NUREG/IA-0516

LOCAs With Loss of One Active

Emergency Cooling System Simulated

by RELAP5

Prepared by:

Andrej Prošek

Jožef Stefan Institute

Jamova cesta 39

SI-1000, Ljubljana, Slovenia

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: November 2019

Date Published: January 2020

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
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NUREG/IA-0516

LOCAs With Loss of One Active 

Emergency Cooling System Simulated

by RELAP5

Prepared by:

Andrej Prošek

Jožef Stefan Institute

Jamova cesta 39

SI-1000, Ljubljana, Slovenia

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: November 2019

Date Published: January 2020

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

The second generation reactors were designed and built to withstand without loss to the 

structures, systems, and components necessary to ensure public health and safety during 

design basis accidents. There are also accident sequences that are possible but were judged to 

be too unlikely and therefore were not fully considered in the design process of second 

generation reactors. In that sense, they were considered beyond the scope of design-basis 

accidents that a nuclear facility must be designed and built to withstand. They were called 

beyond design basis accidents. After Fukushima-Daiichi in the Europe the design extension 

conditions were introduced as preferred method for giving due consideration to the complex 

sequences and severe accidents without including them in the design basis conditions. In the 

study, the analysis was performed to see if the plant design can prevent spectrum of loss of 

coolant accidents together with the complete loss of one emergency core cooling function (e.g. 

high pressure injection or low pressure injection). The analyzed break spectrum ranged from 

1.27 cm to 30.48 cm. For calculations the latest RELAP5/MOD3.3 Patch 5 has been used and 

the RELAP5 input model of Krško nuclear power plant. 

v 

TABLE OF CONTENTS

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

TABLE OF CONTENTS ............................................................................................................. v

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

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

EXECUTIVE SUMMARY ...........................................................................................................xi

ACKNOWLEDGMENTS..........................................................................................................xiii

ABBREVIATIONS AND ACRONYMS ......................................................................................xv

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

2 METHODS USED .................................................................................................................. 5

2.1 LOCA Scenario Description ..........................................................................................5

2.2 RELAP5 Input Model Description ..................................................................................5

2.3 Initial and Boundary Conditions.....................................................................................6

2.4 Simulated LOCA Break Cases ......................................................................................7

3 RESULTS .............................................................................................................................. 9

3.1 Results of LOCA Break Size Spectrum Without HPIS (noHP)......................................9

3.2 Results of LOCA Break Size Spectrum Without LPIS (noLP).....................................15

3.3 Results of LOCA Break Size Spectrum 1.27 cm to 5.08 cm Without HPIS and 

Using AM1 Strategy (noHP and AM1).........................................................................22

3.4 Results of LOCA Break Size Spectrum 1.27 cm to 5.08 cm With One ECCS 

and Using AM2 Strategy (noHP and AM2) ..................................................................25

3.5 Results of LOCA Break Size Spectrum 1.27 cm to 5.08 cm With One ECCS 

and Using AM3 Strategy (noHP and AM3) ..................................................................28

4 CONCLUSIONS...................................................................................................................33

5 REFERENCES.....................................................................................................................35

vii

LIST OF FIGURES

Figure 2-1 RELAP5 Krško NPP Hydraulic Components View ...............................................6

Figure 3-1 Pressurizer Pressure – noHP (1.27 cm to 7.62 cm) .............................................9

Figure 3-2 SG1 Pressure – noHP (1.27 cm to 7.62 cm) ......................................................10

Figure 3-3 Core Collapsed Liquid Level – noHP (1.27 cm to 7.62 cm) ................................10

Figure 3-4 Cladding Temperature – noHP (1.27 cm to 7.62 cm) .........................................10

Figure 3-5 Primary System Mass – noHP (1.27 cm to 7.62 cm)..........................................11

Figure 3-6 Total Integrated Break Flow – noHP (1.27 cm to 7.62 cm).................................11

Figure 3-7 Total Integrated Accumulator Flow – noHP (1.27 cm to 7.62 cm).......................11

Figure 3-8 Total Integrated HPIS Flow – noHP (1.27 cm to 7.62 cm)..................................12

Figure 3-9 Total Integrated LPIS Flow – noHP (1.27 cm to 7.62 cm)...................................12

Figure 3-10 Pressurizer Pressure – noHP (10.16 cm to 30.48 cm) .......................................12

Figure 3-11 SG1 Pressure – noHP (10.16 cm to 30.48 cm) ..................................................13

Figure 3-12 Core Collapsed Liquid Level – noHP (10.16 cm to 30.48 cm) ............................13

Figure 3-13 Cladding Temperature – noHP (10.16 cm to 30.48 cm) .....................................13

Figure 3-14 Primary System Mass – noHP (10.16 cm to 30.48 cm) ......................................14

Figure 3-15 Total Integrated Break Flow – noHP (10.16 cm to 30.48 cm).............................14

Figure 3-16 Total Integrated Accumulator Flow – noHP (10.16 cm to 30.48 cm)...................14

Figure 3-17 Total Integrated HPIS Flow – noHP (10.16 cm to 30.48 cm)..............................15

Figure 3-18 Total Integrated LPIS Flow – noHP (10.16 cm to 30.48 cm)...............................15

Figure 3-19 Pressurizer Pressure – noLP (1.27 cm to 7.62 cm)............................................16

Figure 3-20 SG1 Pressure – noLP (1.27 cm to 7.62 cm).......................................................16

Figure 3-21 Core Collapsed Liquid Level – noLP (1.27 cm to 7.62 cm).................................16

Figure 3-22 Cladding Temperature – noLP (1.27 cm to 7.62 cm)..........................................17

Figure 3-23 Primary System Mass – noLP (1.27 cm to 7.62 cm)...........................................17

Figure 3-24 Total Integrated Break Flow – noLP (1.27 cm to 7.62 cm)..................................17

Figure 3-25 Total Integrated Accumulator Flow – noLP (1.27 cm to 7.62 cm) .......................18

Figure 3-26 Total Integrated HPIS Flow – noLP (1.27 cm to 7.62 cm)...................................18

Figure 3-27 Total Integrated LPIS Flow – noLP (1.27 cm to 7.62 cm) ...................................18

Figure 3-28 Pressurizer Pressure – noLP (10.16 cm to 30.48 cm)........................................19

Figure 3-29 SG1 Pressure – noLP (10.16 cm to 30.48 cm)...................................................19

Figure 3-30 Core Collapsed Liquid Level – noLP (10.16 cm to 30.48 cm).............................19

Figure 3-31 Cladding Temperature – noLP (10.16 cm to 30.48 cm)......................................20

Figure 3-32 Primary System Mass – noLP (10.16 cm to 30.48 cm).......................................20
viii

Figure 3-33 Total Integrated Break Flow – noLP (10.16 cm to 30.48 cm)..............................20

Figure 3-34 Total Integrated Accumulator Flow – noLP (10.16 cm to 30.48 cm) ...................21

Figure 3-35 Total Integrated HPIS Flow – noLP (10.16 cm to 30.48 cm)...............................21

Figure 3-36 Total Integrated LPIS Flow – noHP (10.16 cm to 30.48 cm)...............................21

Figure 3-37 Pressurizer Pressure – noHP and AM1 (1.27 cm to 5.08 cm) ............................22

Figure 3-38 SG1 Pressure – noHP and AM1 (1.27 cm to 5.08 cm).......................................22

