NUREG/IA-0174 

International 

Agreement Report 

Study of Unusual Occurrence 

of a Partial Core Uncovery 

in an SBLOCA Scenario 

Prepared by 

C. Pretel* 

E Revent6s** 

*Universidad Politecnica De Catalunya 

ETSEIB, Av. Diagonal, N' 647 

08028 Barcelona, SPAIN 

**Asociaci6n Nuclear Asc6 

Para.lel, 51 

08004 Barcelona, SPAIN 

Office of Nuclear Regulatory Research 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-0001 

February 2000 

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-0174

* International 

Agreement Report 

Study of Unusual Occurrence 

of a Partial Core Uncovery 

in an SBLOCA Scenario 

Prepared by 

C. Pretel* 

E Revent6s** 

*Universidad Politecnica De Catalunya 

ETSEIB, Av. Diagonal, N' 647 

08028 Barcelona, SPAIN 

**Asociaci6n Nuclear Asc6 

Para.lel, 51 

08004 Barcelona, SPAIN 

Office of Nuclear Regulatory Research 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-0001 

February 2000 

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

The Polytechnical University of Catalonia (UPC) has been cooperating since 

1992 with the Asociaci6n Nuclear Asc6 (ANA) improving its plant model. 

Asc6 NPP is a two unit, three loop Pressurized Water Reactor (PWR) 

plant of Westinghouse design operated by ANA. Its model, developed for 

RELAP5/MOD3.2, includes thermal-hydraulics, kinetics and protection and control 

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

transients. 

This report shows the results obtained with the Asc6 NPP model for a 5 inch 

Loss of Coolant Accident using RELAP5/MOD3.2. 

It also includes two different sensitivity studies performed to show the 

singularity of the phenomenology identified in the base case: the first study 

analyzes the influence of Main Coolant Pumps (MCPs) coast-down start; the 

second, the influence of the code itself (of its flow regime prediction).

i
Executive Summary

Asc6 Nuclear Power Plant is a nuclear station with two PWR of 930 MWe of 

Westinghouse design. 

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

has developed a model of the plant using RELAP5/MOD3.2, which has been 

recently improved thanks to the collaboration between ANA and the Polytechnical 

University of Catalonia (UPC). This model includes thermal-hydraulics, kinetics 

and protection and control systems. 

The aim of the study presented in this report is to fullfill UPC's commitment 

with the Code Assessment and Maintenance Program (CAMP). 

The transient selected to be analyzed for this purpose is a Loss of Coolant 

Accident at Asc6 NPP with RELAP5/MOD3.2. 

In the following pages the results obtained for a base case calculation and 

two sensitivity studies will be presented. Both studies help to prove the singularity 

of the phenomenology identified and described in the base case calculation: 

• A first sentitivity study is performed to determine the influence of the Main 

Coolant Pumps (MCPs) coast-down start. 

- A second analysis shows the influence of flow regime transitions on 

RELAP's simulation. 

The main conclusions are as follows: 

- this study presents an unusual phenomenon of SBLOCA scenarios with 

injection of all emergency systems (a partial core dryout) for a full size PWR. 

- the unusual occurrence of this phenomenon has been related to loop seal 

clearing and for this reason the clearing mechanism and its relation to flow 

regime change has been analyzed in more detail. 

- Influence of the code has also been determined. By slightly changing the 

conditions of flow regime transition, the partial dryout disappears. 

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

ii
Table of Contents 

1. Introduction ............................................................................... 1 

2. Plant and transient description ................................................ 3 

2.1. Plant description .............................................................................. 3 

2.2. Transient description ...................................................................... 5 

3. Model Description .................................................................... 6 

3.1. Thermal-hydraulic model ................................................................. 6 

3.2. Kinetic model .................................................................................. 8 

3.3. Control and protection systems model .......................................... 8 

4. Steady state calculation ........................................................ 11 

5. Transient calculation ............................................................... 12 

6. Singularity of the analyzed scenario ...................................... 17 

6.1. Sensitivity study to MCPs coast-down .......................................... 17 

6.2. Sensitivity study to RELAP version .............................................. 20 

7. Run Statistics ....................................................................... 24 

8. Conclusions ............................................................................. 25 

9. References ............................................................................ 26 

10. Annex 1 ................................................................................ 28 

Iil
List of Figures

Figure 1: Primary and secondary pressure ..................................................... 14 

