NUREG/IA-0475

TRACE/RELAP5 Comparative Calculations For

Double-Ended LBLOCA and SBO

Prepared by:

S. Iegan, A. Mazur,

Y. Vorobyov, O. Zhabin, S. Yanovskiy

State Nuclear Regulatory Inspectorate of Ukraine

And State Scientific and Technical Center

for Nuclear and Radiation Safety of Ukraine

9/11 Arsenalna str. Kyiv, Ukraine 01011

K. Tien, NRC Project Manager

Division of Systems Analysis

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

Manuscript Completed: February 2016

Date Published: November 2016

Prepared as part of 

The Agreement on Research Participation and Technical Exchange

Under the Thermal-Hydraulic Code Applications and Maintenance Program (CAMP)

Published by

U.S. Nuclear Regulatory Commission

NUREG/IA-0475

TRACE/RELAP5 Comparative Calculations for 

Double-Ended LBLOCA and SBO

Prepared by:

S. Iegan, A. Mazur,

Y. Vorobyov, O. Zhabin, S. Yanovskiy

State Nuclear Regulatory Inspectorate of Ukraine

And State Scientific and Technical Center

for Nuclear and Radiation Safety of Ukraine

9/11 Arsenalna str. Kyiv, Ukraine 01011

K. Tien, NRC Project Manager

Division of Systems Analysis

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

Manuscript Completed: February 2016

Date Published: November 2016

Prepared as part of 

The Agreement on Research Participation and Technical Exchange

Under the Thermal-Hydraulic Code Applications and Maintenance Program (CAMP)

Published by

U.S. Nuclear Regulatory Commission

iii

ABSTRACT

This report is developed by the State Nuclear Regulatory Inspectorate of Ukraine (SNRIU) and its 

technical support organization, the State Scientific and Technical Center for Nuclear and Radiation 

Safety of Ukraine (SSTC NRS), under Implementing Agreement On Thermal-Hydraulic Code 

Applications And Maintenance Between The United States Nuclear Regulatory Commission and 

State Nuclear Regulatory Inspectorate of Ukraine (signed in 2014) in accordance with Article III, 

Section C, of the Agreement.

The report provides description of quantitative analysis results of two initiating events for 

VVER-1000 reactor system with application of TRACE p4 and RELAP5/MOD3.2 codes. For these 

calculations RELAP5 input model for VVER-1000 was converted to TRACE code. The results 

obtained with both codes were compared to evaluate differences and to identify future needs for 

TRACE input deck enhancement.

v 

CONTENTS

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

LIST OF FIGURES ................................................................................................................................vii

LIST OF TABLES................................................................................................................................. viii

ABBREVIATIONS AND ACRONYMS...................................................................................................ix

INTRODUCTION ............................................................................................................................. 1-1

 BRIEF DESCRIPTION OF TRACE MODEL FOR VVER-1000 .................................................... 2-1

2.1 Reactor................................................................................................................................... 2-1

 2.2 Main Coolant Piping ............................................................................................................... 2-3

 2.3 Pressurizer System ................................................................................................................ 2-5

 2.4 Steam Generator.................................................................................................................... 2-6

 2.5 Make-Up and Let-Down......................................................................................................... 2-8

 2.6 Emergency Gas Removal System......................................................................................... 2-8

 2.7 Main Steam Lines................................................................................................................... 2-9

 2.8 Main Steam Header ............................................................................................................ 2-12

 2.9 Main, Auxiliary and Emergency Feedwater Systems......................................................... 2-13

 2.10 Emergency Core Flooding System..................................................................................... 2-14

 2.11 High Pressure Injection System.......................................................................................... 2-15

 2.12 Low Pressure Injection System........................................................................................... 2-15

 2.13 Safeguards and Control Systems Operation Logic ............................................................ 2-16

 TRACE MODEL STEADY STATE CALCULATIONS................................................................... 3-1

 ANALYSIS RESULTS .................................................................................................................... 4-1

4.1 Station Blackout...................................................................................................................... 4-1

 4.2 Double-Ended LB LOCA........................................................................................................ 4-7

 DISCUSSION................................................................................................................................... 5-1

5.1 SBO ........................................................................................................................................ 5-1

 5.2 LB LOCA................................................................................................................................. 5-1

 5.3 Future Plans............................................................................................................................ 5-1

 CONCLUSIONS.............................................................................................................................. 6-1

 REFERENCES ................................................................................................................................ 7-1

vii

LIST OF FIGURES

Figure 1 Nodalization Diagram of VVER-1000 (V-320).................................................................2-2

Figure 2 Nodalization Diagram of RCS Loop 1 .............................................................................2-3

Figure 3 Nodalization Diagram of RCS Loop 2 .............................................................................2-4

Figure 4 Nodalization Diagram of RCS Loop 3 .............................................................................2-4

Figure 5 Nodalization Diagram of RCS Loop 4, Connecting Pipe and Pressurizer......................2-5