Figure 3-39 Core Collapsed Liquid Level – noHP and AM1 (1.27 cm to 5.08 cm) .................23

Figure 3-40 Cladding Temperature – noHP and AM1 (1.27 cm to 5.08 cm) ..........................23

Figure 3-41 Primary System Mass – noHP and AM1 (1.27 cm to 5.08 cm)...........................23

Figure 3-42 Total Integrated Break Flow – noHP and AM1 (1.27 cm to 5.08 cm)..................24

Figure 3-43 Total Integrated Accumulator Flow – noHP and AM1 (1.27 cm to 5.08 cm)........24

Figure 3-44 Total Integrated HPIS Flow – noHP and AM1 (1.27 cm to 5.08 cm)...................24

Figure 3-45 Total integrated LPIS flow – noHP and AM1 (1.27 cm to 5.08 cm).....................25

Figure 3-46 Pressurizer Pressure – noHP and AM2 (1.27 cm to 5.08 cm) ............................25

Figure 3-47 SG1 Pressure – noHP and AM2 (1.27 cm to 5.08 cm).......................................26

Figure 3-48 Core Collapsed Liquid Level – noHP and AM2 (1.27 cm to 5.08 cm) .................26

Figure 3-49 Cladding Temperature – noHP and AM2 (1.27 cm to 5.08 cm) ..........................26

Figure 3-50 Primary System Mass – noHP and AM2 (1.27 cm to 5.08 cm)...........................27

Figure 3-51 Total Integrated Break Flow – noHP and AM2 (1.27 cm to 5.08 cm)..................27

Figure 3-52 Total Integrated Accumulator Flow – noHP and AM2 (1.27 cm to 5.08 cm)........27

Figure 3-53 Total Integrated HPIS Flow – noHP and AM2 (1.27 cm to 5.08 cm)...................28

Figure 3-54 Total Integrated LPIS Flow – noHP and AM2 (1.27 cm to 5.08 cm) ...................28

Figure 3-55 Pressurizer Pressure – noHP and AM3 (1.27 cm to 5.08 cm) ............................29

Figure 3-56 SG1 Pressure – noHP and AM3 (1.27 cm to 5.08 cm).......................................29

Figure 3-57 Core Collapsed Liquid Level – noHP and AM3 (1.27 cm to 5.08 cm) .................29

Figure 3-58 Cladding Temperature – noHP and AM3 (1.27 cm to 5.08 cm) ..........................30

Figure 3-59 Primary System Mass – noHP and AM3 (1.27 cm to 5.08 cm)...........................30

Figure 3-60 Total Integrated Break Flow – noHP and AM3 (1.27 cm to 5.08 cm)..................30

Figure 3-61 Total Integrated Accumulator Flow – noHP and AM3 (1.27 cm to 5.08 cm)........31

Figure 3-62 Total Integrated HPIS Flow – noHP and AM3 (1.27 cm to 5.08 cm)...................31

Figure 3-63 Total Integrated LPIS Flow – noHP and AM3 (1.27 cm to 5.08 cm) ...................31
ix

Table 1-1

LIST OF TABLES 

Accident Management in PWRs For BDBA (With LOCAs) With NonDegraded Core .................................................................................................... 3

Table 2-2 Initial and Boundary Conditions............................................................................ 7

Table 2-3 LOCA Scenario Cases Without One ECCS (HPIS or LPIS) ................................. 7

Table 2-4 LOCA Scenario Cases Without One ECCS (HPIS or LPIS) and Using 

Accident Management Strategies ........................................................................ 8

xi

EXECUTIVE SUMMARY

The second generation reactors were designed and built to withstand without loss to the 

structures, systems, and components necessary to ensure public health and safety during 

design basis accidents (DBAs). In the transient and accident analysis the effects of single active 

failures and operator errors were considered. There are also accident sequences that are 

possible but were judged to be too unlikely and therefore were not fully considered in the design 

process of second generation reactors. In that sense, they were considered beyond the scope 

of design basis accidents that a nuclear facility must be designed and built to withstand. They 

were called beyond design basis accidents (BDBA). The design extension conditions were 

introduced in the Europe as preferred method for giving due consideration to the complex 

sequences and severe accidents at the design stage of the reactors of third generation without 

including them in the design basis conditions. The Western European Nuclear Regulators 

Association (WENRA) recommended a “design extension” analysis in 2007 and it proposed a 

list of events to be analysed at minimum. After Fukushima Dai-ichi accident also the 

International Atomic Energy Agency (IAEA) adopted the term design extension conditions 

(DEC). WENRA reference levels (RLs) from 2014 introduced DEC term too, following IAEA 

definition. Slovenia implemented WENRA reference level issue F requirements into its Rules on 

radiation and nuclear safety factors in 2016.

The control of DECs is expected to be achieved primarily by features implemented in the design 

(safety features for DECs) and not only by accident management measures that are using 

equipment designed for other purposes. This means that in principle a DEC is such if its 

consideration in the design leads to the need of additional equipment or to an upgraded 

classification of lower class equipment to mitigate the DEC.

In the analysis presented the loss of coolant accidents (LOCA) break spectrum ranging from 

1.27 cm to 30.48 cm has been analysed using the latest RELAP5/MOD3.3 Patch 5 and 

RELAP5 input model of Krško nuclear power plant, which is a two loop pressurized water 

reactor (PWR). The analysis was performed to see if the plant design can prevent spectrum of 

LOCAs together with the complete loss of one emergency core cooling function (e.g. high 

pressure injection or low pressure injection). Besides this for smaller breaks ranging from 1.27 

cm to 5.08 cm leading to core heatup (being design extension conditions as new equipment is 

needed) accident management strategies to depressurize the primary system in order to enable 

injection of accumulators and low pressure safety injection system have been also studied.

xiii

ACKNOWLEDGMENTS

The author acknowledge the research core funding no. P2-0026 that was financially supported 

by the Slovenian Research Agency. The Krško nuclear power plant and Slovenian Nuclear Safety 

Administration (SNSA) supported this research through CAMP project no. POG-U3-KER4/104/12 (NEK no. 3180019).

xv

ABBREVIATIONS AND ACRONYMS

AFW auxiliary feedwater

AM accident management

ATWS anticipated transients without scram

BDBA beyond design basis accidents

CVCS chemical and volume control system

DBA design basis accident

DEC design extension conditions

ECCS emergency core cooling system

IAEA International Atomic Energy Agency

LOCA loss of coolant accident

HPI high pressure injection

HPIS high pressure injection sytem

HPSI high pressure safety injection

LSTF Large Scale Test Facility

LPI low pressure injection

LPIS low pressure injection system

LPSI low pressure safety injection

MD motor driven

MFW main feedwater

MSIV main steam isolation valves

NPP Nuclear Power Plant

NR narrow range

PORV power operated relief valve

PRZ pressurizer

PWR pressurized water reactor

RCP reactor coolant pump

RL reference level

RPV reactor pressure vessel

SBO station blackout

SD steam dump

SI safety injection

SL surge line

SG steam generator
xvi

TD turbine driven

UOP Ultimate Operating Procedure

WENRA Western European Nuclear Regulators Association
1

1 INTRODUCTION

The second generation reactors were designed and built to withstand without loss to the 

structures, systems, and components necessary to ensure public health and safety during 

design basis accidents (DBAs). In the transient and accident analysis the effects of single active 

failures and operator errors were considered. There are also accident sequences that are 

possible but were judged to be too unlikely and therefore were not fully considered in the design 

process of second generation reactors. In that sense, they were considered beyond the scope 

of design basis accidents that a nuclear facility must be designed and built to withstand. They 

were called beyond design basis accidents (BDBA). 

