NUREG/IA-0193

*International

Agreement Report

Assessment of Single Recirculation

Pump Trip Transient in Santa Maria

de Garona Nuclear Power Plant With

TRAC-BF1/MOD1, Version 0.4

Prepared by

C. Garcia de ]a Rfia, J. Garcia Garcia, J.V L6pez Montero

NUCLENOR, S.A.

C/ Herman Cortes 26

39080 Santander

SPAIN

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

January 2001

Prepared as part of

The Agreement on Research Participation and Technical Exchange

under the International Code Application and Maintenance Program (CAMP)

Published by

U.S. Nuclear Regulatory Commission
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NUREG/IA.-,193

International

Agreement Report

Assessment of Single Recirculation

Pump Trip Transient in Santa Maria

de Garona Nuclear Power Plant With

TRAC-BF1/MOD1, Version 0.4

Prepared by

C. Garcia de la Rda, J. Garcia Garcia, J.V. L6pez Montero

NUCLENOR, S.A.

C/ Herman Cortes 26

39080 Santander

SPAIN

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

January 2001

Prepared as part of

The Agreement on Research Participation and Technical Exchange

under the International Code Application and Maintenance Program (CAMP)

Published by

U.S. Nuclear Regulatory Commission

ABSTRACT

This report has been prepared by NUCLENOR in the framework of the

CAMP/SPAIN Project. It represents one of the application calculations submitted in

fulfillment of the bilateral agreement for cooperation in thermalhydraulic a'ctivities

between the Consejo de Seguridad Nuclear of Spain (CSN) and the United States

Nuclear Regulatory Commission (USNRC).

The work consisted in using the TRAC-BF1 code to reproduce a transient that

took place at Santa Maria de Garofia Nuclear Power Plant (NPP) on June 9, 1993. The

event was originated by a transformer failure that led to a loss of generator excitation

and consequently a recirculation pump trip.

Santa Maria de Garofia NPP is a 1381 MWth General Electric Boiling Water

Reactor 3 (GE BWR/3) owned by NUCLENOR, S. A., a Spanish utility that participates

in the CAMP Program as a member of UNIDAD ELECTRICA S. A. (UNESA).

The simulation has been carried out with the TRAC-BFI/MOD1, code, version

0.4, running on a Workstation Hewlett Packard cI 80u under HP-UX operating system.

As a result of the analysis, it is felt that TRAC-BF1 is capable of reproducing the

plant behaviour with an acceptable degree of accuracy. The main phenomena of the

transient have been calculated correctly.

iii

CONTENTS

A B STRA CT ............................................................... iii

CON TEN TS ............................................................... v

TA B LES ............ , ................................................... vii

FIG U RES ................................................................ vii

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

I. INTRODUCTION ........................................................ 1

II. NUCLEAR PLANT DESCRIPTION ......................................... 2

III. TRANSIENT DESCRIPTION AND PLANT RESPONSE ....................... 7

IV. CODE INPUT MODEL DESCRIPTION .................................... 14

V. STEADY STATE CALCULATIONS ........................................ 20

VI. CALCULATION RESULTS .............................................. 21

VII. CONCLUSIONS ....................................................... 37

VIII. REFERENCES ........................................................ 38

V

TABLES

TABLE 1.- POWER-FLOW REFERENCE PLANT DATA IN THE OPERATING MAP .......... 10

TABLE 2.- COMPONENTS OF GAROfqA NPP INPUT DECK ................................................ 17

TABLE 3.- VALUES FOR KEY FEEDBACK PHENOMENA .................................................. 19

TABLE 4.- STEADY STATE RESULTS AT NOMINAL POWER ........................................... 20

FIGURES

FIGURE 1.- SANTA MARIA DE GARO9A NPP FUNCIONAL DIAGRAM .......................... 4

FIGURE 2.- POW ER-FLOW M AP ............................................................................................ II

FIGURE 3.- SANTA MARIA DE GARO9A NPP NODALIZATION ..................................... 14

FIGURE 4.- LEVEL DISTRIBUTION IN THE REACTOR VESSEL ..................................... 15

FIGURE 5.- VESSEL RADIAL NODALIZATION ................................................................. 16

FIGURE 6.- AVERAGE VOID FRACTION CALCULATED WITH TRAC-BF1I ................... 22

FIGURE 7.- AVERAGE MODERATOR TEMPERATURE, TRAC-BFI ................................ 22

FIGURE 8.- AVERAGE FUEL TEMPERATURE, TRAC-BFI ............................................... 23

FIGURE 9.- CORE THERMAL POWER, TRAC-BFI CALCULATION ................................ 23

FIGURE 10.- VOID FRACTION REACTIVITY, TRAC-BFI .................................................. 24

FIGURE 11.- DOPPLER REACTIVITY, TRAC-BFI ............................................................... 24

FIGURE 12.- EVOLUTION OF FEEDWATER TEMPERATURE ......................................... 25

FIGURE 13.- MODERATOR TEMPERATURE REACTIVITY, TRAC-BF1 ......................... 26

FIGURE 14.- TOTAL REACTIVITY, TRAC-BFI ................................................................... 26

FIGURE 15.- TOTAL REACTOR POWER, TRAC-BF1 AND PLANT DATA ............ 27

FIGURE 16.- PRESSURE AT THE SENSOR LINE AND REGULATOR SET-POINT ...... 2g

FIGURE 17.- STEAM DOME PRESSURE, TRAC-BFI AND PLANT DATA ...................... 29

FIGURE 18.- STEAM FLOW EVOLUTION, TRAC-BFI AND PLANT DATA .................... 30

FIGURE 19.- COMPARISON BETWEEN CALCULATED AND LEVEL PLANT DATA ........ 31

FIGURE 20.- FLOW RATE PUMP "B", TRAC-BFI AND PLANT DATA ............................ 32

FIGURE 21.- FLOW RATE PUMP "A", TRAC-BFI AND PLANT DATA ............................ 33

FIGURE 22.- CORE FLOW, TRAC-BFI AND MEASURED PLANT SIGNAL ..................... 34

FIGURE 23.- CORE FLOW, TRAC-BFI AND CALCULATED PLANT DATA ................... 35

FIGURE 24.- TOTAL FEEDWATER FLOW, TRAC-BFI AND PLANT DATA ................... 36

vii

EXECUTIVE SUMMARY

This work shows the results of the analysis with TRAC-BF1 of an aclual event

that took place in Santa Maria de Garofia Nuclear Power Plant (NPP) on Juný- 9, 1993

The event was originated by a single recirculation pump trip. Santa Maria de Garofia

NPP is a 1381 MWt General Electric Boiling Water Reactor 3 (GE BWR/3).