Figure 6 Nodalization Diagram of the Pressurizer Spray Line ......................................................2-6

Figure 7 Nodalization Diagram of SG Primary Side ......................................................................2-6

Figure 8 Nodalization Diagram of SG Secondary Side .................................................................2-7

Figure 9 Nodalization Diagram of Makeup and Let-down .............................................................2-8

Figure 10 Nodalization Diagram of Emergency Gas Removal System..........................................2-9

Figure 11 Nodalization Diagram of Steam Line No. 1, BRU-А and SG SRV.............................. 2-10

Figure 12 Nodalization Diagram of Steam Line No. 2, BRU-А and SG SRV.............................. 2-10

Figure 13 Nodalization Diagram of Steam Line No. 3, BRU-А and SG SRV.............................. 2-11

Figure 14 Nodalization Diagram of Steam Line No. 4, BRU-А and SG SRV.............................. 2-11

Figure 15 Nodalization Diagram of MSH and BRU-K .................................................................. 2-12

Figure 16 Nodalization Diagram of the Main and Auxiliary Feedwater Systems......................... 2-13

Figure 17 Nodalization Diagram of the Emergency Feedwater System...................................... 2-13

Figure 18 Nodalization Diagram of the System of ECFS and LPIS Trains.................................. 2-14

Figure 19 Nodalization Diagram of HPIS Trains .......................................................................... 2-15

Figure 20 Logical Diagram of HPIS Operation............................................................................. 2-17

Figure 21 Logical Diagram of LPIS Operation.............................................................................. 2-18

Figure 22 Logical Diagram of Control and Protection System (Part 1......................................... 2-19

Figure 23 Logical Diagram of Control and Protection System (Part 2)........................................ 2-20

Figure 24 Logical Diagram of Control and Protection System (Part 3)........................................ 2-21

Figure 25 Logical Diagram of Control and Protection System (Part 4)........................................ 2-22

Figure 26 Logical Diagram of PRZ Level Control by Makeup and Let-down............................... 2-23

Figure 27 Logical Diagram of Make-up Pressure Difference Controller ...................................... 2-24

Figure 28 Logical Diagram of MFW Flow Rate Controller (Part 1) .............................................. 2-25

Figure 29 Logical Diagram of MFW Flow Rate Controller (Part 2) .............................................. 2-26

Figure 30 Logical Diagram of SG Level Controller....................................................................... 2-27

Figure 31 Logical Diagram of BRU-A Controller .......................................................................... 2-28

Figure 32 Logical Diagram of Turbine Control System ................................................................ 2-29

Figure 33 Reactor Power.................................................................................................................4-3

Figure 34 Primary Pressure.............................................................................................................4-3

Figure 35 SG Pressure ....................................................................................................................4-4

Figure 36 Pressurizer Coolant Level ...............................................................................................4-4

Figure 37 Steam Generators Water Level.......................................................................................4-5

Figure 38 Cold Legs Temperature...................................................................................................4-5

Figure 39 Hot Legs Temperature.....................................................................................................4-6

Figure 40 Max Cladding Temperature.............................................................................................4-6

Figure 41 Reactor Power (Fragment)..............................................................................................4-9

Figure 42 Hot Channel Collapsed Level (Fragment).......................................................................4-9

Figure 43 Hot Channel Mass Flow (Fragment)............................................................................ 4-10

Figure 44 Hot Channel Max Cladding Temperature .................................................................... 4-10

Figure 45 Hot Channel Max Fuel Temperature............................................................................ 4-11

Figure 46 RCS Pressure............................................................................................................... 4-11

Figure 47 HPIS Flow..................................................................................................................... 4-12

Figure 48 LPIS Flow...................................................................................................................... 4-12

Figure 49 HA Water Volume......................................................................................................... 4-13
viii

Figure 50 Maximal Cladding Temperature................................................................................... 4-13

Figure 51 Hot Channel Collapsed Level....................................................................................... 4-14

LIST OF TABLES

Table 1 Steady State Calculation Results .......................................................................................3-1

Table 2 Initial Conditions for SBO....................................................................................................4-1

Table 3 Timing of Main Events for SBO Calculation .......................................................................4-2

Table 4 Initial Conditions for LBLOCA Calculation..........................................................................4-7

Table 5 Timing of Main Events of LBLOCA Calculation..................................................................4-8
ix