The term “design extension conditions” has been introduced to define some selected accident 

sequences due to multiple failures by the European Utility Requirements document [1]. The 

design extension conditions were introduced as preferred method for giving due consideration 

to the complex sequences and severe accidents at the design stage of the reactors of third 

generation without including them in the design basis conditions. The Western European 

Nuclear Regulators Association (WENRA) recommended a “design extension” analysis in 

2007 [2] and they proposed a list of events to be analysed at minimum. By its meaning this list 

corresponds to DEC without core melt. After Fukushima Dai-ichi accident also the International 

Atomic Energy Agency (IAEA) adopted the term design extension conditions (DEC) [3]. WENRA 

reference levels (RLs) from 2014 [4] introduced DEC term. The WENRA guidance document for 

issue F for existing reactors (second generation) explains [5] explains that DEC in WENRA RLs 

are consistent with the definition of DEC in IAEA SSR-2/1 [3], published in 2012: "Accident 

conditions that are not considered for design basis accidents, but that are considered in the 

design process of the facility in accordance with best estimate methodology, and for which 

releases of radioactive material are kept within acceptable limits. Design extension conditions 

could include severe accident conditions." DEC are more complex and/or more severe than 

conditions postulated as design basis accidents [5].

Finally, the new revision 1 of IAEA SSR-2/1 [6] has been released in 2016, redefining the DEC 

term, which definition is now as follows: “Postulated accident conditions that are not considered 

for design basis accidents, but that are considered in the design process for the facility in 

accordance with best estimate methodology, and for which releases of radioactive material are 

kept within acceptable limits. Design extension conditions comprise conditions in events without 

significant fuel degradation and conditions in events with core melting”. WENRA did not follow 

the new IAEA definition of DEC in spite of the fact that modification is significant, as DEC are

now defined as postulated accident conditions.

The DEC concept by IAEA and WENRA (DEC with prevention of core melt is called DEC A) is 

not completely new, since in some countries selected multiple failures have already been 

considered in the design (through back fitting process), for example anticipated transients 

without scram (ATWS) and station blackout (SBO). Also, the research for beyond design basis 

accidents with non-degraded core (i.e. DEC A) for existing reactors has been already done in 

80’s and 90’ of the previous century.

Slovenia implemented WENRA reference levels issue F requirements into its Rules on radiation 

and nuclear safety factors [12]. It is prescribed that the selection process for design extension 

conditions A shall start by considering those events and combinations of events, which cannot 

be considered with a high degree of confidence to be extremely unlikely to occur and which may 

lead to severe fuel damage in the core or in the spent fuel storage. It shall cover: a) events 
2

occurring during the defined operational states of the plant; b) events resulting from internal or 

external hazards; and c) common cause failures. In the presented analysis two such DEC A 

scenarios were simulated, loss of coolant accident (LOCA) in cold leg together with complete 

loss of high pressure injection (HPI) system and LOCA with complete loss of low pressure 

injection (LPI) system. This is in line with IAEA and WENRA, as can be seen in the following.

WENRA guidance document [5] for issue F also provides this scenario as an example of DEC 

A:

• Loss of coolant accident (LOCA) together with the complete loss of one emergency core 

cooling function (e.g. high pressure injection (HPI) or low pressure injection (LPI)).

Also IAEA TECDOC-1791 document [13] also provides this scenario as an example of DEC 

derived from probabilistic safety assessment (PSA):

• LOCA plus loss of one emergency core cooling system (either the high pressure or the 

low pressure emergency cooling system).

The analysis presented will show if the plant design can prevent above LOCA scenarios with 

existing safety systems or not (in this case additional safety features are needed). Following 

document [13], the control of DECs is expected to be achieved primarily by features 

implemented in the design (safety features for DECs) and not only by accident management 

measures that are using equipment designed for other purposes. This means that in principle a 

DEC is such if its consideration in the design leads to the need of additional equipment or to an 

upgraded classification of lower class equipment to mitigate the DEC.

Similar experiments has already been investigated in the past dealing with accident 

management. Therefore, the results of these experiments have been first reviewed. Table 1-1

shows selected tests performed on BETHSY and LOFT test facilities for accident management 

in pressurized water reactors (PWRs) for loss of coolant accidents (LOCAs), in which operator 

actions were studied for BDBA with non-degraded core (in terms of WENRA such accidents are 

called DEC A). Experiments were mainly selected from cross-reference matrix for accident 

management for non-degraded core, which has been created in the frame of OECD/NEA [7].
3

Table 1-1 Accident Management in PWRs For BDBA (With LOCAs) With Non-Degraded 

Core

Test No. Test type Brief description

BETHSY 

6.2TC

6” cold leg break without 

HPIS and LPIS

BETHSY 6.2TC test was a 15.24 cm (6 inch) cold leg break in the 

loop one without available high pressure and low pressure safety 

injection system. Accumulators were available in the intact loops. 

The main aims of this test were to compare the counterpart test 

data from BETHSY and Large Scale Test Facility (LSTF) facilities 

and qualification of CATHARE 2 computer code. Further 

information on BETHSY 6.2TC could be find in [8].

BETHSY 

9.1b

2" cold leg break without 

HPIS and with delayed 

ultimate procedure

BDBA involves two failures: a break on the cold leg together with 

a complete failure of the HPIS, combined with a human error 

regarding the conditions in which the operators start the Ultimate 

Operating Procedure (UOP). The UOP then consists in 

depressurizing the primary circuit by means of a full opening of 

the 3 SG atmospheric steam dumps. Further information on 

BETHSY 6.2TC could be find in [9].

LSTF

PWR Cold-Leg smallbreak LOCA with total HPI 

failure

Cold-leg break tests were conducted at the LSTF for five break 

areas 0.5%, 1%, 2.5%, 5 and 10% of the scaled cold-leg flow 

area, with totally failed HPI [10].

LSTF

0.5% cold leg small-break 

LOCA total failure of the 

HPI and auxiliary 

feedwater (AFW) systems

The depressurization procedure was simulated in a 0.5% cold-leg 

break LOCA experiment [11].

In the scenario LOCA in cold leg together with complete loss of HPI system, the results of 

BETHSY 9.1b, which is 5.08 cm (2 inch, 0.5%) break on the cold leg together with a complete 

failure of the HPI system (HPIS), showed that core heatup is obtained without performing 

depressurization of the primary side through the secondary side. Test data of LSTF and model 

calculations showed that intentional primary system depressurization with use of the pressurizer 

power operated relief valves (PORVs) is effective for break areas of approximately 0.5% or less, 

is unnecessary for breaks of approximately 5% or more, and might be insufficient for 

intermediate break areas to maintain adequate core cooling. It was also shown that there might 

be possibility of core dryout after accumulator injection and before LPI system (LPIS) injection 

for break areas less than approximately 2.5%.

In the scenario LOCA in cold leg together with complete loss of LPI system, the results of 

BETHSY 6.2TC, which is 15.24 cm (6 inch, ≈5%) break on the cold leg together with a complete 

failure of the HPIS and LPIS, showed that core heatup is obtained before accumulator injection 

is started and after accumulator injection is terminated.