Figure 2: Loop seal density ............................................................................. 14 

Figure 3: Collapsed liquid level in the vessel .................................................... 15 

Figure 4: Evolution of the clad temperature ...................................................... 15 

Figure 5: Vapor velocity in the core ................................................................... 16 

Figure 6: Liquid velocity in the core ................................................................... 16 

Figure 7: Evolution of the clad temperature. MCPs at 90s ................................ 18 

Figure 8: Evolution of the clad temperature. MCPs at 30s ............................... 18 

Figure 9: Collapsed liquid level. MCPs at 30 and 60s ...................................... 19 

Figure 10: Loop seal density. MCPs at 30s ...................................................... 19 

Figure 11: Fluid density at the break. Standard RELAP ................................... 21 

Figure 12: Void fraction in the loop seal. Standard RELAP ............................. 22 

Figure 13: Loop seal density. Modified RELAP ............................................... 22 

Figure 14: Evolution of the clad temperature. Modified RELAP ........................ 23 

Figure 15: Collapsed liquid level. Modified and standard RELAP ..................... 23

iv
List of Tables

Table 1: Main Characteristics of Asc6 I and II Nuclear Station ........................... 4 

Table 2: Sequence of events for a 5" Loss of Coolant Accident ......................... 5 

Table 3: Comparison Steady State between model and actual data plant ..... 11 

Table 4: R un Statistics ....................................................................................... 24

V
List of Diagrams

Diagram 1. Asc6 NPP Nodalization Diagram. (Annex 1) 

Diagram 2. Main and Auxiliary FW Nodalization Diagram. (Annex 1) 

List of Abreviations

ANA: 

CAMP: 

EOPs: 

FW: 

HPIS: 

MCPs: 

MOL: 

NPP: 

PRA: 

PRZ: 

PWR: 

SBLOCA: 

SG: 

UPC:

Asociaci6n Nuclear Asc6 

Code Application and Maintenance Program 

Emergency Operating Procedures 

Feed Water 

High Pressure Injection System 

Main Coolant Pumps 

Middle Of Life 

Nuclear Power Plant 

Probabilistic Risk Assessment 

Pressurizer 

Pressurized Water Reactor 

Small Break Loss Of Coolant Accident 

Steam Generator 

Universitat Politbcnica de Catalunya

vi
1. Introduction

Since 1992 the Polytechnical University of Catalonia (UPC) has been 

cooperating with the Asociaci6n Nuclear Asc6 (ANA) in different activities of plant 

and core thermal-hydraulic analysis. The main objective of the combined team 

has been to improve the capabilities of Asc6 NPP models. The most important 

cooperation studies performed are as follows: 

1,, Update and review of plant thermal-hydraulic models based on best

estimate criteria /1/. 

2. Support to ANA's participation in the International Code Assessment and 

Application Program /2/. 

3, Participation as part of the coordinated team UPC-ANA, in LOBI Data 

Users Group. During this period different analyses have been performed in 

two main directions: post-test analysis of LOBI experiments /3//4/ and 

calculations of equivalent scenarios in full size reactors /5//6/. 

4o Collaboration in the migration of RELAP5/MOD2 models to 

RELAP5/MOD3.2 as well as the qualification process of the final model /7//8/. 

5o Calculation of an extensive set of Small Break LOCA scenarios /9/ for 

Asc6 NPP taking into account: 

a) Break sizes. 

b) Emergency operating procedures /10/. 

c) Availability of emergency systems. 

d) Manual actuations/1 1/. 

6e Participation in the Code Assessment and Maintenance Program (CAMP). 

Once the adjustment and qualification process is considered to be completed 

following the above mentioned sets of calculations, 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.

1
This report refers to the results obtained with the Asc6 NPP model for a 5 inch 

Loss of Coolant Accident using RELAP5/MOD3.2. 

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

plant 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. 

Chapter 5 includes a 5" LOCA base case calculation using the standard 

RELAP5/MOD3.2 version. Two different sensitivity studies are presented in chapter 

6: section 6.1 shows a first study performed to determine the influence of the Main 

Coolant Pumps (MCPs) coast-down start (using the standard RELAP5/MOD3.2 

version). Section 6.2 presents the results obtained for a second sensitivity study, 

using a modified version of RELAP which includes slight changes in the prediction of 

flow regime transitions. 