ABBREVIATIONS AND ACRONYMS

BRU-A Steam Dump Valves to the Atmosphere

BRU-K Steam Dump Valves to the Condenser

CAMP Code Applications and Maintenance

ECFS Emergency Core Flooding System

FA Fuel Assembly

HA Hydroaccumulator

HC Hydraulic Component

HPIS High Pressure Injection System

HS Heat Structure

LB LOCA Large Break Loss-of-Coolant Accident

LPIS Low Pressure Injection System

MCP Main Coolant Piping

MSH Main Steam Header

MFW Main Feedwater

PORVs Pilot Operated Relief Valve

PRZ Pressurizer

RCP Reactor Coolant Pumps

RCS Reactor Coolant System

SBO Station Blackout

SG Steam Generator

SNRIU State Nuclear Regulatory Inspectorate of Ukraine

SSTC NRS State Scientific and Technical Center for Nuclear and Radiation Safety

SRV Steam Relief Valve

USNRC United States Nuclear Regulatory Commission

VVER Pressurized Water Reactor, Russian design

1-1 

INTRODUCTION

At the end of 2014 the United States Nuclear Regulatory Commission (USNRC) and the State 

Nuclear Regulatory Inspectorate of Ukraine (SNRIU) signed Implementing Agreement On 

Thermal-Hydraulic Code Applications And Maintenance (CAMP). In accordance with Article III, 

Section C, of the Agreement, SNRIU shall submit to the USNRC the in-kind contribution reports 

providing the code assessment results or other activities results of equivalent value.

In the framework of the Agreement SNRIU and SSTC NRS obtained the state-of the-art TRACE 

code which provides advanced capabilities for modeling thermal-hydraulic processes and 

components, control systems and allows coupling with PARCS neutron kinetics code. In 2015 

SSTC NRS initiated activities on TRACE code application for evaluation of safety assessment 

results developed for Ukrainian NPPs. The existing SNRIU/SSTC NRS RELAP5 model for 

VVER-1000 was converted to TRACE code using Symbolic Nuclear Analysis Package (SNAP). 

The control systems operation logic was converted to TRACE model manually. RELAP and 

TRACE model descriptions are provided in [1, 2]. Brief description of TRACE model is provided in 

Section 2 to this report. Comparison of steady state results with nominal valves is provided in 

Section 3 . 

After preparation of TRACE model and steady-state calculation the transient calculations were 

performed with application of RELAP5 and TRACE codes. The main objective of these calculations 

is initial evaluation of TRACE model developed in order to identify further steps for its improvement 

and qualification.

The initiating events analyzed are:

• Total loss of off-site and on-site power supply (total station blackout, SBO);

• Large break LOCA (850mm double-ended cold leg break) scenario with simultaneous loss of 

power.

2-1 

BRIEF DESCRIPTION OF TRACE MODEL FOR VVER-1000

The TRACE model was developed by conversion of existing RELAP5 model for VVER-1000. The 

conversion was performed in two stages, namely: 

• conversion of thermal hydraulic part of the model (hydraulic components and heat structures);

• conversion of logical part (control systems).

Conversion was performed with application of Symbolic Nuclear Analysis Package. Hydraulic 

elements obtained resulting from conversion almost do not require further manual correction and 

processing. Heat structures, which were connected to appropriate hydraulic components in RELAP 

model, and thermal and physical properties of constructional materials do not require correction

as well. 

However automatic conversion of control systems logic results in a number of errors and 

inconsistencies. Therefore, the safeguards and control systems logic were manually converted 

from RELAP deck to TRACE code.

Description of the main hydraulic components of TRACE model for VVER-1000 is provided below.

2.1 Reactor

Reactor model is 4-sectoral and has cross-links to simulate crossflows between the sectors. The 

area of inlet and outlet nozzles is divided into 8 equal parts modeling annular gaps between the 

core barrel and the reactor pressure vessel. This allows proper flow distribution in scenarios with 

partial number of reactor coolant pumps (RCP) in operation. 

The core region is divided into 4 sectors (individual sector for each of 4 reactor coolant system 

loops)and modeled by hydraulic components (HC) 5029, 5032, 5030, 5031, 5033, 5034, 5035, 

5036. Radial division of core region is not envisaged. Each sector has two channels: for "average" 

fuel assemblies (FA) and for "hot" fuel assembly.

Core bypasses are simplistically joined and presented by HC 5259, 5352 and 5271.

HC 5253, 5254, 5250, 5251, 5255, 5256, 5252 model downcomer parts at the cold legs nozzles 

region. HC 5245, 5246, 5002, 5003, 5247, 5248, 5004, 5001 model downcomer part below the 

RCS nozzles region. HC 5009, 5010, 5011, 5012 model a gap between reactor bottom and core 

barrel. 

Lower plenum is modeled by HC 5020, 5017, 5019, 5018, 5016, 5015, 5014, 5013, 5021 5022 

5023, 5024, 5025, 5026, 5027, 5028.

Upper plenum is modeled by a range of HC 5037, 5038, 5039, 5040, 5081, 5242, 5243, 5244, 

5260, 5257.
2-2 

Upper head is modeled by HC 5258. Reactor nodalization diagram is shown on figure below.

Figure 1 Nodalization Diagram of VVER-1000 (V-320)
2-3 

HS in the core are modeled by “Fuel Rod” type and have cylindrical geometry reflecting FE 

geometry with the height of 354 cm. Eight heat structures were modeled in the core, and according 

to core division, each heat structure has an appropriate surface multiplier.