The report is organized as follows. In Section 2 the methods used are described. First the LOCA 

scenarios typical sequence is described. Then the RELAP5 thermal-hydraulic system computer 

code and input model description are briefly described, followed by the initial and boundary 

conditions, resulting from steady state calculations. Last, the simulated scenarios include eight 

break sizes, ranging from 1.27 cm to 30.48 cm equivalent diameter break size in cold leg. Both, 

LOCA with complete loss of HPIS and LOCA with complete loss of LPIS scenarios were 

simulated. The simulations include also some scenarios using accident management measures. 

In Section 3 the results of the LOCA calculations are presented, including discussion of the 

result. Finally, main conclusions are drawn.

5

2 METHODS USED

2.1 LOCA Scenario Description

In the LOCAs simulated at the beginning of transient from the emergency core cooling systems 

(ECCSs) two accumulators were assumed available and either the high pressure or the low 

pressure emergency core cooling system. The initiating event is opening of the valve simulating 

the break in the cold leg with reactor operating at 100% power. The reactor trip on 

(compensated) low pressurizer pressure (12.99 MPa) further causes the turbine trip. The safety 

injection (SI) signal is generated on the low-low pressurizer pressure signal at 12.27 MPa. On SI 

signal active safety systems either high pressure safety injection (HPSI) pumps or low pressure 

safety injection (LPSI) pumps and motor driven (MD) auxiliary feedwater (AFW) pumps start. 

The HPSI pumps started to inject, when pressure is lower than 15.14 MPa. When pressure 

drops below 49.55 bars, both accumulators start to inject. Larger is the break size, faster is the 

accumulator discharge. When pressure drops below 1.13 MPa, the LPSI pumps start to inject.

In the case of smaller breaks, the high primary pressure can prevent accumulator and LPSI 

pumps injection.

2.2 RELAP5 Input Model Description

For calculations the latest RELAP5/MOD3.3 Patch 5 [14] has been used and the RELAP5 input 

model of Krško Nuclear Power Plant (NPP), used also in study [15]. Krško NPP is a two loop 

pressurized water reactor (PWR), Westinghouse type, with reactor power uprated to 1994 MW

and thermal power 2000 MW. The base model consists of 469 control volumes, 497 junctions 

and 378 heat structures with 2107 radial mesh points. In terms of SNAP this gives 304 hydraulic 

components and 108 heat structures. Hydraulic components in SNAP consist of both volumes 

and junctions, where pipe with more volumes is counted as one component. Each heat structure 

in SNAP connected to pipe is also counted as one component and not as many heat structures 

as pipe has volumes like it is counted in the RELAP5 output file. This explains the difference in 

the number of heat structures in Figure 2-1 and that reported in RELAP5 output file.

Modeling of the primary side includes the reactor pressure vessel (RPV), both loops (LOOP 1 

and 2), the pressurizer (PRZ) vessel, pressurizer surge line (SL), pressurizer spray lines and 

valves, two pressurizer power operated relief valves (PORVs) and two pressurizer safety 

valves, chemical and volume control system (CVCS) charging and letdown flow, and reactor 

coolant pump (RCP 1 and 2) seal flow. Emergency core cooling systems (ECCSs) piping 

includes two high pressure safety injection (HPSI) pumps, two accumulators (ACC 1 and 2), and 

two low pressure safety injection (LPSI) pumps. The secondary side consists of the two steam 

generators (SGs) - secondary side, main steam line, main steam isolation valves (MSIV 1 and 

2), SG1 and SG2 relief and safety valves, and main feedwater (MFW1 and MFW2) piping. The 

turbine valve is modeled by the corresponding logic. Steam dump (SD) is also modeled by the 

corresponding logic. The turbine is represented by time dependent volume. The MFW and AFW 

(auxiliary feedwater) motor driven (MD) and turbine driven (TD) pumps are modeled as time 

dependent junctions.
6

Figure 2-1 RELAP5 Krško NPP Hydraulic Components View

2.3 Initial and Boundary Conditions

Table 2-2 shows initial and boundary conditions at the beginning of simulation. Initial values and 

boundary conditions are given for both loops (where applicable). It can be seen that RELAP5 

initial and boundary conditions are close to reference PWR values with the exception of steam 

generator pressures, where deviation is about 2.5%. Nevertheless, in LOCA calculations the 

influence of the secondary side is typically smaller than in the secondary side initiated transients 

because due to break the primary side empties and the natural circulation is terminated.
7

Table 2-2 Initial and Boundary Conditions

Parameter (unit) Reference PWR RELAP5/MOD3.3

Core power (MW) 1994 1994

Pressurizer pressure (MPa) 15.513 15.513

Pressurizer level (%) 55.7 55.8

Average RCS temperature no. 1 (K) 578.15 578.15

Average RCS temperature no. 2 (K) 578.15 578.06

Cold leg temperature no. 1 (K) 558.75 559.51

Cold leg temperature no. 2 (K) 558.75 559.32

Hot leg temperature no. 1 (K) 597.55 596.79

Hot leg temperature no. 2 (K) 597.55 596.79

Cold leg flow no. 1 (kg/s) 4694.7 4721.2

Cold leg flow no. 2 (kg/s) 4694.7 4719.6

Steam generator pressure no. 1 (MPa) 6.281 6.438

Steam generator pressure no. 2 (MPa) 6.281 6.415

Steam generator NR level no. 1 (%) 69.3 69.3

Steam generator NR level no. 2 (%) 69.3 69.3

Steam flow no. 1 (kg/s) 544.5 541.3

Steam flow no. 2 (kg/s) 544.5 544.5

Main feedwater temperature (K) 492.6 492.8

2.4 Simulated LOCA Break Cases

The breaks simulated were 1.27 cm (0.5 inch), 2.54 cm (1 inch), 5.08 cm (2 inch), 7.62 cm 

(3 inch), 10.16 cm (4 inch), 15.24 cm (6 inch), 20.32 cm (8 inch) and 30.48 cm (12 inch) 

equivalent diameter cold leg breaks. For each break size two simulations were performed, one 

without HPSI pumps and one without LPSI pumps, with other systems available as can be seen 

from Table 2-3. In all simulations default values for break flows were used for Ransom Trapp 

critical flow model.

Table 2-3 LOCA Scenario Cases Without One ECCS (HPIS or LPIS)

Break size diameter case without HPSI pumps case without LPSI pumps

1.27 cm (0.5 inch) sb0.5_noHP sb0.5_noLP

2.54 cm (1 inch) sb1_noHP sb1_noLP

5.08 cm (2 inch) sb2_noHP sb2_noLP

7.62 cm (3 inch) sb3_noHP sb3_noLP

10.16 cm (4 inch) sb4_noHP sb4_noLP

15.24 cm (6 inch) sb6_noHP sb6_noLP

20.32 cm (8 inch) sb8_noHP sb8_noLP

30.48 cm (12 inch) sb12_noHP sb12_noLP

In addition to scenarios (i.e. DEC), which could not be mitigated, scenarios with accident 

management (AM) measures were also simulated, shown in Table 2-4. The first accident 

management strategy AM1 was the use of pressurizer (PRZ) power operated relief valve 

(PORV) to depressurize reactor coolant system (RCS) to 1.1 MPa. Second AM strategy (AM2) 

was to depressurize the RCS through the secondary side, using SG PORVs to depressurize 
8

SGs to 1.2 MPa. Third AM strategy (AM3) was to depressurize the RCS through the secondary 

side, using SG PORVs to depressurize SGs to 0.7 MPa.