Immediately after the recirculation pump "A" trip, the jet pump difiuser and

drive flows reverse in the tripped loop. The jet pump flow increases in the active loop

because of the decreased core flow and core pressure drop. The decreased core flow

initially causes the core void fractions to increase, resulting in an increase in reactor

water level and a decrease in power. In this situation, proper core flow measurement is

required to assure operation within the power-flow map and to avoid the region of

potential thermalhydraulic instabilities. The plant can operate with a single loop with

appropriately modified procedures and technical specifications (TS).

The main phenomena are reproduced in the simulation and the discrepancies are

justified. A general-purpose nodalization of the plant for TRAC-BF1 has b:en used.

This work is an additional part of the validation set for this TRAC-BF1 noctalization

which has been developed for thermalhydraulic applications.

ix

I. INTRODUCTION

NUCLENOR is a Spanish Electrical Utility, which owns the GE BWR/3 Santa

Maria de Garofia Nuclear Power Plant.

NUCLENOR, in the framework of the "Code Applications and Maintenance

Program" (CAMP), has taken on the task of selecting an application case a:nong the

different transients of the life of the plant to analyse with TRAC-BFl/MOD1 code,

version 0.4, and to evaluate the agreement.

In this case, the analysis of a single recirculation pump trip was selected and a

comparison between measured data and TRAC-BF 1 data has been carried out.

The trip of one recirculation pump does not normally cause scram. In fact, GE

BWRs were licensed specifically for this operating condition, which is generaltly called

single loop operation (SLO). During SLO, proper core flow measurement is required to

assure operation within the power-flow map and to avoid the region of potential

thermalhydraulic instabilities. Additional considerations are necessary because the

inactive jet pumps may be operating with reverse flow when the active loop .jet pump

flow is above 40 % of rated core flow. The jet pump core flow measurement s;ystem is

calibrated to measure core flow when both sets of jet pumps are in forward flow and the

total core flow is the sum of the indicated loop flows.

A summary of the Santa Maria de Garofia NPP is presented in Section II and the

simulated transient characteristic in Section III. Section IV describes the model of the

plant and the main features of the input deck. The steady state results are explained in

Section V. Calculation results and discussions of the main phenomena are pre:;ented in

Section VI. Finally, the main conclusions obtained are summarised in Section VII.

TRAC-BFI/MOD1 v. 0.4 is implemented in a Workstation Hewlett Packard hpcl 80u under HP-UX operating system, where all the calculations have been canied out.

I
II. NUCLEAR PLANT DESCRIPTION

The Santa Maria de Garofia NPP is a General Electric Boiling Water Reactor 3

(BWR-3), with a Mark I primary containment type. The plant is operated by

NUCLENOR and was connected to the grid in 1971. The plant is rated at 1381 MW

(thermal) and is located in the province of Burgos (Spain).

The Nuclear Boiler assembly consists of the reactor pressure vessel and internal

reactor components such as the core structure, steam dryer assembly, fuel supports and

control guide tubes. The reactor core is made up of 400 fuel assemblies and 97 control

rod blades. At present, it is loaded .with GEl 1 (9x9) elements. However the transient,

that we will compare with a TRAC-BF I analysis, took place in 1993 (cycle 17) and the

reactor core was loaded with:

- GE7B (8x8) elements

- GE8B (8x8) elements

- GEl 0 (8x8) elements

Each control rod blade consists of sheathed cruciform array of vertical absorber

rods made of boron carbide (B4 C).

The Recirculation system provides the hydraulic energy required to force

coolant through the reactor core and provide forced convection cooling of the reactor

core. The recirculation system consists essentially of two recirculation piping loops

located outside the reactor pressure vessel, in the Drywell area, and includes twenty jet

pumps located inside the reactor pressure vessel between the reactor pressure vessel

wall and the core shroud.

The flow from the recirculation pump is the driving force for the jet pump. The

recirculation flow is mixed with the feedwater and steam separator water flow and the

total is discharged into the plenum area below the core. The coolant flows upward

around the individual fuel rods inside the fuel channel, where it is heated and becomes a

two-phase steam-water mixture. The steam-water mixture leaves the fuel bundles at the

2
top and enters a plenum located directly above the core. The plenum allows the flow to

be equally spread out into the array of steam separators located above the plenum

region. The steam is separated from the water and passes through a dryer where the

content of water in the steam is minimised. The saturated steam exits through :nozzles at

the top of the vessel body.

Water collected below the dryer is routed through drain lines joins the water

leaving the separators, and flows downward in the annulus between the core shroud and

vessel wall. Feedwater is added to the system through spargers at the top of this annulus

and joins the downward flow of water. A portion of this downward flow exits to the

recirculation pumps.

The primary function of the reactor recirculation system is to permit react:or power

level changes without changing the position of the reactor control rods. In this mode,

reactor power can be changed up to 30% per minute over a nominal 25% range. For

instance, starting at 100% power, the range is between 75% and 100% and the rate is

30% per minute.

The primary purpose of the feedwater system is to maintain the water level in the

reactor vessel within a programmed range during all modes of plant operation. In

normal operation, the level of water in the reactor is controlled by a feedwater controller

which receives inputs from reactor vessel water level, steam-mass flow rate and

feedwater-mass flow rate transmitters.

The feedwater control system generates signals that regulate the opening of the

flow control valves. The rate of flow is thus controlled, maintaining the reactor level at

the desired level. During steady-state operation, feedwater mass flow rate matches the

steam mass flow exactly and the water level is maintained. A change in the steam mass

flow of the control valves is immediately sensed and the system adjusts the opening of

the feedwater mass flow of the control valves to balance the two mass flow rates,

maintaining the water level.

3
Figure 1.- Santa Maria de Garofia NPP funcional diagram.

The main steam system consists of four lines that provide steam to the turbine

from the reactor vessel. Steam lines run downward, parallel to the vertical axis of the

vessel, until they reach the elevation at which they emerge from the containment. Two

air-operated isolation valves are installed on each steam line, one inboard and one

4
outboard of the primary containment penetration. A flow-restricting nozzle i 3 included

in each steam line as an additional engineered safeguard to protect against rapid

uncovering of the core in case of a main steam line break.

Three relief valves (RV's) and three safety/relief valves (SRV's) discharging into

the suppression pool, and seven safety valves (SV's) discharging into the drywell, are

installed on the steam lines. The main function of these valves is protection against

overpressure of the reactor primary system.