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

9 while diagrams showing the nodalization used are in annex 1 .

2
2. Plant and transient description

2.1. Plant description 

Asc6 Nuclear Power Plant is a two unit station of three loop 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 II in its eleventh of normal operation. 

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

157 fuel assemblies of (17x17 -25) fuel rods. 

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

I and in October of 1996 for Unit II. 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.

3
Main Characteristics of Asc6 I and II Nuclear Stations

Thermal reactor power 2686 MWth 

Fuel UO2 

Number of assemblies 157 

Fuel rods per fuel assembly (17x17 - 25) = 264 

Active length of fuel rods 3.657 m 

Outside diameter of fuel rods 9.5e" m 

Cladding tube material Zr-4 

Cladding tube wall thickness 0.5715e"3 m 

Average linear heat generation rate 17750 W/m 

Absorber rods per control assembly 24 

Absorber material Ag - In - Cd 

Number of coolant loops 3 

Reactor operating pressure (PRZ) 15.51 MPa 

Coolant Average Temperature 581.3 K 

Coolant flow rate 14287 Kg/s 

Steam Generator 

Type 61W/D 3 

NumberI 

Reactor coolant pumps 

Type Westinghouse 93-DS 

Discharge head 86.25 m 

Design flow rate 5.928 m3/s 

Speed 155 rad/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 

Steam/Power 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 

-. . .. .. . . t i I I ~ t ,I tt_.. J!. • - ..--

4

Table 1: Main Characteristics ot Asco I ana 11 Nuclear Stations.
2.2. Transient description

The transient simulates a 5 inch break located at the cold leg of the first loop 

which causes a quick depressurization of the primary system and reactor trip. Due 

to the rapid evolution of the transient no significant intervention of the plant 

Emergency Operating Group has been simulated. 

Table 2 shows a list of the main events identified in the RELAP calculation:

Table 2: Sequence of events for a 5" Loss of Coolant Accident.

5

Event Time (seconds) 

Break 0 

HPIS starts to inject 7.5 

Saturation conditions 26 

MCPs trip 60 

Loss of natural circulation -170 

Loop seal clearing (loops 2 and 3) 260/240 

Break uncovery -275 

Primary pressure falls bellow secondary pressure 325 

Beginning of core dryout 460 

Accumulator injection 470 

Maximum Clad temperature (640 2K) 550 

Core rewetting 565
3. Model Description

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

includes the new SGs as well as the new controls. 

Diagrams 1 and 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 junctions, 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. Thermal-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 plenum. Loop 3 (volume 

410) is connected with the pressurizer through volume 510. The pressurizer is 

divided into 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 431 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

6
267, 367, 467 model the high pressure injection pumps by means of the pump 

characteristic curves. Accumulators are also modeled. 

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 810. 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 883 to 

the steam collector, volumes 900 and 904. Finally valves from 931 to 938, 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.

7
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 the 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 from plant design information /12/. 

3.2. Kinetic model 

The kinetic model /13/ 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 burn-up condition of each transient. 

3.3. Control and protection systems model 

The protection and control systems are modeled using RELAP5/MOD3.2 

control blocks and following specific setpoint studies, logical diagrams and technical 

specifications of the plant /13/, /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

8
* 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 

b) High pressure Injection System 

Using the following signals: 

* Low pressurizer pressure 

• High steam line differential 

eHigh steam line flow coincident with Low steam line pressure or Low-Low 

average teperature. 

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 run-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.

9
d) Feed Water Control System 

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

has 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 RELAP5/MOD3.2 

controls both the primary average temperature and the primary-secondary power 

mismatch. 

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. 

I) Permissive and interlock circuits. 

m) Low Pressure Injection System.

10
4. Steady state calculation

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

and kinetic parameters corresponding to MOL /16/ for unit 1. The objective is to 

obtain a stable condition to start transients. 

Table 3 compares the results obtained in the RELAP5 calculation for the main 

variables with the corresponding plant data.

Table 3: Comparison Steady State between model and actual data plant.

