NUREG/IA-0 150

International

Agreement Report

Study of Transients Related to

.AMSAC Actuation,

Sensitivity Analysis

Prepared by

J.M. Posada, P.M.S.A., F. Revetos, A.N.A., J. Sanchez-Baptista, A.N.A.

A. Perez-Navas, A.N.A., P. Moreno, P.M.S.A.

P.M.S.A.

Av. Vifluelas, N*33, 1*D

Tres Cantos - 28760 MADRID

SPAIN

Asociacion Nuclear ASCO (A.N.A)

Paral.lel, 51

08004 BARCELONA

SPAIN

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

November 1998

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-0150

International

Agreement Report

Study of Transients Related to

AMSAC Actuation,

Sensitivity Analysis

Prepared by

J.M. Posada, P.M.S.A., F. Revetos, A.N.A., J. Sanchez-Baptista, A.NA

A. Perez-Navas, A.N.A., P. Moreno, P.M.S.A

P.M.S.A.

Av. Vifittelas, N*33, 1*D

Tres Cantos - 28760 MADRID

SPAIN

Asociacion Nuclear ASCO (A.N.A)

Paral.lel, 51

08004 BARCELONA

SPAIN

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

November 1998

Prepared as part of

The Agreement on Resarch Participation and Technical Exchange under the International Code Application and Maintenance Program (CAMP)

Published by

U.S. Nuclear Regulatory Commission

ABSTRACT

The Asociaci6n Nuclear Asc6 (ANA) has prepared a model of Asc6 NPP using,

RELAP5/MOD3 .2. This model, which includes thermal-hydraulics, kinetics and protection

and control systems, has been qualified in previous calculations of several actual plant

transients.

Asc6 NPP is a two unit station of three loop Pressurized Water Reactors (PWR) of

Westinghouse design operated by ANA. ANA is a spanish utility that contributes to the

Code Application and Maintenance Project (CAMP) as a member of UNIDAD

ELECTRIICA S. A. (UNESA).

This report summarizes the results obtained with Asc6 NPP model for a loss of

normal feedwater ATWS event and presents a sensitivity analysis to kinetic parameters for

the same transient.

The phenomenology prediction has been useful from the operation and safety point

of view.

iii

TABLE OF CONTENTS

1. INTRODUCTION 1.............

2. PLANTAND TRANSIENT DES CRPPTION ............................................. 5

2.1. PLANT DESCRIPTION........................................................... 5

2.2. TRANSIENT DESCRIPTION......................................................... 7

3. MODEL DESCRIPTION ................................................................... 9

3.1. THERMAL-HYDRAULIC MODEL ............................................ 9

3.2. KINETIC MODEL................................................................. 11

3.3. CONTROL AND PROTECTION SYSTEMS MODEL ..................... 11

4. STEA4DYSTATE CALCULATION ....................................................... 15

5. TRANSIENT CALCULA TIONA TBOL CONDITIONS ..................... 17

6. SENSITIVTYANAL YSIS TO KINETIC PARAMETERS .................... 31

7. UNSTA TIS TICS .......................................................................... 39

8. CONCLUSIONS............................................................................. 41

9. RE.FERENCES............................................................................... 43

10. 44IiEX I ...................................................................................... 45

V
LIST OF FIGURES

Figure 1. Reactor Power......................................................................... 20

Figure 2. Average Coolant Temperature ....................................................... 21

Figure 3. PRZ Water Volume................................................................... 21

Figure 4. PRZ Pressure.......................................................................... 22

Figure 5. Primary relief and safety valves flow ............................................... 22

Figure 6. SG Pressure............................................................................ 23

Figure 7. Total Steam-dump flow .............................................................. 23

Figure 8. Secondary safety valves flow........................................................ 24

Figure 9. SG Water Mass ....................................................................... 24

Figure 10. SG Narrow Range Level and Turbine Power...................................... 25

Figure 11. Heat Transfer from primary to secondary side .................................... 25

Figure 12. Auxiliary FW flow .................................................................. 26

Figure 13. Core Flow............................................................................ 26

Figure 14. FW Flow ............................................................................. 27

Figure 15. Total Steam Flow.................................................................... 27

Figure 16. Total Reactivity ..................................................................... 28

Figure 17. Temperatures. Volume 670......................................................... 28

Figure 18. Vapor Generation. Volume 670 .................................................... 29

Figure 19. Reactor Power ....................................................................... 32

Figure 20. Turbine Power....................................................................... 33

Figure 21. Average Coolant Temperature...................................................... 33

Figure 22. PR.Z Water Volume ................................................................. 34

Figure 23. PRZ Pressure........................................................................ .34

Figure 24. SG Pressure .......................................................................... 35
Figure 25. SG Water Mass ...................................................................... 35

Figure 26. Core Flow............................................................................ 36

Figure 27. FW Flow ............................................................................. 36

Figure 28. Total Steamn Flow.................................................................... 37

Figure 29. Total Reactivity..................................................................... .3 7

vii
LIST OF TABLES

Table 1. Main Characteristics of Asc6 I and II Nuclear Station ............................... 6

Table 2. Sequence of events of the transient .................................................... 7

Table.3). Steady State Parameters .............................................................. 15

Table 4. Run Statistics .......................................................................... 3)9

Viii
LIST OF DIAGRAMS

Diagram 1. Asc6 NPP Nodalization Diagram. (ANNEX 1) 46

Diagram 2. Main and Auxiliary FW Nodalization Diagram. (ANNEX 1) 47

ix

EXECUTIVE SUMMARY

Asc6 Nuclear Power Plant is a two unit station of three loop Pressurized Water

Reactors ( PWR) of Westinghouse design.

The Thermal-hydraulic analysis group of the Asociaci6n Nuclear Asc6 (ANA) has

prepared a model of the plant using RELAP5/MOD3.2. This model includes thermalhydraulics, kinetics and protection and control systems.