HS of the reactor pressure vessel models the heat transfer from the coolant through vessel wall to 

the environment. 

HS of reactor internals such as core barrel and baffle model the heat exchange between 

components inside the reactor.

2.2 Main Coolant Piping

Main coolant piping (MCP) consists of four loops. Each loop has a cold and a hot leg. Nodalization 

diagrams of loops 1-4 are given on Figures 2 – 5.

Figure 2 Nodalization Diagram of RCS Loop 1
2-4 

Figure 3 Nodalization Diagram of RCS Loop 2

Figure 4 Nodalization Diagram of RCS Loop 3
2-5 

Figure 5 Nodalization Diagram of RCS Loop 4, Connecting Pipe and Pressurizer

2.3 Pressurizer System

Pressurizer (PRZ) model is implemented by PRIZER HC 50001. Steam discharge piping (PIPE 

HC 50002 and 50003) is connected to the pressurizer top part. HC 605, 607 (VALVE) model 

pressurizer pilot operated relief valves (PORVs) (control and the main PORVs, respectively).

Pressurizer surge line is connected to RCS hot leg 4.

Pressurizer steam discharge includes three PORVs YP21,22,23S10. Valve operation algorithm is 

implemented according to opening/closure setpoints of the main and control PORVs. The setpoints 

are changed at station blackout after depletion of batteries to simulate spring-controlled PORV 

operation. 

Nodalization diagram of the pressurizer system is provided in Figure 5 and Figure 6. 
2-6 

Figure 6 Nodalization Diagram of the Pressurizer Spray Line

2.4 Steam Generator

Nodalization diagram of steam generator (SG) primary side is given on Figure 7. SG primary side 

consists of two SG manifolds (cold and hot) and a tube bundle.

Figure 7 Nodalization Diagram of SG Primary Side

SG heat transfer surface consists of manifolds, 11000 tubes and tube support structures. Heat 

exchange U-shaped tubes are assembled into tube bundle and staggered with vertical spacing of 

19 mm and horizontal spacing of 23 mm. Tubes are assembled in two U-shaped bundles, each 

bundle has three vertical corridors ensuring controlled hydrodynamics of circulating water.
2-7 

Tube bundle model is implemented by PIPE hydraulic components. Resulting from the division 

diagram of SG secondary side, five layers of tube bundle are modeled; each of them has two Ushaped parts.

SG secondary side is modeled using quasi-3D approach. Such approximation was chosen for 

correct distribution of thermal load between SG secondary side volumes. Nodalization diagram of 

SG model is presented on Figure 8. 

Figure 8 Nodalization Diagram of SG Secondary Side

HC 563,561 and HC 562,584 model SG secondary volumes of the tube bundle region 

corresponding to the straight and bended (U-shaped) parts of SG-tubes, respectively. HC 575, 567 

(up to the 4th volume) model the secondary side volumes between SG tube bundle and SG vessel, 

and between external and internal tube bundle packs corresponding to the straight portions of SG 

tubes. 5th volume of these HC is somewhat smaller and is bounded by submerged perforated plate 

(SPP) side walls. HC 586,548 model similar SG secondary side volumes which correspond to the 

bended portion of SG tubes. HC 549 represents the central part of SG secondary side between 

SG tubes packs. HC 540, 571, 582, 573 model the secondary side volume between SPP side walls 
2-8 

and SG vessel. The steam volume of SG secondary side and of SG steam header are represented 

by HC 585, 596, 577, 568 and HC 282, respectively.

2.5 Make-Up and Let-Down

The make-up and let-down are modeled at a functional level. This system consists of two 

subsystems:

• makeup is connected to the cold legs in all four loops between SG cold manifold and RCP;

let-down is connected to the cold legs of loop 2 and 3 downstream RCP outlet nozzle.

Makeup subsystem provides makeup water supply, and let-down ensures coolant removal from 

the primary system. Nodalization diagram of the makeup and let-down is given on Figure 9. 

Figure 9 Nodalization Diagram of Makeup and Let-down

2.6 Emergency Gas Removal System

The emergency gas removal system is intended to remove steam-gas mixture from the primary 

system during accidents. The system consists of connecting pipes with cut-off valves installed, 

which connect reactor coolant system elements with bubbler tank.
2-9 

Nodalization diagram of the emergency gas removal system is presented on Figure 10. 

Figure 10 Nodalization Diagram of Emergency Gas Removal System

2.7 Main Steam Lines

Main steam lines system is designed to transfer saturated steam from SG to turbine high pressure 

cylinder. Each steam line model includes steam dump valves to the atmosphere (BRU-A), SG 

steam relief valves (SRV), main steam isolation valves, turbine stop and control valves.

Steam line model is implemented using PIPE HC (1-4)50, HC (1-4)57, TEE HC (1-4)51, HC (1-

4)58, HC (1-4)53, VALVE HC (1-4)52, HC (1-4)54.