11

Steady State

VARIABLE MODEL(RELAP5/MOD3.2) PLANT 

Vessel AT (K) 33.3 33.7 

Primary Pressure (MPa) 15.517 15.516 

Primary Avg. Temperature (K) 581.5 581.5 

Recirculation ratio 3.64 3.65 

SG Narrow Range Level (%) 50.6 50.6 

Secondary Pressure (MPa) 6.82 6.733 

Steam Mass Flow (kg/s) 495.1 496.1
5. Transient calculation

In this section, the results of a base case calculation with ANA's 

RELAP5/MOD3.2 model will be presented. The calculation has been performed 

making following assumptions: 

- Initial normal full power operation in the middle of core life. 

* the scenario chosen corresponds to a 5" cold leg break 

• coast-down of the main coolant pumps starting 60s after scram 

• HPIS and accumulator injection 

• following the old Emergency Operating Procedures (EOPs).' 

Figure 1 shows the evolution of primary and secondary pressure during the 

transient and some of the most relevant events. 

Shortly after the break and due to the low upper plenum pressure, the reactor 

is scrammed automatically. A few seconds later injection of the HPIS starts. 

As saturation conditions are reached the main coolant pumps are tripped 

(about 1 minute after scram) and some minutes later loss of natural circulation 

takes place. 

Degradation of the heat transfer between primary and secondary system 

causes a sharp increase in the void fraction in hot legs. This situation ends with 

the break uncovery and two of three loop seal clearings. 

The quicker depressurization rate of the primary system contributes to a 

further degradation of the heat transfer and produces decoupling of both systems. 

Before the accumulator set point pressure is reached (about 470 s after the 

break) a partial core uncovery is observed. 

1 The old recovery procedures do not include any reference to the core level 

measurement system of Asc6.

12
Loop seal formation and clearing are closely related to this core uncovery. As 

shown in figure 2, loop seal clearing takes place in loops 2 and 3 at about 250s, 

whereas for the broken loop liquid remains trapped in it almost during the whole 

transient. 

Analyzing the collapsed liquid level in the vessel (figure 3), after a quick initial 

decrease a slight recovery is observed, which corresponds to both loop seal 

clearings. Later, the collapsed level depresses further reaching a value close to 

the 30%. 

The accumulator injection allows a new increase of the level and rewetting of 

the core. 

Heater rod temperature in the last two of six nodes increases reaching a 

value of about 605 2K and 644 2K, respectively, and decreases again thanks to the 

accumulator injection (figure 4). 

In some small break LOCA scenarios with partial core uncovery, this 

phenomenon of partial uncovery is not accompanied by cladding temperature 

excursions. In those cases dryout is mitigated by liquid of the vessel's bottom 

being dragged by the vapor generated. 

In the present scenario this drag mechanism is not enough to avoid the 

temperature increase, as observed in figures 5 and 6. Figure 5 shows the vapor 

velocity in the core. Although there is an increase after the break uncovery, vapor 

velocity does not reach a sufficiently high value in order to have a significant drag. 

The same conclusion can be drawn observing figure 6 in which liquid velocity in 

the inlet and outlet junctions of the fifth node are represented.

13
5" LOCA 

MOL, 60s

200 400 

Time (s)

600

I-- primary - secondary I

Figure 1: Primary and secondary pressure. 

5" LOCA 

MOL, 60s

1000 

E 

- 800 

"• 600 

-• 400 

o 200 

0 

0

0 200 400 600 800 

Time (s)

I- loop 1 ...... loop 2 - • loop 3] 

Figure 2: Loop seal density.

14

16 

14

@'12

210

81 

68

0. 4. 

2

0

0

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

.. . . . . .. 

.. 

. .

. .

. .

. .

.

. .

.. 

.

.

.. 

. .

.

.

.

.

.

.. 

. .. .. . 

800
5" LOCA 

MOL, 60s

1~ 

70.8 

0.6 

>0.4

Cr O.2 

0 -

0 200 400 

Time (s)

600

Figure 3: Collapsed liquid level in the vessel.

5" LOCA 

MOL, 60s

650 

600 

0) 

_L550 E 

c 500 

450

0 200 400 600 800

Time (s) 

-node 6-- node 5 -- node 4 

Figure 4: Evolution of the clad temperature

15

800

1000

| | | | | | I I |

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

---------------------------------- --- -----

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

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

I

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

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

..............................
5" LOCA 

MOL, 60s

200 400 

Time (s)

600

r- node 5 inlet - node 5 outle

Figure 5: Vapor velocity in the core.