The aim of the study presented in this report is to fulfill ANA's commitment with the

Code Assessment and Maintenance Program (CAMP).

The transient selected to be analyzed for this purpose is the Loss of feedwater ATWS

event at Asc6 NPP with RELAP5/MOD_3.2.

In the following pages the results obtained under two points of view will be

presented:

1. Best-estimate results using design assumptions.

2. Sensitivity analysis to kinetic parameters.

The main conclusions are as follows:

" The maximum peak pressure in the reactor coolant system is 19.0 MPa. It is well

below the ASME Boiler and Pressure Vessel Code Level for BOL conditions

(service limit stress criterion of 3200 psig (22.16 MPa).) /16/.

" The maximum peak pressure in the reactor coolant system depends significantly

on the kinetic parameters.

* After the loss of feedwater, the pipe between the feedwater isolation valve and

the header of auxiliary and main feedwater, works as a water tank. This stored

water mass helps to cool the primary and produces the isolation of the steam

generator as well as the start of the High Pressure Injection System (HPIS).

* RELAP5/MOD3.2 is a valuable tool to analyze plant transients.

Xi

ANA:

AMSAC:

ASME:

ATWS:

BOL:

CAMP:

CSN:

EW:

HPIS:

MOL:

MSR:

NPP:

PRA:

PRZ:

PWR:

SBLOCA:

SG:

UNESA:

LIST OF ABREVIATIONS

Asociaci6n Nuclear Asc6

ATWS Mitigating System Actuation Circuitry

American Standard for Mechanical Engineers

Anticipated Transients Without Scram.

Begining Of Life

Code Application and Maintenance Program

Consejo Seguridad Nuclear

Feed Water

High Pressure Injection System

Middle Of Life

Moisture Separator Reheater

Nuclear Power Plant

Probabilistic Risk Assessment

Pressurizer

Pressurized Water Reactor

Small Break Loss Of Coolant Accident

Steam Generator

Unidad Eldctrica S.A.

xiii

1. INTRODUCTION

In 1986 the Asociaci6n Nuclear Asc6 (ANA) created a group for plant and core

thermal-hydraulic analysis. The objectives of the group are as follows:

1) Create and update core and plant thermal-hydraulic models based on bestestimate criteria.

2) Provide off-line engineering support to the different technical branches of ANA

(i.e., technical services, reactor operation):

a) Analyze operating events that result in event reports.

b) Assess plant systems and/or equipment modifications as well as plant

operating procedures and emergency instructions.

c) Analyze plant behavior under incident or accident conditions in the above

mentioned cases.

d) Scenarios and core damage evaluation for probabilistic risk assessment.

3) Review final safety analysis report transients and accidents based on bestestimate criteria.

4) In the future and if appropriate, participate in Asc6 individual plant examination.

The plant analysis activities developed so far include the following:

1) Implementation of RELAP5/MOD2 and migration to RELAP5/MOD_3.2/1/.

2) Preparation of the thermal-hydraulic model of both the primary and secondary

system /2/.

3) Preparation of the kinetic model specifically adapted to Asc6.

4) Simulation of control and protection systems /3/.

5) Revision and detailed study of all start-up tests /20/ and every transient that has

occurred in each unit along these twelve years (a total of 65 actual cases were

studied with the model of plant for RELAP5/MOD2 and RELAP5/MOD3.2).

6) Selection of transients suitable for assesment.

a) Black-out.

I
b) Turbine power step.

c) Run-back of turbine because of one feedwater pump trip

d) Turbine trip with all systems available.

e) Reactor Trip because of low-low level in one SG as a result of a closure of the

isolation feedwater valve.

f) Loss of load 100-50%

g) Transition mode of the feedwater valves control system

7) Simulation of the above seven transients and adjustment of control parameters

8) Participation in the International Code assessment and Application Program /11l/.

9) Participation in the Code Assessment and Maintenance Program .

10) Analysis of transients such as small-break-loss-of-coolant accident (SBLOCA)

/21/, /22/, anticipated transient without scram (ATWS), and others for PRA

studies.

At this point, the adjustment and qualification process is considered to be completed.

Sufficiently accurate predictions with meaningful sets of measured data provide validation of

both the model and the procedures to be used in the future to analyze various transient and

accident scenarios of general interest.

The analysis presented in this report has to do with transients related to AMSAC

(ATWS Mitigating System Actuation Circuitry) actuation.

Early in 1989, when AMSAC system was installed in Asco NPP, a set of

calculations were performed /23/ with the purpose of getting a general knowledge of the

phenomenology of related transients. At this time the model of the old configuration of the

plant was used (old SGs and old controls).

Today, as the plant has changed (see section 2.1 'Plant Description') and the model

has improved its level of qualification, it is interesting to re-study the same subject with a

larger scope of analysis.

A transient explanation is given in section 2.2 after a brief description of the plant

2
itself (section 2.1).

Chapter 3 is a summary of what is included in the model and chapter 4 explains the

steady state calculation for nominal conditions.

As the transient will be analyzed under two different points of view, chapters 5

and 6 will explain each one of them. The first one consists of a calculation with a bestestimate code (RELAP5/MOD3.2) using the same boundary conditions and setpoints than

those defined by the vendor in its calculations /12/. The second one has to do with a

sensitivity analysis of the transient under different kinetic parameters (BOL vs MOL).

Run statistics, conclusions and references can be found in chapters 7, 8 and 9 while

diagrams showing the nodalization used are in Annex I .

3

2. PLANTAND TRANSIENT DESCRIPTION

2.1. PLANT DESCRIPTION

Asc6 Nuclear Power Plant is a two unit station of three ioop Pressurized Water

Reactors (PWR) of Westinghouse design. The first criticality was reached in June 1983 in

Asc6 I and in September 1985 in Asc6 II. Today, unit I is in its twelfth cycle and unit 11 in its

eleventh of normal operation.