BRU-A is presented by VALVE HC (1-4)59, BREAK HC (1-4)60, which model at-mospheric 

conditions. BRU-A flow rate is 900 t/h. Time for full opening/closure of BRU-A is 18 sec. SG SRV 

are modeled similarly to BRU-А. The model implements one control SG SRV and one double main 

SG SRV.

Turbine is modeled by boundary conditions for each steam line (downstream turbine stop valve) 

at functional level and presented by BREAK HC 574,635,576,587. Turbine stop valves are 

modeled by HC 566, 666, 756, 856 with identical boundary conditions.

Nodalization diagram of steam piping, BRU-A and SG SRV is shown on Figures 11–14.
2-10

Figure 11 Nodalization Diagram of Steam Line No. 1, BRU-А and SG SRV

Figure 12 Nodalization Diagram of Steam Line No. 2, BRU-А and SG SRV
2-11

Figure 13 Nodalization Diagram of Steam Line No. 3, BRU-А and SG SRV

Figure 14 Nodalization Diagram of Steam Line No. 4, BRU-А and SG SRV
2-12

2.8 Main Steam Header

Main steam header (MSH) includes two semi-headers, connection lines between them and steam 

dump valves to the condenser (BRU-К).

MSH is modeled by PIPE HC500, 501.

TEE HC552, 556 model connection lines between two semi-headers. The connection are modeled 

as double components with appropriate equivalent hydraulic diameters assigned. VALVE HC553, 

557 model BRU-K. One valve models two valves with equivalent cross section. BREAK HC554, 

558 model turbine condenser. Nodalization diagram of MSH and BRU-K is presented in Figure 15. 

Figure 15 Nodalization Diagram of MSH and BRU-K 
2-13

2.9 Main, Auxiliary and Emergency Feedwater Systems

The nodalization diagrams of the main, auxiliary and emergency feedwater systems are presented 

on Figure 16 and Figure 17. 

Figure 16 Nodalization Diagram of the Main and Auxiliary Feedwater Systems

Figure 17 Nodalization Diagram of the Emergency Feedwater System
2-14

2.10 Emergency Core Flooding System

Emergency core flooding system (ECFS) consists of four independent functional groups, each of 

them includes:

• hydroaccumulator (HA);

• cut-off valve;

• connecting pipe.

Two functional groups supply boric acid solution to the reactor upper plenum, and other two groups 

supply it to the lower plenum. ECFS model is presented by PIPE HC610, 620, 630, 640 (option – 

Accumulator).

The nodalization diagram of ECFS and three LPIS trains is demonstrated on Figure 18. 

Figure 18 Nodalization Diagram of the System of ECFS and LPIS Trains
2-15

2.11 High Pressure Injection System

High Pressure Injection System (HPIS) includes two subsystems: high pressure injection system 

TQ13,23,33 and high pressure boron injection system TQ14,24,34.

TQ13,23,33 system model consists of three independent trains (TQ13, TQ23, TQ33), each train 

includes emergency storage tank of boric acid solution (TQ13B01, TQ23B01, TQ33B01), 

emergency boron injection pump (TQ13D01, TQ23D01, TQ33D01), piping, and valves.

TQ14,24,34 system model consists of three independent trains (TQ14, TQ24, TQ34), each train 

includes emergency storage tank of boric acid solution (TQ14B01, TQ24B01, TQ34B01), high 

pressure boron injection pump (TQ14D01, TQ24D01, TQ34D01) , piping, and valves.

HPIS model is implemented at functional level by FILL HC 913, 914, 923, 924, 933, 934. 

Connecting pipes are modeled by VALVE components on which valves are modeled. Nodalization 

diagram of HPIS trains is presented on Figure 19. Logic diagram of HPIS operation is shown on 

Figure 20. 

Figure 19 Nodalization Diagram of HPIS Trains

2.12 Low Pressure Injection System

Low Pressure Injection System (LPIS) performs both safety function (injection to the reactor 

coolant system and heat removal during the accidents) and normal operation function (namely, 

residual heat removal at low RCS pressures during normal plant shutdown) LPIS model consists 

of three independent trains (TQ12, TQ22, TQ32), each train is connected to sump tank (TQ10B01, 

TQ20B01, TQ30B01) and includes LPIS pump (TQ12D01, TQ22D01, TQ32D01), piping, and 

valves. Two trains are connected to HA–reactor connection lines to supply boric acid solution the 

reactor upper and lower plenum. TQ12 is connected to cold and hot legs of loop 1.

LPIS model is implemented at functional level and presented by TEE HC 6(1–4)2 (side 

connections), VALVE HA8(3,4) (4,5), FILL HC812, 822, 832. LPIS trains 2 and 3 are similar. 