5" LOCA 

MOL, 60s

200 400 600 800 

Time (s) 

- node 5 inlet - node 5 outlet

Figure 6: Liquid velocity in the core.

16

2 

0 o1.2

0 

0.4 

0 0*

0 800

2 

0 

> 0 .5( 

0 

-2

0

S| I i

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

------ ----------------- 

J ------------- -----

. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
6. Singularity of the analyzed scenario

6.1. Sensitivity study to MCPs coast-down 

The singularity of this particular scenario is due to the liquid retained in the 

loop seal of the broken loop, which affects considerably the later evolution of the 

transient, specifically in the temperature excursion. 

A proof of the singular and unusual of this transient is the fact that the same 

scenario with a different time of MCPs coast-down start (earlier as well as later) 

does not show any temperature excursion 

Figure 7 and 8 correspond to a 5" LOCA scenario in which pumps are tripped 

90 and 30 seconds after scram, respectively. Both show a continuosly decreasing 

evolution of the heater rod temperature all through the transient. 

The vessel's collapsed liquid level (figure 9) decreases in a similar way as in 

the previous scenario, but now core uncovery is compensated by the accumulator 

injection prior to the temperature increase. 

The cause of this different behaviour is again the loop seal clearing (figure 

10). In this new transient loop seal clearing occurs for all three loops, in such a 

way that liquid retained previously in the broken loop is able now to reach the core, 

delaying its uncovery and allowing accumulators injection before the temperature 

increase. 

Even now drag of liquid is not a significant phenomenon.

17
5" LOCA 

MOL, 90s

700 

S600

CL 

E 

a 500

40 

400-

400 

Time (s)

600

-- node 6-- • node 5 - node 4 1 

Figure 7: Evolution of the clad temperature. MCPs at 90s. 

5" LOCA 

MOL, 30s

620

2£600 

S580 

-560 

a, 

E 540 

"0520 

0500 

480

0 100 200 300 

Time (s)

400 500

-node 4 ....... node 5 - - • node 6j 

Figure 8: Evolution of the clad temperature. MCPs at 30s.

18

0 200

I I I

& i I

600

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

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

-----* --------- ------ ------ --- ----- -------- -------------------------------------------

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

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

..... .. ........ ... ..... ... .... .. .. .. I -- -- -- --

.......................................
5" LOCA 

MOL, 30s

50.8 

S0.6 

6 0.4 

ir 0.2 

0 

0 200

Time (s)

400

I- MCPs 30s - MCPs 60s I 

Figure 9: Collapsed liquid level. MCPs at 30 and 60s.

5" LOCA 

MOL, 30s

1000 

E 

• 800 

z 600 

ca 400 

co 

o 200 

0 

0 

0

100 200 300 

Time (s)

400 500

I,-loop 1 - loop 2 -- -'loop3 

Figure 10: Loop seal density. MCPs at 30s.

19

600

600

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

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

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

KI - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ----

I I I ! ' I I I

m

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

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

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

-- -- -- ---- -- ------ - ------- ---
6.2. Sensitivity study to RELAP version 

As mentioned before, the singularity of the base case calculation is strongly 

related to the loop seal clearing. A deeper analysis of this phenomenon proves its 

relation not only to the break uncovery, but also to the pressure generated in the 

hot legs (delta-P in the loop seals) and even in some way to the code itself. 

While flow through the break is basically liquid, vapor accumulates in the hot 

legs producing a pressure increase that favours loop seal clearing. However, if 

break uncovery occurs before the clearing, this reasoning results more complex. 

Figure 11 presents fluid density at the break and shows that break uncovery 

does not occur instantaneously: after a quick initial decrease, density continues 

decreasing slowly proving the presence of a two-phase mixture in the cold leg of 

this loop. In such a situation pressure does not favour loop seal clearing clearly 

enough to expain it from a strictly mechanical point of view. This may be, to some 

extent, the reason of the arbitrariness of this phenomenon. 

In this specific scenario the evolution of the loop seal void fraction (figure 12) 

shows some situations of initiation of loop seal clearing that do not progress. In 

order to progress, and here is where the code can influence, it is not only 

necessary to have appropriate pressure conditions, but also the suitable flow 

regime. 