The primary containment in Santa Maria de Garofia NPP is of the Mark I type.

The steel "Light-bulb shaped" Drywell is a spherical shell intersected by a cylinder

(Figure 1). A bolted head closes the top of the cylinder. The pressure-suppression

chamber, or the Wetwell, is a toroidal steel vessel that surrounds the lower portion of

the Drywell. Eight circular vent pipes interconnect the Wetwell and the Drywell. The

containment is enclosed by the reactor building, which also contains the refuei.ling area,

fuel storage facilities and other auxiliary systems.

Figure 1 gives the reader an insight into the arrangement and interfac:es of the

Santa Maria de Garofia NPP Isolation Condenser System (IC) and Emergency Core

Cooling System (ECCS).

The Isolation Condenser system (IC) is designed to provide emergenc~y reactor

core cooling without loss of water when the reactor becomes isolated from the turbine

and the main condenser by closure of the main steam line isolation valves.

The IC system consists of one elevated condenser containing two tube bundles,

immersed in cooling water, and the necessary piping and valves to condense the reactor

steam and return the condensate by gravity to one of the recirculation loops. The IC unit

is located in the Reactor Building at an elevation higher than the reactor vessel, in order

to provide the necessary operating pressure head.

The Santa Maria de Garofia ECCS, consists of the following systems:

5
- High Pressure Injection System (HPCI)

- Low Pressure Cooling Injection System (LPCI)

- Core Spray System (CS)

The HPCI system consists of a turbine-driven pump assembly and the necessary

elements to pump water from the Condensate Storage Tank or Suppression Pool in the

reactor vessel. The HPCI system provides emergency core cooling in the event of a

small break in a process line that does not Fesult in significant depressurisation of the

vessel. The primary steam line provides steam from the pressure vessel to operate the

HPCI turbine. Exhaust steam from the turbine is piped to the Suppression Pool where it

is condensed.

The Low-Pressure Cooling Injection system (LPCI) consists of two identical and

completely independent cooling loops. Each loop has two pumps in parallel. The two

loops are arranged to discharge water into the reactor recirculation loops. A cross

connection exists between the pump discharge headers of each loop. Containment

cooling water from the Suppression Pool is cooled by either of the two, or both, shelltube heat excharigers. One heat exchanger is provided in each of the two LPCI loops.

The Core Spray system consists of two independent pumping subsystems, each

capable of fulfilling the cooling requirements of the reactor core in the event of failure

in the primary system. Operation mode -involves taking water from the suppression

chamber and pumping it to the reactor vessel by core spray nozzles.

Both LPCI and CS start automatically when either a Drywell high pressure or

Low Low vessel signal exists.

6
Il1. TRANSIENT DESCRIPTION AND PLANT RESPONSE

Santa Maria de Garofia NPP, as well as other GE BWRs occasionally find it

necessary to operate with one recirculation loop (SLO, Single Loop Operation) out of

service. The SLO follows mechanical or electrical failures, which affect one of the

recirculation loops. Many such failures (e.g. a circuit breaker trip) can be corrected

quickly by plant personnel [1], [2]. The initial design of GE BWRs generally considered

the mechanical, structural and thermal effects of limited SLO and were licensed

specifically for this operating condition. Additionally, some GE BWRs, including Santa

Maria de Garofia NPP, have also been licensed specifically for extended SLO with

appropriately modified procedures and technical specifications. Extended SLO is

required when a more significant mechanical or electrical failure (e.g. motor-generator

set failure) cannot be corrected quickly.

Transient description

On June 9th, 1993 at 18:58 h., the plant was operating at 1381 MWt in stable

conditions when a transformer failure led to a loss of generator excitation and,

consequently, a motor-generator group "A" trip. As a consequence, the rec'Xculation

pump "A" trip [1].

Control room alarms on panel 904, together with the recirculation flow and

power reduction, forced the operating crew to follow the Abnormal Operating

Procedure (AOP) 202-2 "Loss of a Recirculation Pump". There were no automatic

actions and the crew had to carry out the AOP immediate actions set, whichi will be

summarised next:

Immediately after the recirculation pump "A" trip, loop A started to backflow and

loop "B" flow rate was increased due to the reactor pressure decrease. In this situation,

the crew had to close the loop "A" discharge valve to protect the tripped pump impeller

*from rotating backwards. This valve had to remain closed about five minutes.

7
Subsequently, the crew had to isolate pump "A" seals of the Control Rod Drive

Hydraulic System (CRDH) water make up to avoid an excessive loop "A" cooling.

It is well known that the jet pump core flow measurement system is calibrated to

measure core flow when both sets of jet pumps are in forward flow and the total core

flow is the sum of the indicated loop flows. However, for SLO, the inactive jet pumps

start reverse flow, so that the measured flow in the reverse flow jet pumps must be

subtracted from the measured flow in the active loop. In addition, the jet pump flow

coefficient is different in reverse flow than forward flow (the reverse flow coefficient is

about 95% of the forward flow coefficient), and the measurement of reverse flow must

be modified to account for this fact.

The operating crew ended immediate actions by checking pressure, reactor

power and off-gas activity. The subsequent actions were intended to fulfil Technical

Specifications (TS):

- Maintaing idle loop temperature. If the idle loop is permitted to cool down,

restart of the idle recirculation pump may be prohibited because of TS. If the idle loop

cannot be warmed, these operating restrictions can make it necessary to shut down and

cool down the RPV before the idle recirculation pump can be restarted.

- Inserting control rods within a limit of 30 minutes, until nuclear power is 5 %

lower than 70 % rod pattern line.

- Reduce the active pump speed to 53 %. If the idle loop is not isolated and if the

operating loop recirculation pump is providing more than about 40 % of rated core flow,

the reverse flow through the idle jet pumps will force some reverse flow through the

idle loop and will tend to keep the idle loop warm. At lower operating flows, reverse

flow will decrease and a cool down will occur. In this way, a compromise is maintained

between a reverse flow high enough to keep the loop warm and, at the same time, weak

enough not to damage the repaired pump when restarting, if this is possible.

8
The plant was operated about 23 h. with a single loop, until the olher pump

could be restarted.

During the whole transient, proper core thermal power and flow measurement

were required to assure operation within the power-flow map. If near in instability

region, a transition into POA 202-4 "Prevention of core thermal-hydraulic instabilities"

might be required. However, this was not necessary on this occasion.

Figure 2, shows the power-flow map of Santa Maria de Garofia NPP. This

diagram shows the relation between core thermal power versus core flow rate for

various operating conditions.

