for GENE AS2B steam separator, shroud

cover and core flooder distributors

29

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

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

NRCMO 3.7 and Addendum Numbers, If any.)

BIBLIOGRAPHIC DATA SHEET

(See instructions on the reverse) NUREG/IA-0204

2. TITLE AND SUBTITLE 3. DATE REPORT PUBLISHED

OLKILUOTO 2- RELAP5/MOD3.2.1.2 Analysis of the Reactor Scram on June 13,1997 MONTH YEAR

April 2007

4. FIN OR GRANT NUMBER

N/A

5. AUTHOR(S) 6. TYPE OF REPORT

Heikki Sjovall

7. PERIOD COVERED (inclusive Dates)

N/A

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

provide name and maiing address.)

Teollisudden Voima Oy

Fin-21760 Olkiluoto

Finland

9. SPONSORING ORGANIZATION - NAME AND ADDRESS (If NRC, type "Same as above'; I contractor, provide NRC Division, Office or Region. US. Nuclear Regulatory Commission,

and mailing address)

Division of Risk Assessment and Special Projects

Office of Nuclear Regulatory Research

U.S Nuclear Regulatory Commission

Washington, DC 20555-001

10. SUPPLEMENTARY NOTES

Michael B. Rubin, NRC Proiect Manager

11. ABSTRACT (200 words or less)

The reactor scram, which occured at Olkiluoto 2 June 13 in 1997 has been analysed with the RELAP5/MOD3.2.1.2 computer

program. The RELAP5 model of the Olkiluoto reactors is based on the model prepared for analysing PSA level 1 success

criteria. The purpose of the analysis was to investigate how well RELAP5 is able to calculate the behaviour of process

parameters after scarm and validate the RELAP5 model of the Olkiluoto reactors.

The main plant features connected with the event are briefly presented. Modelling aspects are discussed regarding the

RELAP5 analysis.

The RELAP5 result agrees with the measured reactor water level at the beginning of the transient when the level drops due to

the collapse of steam bubbles. Later the level rise due to supply of feed water and the warming up of the reactor water

inventory is slower in the RELAP5 analysis than according to the measurements.

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

RELAP5 unlimited

reactor 14. SECURITY CLASSIFICATION

core (This Page)

unclassified

(This Report)

unclassified

15. NUMBER OF PAGES

16. PRICE

NRC FORM 335 (9.2004) PRINTED ON RECYCLED PAPER
Federal Recycling Program

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ANALYSIS OF THE REACTOR SCRAM ON JUNE 13, 1997

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UNITED STATES

NUCLEAR REGULATORY COMMISSION

WASHINGTON, DC 20555-0001

OFFICIAL BUSINESS