In low flow conditions, as it is the case in this transient, RELAP sets a void 

fraction of 25% for the transition from bubbly to slug regime in the considered 

vertical volume. This minimum 25% necessary to change into the flow regime that 

will enable loop seal clearing is reached in loops 2 and 3, but not in the first (the 

broken loop), where the maximum void fractions in this initial period are about 

20%. The code influence seems thus to be determinant in the transient's evolution. 

According to what is mentioned in the code's manual /17/, choosing this 

particular transition value is somewhat of arbitrary although it lies in the range of 

values suggested by different studies (from 18% up to 30%). 

All this suggests the possibility of changing this transition value in the 

corresponding RELAP routine, to check its effect in the transient's evolution. 

In the study presented next the void fraction value for the transition from

20
bubbly to slug flow regime has been set to 18%.

Using this modified recompiled version of RELAP a new calculation has been 

performed simulating again a 5" LOCA with MCPs tripped 60 s after scram. 

The reduction of the transition value has an impact in the timing of all three 

loop seal clearings (see figure 13). In this case, even the third loop seal clears just 

short before the accumulator injection. 

As a result of this, the vessel's collapsed level decreases less than in the 

base case calculation, so that no temperature excursion is observed (figures 14 

and 15). 

All these results help to prove the suitability of the new EOPs of Asc6: 

following the new procedures, the operator would have started cooling at a 

maximum rate when detecting a vessel's collapsed level bellow 39%, avoiding 

then the temperature increase. 

5" LOCA 

MOL, 60s 

1000 

S800 --- --------------* . . . . ... . . . . . . . . .. . . . . . . . . . . -E 

>, 600 -- - - - -- . . . . .. . . . . .. . . . . 

Cz 

(400 

a., 

0 100 200 300 400 500 600 

Time (s) 

Figure 11: Fluid density at the break. Standard RELAP.

21
5" LOCA 

MOL, 60s

1

0 

O0.8 

-0.6

0 

(D0.4 

C

0.2 

0.

400

I-loop 1 -loop 2 -- -loop3

Figure 12: Void fraction in the loop seal. Standard RELAP. 

5" LOCA 

MOL, 60s, modified RELAP

1000 

E 

C 800 

S600 

-• 400 

U,o 

•" 200 

0 

-J 

0

0 200 400 

Time (s) 

I-loop 1 ....... loop2.-loop3

600

Figure 13: Loop seal density. Modified RELAP.

22

200 

Time (s)

0

. . ,-.-.. .. . . . . . . . . . . .. ... 

- - - - -°

:: "-1

--------------------

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

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

.....................
5" LOCA 

MOL, 60s, modified RELAP

620 

7'600 

1;580 

560 

E 540 

"-0520 

0500 

480

0 200

Time (s)

400 600

I- node 4-- node 5 - -- node 6 

Figure 14: Evolution of the clad temperature. Modified RELAP. 

5" LOCA 

MOL, 60s

1 

> 0.8 

I 0.6 

0) S0.4 

r0.2 

0

0 200 400 

Time (s)

600 800

I--modified RELAP - standard RELAP I

Figure 15: Collapsed liquid level. Modified and standard RELAP.

23

I I I I

4

-- ------------------------------------------

--------------------------------------- ---- 

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

-------------------------------------------

-------------------------------------------

---------- ---- -- --- ----------- ------ 

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

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

.........................................
7. Run Statistics

Calculations were carried out on a DecAipha with UNIX OSF/1 AXP. 

RELAP5/MOD3.2 was used in all the calculations, with minor changes in the 

second sensitivity study.

TRANSIENT TIME (s) 900 

CPUTIME (s) 3272,1 

Total number of time steps 19836 

CPUTIMEiTRANSIENT TIME 3,64 

Sensitivity Study 1 

MOL, MCPs at 30s, stardard RELAP version 

TRANSIENT TIME (s) 700 

CPUTIME (s) 2749,2 

Total number of time steps 15921 

CPUTIME/TRANSIENT TIME 3,93 

Sensitivity Study 1 

MOL, MCPs at 90s, stardard RELAP version 

TRANSIENT TIME (s) 700 

CPUTIME (s) 2577,5 

Total number of time steps 15902 

CPUTIME/TRANSIENT TIME 3,68 

Sensitivity Study 2 

MOL, MCPs at 60s, modified RELAP version 

TRANSIENT TIME (s) 700 

CPUTIME (s) 2866,0 

Total number of time steps 16528 

CPUTIME/TRANSIENT TIME 4,09

Table 4: Run Statistics.