Table 2-4 LOCA Scenario Cases Without One ECCS (HPIS or LPIS) and Using Accident 

Management Strategies

Break size 

diameter

accident management (AM) measures case without 

HPIS pumps

1.27 cm (0.5 inch) AM1: using PRZ PORV to depressurize RCS to 1.1 MPa sb0.5_noHPd

2.54 cm (1 inch) AM1: using PRZ PORV to depressurize RCS to 1.1 MPa sb1_noHPd

5.08 cm (2 inch) AM1: using PRZ PORV to depressurize RCS to 1.1 MPa sb2_noHPd

1.27 cm (0.5 inch) AM2: using SG PORVs to depressurize SGs to 1.2 MPa sb0.5_noHPdsg

2.54 cm (1 inch) AM2: using SG PORVs to depressurize SGs to 1.2 MPa sb1_noHPdsg

5.08 cm (2 inch) AM2: using SG PORVs to depressurize SGs to 1.2 MPa sb2_noHPdsg

1.27 cm (0.5 inch) AM3: using SG PORVs to depressurize SGs to 0.7 MPa sb0.5_noHPdsg7

2.54 cm (1 inch) AM3: using SG PORVs to depressurize SGs to 0.7 MPa sb1_noHPdsg7

5.08 cm (2 inch) AM3: using SG PORVs to depressurize SGs to 0.7 MPa sb2_noHPdsg7
9

3 RESULTS

The results of LOCA break spectrum calculations are shown in Figures 3-1 through 3-63. For 

each break size the following nine parameters are shown: pressurizer pressure, SG1 pressure, 

core collapsed liquid level, cladding temperature, RCS mass, integrated break flow, 

accumulators integrated break flow, HPIS integrated break flow and LPIS integrated break flow.

The pressurizer pressure is shown to know when we are below the setpoint for ECCSs injection. 

Secondary pressure is important when accident management strategies are performed. Core 

collapsed liquid level gives information on core uncovery and cladding temperature about core 

heatup. Finally, masses are shown for RCS inventory, mass inventory discharged through the 

break and mass injected by accumulators, HPIS and LPIS.

3.1 Results of LOCA Break Size Spectrum Without HPIS (noHP)

The results for simulations of LOCA spectrum ranging from 1.27 cm to 30.48 cm with assumed 

loss of HPI system are shown in Figures 3-1 to 3-18. The simulations were performed for 24 h 

(86400 s), if simulation was not terminated earlier due to significant heatup. It is shown that in 

case of loss of HPI system the breaks smaller than 7.62 cm lead to core heatup of average fuel 

rod cladding above 1100 K and/or low primary system mass with LPIS not injecting due to primary 

pressure being above the injection setpoint. This mean, that smaller break LOCAs are DEC for 

selected PWR. For breaks equal or larger 10.16 cm it was shown that cooling through the break 

is sufficient and LPI system has sufficient capacity in long term. This is in accordance with results 

obtained from BETHSY 9.1b experiment [9].

Figure 3-1 Pressurizer Pressure – noHP (1.27 cm to 7.62 cm)

0

2

4

6

8

10

12

14

16

18

0 10000 20000 30000 40000

Pressure (MPa)

Time (s)

sb0.5_noHP

sb1_noHP

sb2_noHP

sb3_noHP
10

Figure 3-2 SG1 Pressure – noHP (1.27 cm to 7.62 cm)

Figure 3-3 Core Collapsed Liquid Level – noHP (1.27 cm to 7.62 cm)

Figure 3-4 Cladding Temperature – noHP (1.27 cm to 7.62 cm)

0

1

2

3

4

5

6

7

8

9

0 10000 20000 30000 40000

Pressure (MPa)

Time (s)

sb0.5_noHP

sb1_noHP

sb2_noHP

sb3_noHP

0

1

2

3

4

0 10000 20000 30000 40000

Level (m)

Time (s)

sb0.5_noHP

sb1_noHP

sb2_noHP

sb3_noHP

0

200

400

600

800

1000

1200

1400

0 10000 20000 30000 40000

Temperature (K)

Time (s)

sb0.5_noHP

sb1_noHP

sb2_noHP

sb3_noHP
11

Figure 3-5 Primary System Mass – noHP (1.27 cm to 7.62 cm)

Figure 3-6 Total Integrated Break Flow – noHP (1.27 cm to 7.62 cm)

Figure 3-7 Total Integrated Accumulator Flow – noHP (1.27 cm to 7.62 cm)

0

20

40

60

80

100

120

140

160

180

0 10000 20000 30000 40000

Mass (t)

Time (s)

sb0.5_noHP

sb1_noHP

sb2_noHP

sb3_noHP

0

1000

2000

3000

4000

5000

0 10000 20000 30000 40000

Mass (t)

Time (s)

sb0.5_noHP

sb1_noHP

sb2_noHP

sb3_noHP

0

10

20

30

40

50

60

70

80

0 10000 20000 30000 40000

Mass (t)

Time (s)

sb0.5_noHP

sb1_noHP

sb2_noHP

sb3_noHP
12

Figure 3-8 Total Integrated HPIS Flow – noHP (1.27 cm to 7.62 cm)

Figure 3-9 Total Integrated LPIS Flow – noHP (1.27 cm to 7.62 cm)

Figure 3-10 Pressurizer Pressure – noHP (10.16 cm to 30.48 cm)

0

2

4

6

8

10

0 10000 20000 30000 40000

Mass (t)

Time (s)

sb0.5_noHP

sb1_noHP

sb2_noHP

sb3_noHP

0

1000

2000

3000

4000

5000

0 10000 20000 30000 40000

Mass (t)

Time (s)

sb0.5_noHP

sb1_noHP

sb2_noHP

sb3_noHP

0

2

4

6

8

10

12

14

16

18

0 20000 40000 60000 80000

Pressure (MPa)

Time (s)

sb4_noHP

sb6_noHP

sb8_noHP

sb12_noHP
13

Figure 3-11 SG1 Pressure – noHP (10.16 cm to 30.48 cm)

Figure 3-12 Core Collapsed Liquid Level – noHP (10.16 cm to 30.48 cm)

Figure 3-13 Cladding Temperature – noHP (10.16 cm to 30.48 cm)

0

1

2

3

4

5

6

7

8

9

0 20000 40000 60000 80000

Pressure (MPa)

Time (s)

sb4_noHP

sb6_noHP

sb8_noHP

sb12_noHP

0

1

2

3

4

0 20000 40000 60000 80000

Level (m)

Time (s)

sb4_noHP

sb6_noHP

sb8_noHP

sb12_noHP

0

100

200

300

400

500

600

700

800

0 20000 40000 60000 80000

Temperature (K)

Time (s)

sb4_noHP

sb6_noHP

sb8_noHP

sb12_noHP
14

Figure 3-14 Primary System Mass – noHP (10.16 cm to 30.48 cm)

Figure 3-15 Total Integrated Break Flow – noHP (10.16 cm to 30.48 cm)

Figure 3-16 Total Integrated Accumulator Flow – noHP (10.16 cm to 30.48 cm)

0

20

40

60

80

100

120

140

160

0 20000 40000 60000 80000

Mass (t)

Time (s)

sb4_noHP

sb6_noHP

sb8_noHP

sb12_noHP

0

5000

10000

15000

20000

25000

0 20000 40000 60000 80000

Mass (t)

Time (s)

sb4_noHP

sb6_noHP

sb8_noHP

sb12_noHP

0

10

20

30

40

50

60

70

80

0 20000 40000 60000 80000

Mass (t)

Time (s)

sb4_noHP

sb6_noHP

sb8_noHP

sb12_noHP
15

Figure 3-17 Total Integrated HPIS Flow – noHP (10.16 cm to 30.48 cm)

Figure 3-18 Total Integrated LPIS Flow – noHP (10.16 cm to 30.48 cm)

3.2 Results of LOCA Break Size Spectrum Without LPIS (noLP)

The results for simulations of LOCA spectrum ranging from 1.27 cm to 30.48 cm with assumed 

loss of LPI system are shown in Figures 3-19 to 3-36. The simulation time was 24 h (86400 s).