The main characteristics of both units are given in Table 1. The core contains 157 fuel

assemblies of (17x 17 -25) fuel rods.

The steam generators have been recently changed, in October of 1995 for Unit I and in

October of 1996 for Unit HI. The new ones are SIEMENS-FRAMATOM (61 W/D3).

At the same time a deep upgrade of the original control systems has been performed.

The new controls are fully digital and allow to establish sets of data more suitable for

assessment.

All other major components are standard Westinghouse components.

Main Characteristics of Asc6 I and II Nuclear Station

Thermal reactor power 2686 Mwth

Fuel U02

Number of assemblies 157

Fuel rods per fuel assembly (17xl7 - 15) =264

Active length of fuel rods 3.657 m

Outside diameter of fuel rods 4.75e' m

Cladding tube material Zr-4

Cladding tube wall thickness 0.655e' m

Average linear heat generation rate 17750 W/m

Absorber rods per control assembly 24

Absorber material Ag - n -Cd

Number of coolant loops 3

Reactor operating pressure (PRZ) 15.51 MPa

Coolant Average Temperature 581.3 K

5
Coolant flow rate 114287 Kg/s

Steam Generator

Type 61 WID3

Number 3

Reactor coolant pumps

Type Westinghouse 93-DS

Discharge head 86.25 m

Design flow rate 5.928 ma/s

Speed 155 i-ad/s

Pressurizer

Height 12.835 m

Diameter (inner) 2.134 m

Volume 39.64 m"

Operating saturation pressure 15.51 MPa

Heating power of the heater rods 1.40 Mwe

SteanifPower Conversion Plant

Feed Water flow rate 497.5 Kg/s (loop)

Main steam flow rate 1492 Kg/s

Steam moisture at steam generator outlet 0.25%

Feedwater temperature 497.05 K

Table 1. Main Characteristics of Asc6 I and 11 Nuclear Station.

6
2.2. TRANSIENT DESCRIPTION

After a loss of normal feedwater, the reactor protection systemn and the average

temperature control system are assumed to fail. So there is no reactor trip signal as wvell as no

rod insertion. At this point, the AMSAC (ATWS Mitigating System Actuation Circuitry)

generates the turbine. trip signal and starts the auxiliary feedwater pumps to mitigate the

transient.

In table 2 a description of the main events are presented for the RELAP study.

Sequence of events of the transient

Event Time (seconds)

Loss of main feedwater 0.

Main feedwater flow completely lost 4.

Pressurizer relief valves lift 31.

Low SG level AM SAC setpoint reached 41.

Turbine Trip via AMSAC signal 42.

Auxiliary feedwater pumps start via AMSAC signal 71.

Maximum RCS pressure reached 94.

Primary pumps trip 150.

Main steam isolation signal and HPIS start 150.

Table 2. Sequence of events of the transient.

7

3. MODEL DESCRIPTION

The model corresponds to the plant configuration explained in section 2.1. It includes

the new SGs as wvell as the new controls.

Diagrams 1 y 2 show the model used to simulate the plant. It consists of 152 volumes

with its correspondent junctions, 51 valves, 27 time dependent volumes, 8 time dependent

junction, 6 heat structures, 466 control variables and 351 trips. The model includes the

vessel, the three primary loops, the pressurizer, the three steam generators, the three

secondary loops, and the steam lines. The turbine, condenser, moisture separator reheaters,

suction lines of main and auxiliary feedwater pumps are modeled as time dependent

volumes.

3.1. THERMA4L-HYDRAULIC MODEL

a) Primary system

The model of the primary system (diagram 1) includes the main components of the

plant. The core is modeled by volume 120 and the proper heat structures. Volume 130

simulates the by-pass region between the core baffle and the core barrel. Volume 140 models

the upper plenum and volumes 150 and 160 the vessel upper head. The three hot lines depart

from the core upper plenumn. Loop 3 (volume 410) is connected with the pressurizer through

volume 510. The pressurizer is divided in 2 volumes.

The lower one (volume 520) is divided into 19 nodes. Heat structures, simulating the

pressurizer actual heaters, are attached to the first three nodes. Volume 525 is a branch in

order to model the junctions connecting the pressurizer with the safety and relief valves and

with the spray system.

Volume 420 models the remaining hot leg. Volumes 430 and 440 simulate the water

boxes of the steam generator. The primary side of the steam generator is modeled by volume

43 1 divided into 20 nodes. Volume 460 models the primary coolant pump. Proper

homologous curves, given by the vendor, have been used for this purpose. Volumes 268, 368,

468 model the HPIS. Time dependent junctions 267, 367, 467 model the high pressure

injection pumps by means of the pump characteristic curves. Accumulators are also

modeled.

9
b) Secondary system

The model of the secondary system (diagram 2) starts with time dependent volumes 662

and 654, that represent the suction lines of the main feedwater pumps. Volumes 679, 693 , and

682 represent the suction lines of the auxiliary feedwater pumps.

Volumes 664 and.656 model the main feedwater pumps. Proper homologous curves,

given by the vendor, have been used for this purpose.

Time dependent junctions 690, 695, 694 model the auxiliary feedwater pumps by

means of the pump characteristic curves.

Volumes 665, 666 and 657, 658 represent lines from pumps to hot-collector (volume

668). The feedwater high pressure heater is modeled by a heat structure between the first node

of volumes 666 and 658 with volume 942. This last volume is a time dependent volume with

a temperature function of the turbine power.

Volumes 869, 870, 872, 875 (loop. 3) model lines of main feedwater to the steam

generator. Volumes 888, 852 model lines from auxiliary feedwater pumps to main feedwater

connection. Their correspondent check, control and isolation valves are also modeled.