Nodalization diagram of LPIS (TQ12,22,32) is presented on Figure 18. Logic diagram of LPIS 

operation is shown on Figure 21. 
2-16

2.13 Safeguards and Control Systems Operation Logic

Figures 20–32 present logical diagrams for the main safeguards and control systems. The 

operation logic for the following systems and elements is provided:

• HPIS and LPIS;

• reactor control and protection system;

• PRZ level control;

• make-up pressure difference control;

• MFW flow rate controller;

• SG level controller;

• BRU-A controller;

• turbine control system.
2-17

Figure 20 Logical Diagram of HPIS Operation
2-18

Figure 21 Logical Diagram of LPIS Operation
2-19

Figure 22 Logical Diagram of Control and Protection System (Part 1) 
2-20

Figure 23 Logical Diagram of Control and Protection System (Part 2)
2-21

Figure 24 Logical Diagram of Control and Protection System (Part 3)
2-22

Figure 25 Logical Diagram of Control and Protection System (Part 4)
2-23

Figure 26 Logical Diagram of PRZ Level Control by Makeup and Let-down
2-24

Figure 27 Logical Diagram of Make-up Pressure Difference Controller
2-25

Figure 28 Logical Diagram of MFW Flow Rate Controller (Part 1)
2-26

Figure 29 Logical Diagram of MFW Flow Rate Controller (Part 2)
2-27

Figure 30 Logical Diagram of SG Level Controller
2-28

Figure 31 Logical Diagram of BRU-A Controller
2-29

Figure 32 Logical Diagram of Turbine Control System

3-1 

TRACE MODEL STEADY STATE CALCULATIONS

The results of TRACE model steady-state calculation for full power unit operation and 

correspondent nominal parameters are provided in table below. 

Table 1 Steady State Calculation Results

Parameter Dimension Parameter value

Nominal value Calculated 

value

Reactor thermal power MW

(%)

3000±120 (100±4%) 3000.0 (100%)

Reactor outlet pressure kgf/cm2

(MPа)

160 ± 3 (15.7 ± 0.3) 159 (15.69)

Pressurizer level m 8.770±0.15 8.77

Coolant flow rate through the reactor m3/h 87500 85600

Coolant temperature at inlet to the 

reactor

°C 290 291.9

Coolant temperature at outlet from the 

reactor

°C 320 321.9

Coolant heat-up in the core °C 30.3 30.0

Fuel cladding maximal temperature °C 350 329.15

SG pressure kgf/cm2

(MPа)

64 ± 2 (6.28 ± 0.2) 63.01

MSH pressure kgf/cm2

(MPа)

61 ± 1 (6 ± 0.1) 61.03

Water level in SG (wide range level 

gauge)

m 2.25 2.20-2.27

Water level in SG (narrow range level 

gauge)

m (0.27–0.32) ±0.015 0.312

Feedwater temperature °C 220±5 220
3-2

Parameter Dimension Parameter value

Nominal value Calculated 

value

SG steam flow rate kg/sec 408.3 ± 16.7 408.0

SG total steam flow rate t/h 5880 ± 240 5800

Water temperature in HA °C 20–60 60

Water temperature in ECCS tanks 

(HPIS, LPIS)

°C 55–60 60
4-1 

ANALYSIS RESULTS

4.1 Station Blackout

The calculation simulates total SBO scenario (i.e., loss of off-site and on-site power supply 

including a failure of emergency diesel generators) without operators recovery actions. The battery 

discharge time is assumed to be one hour (3600 sec.) after which the secondary side BRU-A are 

stuck in intermediate position. The feedwater control valves are assumed to stuck simultaneously 

with initiating event occurrence. 

Initial conditions for SBO calculation are listed in Table 2 below. 

Table 2 Initial Conditions for SBO

Parameter RELAP5 Value TRACE Value

Reactor Power 3000 (MWt)/ 100% 3000 (MWt)/ 100%

Pressure in primary circuit 159.1 (kgf/cm2

) 158.6 (kgf/cm2) 

Level in Pressurizer 8.70 (m) 8.72 (m)

Reactor Flow 84800 (m3/h) 84700 (m3/h)

Cold leg temperature 288°C 289°C

Hot leg temperature 320°C 321°C

Maximal fuel cladding temperature 355°C 357°C

Pressure in Steam Generators 62-63 (kgf/cm2

) 62-64 (kgf/cm2)

Level in Steam Generators (1-m base) 0.274 (m) 0.300 (m)

After reactor scram and closure of turbine stop valves the steam dump from all SG begins via BRUAs operating in SG pressure maintenance mode. Condenser steam dump is not available because 

of loss of power. The decay heat is removed from reactor in natural circulation mode by evaporation 

of SGs water inventory. After decrease of SG level the primary to secondary circuit heat transfer 

degrades leading to an increase of RCS pressure with subsequent operation of PRZ PORVs 

powered from emergency power supply batteries. After batteries depletion (at 3600 sec.) PRZ 

PORVs operation is controlled by the spring setpoints which are slightly higher than normal ones.