24

Base case 

MOL, MCPs at 60s, standard RELAP version
8. Conclusions

This study presents an unusual phenomenon (a partial core dryout) of 

SBLOCA scenarios with injection of all emergency systems for a full size PWR: 

Asc6. 

. The unusual occurrence of this phenomenon has been related to loop seal 

clearing and for this reason the clearing mechanism and its relation to flow 

regime change has been analyzed in more detail. 

. Influence of the code has also been determined. By slightly changing the 

transition value a new phenomenology is predicted and the partial dryout 

disappears. 

- The results obtained in the study of this transient prove the suitability of the 

new EOPs of Asc6. 

• The study of this transient helps to improve the knowledge of plant dynamics. 

* The model of Asc6 using RELAP5/MOD3.2 is a valuable tool to analyze plant 

transients and to provide engineering support to plant operation.

25
9. References

/11 F. PRELLEZO, F. REVENTOS, C. TAPIA, F. CALVINO. "Actualizaci6n de la 

matriz de validaci6n del Modelo Termohidrdulico de la C.N. Asc6 para el apoyo y 

seguridad de la explotaci6n". July 1992. 

/2/ F. REVENTOS. "Experiencia adquirida como resultado del proyecto ICAP

Esparha en el c6digo RELAP5/MOD2". ICAP-SPAIN 2 ond Technical Meeting. 

November 19-21, 1991. 

/3/ C. PRETEL, F. REVENTOS. " Post-test Analysis of LOBI-MOD2 Test BL-30 

using RELAP5/MOD2.5". CEC-JRC. Technical Note 1.93.64. May 1993. 

/4/ C. PRETEL, F. REVENTOS. " Post-test Analysis of LOBI-MOD2 Test BL-34 

using RELAP5/MOD2.5". CEC-JRC. Technical Note 1.94.139. September 1994 

/5/ F. REVENTOS, C. PRETEL. "Considerations on Core Dryout Phenomena in a 

Full Size PWR ". LOBI Data Users Club Meeting. KFKI Atominstitut (Budapest), 

October 1994. 

/6/ F. REVENTOS, C. PRETEL. " Explotaci6n de Datos de Experimentos 

Termohidr.ulicos. Experiencias en el perfodo 92-95". XXI Spanish Nuclear Society 

Annual Meeting. Reus (Tarragona), October 1995. 

/7/ C. PRETEL, F. REVENTOS. " Small Break LOCA Analysis in a Full Size PWR 

using RELAP5/MOD3.2 and RELAP5/MOD2.5". CAMP Meeting. Madrid, May 6-8 

1996. 

/8/ F. REVENTOS. "Start-up Test of Asc6 NPP with new Steam Generators using 

RELAP5/MOD3.2". CAMP Meeting. Madrid, May 6-8, 1996. 

/9/ C. PRETEL. "Analysis of LOCA in Pressurized Water Reactors and Accident 

Management Optimization". PhD. July 1997. 

/10/ C. PRETEL, F. REVENTOS. "Impact of the Initiation of Secondary System 

Cooling Actions in a SBLOCA Scenario for the Asc6 NPP". RELAP5 International 

User's Seminar. College Station. May 17-22,1998. 

/11/ C. PRETEL, F. REVENTOS. " Prediction of Unusual Occurrence of Core 

Dryout Phenomena in SBLOCA Scenarios of the Asc6 NPP". 6 Th International 

Conference on Nuclear Engineering. May 10-15, 1998.

26
/12/ 'Final Safety Analysis Report-Asc6 I', June 1983, 'Final Safety Analysis 

Report-Asc6 !1', September 1985, and later revisions, Asociaci6n Nuclear Asc6. 

/13/ F. REVENTOS, J. SANCHEZ-BAPTISTA, J.M. POSADA and P. MORENO, 

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

Nuclear Asc6, March 1998. 

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

Corporation (March 1976). 

/15/ 'Precautions, Limitations and Setpoints of Asc6 Unit I', Revision N2 14, 

February 1996. 