It is shown that in case of loss of LPI system the HPI system has sufficient capacity in long term 

for all break sizes for the selected PWR.

0

2

4

6

8

10

0 20000 40000 60000 80000

Mass (t)

Time (s)

sb4_noHP

sb6_noHP

sb8_noHP

sb12_noHP

0

5000

10000

15000

20000

25000

0 20000 40000 60000 80000

Mass (t)

Time (s)

sb4_noHP

sb6_noHP

sb8_noHP

sb12_noHP
16

Figure 3-19 Pressurizer Pressure – noLP (1.27 cm to 7.62 cm)

Figure 3-20 SG1 Pressure – noLP (1.27 cm to 7.62 cm)

Figure 3-21 Core Collapsed Liquid Level – noLP (1.27 cm to 7.62 cm)

0

2

4

6

8

10

12

14

16

18

0 20000 40000 60000 80000

Pressure (MPa)

Time (s)

sb0.5_noLP

sb1_noLP

sb2_noLP

sb3_noLP

0

1

2

3

4

5

6

7

8

9

0 20000 40000 60000 80000

Pressure (MPa)

Time (s)

sb0.5_noLP

sb1_noLP

sb2_noLP

sb3_noLP

0

1

2

3

4

0 20000 40000 60000 80000

Level (m)

Time (s)

sb0.5_noLP

sb1_noLP

sb2_noLP

sb3_noLP
17

Figure 3-22 Cladding Temperature – noLP (1.27 cm to 7.62 cm)

Figure 3-23 Primary System Mass – noLP (1.27 cm to 7.62 cm)

Figure 3-24 Total Integrated Break Flow – noLP (1.27 cm to 7.62 cm)

0

100

200

300

400

500

600

700

0 20000 40000 60000 80000

Temperature (K)

Time (s)

sb0.5_noLP

sb1_noLP

sb2_noLP

sb3_noLP

0

50

100

150

200

0 20000 40000 60000 80000

Mass (t)

Time (s)

sb0.5_noLP

sb1_noLP

sb2_noLP

sb3_noLP

0

1000

2000

3000

4000

5000

6000

7000

0 20000 40000 60000 80000

Mass (t)

Time (s)

sb0.5_noLP

sb1_noLP

sb2_noLP

sb3_noLP
18

Figure 3-25 Total Integrated Accumulator Flow – noLP (1.27 cm to 7.62 cm)

Figure 3-26 Total Integrated HPIS Flow – noLP (1.27 cm to 7.62 cm)

Figure 3-27 Total Integrated LPIS Flow – noLP (1.27 cm to 7.62 cm)

0

10

20

30

40

50

60

70

80

0 20000 40000 60000 80000

Mass (t)

Time (s)

sb0.5_noLP

sb1_noLP

sb2_noLP

sb3_noLP

0

1000

2000

3000

4000

5000

6000

7000

0 20000 40000 60000 80000

Mass (t)

Time (s)

sb0.5_noLP

sb1_noLP

sb2_noLP

sb3_noLP

0

1000

2000

3000

4000

5000

0 20000 40000 60000 80000

Mass (t)

Time (s)

sb0.5_noLP

sb1_noLP

sb2_noLP

sb3_noLP
19

Figure 3-28 Pressurizer Pressure – noLP (10.16 cm to 30.48 cm)

Figure 3-29 SG1 Pressure – noLP (10.16 cm to 30.48 cm)

Figure 3-30 Core Collapsed Liquid Level – noLP (10.16 cm to 30.48 cm)

0

2

4

6

8

10

12

14

16

18

0 20000 40000 60000 80000

Pressure (MPa)

Time (s)

sb4_noLP

sb6_noLP

sb8_noLP

sb12_noLP

0

1

2

3

4

5

6

7

8

9

0 20000 40000 60000 80000

Pressure (MPa)

Time (s)

sb4_noLP

sb6_noLP

sb8_noLP

sb12_noLP

0

1

2

3

4

0 20000 40000 60000 80000

Level (m)

Time (s)

sb4_noLP

sb6_noLP

sb8_noLP

sb12_noLP
20

Figure 3-31 Cladding Temperature – noLP (10.16 cm to 30.48 cm)

Figure 3-32 Primary System Mass – noLP (10.16 cm to 30.48 cm)

Figure 3-33 Total Integrated Break Flow – noLP (10.16 cm to 30.48 cm)

300

400

500

600

700

800

0 20000 40000 60000 80000

Temperature (K)

Time (s)

sb4_noLP

sb6_noLP

sb8_noLP

sb12_noLP

0

20

40

60

80

100

120

140

160

0 20000 40000 60000 80000

Mass (t)

Time (s)

sb4_noLP

sb6_noLP

sb8_noLP

sb12_noLP

0

1000

2000

3000

4000

5000

6000

7000

0 20000 40000 60000 80000

Mass (t)

Time (s)

sb4_noLP

sb6_noLP

sb8_noLP

sb12_noLP
21

Figure 3-34 Total Integrated Accumulator Flow – noLP (10.16 cm to 30.48 cm)

Figure 3-35 Total Integrated HPIS Flow – noLP (10.16 cm to 30.48 cm)

Figure 3-36 Total Integrated LPIS Flow – noHP (10.16 cm to 30.48 cm)

0

10

20

30

40

50

60

70

80

0 20000 40000 60000 80000

Mass (t)

Time (s)

sb4_noLP

sb6_noLP

sb8_noLP

sb12_noLP

0

1000

2000

3000

4000

5000

6000

7000

0 20000 40000 60000 80000

Mass (t)

Time (s)

sb4_noLP

sb6_noLP

sb8_noLP

sb12_noLP

0

2000

4000

6000

8000

10000

12000

0 20000 40000 60000 80000

Mass (t)

Time (s)

sb4_noLP

sb6_noLP

sb8_noLP

sb12_noLP
22

3.3 Results of LOCA Break Size Spectrum 1.27 cm to 5.08 cm Without HPIS and 

Using AM1 Strategy (noHP and AM1)

In section 3.1 it is shown that in case of loss of HPI system the breaks smaller than 7.62 cm lead 

to final core uncovery, because primaty pressure is above injection setpoint of the accumulators 

and LPI system. Therefore AM1 accident management strategy to reduce the primary pressure 

by one PRZ PORV has been assumed in the simulations. The results are shown in Figures 3-37

to 3-45. As can be seen the strategy AM1 using one PRZ PORV to depressurize the primary 

system has not been successful for break spectrum from 1.27 cm to 5.08 cm.