The SG downcomer (diagram 1) is simulated by means of volumes 800, 801, 822 and

825. The steam generator riser is modeled by volume 8 10. The steam separator with volume

820. Volumes 830, 840 and 850 model the steam dryer and the dome of the steam generator.

The steam line starts at volume 880. Safety and relief valves (components 886 and 884) are

connected to volume 881. Component valve 885 models the isolation valve. Time dependent

volumes 994 and 999 represent the free atmosphere. The steam is conducted throughout

volume 8 83 to the steam collector, volumes 900 and 904. Finally valves from 93)1 to 93 )8, 903

and 907 model the by-pass to condenser valves, the turbine stop and control valves

respectively. And valve 917 represents the steam flow to moisture separator reheater.

Proper heat structures are used to thermally connect the primary side of the steam

generator with secondary side. Actual values are used for all the variables. Fouling factor

coefficient is introduced in order to achieve de actual heat transfer rate without any change in

primary average temperature and secondary pressure.

The data used to model volumes, junctions and valves as well as heat structures were

taken form plant design information /13/.

10
3.2. KINETIC MODEL

The kinetic model /3/ was prepared using the RELAP5/MOD3).2 space-independent

reactor kinetic option with data from the ANA Nuclear Analysis Group. The model includes a

scram table of reactivity versus time. The total control rod drop time is the actual value

measured at plant. This table is activated by reactor trip.

The control model supplies the reactivity of the C and D control rod banks.

This control reactivity is added to feedback reactivities calculated by the kinetic model

from the data supplied for the specific bum-up condition of each transient.

3.3. CONTROL AND PROTECTION SYSTEMS MODEL

The protection and control systems are modeled using RELAP5IMOD3).2 control

blocks and following specific setpoint studies, logical diagrams and technical specifications of

the plant /3/, /14/, /15/.

The model includes the following systems:

a) Reactor trip system.

The reactor can trip in the RELAP5/MOD3.2 model because of the following effects:

" Power range, low range, high level

" Power range, high range, high level

* Overtemperature

* Over power

" High pressurizer pressure

* High pressurizer water level

" Low pressurizer pressure

" Loss of primary coolant flow

" Low-low steam generator water level

" Turbine trip

* Safety injection

" Manual reactor trip

I1I
b) High pressure Injection System.

Using the following signals:

" Low pressurizer pressure

" High steam line differential

* High steam line flow coincident with Low steam line pressure or Low-Low Tavg.

c) Turbine Trip and Control System

The position of the turbine control valve is controlled as a function of the

difference between the required power and actual Power, with the proper control block to

model the actual logic of the plant. Turbine rmn-back signals have also been modeled.

The Turbine Trip (closure of the turbine stop valve) is modeled. The signals of Safety

Injection, very high steam generator level, reactor trip, AMSAC signal and manual signal are

used to trip the turbine.

d) Feed Water Control System

The feed water control system of the main and by-pass valves and the pumps have been

modeled as shown in diagram 2. The auxiliary feed water system is also included in the

model.

e) Pressurizer level and pressure control system

The model of the pressure control system actuates upon heaters, spray valve and

charging pump. Pressurizer safety and relief valves and level control systems are also

simulated.

f) Average Temperature Control System

The average temperature control system modeled with RELAP5IMOD3.2 controls

both the primary average temperature and the primary-secondary power mismatch.

12
Other systems modeled are:

g) Steam Dump Control System

h) Steam line isolation logic.

i) Safety and Relief valves of the secondary.

j) Primary pumps trip.

k) General Variables.

1) Permissive and interlock circuits.

13

4. STEADYSTA TE CALCULATION

A steady state calculation was performed with the plant at 100% rated condition. The

objective is to obtain a stable condition to start transients.

In Table 3 a comparison between the model results and the plant data is given for the

main plant variables.

Steady State Parameters

VARIABLE MODEL(RELAP5IMOD3.2) PLANT(/18/)

Primary mass flow rate (Kg/s) 4760.8 4761.2

Vessel AT (K) 32.6 32.7

Primary Pressure (MPa) 15.513 15.5 16

Primary Avg. Temperature (K) 581.5 581.5

Recirculation ratio 3.68 3.65

SG Narrow Range Level ()50.6 50.6

Secondary Pressure (MPa) 6.746 6.733

Steam Mass Flow (Kg/s) 494.5 496.1

Table 3. Steady State Parameters.

15

5. TRANSIENT CALCULA TIONA TBOL CONDITIONS.

According to what is establish by the Spanish Nuclear Authority (CSN) ANA must

demonstrate that the setpoints implemented in the plant for the AMSAC system are still

valid for the new steam generators.

The plant vendor issued a report, analyzing the transient with a design code /12! and

fulfilling the regulatory requirements.

In this section, the results of a base case calculation with ANA's RELAP5/MOD3).2 /I/

model will be presented. The calculation has been performed using, the same boundary

conditions and setpoints than those defined by the vendor in its report /12/.

The following assumptions are made in order to have the same vendor boundary

conditions:

" No credit for automatic reactor trip.

" Initial normal full power operation early in core life.

" No credit for automatic control rod insertion as reactor coolant temperature rises.

" No credit for pressurizer spray in pressure reduction.

" Pressurizer pressure relief through all power operated and spring loaded relief

valves available.

" Main feed water falls to 0 % of nominal flow in the first four seconds of the

transient.

" The AMSAC signal is actuated when the steam generator water level reaches

12% of narrow range span. Following the signal, a one second delay is assumed

for turbine trip.

" Auxiliary feedwater starts 30 seconds after the AMSAC setpoint is reached, at

the temperature of 54.4 0G and a total minimum flow of 259 m3/h, equally

delivered to the three steam generators.

* The moderator temperature coefficient is -12.5 pcmI0C corresponding to a

conservative value during more than 95% of the reactor time life (Asc6 Unit I

Cycle 12)

17
The vendor also makes an additional assumption which is the following:

*Primary to secondary heat transfer area is reduced as the steam generator shellside water inventory falls below the value necessary to cover the tube bundle.