/16/'Informe de Diseho Nuclear de ASCO I ciclo 12' ITEC-484 revision 1 ENUSA. 

Octubre 1995. 

/17/ RELAP5/MOD3.2 CODE MANUAL, EGG-2596, June 1995.

27
10. Annex 1

- Nodalization Diagrams for Asc6: 

(1) Plant Diagram 

(2) Main Feedwater and Auxiliary Feedwater Systems

28
VAI NODALIZATION DIAGRAM FOR ASCO (1) 

RELIE 098006 SFETYVALV VALVALVE 

2E1 

725 72077267826 0082 

MFW 722 ~~ 2 MFW 87pump2238

LOOP 1 LOOP 2 LOOP 3
NOD.ALIZATION OF MFW AND AFW LINES. DIAGRAM 2 

(i 682 So986 

38.368 78 430074664i 

MFW$7 T-00BE

TUBES
NRC FORM 336 U.S. NUCLEAR REGULATORY COMMISSION 1. REPORT NUMBER 

(2-89) (Assigned by NRC, Add Vo., Supp., Rev., NRCM 1102. BIBUOGRPHIC DATA SHEET and Addendum Numbers, If any.) 

3201,3202 IUGAHCDTSET 

(See rnsfrzr-azos on me• wvve) 

2. TITLE AND SUBTITLE NUREG/IA-0174 

Study of Unusual Occurrence of a Partial Core Uncovery 

in a SBLOCA Scenario 3. DATE REPORT PUBLISHED 

MONTH YEAR 

February 2000 

4. FIN OR GRANT NUMBER 

5. AUTHOR(S) 6. TYPE OF REPORT 

C. Pretel*, F. Reventos** Technical 

7. PERIOD COVERED (nciusive Deo) 

S. PERFORMING ORGANIZATION - NAME AND ADDRESS (if NRC, provide Division, Ofce orRegion. U.S. NuclwRegulatcry Commission, and maitng address; if conbactov 

provide nano and mailing addnss) 

*Universidad Politecnica De Catalunya **Asociacion Nuclear Asco 

ETSEIB, Av. Diagonal, No 647 Para.lel, 51 

08028 Barcelona, SPAIN 08004 Barcelona, SPAIN

9. SPONSORING ORGANIZATION - NAME AND ADDRESS (f f RCpe 'Same as above if ontractr, provide NRC Division, Offie orRegion, U.S. Nuclear Regulatory Commission, and mailing addrass.)

Division of System Analysis and Regulatory Effectiveness 

Office of Nuclear Regulatory Research 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-0001

1U. SUPP"LEMENT IAR.Y NU O I 

11. ABSTRACT (2400 wods orins) 

The Polytechnical University of Catalonia (UPC) has been cooperating since 1992 with the Asociacion Nuclear Asco (ANA) improving its plant model. Asco NPP is a two unit, three loop Pressurized Water Reactor (PWR) plant of Westinghouse design operated by ANA. Its model, developed for RELAP5/MOD3.2, includes thermal-hydraulics, Idnetics and protection and control 

systems and has been qualified in previous calculations of several actual plant transients. This report shows the results obtained 

with the Asco NPP model for a 5 inch Loss of Coolant Accident using RELAP5/MOD3.2. It also includes two different sensitivity studies performed to show the singularity of the phenomenology identified in the base case: the first study analyzes the influence 

of Main Coolant Pumps (MCPs) coast-down start; the second, the influence of the code itself (of its flow regime prediction).

12. KEY WORDS/DESCRIPTORS (List i•ads or phrases met will assisteusearrl in locating herepiot.) 13. AVAILABILITY STATEMENT 

SBLOCA unlimited 

RELAP5 14. SECURJTY CLASSIFICATION 

(his Page) 

unclassified 

(This Repot) 

unclassified 

15. NUMBER OF PAGES 

16. PRICE

NRC FORM 335 (2489)
Federal Recycling Program
NUREG/IA-0174

UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. 20555-0001

FEBRUARY 2000

SPECIAL STANDARD MAIL 

POSTAGE AND FEES PAID 

USNRC 

PERMIT NO. G-67

STUDY OF UNUSUAL OCCURRENCE OF A 

PARTIAL CORE UNCOVERY IN AN SBLOCA SCENARIO