Line A is called the natural circulation line and shows the power versus flow with

no recirculation pumps running. Line B is for minimum recirculation pump .5peed. As

the core power increases along lines A and B, the core flow rate tends to a limit. This is

a result of the additional flow resistance created by the increase in two-phase flow.

The shape of line C will be followed if the core power is changed with no change

in the recirculation pump speed (e.g., control rod movement or end-of-cycle

coastdown). For a power reduction, the flow increase is due to a reduction in channel

void fraction and a subsequent reduction in two-phase flow resistance. The :reverse is

true for a power increase.

Line D is called the 100% flow control line, the 100% rod pattern line or the

100% load line. It is based on balanced xenon conditions and is defined by the core

configuration that will result in rated core thermal power when the core flow is

increased to rated flow. Line D can be derived by operating the plant at full power and

full flow until balanced xenon conditions are established and then reducing recirculation

pump speed to minimum speed in a relatively short period. Line D is slightly concave

because the core inlet subcooling decreases somewhat as core flow rate increases. Thus,

9
the core thermal power is slightly less at full flow than it would be, if the inlet

temperature did not change.

Line E is a flow interlock line to prevent cavitation of the jet pumps and the

recirculation pumps. The recirculation pumps cannot be operated above minimum pump

speed until feedwater flow is greater than 20% of rated flow. Similarly, the recirculation

pump will run back to minimum speed if the feedwater flow becomes less than 20%.

This assures adequate subcooling for all normal modes of pump operation.

Table I shows the power-flow pairs belonging to the initial steady-state (point

1), recirculation pump trip (point 2) and final state (point 3). In any case, the plant

conditions are outside the instability region. The total core flow along the transient is

out of instability region, taking into account that the uncertainty in measured core flow

(6%) has also been considered. Thus, the plant can operate with a single loop with

appropriately modified procedures and technical specifications (TS).

Time (s) 0-55 ~60 300

Power (%) 100 % -53 % -70 %

Core Flow (T/H) -20190 -10700 -10700

Icon in power-flow map (1) (2) (3)

Table 1.- Power-flow reference plant data in the operating map.

The core flow, after the trip, has been calculated using the formula [3],

(Total Core Flow) = (Active Loop Indicated Flow) - C (Inactive Loop Indicated Flow)

where the factor C = 0.95, is the recommended (conservative) value for measurement of

core flow during single loop operation.

10
C"l

POWER-FLOW MAP

0,.

A2 i3 3* w*I- (KIM"II

-T WT (%) so0.

C~k FLU*

M

Figure 2.- Power-flow map.

II
Core response to recirculation flow changes

Coolant flow-rate control in Santa Maria de Garofia NPP is accomplished by

varying the recirculation pump motor speeds. As a consequence of changes in the

recirculation flow, the core power will change.

Basically, the phenomenology of the process can be explained taking into account

that the core power response to recirculation flow changes in BWR consists of an

interaction of four basic variables: voids, Doppler effect, inlet temperature and xenon

effect.

The first response to a change in coolant flow is a change in the core void fraction.

A core flow decrease will result in a quick increase in the void fraction at any elevation

of the core. As BWRs have a negative void coefficient of reactivity, the total core power

decreases.

The fuel temperature goes down and causes a reactivity increase due to the

Doppler effect. This increase in reactivity, must be compensated by an equal decrease

due to increased void fraction. The Doppler coefficient of reactivity is much smaller

than the void coefficient.

The third effect of changing flow is the change in core inlet temperature. For a

given core flow decrease, the feedwater flow decrease is proportionally smaller. Thus, a

reduction in core flow creates an increase in subcooling as a consequence of a reduction

in the extraction steam to the feedwater heaters. This low inlet temperature increases the

coolant density in the lower part of the core and thus results in an increase in reactivity.

The net immediate effect of all these variables is to reduce the power level while

leaving the power shape essentially unchanged. The second order effects due to

Doppler, void fraction, and inlet temperature, tend to shift the power slightly towards

the top of the core.

12
The fourth effect is the evolution of the xenon concentration. The xenon

redistribution after a flow change is very slow compared to the rapid responses of void,

Doppler, and inlet temperature. As xenon and iodine concentration decreas;e thermal

power will slowly rise.

13
IV. CODE INPUT MODEL DESCRIPTION

The TRAC-BF1 input is made up of 51 components, including FILL and

BREAK components, as well as a very detailed model of plant controls and logic [4].

[5]. The development of TRAC-BF1 input deck for this analysis was based on data

taken from drawings [6] and specific technical documents related with the nuclear fuel

design [7]. The main features of the nodalization can be seen in Figure 3. Figure 4

shows the diagram of level distribution in the reactor vessel. Additionally, components

are listed in Table 2.

VESSEL

4 3 2 1 2 3 4

RINGS

Figure 3.- Santa Maria de Garofia NPP nodalization.

14
The Component Vessel is divided into 12 axial segments (levels), fbur radial

segments (rings) and only one azimuthal segment. The downcomer upper level

corresponds to level 8.

84847 rnLEVEL12

- 16,454 mLEVEL11

LEVELIO

- 12,221 m -

LEVEL 9

- 11,717mLEVEL I

-9,390mLEVEL7

-9,147m--.

LEVEL6

- 758 m

LEVEL5

- 4,807 M

LEVEL4

-- 3,567 m

LEVEL3

- Z,361 m

LEVEL2

-0,933mLEVEL1

Figure 4.- Level distribution in the reactor vessel.

The radial segments are distributed such that the inner three rings extend over the

core region with the first ring containing 28 fuel channels, the second ring containing

288 fuel channels and the third ring containing 84 fuel channels. The fourth radial ring

models the reactor vessel downcomer that extends from the top of level 2 to the top of

level 8. Figure 5 shows the radial division of the vessel.

15
RING 4

Figure 5.- Vessel radial nodalization.

The first level extends from the vessel bottom to the top control rod drive

housings. The second one ends at the jet pumps discharge support ring. The third and

fourth ones go from this support ring up to the core bottom.

The Core is divided into three axial levels of different lengths (level 5 to level 7).

The neutronic core region, in the CHAN component, is separated into 9 hydraulics

levels. Note that, the top of the jet pumps (end of fifth level) is located at 2/3 of the total

core axial active length.

Instead of individual steam separators and dryer components, a perfect separator

option is used for axial level 9. This component allows the vapour to continue upward

into the axial level 10 and liquid to drain radially outward in the downcomer region.