This assumption is not required in RELAP calculations since the code calculates the

heat transfer between primary and secondary in all conditions.

After a through analysis of the vendor report, two supplementary assumptions have

been made in R.ELAP5IMOD3.2 study.

*No credit to turbine run-back .

*Secondary relief valves remain closed.

The results of RELAP5IMOD3.2 calculations are given in the following figures:

" Figure 1. Reactor Power.

" Figure 2. Average Coolant Temperature.

" Figure 3. PR.Z Water Volume.

* Figure 4. PRZ Pressure.

* Figure 5. Primary relief and safety valves flow.

" Figure 6. SG Pressure.

* Figure 7. Total Steam-dump flow.

* Figure 8. Secondary safety valves flow.

" Figure 9. SG Water Mass.

* Figure 10. S5G Narrow Range Level and Turbine Power.

" Figure 11. Heat Transfer from primary to secondary side.

" Figure 12. Auxiliary FW flow.

" Figure 13. Core Flow.

* Figure 14. FW Flow.

* Figure 15. Total Steam Flow.

18
* Figure 16. Total Reactivity.

* Figure 17. Temperatures. Volume 670.

" Figure 18. Vapor Generation. Volume 670.

At time zero the loss of feedwater takes place and the main feedwater flow is

completely lost in 4 seconds (table 2). As a result of this the SG water mass (figure 9) and

the SG Narrow Range Level decrease (figure 10).

When two of the three SG Narrow Range Level signals reach the value of 12%,

the AMSAC signal is produced and 1 second later the turbine is tripped (figure 10).

With the turbine trip, the secondary pressure increases (figure 6) and the heat

transfer from primary to secondary side (figure 11) falls.

Due to heat transfer decrease, the primary average temperature (figure 2)

increases and generates a negative feedback reactivity (figure 16) which results in a reactor

power decrease (figure 1).

Figure 12 shows auxiliary feedwater mass flows. According to boundary

conditions auxiliary feedwater starts with AM SAC signal instead of main feedwater

closure or low-low Narrow Range Level signal (17.6 %) (table 2). The same figure also

shows a flow decrease (320 s.) due to secondary side pressure increase (figure 6). At that

pressure the auxiliary feedwater pumps deliver a flow that corresponds to the boundary

condition of 259m3/h.

The pressurizer pressure predicted by RELAP can be seen in figure 4. When

primary pressure reaches the corresponding setpoint relief and safety valves open (figure 5)

to control the pressure peak. This pressure peak is smaller than that predicted by the vendor

in about 2 MPa. This is due to a larger heat transfer to the secondary side during the first

100 seconds of the transient predicted by RELAP. The calculation considers not only

steam-dump capacity (figure 7) and the secondary relief valves (figure 8), but also the

steam flow to MSR (Moisture Separator Reheater), auxiliary feedwater pumps and others

(see figure 15). This larger heat transfer from primary to secondary is also due to a lower

pressure in SG's (figure 6).

As it can be observed in figure 14, RELAP5/MOD3).2 predicts several peaks in

feedwater flow after second 150. These are due to flashing in the pipe between the

19
feedwater isolation valve and the header of auxiliary and main feedwater. Once the

isolation valve is closed (in the first four seconds of the transient) the water inside this pipe

upstream of the check valve remains at the temperature of main feedwater (497 K). So

when secondary pressure falls below the saturation pressure for that temperature (figure

17), the amount of vapor generated (figure 18) leads to a peak in feedwater flow (figure

14) and a sudden increase of heat transfer (figure 2). This increase produces two main

effects:

* The first one is a rapid fall of the primary pressure (figure 4), which almost

reaches the saturation pressure at the main pump suction line. This leads to the

main pumps trip due to cavitation induced vibration (figure 13). Pump is assumed

to cavitate when cold leg temperature is minus or equal to 6 TF below saturation

temperature /16/.

* The second effect consists of a quick evaporation of the water within the

steam generator, which activates the high steam line flow coincident with low

steam line pressure signal. This produces the closure of the main steam valves and

the start of high pressure injection.

The above explained flashing phenomenon has also been predicted for other plants

/19/.

BOL

.12

0- - - - - - -

OA - I I

0

0 IJO 200 300 400 Boo 600

Tirm (s)

Figure 1. Reactor Power.

20
BOL

625

t, 600 ------

59 - - - - - - - - - - - --

0 100 200 300 400 500 600

Time (s)

Figure 2. Average Coolant Temperature.

BOL

40

~30 ---- ----

M I 

N 10 - - - I 

5II

0 1D0 200 300 400 500 600

Time (s)

Figure 3. PRZ Water Volume.

21
BOL

20000000

19000000 - - ------- ---------

18000000 -1 -L ---- ----- -- ----

~. 17000000 - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -

1400000 - - -- - -

14000000 -

.10000000-

0 100 200 300 400 500 600

Timre (s)

Figure 4. PRZ Pressure.

BOL

EI

0 1 1

N 0.

I-- --

0.

0 100 200 300 400 50)

Tire (s)

Figure 5. Primary relief and safety valves flow.

22
BOL

9000000

8000000-- --- --- - - -

60~00000 - -- - -J 

~4000000 -

~2000000 -- - --------------- -

1000000 --I

0 100 200 300 400 500 600

Time (s)

Figure 6. SG Pressure.

Figure 7. Total Steam-dump flow.

23
BOL

2M50 -------

)200. --

0 10 20 30 40 0 0

Tim (s)

Figure 8. Secondary safety valves flow.

BOL

45000

40000 - - -I

S30000 - ---------------

w 20000 - - - - , - ------- -----

>~~~~~~~~~~~~~~ 1I0

000

0 'IDO 200 300 400 500 600

Time (s)

Figure 9. SG Water Mass.