By 4000 sec. the PRZ level reaches PRZ top due to primary coolant thermal expansion and boiling 

in the core. At 13000 sec. RELAP5 calculation shows initiation of cladding temperature increase 

which is temporarily interrupted by a drop due to loop seal clearance.
4-2

TRACE code gives later initiation time of cladding temperature increase (17000sec). Peak cladding 

temperature of a 1200°C is reached at about 26000 sec. that is 5000 sec. later than RELAP case.

The main events of the SBO calculations for RELAP5 and TRACE are listed in Table 3 below.

Table 3 Timing of Main Events for SBO Calculation

Event Relap5 Timing [sec.] TRACE Timing [sec.]

Loss of external power 0 0

BRU-As open 120 120

First PRZ PORV Opening 1050 1300

BRU-As stuck in intermediate position 3600 3600

Pressurizer is filled 4600 4550

SGs dry out 5900-7000 5900-7000

RELAP PCT is 1200°C 21000 -

End of TRACE calculation - 25000

Plots with the results of RELAP5 and TRACE SBO calculations are given below.
4-3 

Figure 33 Reactor Power

Figure 34 Primary Pressure
4-4 

Figure 35 SG Pressure

Figure 36 Pressurizer Coolant Level
4-5 

Figure 37 Steam Generators Water Level

Figure 38 Cold Legs Temperature
4-6 

Figure 39 Hot Legs Temperature

Figure 40 Max Cladding Temperature
4-7

4.2 Double-Ended LB LOCA

This analysis simulates the double ended guillotine break of the cold leg at the reactor inlet with 

simultaneous loss of power supply. Failure of one emergency diesel generator (DG) is postulated 

as the single failure.

Opening time of the break valves is equal to 0.001 sec. Initial reactor power is 104% and after 

reactor shutdown the decay heat is increased by 10%. The minimal reactor flow is assumed before 

initiating event occurrence. Postulated failures are failure of one HA connected to reactor 

downcomer, failure of one HPIS and LPIS trains. Axial core power profiles are symmetrical with 

peak value Kz = 1.49 and fuel pin peaking factor Kr = 1.74.

It is assumed conservatively that hot channel is isolated completely from the other core (for all 4 

core sectors). Initial mass flow through hot channel during steady state is 96 kg/s (460 m3

/h) in 

both RELAP and TRACE models. The reactor power in TRACE calculation is specified using a 

table from RELAP 5 model to get identical power decay after shutdown.

Initial conditions for LB LOCA calculation are listed in Table 4 below.

Table 4 Initial Conditions for LBLOCA Calculation

Parameter RELAP 5 Value TRACE Value

Core power 3120 (MW)/ 104% 3120 (MW)/ 104%

RCS pressure 159.0 (kgf/cm2) 158.2 (kgf/cm2)

Pressurizer level 8.70 (m) 8.77(m)

Vol. flow through reactor 80000 (m3/h) 80400 (m3/h)

Cold leg coolant temperature 290°C 292°C

Hot leg coolant temperature 323°C 324°C

Maximal fuel cladding temperature 351°C 348°C

Maximal fuel temperature (hot channel) 1967°C 1927°C

Secondary pressure 61 (kgf/cm2) 61 (kgf/cm2)

SG Level (narrow range gauge, 1-m base) 0.27 (m) 0.30 (m)

ECCS Temperature 90°C 90°C

Hot channel inlet mass flow 96 (kg/s) 96 (kg/s)
4-8

At 0 sec. of calculation the break is initiated. After the break the flow reversal occurs causing first 

cladding temperature rise. The temperature is decreased after HAs injection. After DGs start the 

ECCS trains start to inject borated water with 90°C temperature.

The second peak temperature at core reflood phase is reached at 315 sec. for RELAP 5 and at 

355 sec. for TRACE calculation. After that because of continuous ECCS injection the temperature 

begins to decrease.

The main events of the RELAP5 and TRACE calculations of LBLOCA are listed in Table 5 below.

Table 5 Timing of Main Events of LBLOCA Calculation

Event Relap 5 

Timing [sec.]

TRACE Timing 

[sec.] Comment

LOCA opens

Loss of power supply 0 0

Diesel generators start 18 18

First cladding temperature peak 4.8 4.6 TRACE value 778°C

RELAP value 980°C

Start of HPIS and LPIS injection 60 60

Second cladding temperature 

peak 315 355 TRACE value 1127°CRELAP value 973°C

End of calculation 1000 1000

Comparison plots for LB LOCA RELAP and TRACE calculations are shown below.
4-9 

Figure 41 Reactor Power (Fragment)

Figure 42 Hot Channel Collapsed Level (Fragment)
4-10

Figure 43 Hot Channel Mass Flow (Fragment)

Figure 44 Hot Channel Max Cladding Temperature
4-11

Figure 45 Hot Channel Max Fuel Temperature

Figure 46 RCS Pressure
4-12

Figure 47 HPIS Flow

Figure 48 LPIS Flow
4-13

Figure 49 HA Water Volume

Figure 50 Maximal Cladding Temperature
4-14

Figure 51 Hot Channel Collapsed Level
5-1 

DISCUSSION

5.1 SBO

The results demonstrate that SG levels behavior in RELAP and TRACE calculations is in a good 

agreement so the heat transfer from primary to secondary is modeled adequately. But PRZ level 

increase timing is quite different that could be caused by difference in PRZ PORVs operation. 