Figure 3-37 Pressurizer Pressure – noHP and AM1 (1.27 cm to 5.08 cm)

Figure 3-38 SG1 Pressure – noHP and AM1 (1.27 cm to 5.08 cm)

0

2

4

6

8

10

12

14

16

18

0 2000 4000 6000 8000 10000

Pressure (MPa)

Time (s)

sb0.5_noHPd

sb1_noHPd

sb2_noHPd

0

1

2

3

4

5

6

7

8

9

0 2000 4000 6000 8000 10000

Pressure (MPa)

Time (s)

sb0.5_noHPd

sb1_noHPd

sb2_noHPd
23

Figure 3-39 Core Collapsed Liquid Level – noHP and AM1 (1.27 cm to 5.08 cm)

Figure 3-40 Cladding Temperature – noHP and AM1 (1.27 cm to 5.08 cm)

Figure 3-41 Primary System Mass – noHP and AM1 (1.27 cm to 5.08 cm)

0

0.5

1

1.5

2

2.5

3

3.5

4

0 2000 4000 6000 8000 10000

Level (m)

Time (s)

sb0.5_noHPd

sb1_noHPd

sb2_noHPd

0

200

400

600

800

1000

1200

1400

0 2000 4000 6000 8000 10000

Temperature (K)

Time (s)

sb0.5_noHPd

sb1_noHPd

sb2_noHPd

0

50

100

150

200

0 2000 4000 6000 8000 10000

Mass (t)

Time (s)

sb0.5_noHPd

sb1_noHPd

sb2_noHPd
24

Figure 3-42 Total Integrated Break Flow – noHP and AM1 (1.27 cm to 5.08 cm)

Figure 3-43 Total Integrated Accumulator Flow – noHP and AM1 (1.27 cm to 5.08 cm)

Figure 3-44 Total Integrated HPIS Flow – noHP and AM1 (1.27 cm to 5.08 cm)

0

50

100

150

200

0 2000 4000 6000 8000 10000

Mass (t)

Time (s)

sb0.5_noHPd

sb1_noHPd

sb2_noHPd

0

10

20

30

40

50

60

70

80

0 2000 4000 6000 8000 10000

Mass (t)

Time (s)

sb0.5_noHPd

sb1_noHPd

sb2_noHPd

0

2

4

6

8

10

0 2000 4000 6000 8000 10000

Mass (t)

Time (s)

sb0.5_noHPd

sb1_noHPd

sb2_noHPd
25

Figure 3-45 Total integrated LPIS flow – noHP and AM1 (1.27 cm to 5.08 cm)

3.4 Results of LOCA Break Size Spectrum 1.27 cm to 5.08 cm With One ECCS 

and Using AM2 Strategy (noHP and AM2)

In section 3.1 it is shown that in case of loss of HPI system the breaks smaller than 7.62 cm lead 

to final core uncovery, because the primary pressure is above the setpoints to inject with the 

accumulators and LPI system. Therefore AM2 accident management strategy reducing the 

primary pressure through the secondary side has been assumed in simulations. The simulation 

time was 24 h (86400 s). The steam generators were depressurized to 1.1 MPa. The results are 

shown in Figures 3-46 to 3-54. As can be seen the AM2 strategy has been successful for breaks

1.27 cm and 2.54 cm, while in the case of 5.08 cm break it was not successful. In the long term it 

can be seen that the primary system was not sufficiently depressurized to enable LPI system 

injection.

Figure 3-46 Pressurizer Pressure – noHP and AM2 (1.27 cm to 5.08 cm)

0

1000

2000

3000

4000

5000

0 2000 4000 6000 8000 10000

Mass (t)

Time (s)

sb0.5_noHPd

sb1_noHPd

sb2_noHPd

0

2

4

6

8

10

12

14

16

18

0 20000 40000 60000 80000

Pressure (MPa)

Time (s)

sb0.5_noHPdsg

sb1_noHPdsg

sb2_noHPdsg
26

Figure 3-47 SG1 Pressure – noHP and AM2 (1.27 cm to 5.08 cm)

Figure 3-48 Core Collapsed Liquid Level – noHP and AM2 (1.27 cm to 5.08 cm)

Figure 3-49 Cladding Temperature – noHP and AM2 (1.27 cm to 5.08 cm)

0

1

2

3

4

5

6

7

8

9

0 20000 40000 60000 80000

Pressure (MPa)

Time (s)

sb0.5_noHPdsg

sb1_noHPdsg

sb2_noHPdsg

0

1

2

3

4

0 20000 40000 60000 80000

Level (m)

Time (s)

sb0.5_noHPdsg

sb1_noHPdsg

sb2_noHPdsg

400

600

800

1000

1200

1400

1600

0 20000 40000 60000 80000

Temperature (K)

Time (s)

sb0.5_noHPdsg

sb1_noHPdsg

sb2_noHPdsg
27

Figure 3-50 Primary System Mass – noHP and AM2 (1.27 cm to 5.08 cm)

Figure 3-51 Total Integrated Break Flow – noHP and AM2 (1.27 cm to 5.08 cm)

Figure 3-52 Total Integrated Accumulator Flow – noHP and AM2 (1.27 cm to 5.08 cm)

0

20

40

60

80

100

120

140

160

0 20000 40000 60000 80000

Mass (t)

Time (s)

sb0.5_noHPdsg

sb1_noHPdsg

sb2_noHPdsg

0

50

100

150

200

250

0 20000 40000 60000 80000

Mass (t)

Time (s)

sb0.5_noHPdsg

sb1_noHPdsg

sb2_noHPdsg

0

10

20

30

40

50

60

70

80

0 20000 40000 60000 80000

Mass (t)

Time (s)

sb0.5_noHPdsg

sb1_noHPdsg

sb2_noHPdsg
28

Figure 3-53 Total Integrated HPIS Flow – noHP and AM2 (1.27 cm to 5.08 cm)

Figure 3-54 Total Integrated LPIS Flow – noHP and AM2 (1.27 cm to 5.08 cm)

3.5 Results of LOCA Break Size Spectrum 1.27 cm to 5.08 cm With One ECCS 

and Using AM3 Strategy (noHP and AM3)

In previous section 3.4 it was shown that the primary system was not sufficiently depressurized 

to enable LPI system injection. Therefore the steam generators were depressurized to 0.7 MPa. 

Again accident management strategy reducing the primary pressure through the secondary side 

has been assumed in simulations. The simulation time was 24 h (86400 s). The results are shown 

in Figures 3-55 to 3-63. As can be seen the strategy has been successful for breaks 1.27 cm and 

2.54 cm, and also 5.08 cm break with initial heatup. Short term initial peak in cladding temperature 

occurred only in the case of 5.08 cm break size LOCA. The primary system was sufficiently 

depressurized to enable LPI system injection in long term.