24
BOL

60 1.20

50 II ----- 100D

40 i - -- - ------- - 0809Q

> 30C .60~

Z 20 - - - - - - - - - a 0.4().

10 a - a a a a

0 10 20 30 40 50 60 70 80 90 100

Thre (s)

Figure 10. SG Narrow Range Level and Turbine Power.

Figure 11. Heat Transfer from primary to secondarT side.

25
BOL

25 - - - - - - - - - - - - - - - - - - - - - )- - - --

201

-2 15 - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -- - - - - - - - - - -

Li.. 1

U- - 10 I

0 10 20 30 0 0 0

Time (s)

Figure 12. Auxiliary FW flow.

BOL

t2

_I I I

00.

0 20I0 I0 0 0 0

Iir (s)

Figur 13Ioelw

26
BOL

600

400 O - - L -

.2 2 --- 0 -- --- I I

II I

1300

2000 0 3040 0 0

Tirm (s)

F.FIR

Figure 14. FW Flow.

Figure 15. Total Steam Flow.

27
BOL

02

*--0.4 - - -

I-1

0 IJO 200 300 400 500 600

Tirro (s)

Figure 16. Total Reactivity.

BOL

80

570- - --- -- - - --- -- -- -- --- --- - -- -- ---

.5 60 -- -- -- - -- --- -- -- -- -- -- ---

!25 0 - - - - - - - - - - - - - - - - - - - - - - - -

500 - -

490

0 100-------00--.--- -- -- 400- - - -. 60

Iir (s)

- Wte TeI - StIrp. Sem

Figure 17. Temperatures. Volume 670.

28
BOL

0.7 -______-

0.6 _____ _____ _____ __ _

S0.5

-0.4

~0.3

0 0.2

0.1.

0100 200 300 400 500 600

Time (s)

Figure 18. Vapor Generation. Volume 670.

29

6. SENSITIVITYANAL YSIS TO KINETIC PARA4METERS.

In this section, the influence of the kinetic parameters on the peak- of pressure for the

loss of normal feedwater ATWS event will be studied.

For this purpose the kinetic parameters of the BOL (above section), and the

correspondent to MOL /17/ will be used.

All other conditions of the above section remain the same.

The comparison between the results obtained by RELAP5/MOD2;.2 on both

calculations is given in the following figures:

" Figure 19. Reactor Power.

" Figure 20. Turbine Power.

" Figure 2 1. Average Coolant Temperature.

* Figure 22. PRZ Water Volume.

" Figure 23. PRZ Pressure.

* Figure 24. SG Pressure.

" Figure 25. SG Water Mass.

" Figure 26. Core Flow.

* Figure 27. FW Flow.

" Figure 28. Total Steam Flow.

" Figure 29. Total Reactivity.

Figure 23 shows that the PRZ pressure predicted in the MOL case is smaller than that

predicted in the BOL case. This is due to a more negative feedback reactivity producing a

smaller reactor power during the rise of the pressurizer pressure (see figure 19 and 29).

As it can be seen in figure 24 the secondary pressure is the same from the beginning of

the transient until the peak pressure in the pressurizer is reached, for both cases. LSo both cases

have a similar heat transfer. The different kinetic parameters used in each case, lead to a

different reactor power. This power is lower in the MOL case and produces a lower peak

31
pressure by about 1.OM~a.

As explained in the above section, there is flashing phenomenon that leads to isolate the

SGs, to start the HLPIS and to trip the primary pumps.

The sudden fall of the average coolant temperature is also produced by the peak- of

feedwater flow (see figures 21 and 27).

Figure 24 shows how pressure increases until the opening of the SG safety valves.

The evolution of the steam generator water mass is shown in figure 25.

Figure 26 shows that core flow is never zero, meaning that natural circulation keeps

cooling the core after the primary pumps trip.

The volume of water in the pressurizer keeps high because of the start of the I-LP.I.

(figure 22).

BOL vs. MOL

12 0

~0.

02

0 100 200 300 400 500

The (s)

Figure 19. Reactor Power.

`32
BOL vs. MOL

9! 0.8

S0.6 0

IL

0.4

02

I I I I

I I I I I I

I I I I I I I I I

I I . I

I I I I I I I I

I I I I I I I

I I I I I I I I I

I I I I I I I I

I I I I I I I

I I I I

I I I I I I I I

I I I I I I I

I I I I I I I I

I I I I I I I

I I I I I I I I

I I i I I I I I

I I I I i I I I

I I I I I I I I I

I I - -

I I I I I I

I I I I I I I

I I I I I I I I I

I,

0 10 20 30 40 50

Tinme(s)

60 70 80 90 100

IL P.RmýP.Tiy

Figure 20. Turbine Power.

BOL vs. MOL

615

12 605 -- --- -- -- --- --- - -- -- -- -

0 1

5a5 - - -

580W

0 100 200 300 400 500 600

Timre(s)

Figure 21. Average Coolant Temperature.

33
BO0-L vs. MOL

E30

15 -- -- - N

n

10 - - - - - - - - - - - - - - - - - - - - - - - - - - - - -

0

0 100 200 300 400 500

Time (s)

-WaterR -WaterM

Figure 22. PRZ Water Volume..

BOL vs. MOL

20000000

T 1 5 0 0 0 0 0 0 - - - - - - -- -- - - - - - - - - -- - - - - - - - -

S13000000 -- - - -- - - - -- - - -

- -- -I) - - -I - - -- - - -

14 000 - - - I

10000000

0 100 200 300 400 5006

Tirm (s)

Figure 23. PRZ Pressure.

34
B Ox vs. MOL

9000000

80000=00 - - - - -

m m ------- I ------ I

0o -----

0 0 003040 0

Th)s

,I 4 - 0 M -- -- - - - -- - - -- -- - - -- - - - - - -- - - --

Figure 24. SG Pressure.