"Double-sided lab models" account for the heat capacity and transmission within vessel

internals.

16
VESSEL

CLHAN

CHAN

CHAN

FILL

PIPE

PIPE

PIPE

PIPE

PIPE

PIPE

FILL

FILL

FILL

PIPE

PUMP

PUMP

TEE

TEE

JETP

TEE

TEE

TEE

TEE

JETP

TEE

VALVE

VALVE

VALVE

VALVE

VALVE

VALVE

VALVE

VALVE

VALVE

VALVE

BREAK

BREAK

BREAK

BREAK

1

36

38

40

61

60

31

70

87

88

89

82

83

84

92

22

4

20

24

25

50

52

55

6

7

8

21

23

3

5

51

53

54

68

69

32

56

57

58

5

VESSEL

84 PERIPHERAL CHANNELS

288 AVERAGE CHANNELS

28 HOT CHANNELS

FEED WATER FILL

FEED WATER

ISOLATE CONDENSER

PIPE OF CORE SPRAY

CENTRAL GUIDE TUBE

AVERAGE GUIDE TUBE

PERIPHERICAL GUIDE TUBE

GUIDE TUBE ENTRANCE

GUIDE TUBE ENTRANCE

GUIDE TUBE ENTRANCE

HPCI TURBINE DISCHARGE

RECIRCULATION PUMP "B"

RECIRCULATION PUMP "A"

RECIRCULATION SUCTION LINE

RECIRCULATION DISCHARGE LINE

JETPUMP LOOP "B"

MAIN STEAM LINE FROM THE VESSEL

MAIN STEAM LINE FROM THE VESSEL

MAIN STEAM LINE TO THE TURBINE

RECIRCULATION DISCHARGE LINE

JETPUMP LOOP "A"

RECIRCULATION SUCTION LINE "A"

RECIRCULATION SUCTION VALVE

RECIRCULATION DISCHARGE VALVE

RECIRCULATION SUCTION VALVE "A"

RECIRCULATION DISCHARGE VALVE

RELIEF & SAFETY/RELIEF VALVES SET

SAFETY VALVES SET

ISOLATION MAIN STEAM VALVE

BYPASS VALVE

CONTROL VALVE

ISOLATION CONDENSER LINE VALVE

WETWELL BOUNDARY CONDITION

BYPASS BOUNDARY CONDITION

TURBINE BOUNDARY CONDITION

DRYWELL BOUNDARY CONDITION

Table 2.- Components of Garofia NPP input deck.

17
Inside the VESSEL component, axial level 1 and axial level 2, the PIPES

components 87, 88, 89 model the control rod guide tubes. These three pipes represent

97 real control guide tubes. The guide tubes were modelled from the top of the control

rod drive housing to the core plate. The heat transfer between the fluid in the guide

tubes and the lower plenum has also been taken into account.

Fuel bundle modelling is accomplished by using CHAN component simulating

the 84 peripheral bundles by CHAN component number 36, 288 average bundles by

CHAN component number 38 and 28 hot bundles by number 40.

Finally, the main plant control systems have been simulated. Pressure control

allows valve 69 (turbine control valve) position to be controlled in order to regulate

reactor pressure. A very detailed feedwater - level control system is simulated in the

main operational modes. The so called "three elements" level control mode is normally

used and additionally, the possibility of changing to flow control mode is also

contemplated.

Core power model

The TRAC/BF1/MODI code has three methods of simulating core power

response. The first method is a user- supplied table lookup scheme. The second method

employs a space-independent reactor kinetics model with reactivity feedback. This

model is useful for simulating transients where time-dependent spatial variations in the

core power distribution are not significant. The third method employs a onedimensional, two-group, neutron transport model that solves the one-dimensional

steady- state and time-dependent neutron diffusion equations in a rectangular geometry.

To simulate the recirculation pump trip transient, a reactor point kinetics model

with trip initiated reactivity feedback and trip initiated scram reactivity insertion was

turned on to calculate the core power rate. The reactivity feedback model for void,

boron, moderator and fuel temperature currently employs reactivity coefficients that are

polynomial approximations using core-averaged properties. In this case, the reactivity

18
feedback coefficients have been calculated with the PANACEA code. There had been

no boron injection. The set of reactivity feedback coefficients used in this si:nulation is

listed in Table 3.

Parameter Value

Fuel reactivity coefficients -1.729E-5 -1.264E-5 5.298E-6

Moderator temperature reactivity coefficients' -1.158E-4 4.63E-7 -9.69E-10

Void reactivity coefficients -2.OOE-1 7.770E-1 -1.900

Table 3.- Values for key feedback phenomena.

The calculation of reactivity feedback is initially done on a cell-by-cell basis.

The reactor core is partitioned into channel regions (using CHAN component) and

bypass regions (using VESSEL component). Contributions for the moderatoi:, Doppler,

void and boron feedback coefficients are globally summed over all CHAN and VESSEL

cells. The same reactivity coefficient polynomial curve fits are used for the CHAN and

bypass VESSEL cells.

'Default values provided by TRAC-BFI.

19
V. STEADY STATE CALCULATIONS

The calculation of the steady state with TRAC-BF 1 has been developed in two

steps. First of all, a null transient calculation was carried out setting the initial plant

conditions of pressure, level and recirculation flow. Secondly, the use of EXTRACT

subcode allowed us to extract steady state component data and configures a new steady

state input deck. The main measured data of the plant against the steady-state data

obtained with TRAC-BF 1 are shown in Table 4.

Parameter Plant data TRAC-BF1

Thermal power 100% (MW) 1381 1381

Dome pressure (Kg/cm L rel) 70.9 70.9

Reactor level (cm rel ') 64 65

Feedwater flow (Tm/h) FW-A 1250 1 FW-B 1250 2500

Feedwater temperature (Q C) 182 182

Core flow (T/h) 20190 20050

RP speed A (RPM) 1251 1251

RP speed B (RPM) 1315 1315

Recirculation flow A (lfs) 1710 1705

Recirculation flow B (ifs) 1710 1705

Table 4.- Steady - state results at nominal power.

2 Centimeters relative to zero-scale (12.22 m from the bottom head) which corresponds to bottom of

steam separators.

20
VI. CALCULATION RESULTS

The analysis of the main TRAC-BF1 results together with the available plant

data will be carried out in this section. The real data were extracted from the Santa

Maria de Garofia NPP Transient Analyser; a computer system that is able lo store the

main transient data with an interval of twenty milliseconds.