Because of that the water release is also different that causes later core heat-up for the TRACE 

model. Other parameters seem to be in a good agreement.

Comparison of the results suggest that PRZ PORVs model for TRACE code requires mode 

detailed evaluation in order to verify PORVs operation logic, discharge options, PRZ dome 

nodalization and other modeling options.

5.2 LB LOCA

Core flow reversal at the initial phase is clearly seen in both codes calculations. First cladding 

temperature peak in TRACE calculation is lower than the one predicted by RELAP. This can be 

caused by differences in heat transfer correlations. The ECCS flows and HA inventories are in 

good agreement because of nearly identical RCS pressure values after blowdown phase in both 

RELAP and TRACE calculations. 

Second cladding peak temperature in TRACE calculation is higher because of lower coolant 

inventory in the reactor core. This suggests that the break discharge models and flow regimes 

during the reflood phase (especially droplet flows) should be checked. Timing of the max cladding 

temperatures is in good agreement for both codes.

5.3 Future Plans

Based on the results of initial TRACE model evaluation it can be concluded that the VVER-1000 

model developed needs further improvement in the components/models and control logic 

simulation. Areas of further model improvement and qualification include: 

• core modeling (hot channels, gap deformation);

• discharge models adjustment;

• improvement of reactor kinetics modeling;

• model verification

• model validation using actual NPP transient data.

Correspondent activities are foreseen in the framework of Task Order 17 under Basic Ordering 

Agreement #257586 between SSTC NRS and Brookhaven National Laboratory [3].

6-1 

CONCLUSIONS

This report provides brief description of the results of comparative calculations performed for two 

accident scenarios with application of RELAP and TRACE VVER-1000 models. The initiating 

events analyzed include: 

• Total loss of off-site and on-site power supply (total station blackout, SBO);

• Large break LOCA (850 mm double-ended cold leg break) scenario with simultaneous loss of 

power.

The comparison demonstrates good agreement between the results obtained with both codes but 

identified the need for further improvement of TRACE model developed. In particular, the model 

needs to be checked and improved in such areas as core modeling, systems performance 

characteristics and discharge model adjustment, reactor kinetics, etc. TRACE model qualification 

is foreseen to enable its further application in SNRIU/SSTC NRS regulatory activities on evaluation 

of safety assessments performed for Ukrainian NPPs.

7-1 

REFERENCES

1. Development of the Multipurpose Thermal Hydraulic Model of NSSS with WWER-1000/320. 

Specification of the Main Components of the Model. R&D Report. Kyiv, SSTC NRS, 2010.

2. R&D Report. Support in Mastering the Computer Codes Obtained Within the 2014 CAMP and 

CSARP Agreements. Transfer of WWER-1000/320 Thermal Hydraulic Model Developed For 

RELAP5 Code Into TRACE Format, Kyiv, SSTC NRS, 2015.

3. Task Order 17 under BOA #257586 between the Brookhaven National Laboratory and the 

SSTC NRS, 2015.



November 2016 TRACE/RELAP5 Comparative Calculations for Double-Ended LBLOCA and SBO

S. Iegan, A. Mazur,

Y. Vorobyov, O. Zhabin, S. Yanovskiy

State Nuclear Regulatory Inspectorate of Ukraine

And State Scientific and Technical Center

9/11 Arsenalna str. Kyiv, Ukraine 01011

 

Division of Systems Analysis

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

K. Tien, NRC Project Manager

This report is developed by the State Nuclear Regulatory Inspectorate of Ukraine (SNRIU) and its technical support organization, the State 

Scientific and Technical Center for Nuclear and Radiation Safety of Ukraine (SSTC NRS), under Implementing Agreement On ThermalHydraulic Code Applications And Maintenance Between The United States Nuclear Regulatory Commission and State Nuclear Regulatory 

Inspectorate of Ukraine (signed in 2014) in accordance with Article III, Section C, of the Agreement.

The report provides description of quantitative analysis results of two initiating events for VVER-1000 reactor system with application of 

TRACE p4 and RELAP5/MOD3.2 codes. For these calculations RELAP5 input model for VVER-1000 was converted to TRACE code. The 

results obtained with both codes were compared to evaluate differences and to identify future needs for TRACE input deck enhancement.

NUREG/IA-0475

Technical



NUREG/IA-0475 TRACE/RELAP5 Comparative Calculations for Double-Ended LBLOCA and SBO November 2016