BOL vs. MOL

45M0

450000

1000 --------------- :--------4 -------- ---------

-300 10 20 0 005 0

U)G..Ms- IS MMs M I

Figure 25. SG Water Mass.

35
Figure 26. Core Flow.

BOL vs. MOL

600

100 - - - -I

-0 0I

0 10 Io 3W4IO6

~Tir0 (s)

I W F1 I W I 

Figure 27. FW Flow.

36
BOL vs. NML

1.2- -- -- - -- - - - - - - - ---- - -

0~ -- 0. -- - - - - - -- - - -- - - - - - - - -

LL:I

o 0.8 -- -

0. - - - -q 

-0. 1

0 100 200 300 400 500 600

The (s)

Figure 28. Total Steam Flow.

BOL vs. MAOL

0.5

0 1- -- - - -- -- - - -

-2

-2 5- - -I

0 100 20300

Tirb (s)

- - - -- -

Figure 29. Total Reactivity.

337

7. RUNSTATISTICS

Calculations were carried out on a HP 73 5/1 25 with UNIX-HP 10.0

RELAP5/MOD3.2 was used in all the calculations.

Run Statistics

BOL case

TRANSIENT TIME (s) 600.

CPU TIME (s) 1083.95

Total number of time steps 6049

CPU TIME/TRANSIENT TIME 1.81

MOL case

TRANSIENT TIME (s) 600.

CPU TIME (s) 1129.19

Total number of time steps 605-3

CPU TIME/TRANSIENT TIME 1.88

Table 4. Run Statistics.

319

8. CONCL USIONS

A set of best-estimate calculations of transients meant to cause AMSAC actuation

has been performed successfully with ANA's model.

The maximum peak pressure in the reactor coolant system is 19.0 MPa. This is well

below the ASME Boiler and Pressure Vessel Code Level for BOL conditions ('service

limit stress criterion of 3200 psig. (22.16MPa).) /16!.

The maximum peak pressure in the reactor coolant system depends significantly on

the kinetic parameters.

After the loss of feedwater, the pipe between the feedwater isolation valve and the

header of auxiliary and main feedwater, works as a water tank. This stored water mass

helps to cool the primary and produces the isolation of the steam generator asvwell as the

start of the High Pressure Injection System.

Complete models of systems like main and auxiliary feedwater, (usueily

simulated as boundary conditions) help to find out unsuspected phenomena like the

previously mentioned.

The results obtained in the study of this transient improve the knowledge of plant

dynamics.

The model of Asc6 using RELAP5IMOD3.2 is a valuable tool to analyze plant.

transients and to provide engineering support to plant operation.

41

9. REFERENCES.

/I/ RELAP5/MOD'3.2 CODE MANUAL, EGG-2596, June 1995.

/2/ F. REVENTOS, J. SANCHEZ-BAPTISTA, J.M. POSADA and P. MORENO, 'Modelo

Termohidraulico de la C.N. Asc6', Asociaci6n Nuclear Asc6, March 1998.

/3/ F. REVENTOS, 3. SANCHEZ-BAPTISTA, J.M. POSADA and P. MOREN0, 'Modelo

Termohidraulico y de protecci6n y control de la C.N. Asc6', Asociaci6n Nuclear Asc6,

March 1998.

/4/ F. REVENTOS, J. SANCHEZ-BAPTISTA and P. MORENO, 'An6.lisis de transitorios

en A.N.A. con RELAP5/MOD2', presented at 14th Annual Mtg. Sociedad Nuclear

Espafiola, Marbella, Spain, October 26-28 1988.

151 F. REVENTOS, 3. SANCHEZ-BAPTISTA, P. MORENO and A. PEREZ NAVAS,

'Transient Analysis for ASCO Nuclear Power Plant Using RELAP5/MOD2', Proc. Ist. Int.

RELAP5 Users' Seminar, College Station Texas, January '31 - February 2 1989.

/6/ F. REVENTOS, J. SANCHEZ-BAPTISTA, P. MORENO and A. PEREZ NAVAS,

'Transient Analysis in the Asc6 NPP using RELAP5/MOD2', Nuclear Technology,

Volume 90, Number 3, pg. 294-3 07, June 1990.

M7 R. TRUJILLO TARIFA, 'Disefio y verificaci6n de sistemas de protecci6n y control

pama transitorios en centrales nucleares tipo -PWR- disefio Westinghouse'. Tesis Doctoral.

1996.

/8/ F. REVENTOS,, J. SANCHEZ-BAPTISTA, J.M. POSADA and P. MORENO,

'Simulaci6n de los transitorios: ASI-037 y ASI-046 con el modelo de plaiita para el

c6digo RELAP5/MOD3.2-NPA', July 1997.

/91 F. REVENTOS, J. SANCHEZ-BAPTISTA, J.M. POSADA and P. MORENO,

Simulaci6n de la prueba de arranque, Rechazo de carga del 100% al 50% (Grupc' II), con el

modelo completo de Asc6 para el codigo RELAP5/MOD3.2-NPA', September 1997.

/10/ F. REVENTOS, J. SANCHEZ-BAPTISTA, J.M. POSADA and P. MORENO,

'And.lisis del cambio de modo del sistema de control de las vdlvulas de A.A.P.', March

1998.

/11I/ F. REVENTOS, J. SANCHEZ-BAPTISTA, A. PEREZ NAVAS and P. MORENO,

'Assessment and application of blackout transients at Asc6 NPP with RELA:P5/MOD2.

43
ICAP ASSESSMENT REPORT, June 1991

/12/ C.MA'ITHYS, N. PHAN NGOC, T. ROBIN, C. SIMON, J.P. TESCH, 'ASCO Unit 1

Anticipated Transient Without Scram Loss of Normal Feedwater Analysis', October 1995

/13/ 'Final Safety Analysis Report-Asc6 1', June 1983), 'Final Safety Analysis Report-Asc6

IF', September 1985, and later revisions, Asociaci6n Nuclear Asc6.