The initial transient conditions were obtained from the steady state calculation,

which are described in chapter V above. A "null transient" during 55 seconds, from the

steady state reached, was run with the reactivity feedback models activated to verify the

stability of steady state conditions.

Nuclear Power

In the initial state (0 - 55 s), there is no increase in the main variables that affect

the direct behaviour of reactor power. Void fraction (Figure 6), moderator temperature

(Figure 7) and fuel temperature (Figure 8) remain constant; there is no reactivity

insertion.

At 55 s the recirculation pump trip starts and the core power evolution (Figure 9)

is directly related to the abrupt flow rate decrease through the core. This flow reduction

leads to a sudden increase in the average void fraction in the core. The negative

reactivity inserted (Figure 10) causes the power to go down. There is, also, an additional

positive reactivity from the Doppler effect (Figure 11) as results of the fuel cool down.

Subsequently, the core flow starts to increase again because the operating jet pumps

begin to provide more forced circulation flow due to the reactor pressure decrease. For

this reason, the power is stabilised around 67.5 % (TRAC-BFI and plant data) at about

80s.

21
Average void fraction

0.468

0,466

0,463

0,461

•, 0,458

0.458 CL

0 0•,456

0,453

0.451

0,448

0,446

0,443

25 so 75 0oo 225 0SO 175 200 225 250 275 300

Time (s)

Figure 6.- Average void fraction calculated with TRAC-BF1.

Average moderator temperature

285.5

285,0

284.5

284,0

283.5

E

C. 283.0

282.5

282.0

281,5

281.0

0 25 50 75 100 125 150 175 200 225 250 275 300

Time (s)

Figure 7.- Average moderator temperature, TRAC-BF1.

22
Average fuel temperature

470 S

% 460

E 450

I0 25 50 75 100 125 150 175 200 225 250 275 30C

Time (s)

Figure 8.- Average fuel temperature, TRAC-BF1.

Total reactor power

2-

0

0

0-

98. __-=TRAC/BF1T

94

92.

90.

88

86.

84.

82

so___________

78.

76 ___________

74

72 __________

70_____________________

68

66

64

62 A60 -

58 -______________________________

56

54 '

52

0 25 50 75 too 125 ISO

Time (s)

175 200 225 250 275 300

Figure 9.- Core thermal power, TRAC-BF1 calculation.

23
Void reactivity

0,05

0.00

-0.05

-0,10

-0,15

-0,20

" -0.25

-0,30

-0,35

-0,40

-0.45

-0,50

-0,55

25 50 75 100 125 ISO 175 200 225 250 275 300

Time (s)

Figure 10.- Void fraction reactivity, TRAC-BF1.

Doppler reactivity

0.0070

0,0065

0,0060

0,0055

0.0050

0,0045

"0,0040

' 0,0035

S0,0030

0 25 50 75 I00 125 150 175 200 225 250 275 300

Time (s)

Figure 11.- Doppler reactivity, TRAC-BF1.

24
On the other hand, the steam flow to the turbine and the feedwater flow to tfie

reactor are reduced and consequently, the feedwater heater efficiency is modified. For

this reason, the feedwater temperature starts to decrease 20 s after the trip. The

feedwater temperature drops approximately 14 0 C; this variation was introduced in the

input deck as a table, which shows the behaviour of the variable along the actual

transient. Figure 12 displays the measured and the input signal in TRAC-BF 1.

Figure 13 presents the calculated moderator temperature reactivity. The

moderator temperature cool down causes reactor power to begin to increase again due to

positive reactivity insertion. At 250 s the reactor power values are close to 72 / 70 %

(TRAC-BF1/ plant data).

The average void fraction in the core in the final steady state will not be the

same as the initial state. The positive reactivity inserted from the Doppler efficct has to

be balanced by moving down the boiling boundary in the fuel channel.

Feedwater temperature

184

183

182 -TRAC/BF1 -

181 - Plant _

180

'79

178

177

176

175

=174

E 173

I! 172

171

170

169 -

168

167.

166

165 -1

0 25 50 75 100 125 150 175 200 2 250 275 300

Time (s)

Figure 12.- Evolution of feedwater temperature.

25
Moderator temperature reactivity

0.0750

0.0625

0,0500

0,0375

0.0250

0.0125

0,0000

-0,0125

-0.0250

0 25 50 75 100 125 150 175 200 225 250 275 300 Time (S)

Figure 13.- Moderator temperature reactivity, TRAC.-BF1.

Total reactivity

0.10

0.05

0.00

-0.05

-0.10

-0.15

•. -0.20

-0.25

-0.30

-0.35

-0.40

-0.45

-0.50

-0.55

-0.60

I-TRACIBF1 I

0 25 50 75 100 125 150 175 200 225 250 275 300

Time (s)

Figure 14.- Total reactivity, TRAC-BF1.

26
Figure 14 shows total reactivity, in which the main contribution, in the first

seconds of the transient, is the void fraction. The effect of the moderator temperature

cool down starts to affect reactivity from 130 s, when the reactivity void fraction effect

begins to decrease.

Finally, Figure 15 shows the calculated and measured behaviour of the average

power range monitor (APRM) signal during the transient. The TRAC-BF 1 signal is

based on the percent rated core power to represent the APRM signal. A good degree of

accuracy is attained for the nuclear power final value and evolution (72 / 70 %, TRACBF1/ plant data) considering that the solution scheme used in TRAC-BF1 to integrate

point kinetics equations is a Runge-Kutta integration method. Moreover, it is necessary

to take into account that the xenon and iodine effects were not included in the TRAC

simulation.

Total reactor power

P

a.

102-Pa Io2 AjiAm[lhawl W1I• lnt

100 T

98

96

94

92

90

88

86

84

82

80

78

76

74

72

70-

68

66

64

62-L

60 A

58 -

56

54

52

0 25 50 75 100 125 ISO

Time (s)

175 200 225 250 275 300

Figure 15.-Total reactor power, TRAC-BF1 and plant data.

27
Reactor Pressure

When the pressure control is working, the pressure measured in the sensor line

(upstream to the control valve) is compared to the pressure regulator set- point. The

nominal steam flow rate causes a pressure drop in the steam lines of 3,87 kg/cm 2

Therefore, while the nominal reactor pressure is 70,65 kg/cm 2, the sensor line pressure

is 66,78 kg/cm 2. Figure 16 shows steam flow rate-pressure relation.