/14/ F. BALLERINI, 'Setpoint Study Asc6 Units I and II', Westinghouse Electric

Corporation (March 1976).

/15/ 'Precautions, Limitations and Setpoints of Asc6 Unit I', Revision N' 14, February

1996.

/16/ 'Westinghouse Anticipated Transients Without Trip Analysis', August 1974.

/17/ 'Informe de Diseflo Nuclear de ASCO I ciclo 12' ITEC-484 revision 1 ENUSA.

Octubre 1995.

/18/ J.M.POSADA, M.MARTiN, P.MORENO, 'MODIFICACCIONES AL MODELO

DE PLANTA DE ASC6 PARA EL CODIGO RELAP5/MOD3).2-NPA, February 1997.

/19/ 'Perdida de agua de alimentaci6n, alimentando con el deposito de A.A. y

despresurizando con dos vdlvulas de alivio de los GG.VV.'. Informe N0 1 8-EL-2728.

Empresarios Agrupados.

/20/ F. REVENTOS, J.SANCHEZ-BAPTISTA, and A. PEREZ-NAVAS. 'Start-up test of

ASCO NNP with new steam generators' CAMP meeting Madrid. May 1996.

/21/ C. PRETEL, F. REVENTOS, 'Consideraciones acerca de los modos dindmicos

implicados en descubrimientos parciales de nficleo en escenarios de LOCA pequeiio.' I'

Jomnadas T~cnicas de CAMP-Espafia. Barcelona. April 1997.

/22/ C. PRETEL, F. REVENTOS, 'Small Break LOCA Analysis in a full size PWR using

RELAP5/MOD3.2 (MOD2.5)', CAMP Meeting. Madrid. May 1996.

/23/ J.SANCHEZ-BAPTISTA, F. REVENTOS, P.MORENO and A.PEREZ-NAVAS

'Estudio de la secuencia adoptada en C.N.Asc6 en el andlisis del AMSAC' presented at

Annual Mtg. Sociedad Nuclear Espafiola, Alicante, Spain, October 1989.

44
10. ANNEX 1

45
ftfttrp BArETY

DIAGRAM 1. Asc6 NPP Nodalization Diagram

To K-F.W. PUMP[

65 7$6 ACCUMULATORS

LOOP 1
Diagram 2.Main and Auxiliary F.W. Nodalization Diagram.

$4.589 M

38.35m 43.00m

e. [883

-- -- -- 47.764 m

M.F.W.

$59

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

(2.-M) (Assigned bl NRC. Add Vol.. Supp., Rev., NRCM 1102, BB GAPIDTASETand Addendarm Numbers, If any.) 3201.3202 IU GAHCDTSET

(See htsfruction, on e reverse) NIJREGAIA-01 50

2. TITLE AND SUBTITLE

Study of Transients Related to AMSAC Actuation, 3. DATE REPORT PUBLISHED

Sensitivity Analysis O T ]

November 1998

4. FIN OR GR'.NT NUMBER

W6706

5. AUTHOR(S) 6. TYPE OF REPORT

J.M. Posada, P.M.S.A, F. Revetos, A.N.A.I J. Sanchez-Baptista, A.N.A., Technical

A. Perez-Navas, AN.NA, P. Moreno, P.M.S.A.

7. PERIOD COVERED (inclusive Dates)

8. PERFORMING ORGANIZATION - NAME AND ADDRESS (ifla provid Division. Office orRegion, U.S. NucleeRegulatcxy Commrission, and mailiag address; ifcontractor,

provide name and mailing address.)

P.M.S.A. Asociacion Nuclear ASCO (A.N.A.)

AV. Vinuelas, N 33, I D Paral.lel, 51

Tres Cantos -28760 MADRID 08004 BARCELONA

SPAIN SPAIN

9. SPONSORING ORGANIZATION - NAME AND ADDRESS (ifNRC type Same asaeboveif9contracla'. provide NRC Division, Office or Region, U.S. Nu.-JearRegulatofy Comnmission,

and mailing address)

Office of Nuclear Regulatory Research

Division of Systems Technology

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

10. SUPPLEMENTARY NOTES

J. Uhle. NRC Project Manager

11. ABSTRACT rmo words or Ass)

The Asociiacion Nuclear Asco (ANA) has prepared a model of Asco NPP using RELAP5IMOD3.2. This mode, which includes

thermal-hydraulics, kinetics and protection and control systems, has been qualified in previous calculations of several actual plant

transients. Asco NPP is a two unit station of three loop Pressurized Water Reactor (PWR) of Westinghouse design operated by ANA. ANA is a Spanish utility that contributes to the Code Application and Maintenance Project (CAMP) as a member of UNIDAD

ELECTRIC A S.A. (UNESA). This report summarizes the results obtained with Asco NPP model for a loss of normal feedwater

ATWS event and presents a sensivitity analysis to kinetic parameters for the same transient The phenomenology prediction has

been useful from the operation and safety point of view.

12. KEY WORDSIDESCRIPTORS (Lit wards crphresas that will assist reseaercher in Aating fth report.) 13. I.VAILABILrTY STATEMENT

RELAP5IMOD3.2 unlimited

PWR 14. S.ECURITY CLASSIFICATION

Asco (Th s Page)

ATWS unclassified

(Th s Rep*Vi

unclassified

15. IJUMBER OF PAGES

16. PRICE

NRC FORM 335 C2-89) This form was ek-cftnically pioduced by Elite Federal Forms. Inc.
Federal Recycling Program

NUREGIIA-0150 STUDY OF TRANSIENTS RELATED TO AMSAC ACTUATION, SENSITIVITY ANALYSIS NOVEMBER 1998

UNITED STATES

NUCLEAR REGULATORY COMMISSION

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