(kgcmn2)

70.65-

70.30-

(Psig)

1000-

68.89 --

67.49-

66.78-

65.38-

64,67-

64-

55 PSI

PRESSURE STEAM

LINE

3'87 k9Icm2

40 PSI

PRESSURE REGULATION

RANGE

2r8 kVcmt

T I

63.27-

20 25% 40 so 60

%STEAM FLOW

I1

80 10

Figure 16.- Pressure at the sensor line and regulator set -point.

As a result of the recirculation pump trip, the core flow rate was decreased and

therefore, the pressure goes down nearly 2 kg/cm2 due to the reduction in the steam flow

generation rate. At 250 s, the pressure reaches about 68.0 kg/cm 2 (TRAC-BFI/ plant

data) and remains constant the rest of the transient. Figure 17 displays the calculated

28
and measured data for steam dome pressure variation with time during the transient. The

results obtained by TRAC-BF1 are in excellent agreement with the measured data.

The difference in power final level between plant data and the code leads to

steam flow variations between TRAC-BF1 and plant data (Figure 18). However, the

agreement is good considering that an ideal separator-dryer model (included in

VESSEL component) has been selected to perform this analysis.

Reactor Pressure

72,5

72.0

71,5

71,0

70.5

E

70,0

69,5

69,0

68.,5

68,0

67,5

67,0

66,5

F~w-7.

Jh.LJLýAJ V11"RA

0 25 so 75 100 125 IS0

Time (s)

175 200 225 250 275 300

Figure 17.- Steam dome pressure, TRAC-BF1 and plant data.

29
Total steam flow

2800

2700 -_Plant

2600 -TRACIBF1

2500. K

2400

2300

2200

2100

• 2000

LL 1900

1800

1700

1600

1500

1400

1300 I I I

0 25 50 75 100 125 150 175 200 225 250 275 300

Time (s)

Figure 18.- Steam flow evolution, TRAC-BF1 and plant data.

Reactor Level

After the recirculation pump trip, ten jet pumps quickly reduce the amount of

flow suctioned and additional flow rate from the lower plenum starts to enter jet pump

diffusers, due to the pressure decrease inside the jet pumps. Under these conditions the

downcomer water level, of the plant, rises quickly, from 64 to 90 cm. The level control

demands less feedwater flow and the level decrease towards its initial state.

Figure 19 shows the measured and calculated change in the sensed reactor water

level with time. As shown in Figure 19, the water level peak and the time of its

occurrence are predicted remarkably well. In fact, during the initial stage of the

transient, the TRAC-BF1 calculations predict the measured water level very closely. As

30
the plant level starts to return to its original value beyond 80 s, slight discrepancies in

the subsequent TRAC-BFI level evolution can be observed.

Reactor level

0

E

-J

1.00

0.95

0,90

0.85

0.80

0,75

0.70

0.65

0.60

0,55

0.50

0 25 50 75 100 125 150 175 200 225 250 275 300 325 35)

Time (s)

Figure 19.- Comparison between calculated and level plant data.

Recirculation Flow

Recirculation pump "A" trip causes a change in pump "B" working conditions.

The reactor pressure variation induces a slight displacement along the pump

characteristic curve. Lower plenum pressure decreases and the reactor pressure is also

reduced due to the steam flow reduction, as can be seen above.

Figure 20 shows flow rate pump "B" evolution. An increase in the flow can be

observed with regard to steady conditions in response to the pressure drop. The final

31
value of recirculation flow calculated with TRAC is 1770 l's, very close to plant data,

that is about 1760 1/s.

Recirculation flow rate "B"

2000

1950

1900

1850

1800

-TRACIBF1

U

0

cr170FI:r -V I-':

0 25 50 75 t00 125 150

Time (s)

175 200 225 250 275 300

Figure 20.- Flow rate pump "B", TRAC-BF1 and plant data.

Pump "A" flow rate reduction is shown in Figure 21. The plant recirculation

flow instrumentation is not able to measure backflow and the flow measured is positive

throughout the transient. The recirculation discharge valve is completely closed at 120 s

and the recirculation flow is reduced to zero in this loop.

Besides, the inactive jet pumps are backflowing. A fraction of this backflow

enters the recirculation loop until the operating crew closes the discharge valve. On this

occasion, the reaction was very fast and it has been estimated that the valve started to

close 30 s after the recirculation pump trip. However, most backflow directly reaches

the downcomer. Real data about the opening and closing of the discharge valve have

been included in the TRAC-BF1 input deck.

32
Recirculation flow rate "A"

1875 -TRArdBF1

1750 -

1625

1500

1375

1250

1125

S1000

875 0

E 750

625

375

250

125-

oNd 

-I

0 25 50 75 100 125 150

11me (s)

175 200 225 250 275 300

Figure 21.- Flow rate pump "A", TRAC-BF1 and plant data.

Core Flow

The jet pump core flow measurement system is calibrated to measure core flow

when both sets ofjet pumps are in forward flow and the total core flow is the sum of the

indicated loop flows. However, for single loop operation, the inactive jet pumrps will be

backflowing, so the measured flow in the backflowing jet pumps must be subtracted

from the measured flow in the active loop, in order to estimate core flow. This is

because the plant core flow instrumentation is not able to measure backflowv and the

flow measured is positive throughout the transient. Figure 22 shows the ccmparison

between the indicated core flow in the plant and TRAC-BF1.

The total core flow will ordinarily be measured by the formula,

(Total Core Flow) = (Active Loop Indicated Flow) - C (Inactive Loop Indicat,-d Flow)

33
where the factor C = 0.95 is the recommended (conservative) value for measurement of

core flow during Single Loop Operation (SLO).

Figure 23 shows the comparison between TRAC-BF 1 and the plant flow data

corrected with factor C. The uncertainty in measured core flow should be 6% of rated

core flow for single loop operation. The calculated flow in plant is about 10700 T/h; the

6 % of uncertainty in measured core flow means that the measured core flow was

(10050, 11350). Note that the TRAC-BF1 calculation (about 11600 T/h) is slightly

higher than the core flow estimation of the plant data.

Core flow

23000

22000

210 Plant (measured) 21000 •-TRAC/BFI

20000

1900O 18000

•.17000 / - -

S16000-

15000

14000

13000

12000

11000

9000

80OO

0 25 50 75 100 125 ISO 175 200 225 250 275 300

Time (s)

Figure 22.- Core flow, TRAC-BF1 and measured plant signal.

34
Core flow

23000

22000

21000

20000

19000

18000

17000

16000

.215000

0 25 50 75 100 125 150 175 200 225 250 275 300

Time (s)

Figure 23.- Core flow, TRAC-BF1 and calculated plant data.

Feedwater Flow

The feedwater flow is controlled by the level control or feedwater control

system. This control is based on a level error signal and mismatch between steam and

feedwater flows. After the trip, the level control reacts to level rise and leads to a

decrease in the demand of feedwater flow.

Figure 24 is a representation of feedwater flow as a function of time. It can be

seen that the TRAC-BF1 prediction of feedwater flow is in close agreement with the

measured feedwater flow. However, for the second part of the transient, where

feedwater flow accommodates to the new steady state, the TRAC-BF1 predicted flow

differed from the plant data as much as 6 %. Nevertheless, the overall agreement

between calculated and measured feedwater flow rate is adequate.

35
Total feedwater flow

25 50 75 100 125 150 175 200 225 250 275 300

"ime(s)

Figure 24.- Total feedwater flow, TRAC-BF1 and plant data.

36
VII. CONCLUSIONS

A model of Santa Maria de Garofia NPP has been developed and proven to be

adequate for operational transient analysis. The transient of a trip of one recirculation

pump has been reproduced with this model and results have been compared with plant

data.

A good degree of accuracy is attained for the thermal power final value and

evolution, considering the propagation errors inherent to the solution scheme used in

TRAC-BF 1 to integrate point kinetics equations.

The agreement between the results in pressure evolution is good considering that

an ideal separator-dryer model has been selected to perform this analysis.

As for downcomer level, the agreement is good during the increase level stage.

Slight discrepancies in the subsequent level evolution can be observed but it should be

taken into account that the total range of the level instrumentation is 3 meters and the

discrepancies are around 3 to 5 cm (1.6 % of error).

Core flow during single loop operation is measured in a conservative way so that

there are no significant discrepancies between core flow evolution calculated by TRACBF I and the corrected data plant.

Control systems models closely simulate the response of plant cantrollers.

Further improvements or adjustments of feedwater control could more accurately

evaluate behaviour of feedwater flow.

37
VIl. REFERENCES

[1] NUCLENOR, S.A., "Informe disparo Grupo A de recirculaci6n (09/06/93)",

REV.0, July 1993.

[2] NUCLENOR, S.A., "Informe disparo Grupo A de recirculaci6n (13/08/93)",

REV.0, September 1993.

[3] G.H. Chao, D.C. Serell, "Core Flow Measurement and Uncertainty Analysis During

Single Recirculation Pump Operation", NEDE-25263, GENERAL ELECTRIC,

February 1980.

[4] Division of Systems Research, "TRAC-BFJ/MODI: An Advanced Best-Estimate

Computer Program for BWR Accident Analysis", NUREG/CR -4356, Vol.1- Model

Description, Vol.2- User' s Guide, June 1992.

[5] Division of Systems Research, "TRAC-BF1/MODI: Models and Correlation's",

NUREG/CR -4391, August 1992.

[6] NUCLENOR, S.A., "Cuadernos de Cdlculo. NUCLE-Ol", REV.0, November 1994.

[7] ENUSA, S.A., "Informe de diseho nuclear para Santa Maria de Garoha. Ciclo 17".

38
NRC FORM 335 U.S. NUCLEAR REGULATORY COMMISSION 1. REPORT NUMBER

(2-89) (Assigned lhy NRC, Add Vol., Supp., Rev.,

NRCM 1102. and Adden "lum Numbers, If any.) 3201.3202 BIBLIOGRAPHIC DATA SHEET

(See Instructions on the reverse) NUREG(IA-0193

2. TITLE AND SUBTITLE

Assessment of Single Recirculation Pump Trip Transient in Santa Maria

de Garona Nuclear Power Plant With TRAC-BF1/MODI, Version 0.4 3. DATI: REPORT PUBLISHED

MONT-I YEAR

Janu ry 2001

4. FIN OR GRANT NUMBER

5. AUTHOR(S) 6. TYPE OF REPORT

C. Garcia de la Rua, J. Garcia Garcia, J.V. Lopez Montero 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 mail, ng address: if contractor,

provide name and mailing address.)

NUCLEAR, S.A.

C/ Herman Cortes 26

39080 Santander

SPAIN

9. SPONSORING ORGANIZATION - NAME AND ADDRESS (If NRC, type 'Same as above'; f contractor, provide NRC Division, Office or Region, U.S. N,•cear Regulatory Commission,

and mailing address.)

Division of Systems Analysis and Regulatory Effectiveness

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

10. SUPPLEMENTARY NOTES

11. ABSTRACT (200 words or less)

This report has been prepared by NUCLENOR in the framework of the CAMP/SPAIN Project. It representa; one of the

application calculations submitted in fulfillment of the bilateral agreement for cooperation in thermal hydraultc activities between

the Consejo de Seguridad Nuclear of Spain (CSN) and the United States Nuclear Regulatory Commission (USNRC). The work

consisted in using the TRAC-BF1 code to reproduce a transient that took place at Santa Maria de Garona (,4PP) on June 9,

1993. The event was originated by a transformer failure that led to a loss of generator excitation and consequently a

recirculation pump trip. Santa Maria de Garona NPP is a 1381 MWth General Electric Boiling Water Reactor 3 (GE BWR/3)

owned by NUCLENOR, S.A., a Spanish utility that participates in the CAMP Program as a member of UNIOAD ELECTRICA

S.A. (UNESA). The simulation has been carried out with the TRAC-BF1I/MOD1, code, version 0.4, running on a Workstation

Hewlett Packard c180u under HP-UX operating system. As a result of the analysis it is felt that TRAC-BF1 is capable of

reproducing the plant behavior with an acceptable degree of accuracy. The main phenomena of the transient have been

calculated correctly.

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

TRAC-BF1 unlimited

Recirculation Pump Trip 14. SECURITY CLASSIFICATION

BWR (n is Page)

unclassified

(77 is Report)

unclassified

15. NUMBER OF PAGES

16. PRICE

NRC FORM 335 (2-89) This form was electron~icypde bEitFeraFos.I. NRC FORM 335 (2-89) This form was electronically poduced by Elite Federal Forms, Inc.


Fn Printed recycledg r

Federal Recycling Program

NUREG/IA-0193 ASSESSMENT OF SINGLE RECIRCULATION PUMP TRIP TRANSIENT IN SANTA MARIA

DE GARONA NUCLEAR POWER PLANT WITH TRAC-BF1/MODI, VERSION 0.4

JANUARY 2001

UNITED STATES

NUCLEAR REGULATORY COMMISSION

WASHINGTON, DC 20555-0001

OFFICIAL BUSINESS

PENALTY FOR PRIVATE USE, $300