0

2

4

6

8

10

0 20000 40000 60000 80000

Mass (t)

Time (s)

sb0.5_noHPdsg

sb1_noHPdsg

sb2_noHPdsg

0

5000

10000

15000

20000

25000

0 20000 40000 60000 80000

Mass (t)

Time (s)

sb0.5_noHPdsg

sb1_noHPdsg

sb2_noHPdsg
29

Figure 3-55 Pressurizer Pressure – noHP and AM3 (1.27 cm to 5.08 cm)

Figure 3-56 SG1 Pressure – noHP and AM3 (1.27 cm to 5.08 cm)

Figure 3-57 Core Collapsed Liquid Level – noHP and AM3 (1.27 cm to 5.08 cm)

0

2

4

6

8

10

12

14

16

18

0 20000 40000 60000 80000

Pressure (MPa)

Time (s)

sb0.5_noHPdsg7

sb1_noHPdsg7

sb2_noHPdsg7

0

1

2

3

4

5

6

7

8

9

0 20000 40000 60000 80000

Pressure (MPa)

Time (s)

sb0.5_noHPdsg7

sb1_noHPdsg7

sb2_noHPdsg7

0

1

2

3

4

0 20000 40000 60000 80000

Level (m)

Time (s)

sb0.5_noHPdsg7

sb1_noHPdsg7

sb2_noHPdsg7
30

Figure 3-58 Cladding Temperature – noHP and AM3 (1.27 cm to 5.08 cm)

Figure 3-59 Primary System Mass – noHP and AM3 (1.27 cm to 5.08 cm)

Figure 3-60 Total Integrated Break Flow – noHP and AM3 (1.27 cm to 5.08 cm)

300

600

900

1200

1500

0 20000 40000 60000 80000

Temperature (K)

Time (s)

sb0.5_noHPdsg7

sb1_noHPdsg7

sb2_noHPdsg7

0

20

40

60

80

100

120

140

160

180

200

0 20000 40000 60000 80000

Mass (t)

Time (s)

sb0.5_noHPdsg7

sb1_noHPdsg7

sb2_noHPdsg7

0

2000

4000

6000

8000

10000

0 20000 40000 60000 80000

Mass (t)

Time (s)

sb0.5_noHPdsg7

sb1_noHPdsg7

sb2_noHPdsg7
31

Figure 3-61 Total Integrated Accumulator Flow – noHP and AM3 (1.27 cm to 5.08 cm)

Figure 3-62 Total Integrated HPIS Flow – noHP and AM3 (1.27 cm to 5.08 cm)

Figure 3-63 Total Integrated LPIS Flow – noHP and AM3 (1.27 cm to 5.08 cm)

0

10

20

30

40

50

60

70

80

0 20000 40000 60000 80000

Mass (t)

Time (s)

sb0.5_noHPdsg7

sb1_noHPdsg7

sb2_noHPdsg7

0

2

4

6

8

10

0 20000 40000 60000 80000

Mass (t)

Time (s)

sb0.5_noHPdsg7

sb1_noHPdsg7

sb2_noHPdsg7

0

2000

4000

6000

8000

10000

0 20000 40000 60000 80000

Mass (t)

Time (s)

sb0.5_noHPdsg7

sb1_noHPdsg7

sb2_noHPdsg7

33

4 CONCLUSIONS

The RELAP5/MOD3.3 Patch05 computer code calculations of loss of coolant accident (LOCA)

without high pressure injection (HPI) system and LOCA without low pressure injection (LPI)

system for a spectrum of break sizes from 1.27 cm (0.5 inch) to 30.48 cm (12 inch) have been 

performed. When HPI system is not available, depressurization of the primary side through the 

secondary side can be used in case of smaller breaks to mitigate consequences. For larger 

breaks, the pressure drop is faster and LPI systems maintained the core inventory. When LPI 

system is not available, the HPI system has sufficient capacity for safety injection in case of 

simulated LOCA break spectrum. It was shown that accident management strategy using steam 

generator PORVs to depressurize the primary system below LPIS injection setpoint are 

sufficient to mitigate smaller size LOCA without HPI systems operable. However, the control of 

design extension conditions (DEC) is expected to be achieved primarily by features 

implemented in the design (safety features for DECs) and not only by accident management 

measures that are using equipment designed for other purposes. This means that in principle a 

DEC is such if its consideration in the design leads to the need of additional equipment or to an 

upgraded classification of lower class equipment to mitigate the DEC. Therefore it can be 

concluded that safety injection pump with sufficient pressure capability should be introduced to 

prevent smaller break LOCAs in case HPI system is lost.

35

5 REFERENCES

[1] European Utility Requirements, Volumes 1 and 2, Rev C., April 2001.

[2] WENRA, WENRA Reactor Safety Reference Levels, January 2007.

[3] International Atomic Energy Agency (IAEA), Safety of Nuclear Power Plants: Design,

Specific Safety Requirements No. SSR-2/1, January 2012.

[4] WENRA, WENRA Safety Reference Levels for Existing Reactors, September 2014.

[5] WENRA, Guidance Document Issue F: Design Extension of Existing Reactors, 29

September, 2014.

[6] International Atomic Energy Agency (IAEA), Safety of Nuclear Power Plants: Design,

Specific Safety Requirements No. SSR-2/1 (Rev. 1), February 2016.

[7] A. Annunziato, H. Glaeser, J. Lillington, P. Marsili, C. Renault, A. Sjöberg, CSNI Integral

Test Facility Validation Matrix for the Assessment of Thermal-Hydraulic Codes for LWR

LOCA and Transients, NEA/CSNI/R(96)17, July 1996.

[8] A. Prošek, O. A. Berar, Advanced Presentation of BETHSY 6.2TC Test Results

Calculated by RELAP5 and TRACE, Science and Technology of Nuclear Installations,

Vol. 2012, Article ID 812130, 15 pages, 2012.

[9] A. Prošek, RELAP5 Calculations of Bethsy 9.1b Test, Science and Technology of

Nuclear Installations, Vol. 2012, Article ID 238090, 11 pages, 2012.

[10] H. Kumamaru, Y. Kukita, PWR Cold-Leg Small-Break LOCA with Total HPI Failure,

Journal of Nuclear Science and Technology, Vol. 29, No. 12, pp. 1162-1172, 1992.

[11] H. Asaka, Y. Kukita, Intentional Depressurization of Steam Generator Secondary Side

during a PWR Small-Break Loss-of-Coolant Accident, Journal of Nuclear Science and

Technology, Vol. 32, No. 2, 1995.

[12] Rules on radiation and nuclear safety factors (UNOFFICIAL TRANSLATION of the

original published in Slovene languauge in the Official Gazette of the Republic of

Slovenia, No. 74/2016).

[13] International Atomic Energy Agency (IAEA), Considerations on the Application of the

IAEA Safety Requirements for the Design of Nuclear Power Plants, IAEA TECDOC No.

1791, Vienna, 2016.

[14] USNRC, RELAP5/MOD3.3 Code Manual, Patch 05, Vols. 1 to 8, Information Systems

Laboratories, Inc. Idaho Falls, Idaho, prepared for United States Nuclear Regulatory

Commission (USNRC), 2016.

[15] Andrej Prošek, RELAP5 and TRACE Calculations of LOCA in PWR, NUREG/IA-0479

(ML17159A590).

LOCAs With Loss of One Active Emergency Cooling System Simulated 

by RELAP5

NUREG/IA-0516

Andrej Prošek

Jožef Stefan Institute

Jamova cesta 39 

SI-1000, Ljubljana, Slovenia 

Division of Systems Analysis

Office of the Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, D.C. 20555-0001

K. Tien, NRC Project Manager

The second generation reactors were designed and built to withstand without loss to the structures, systems,

and components necessary to ensure public health and safety during design basis accidents. There are also 

accident sequences that are possible but were judged to be too unlikely and therefore were not fully

considered in the design process of second generation reactors. In that sense, they were considered beyond 

the scope of design-basis accidents that a nuclear facility must be designed and built to withstand. They were 

called beyond design basis accidents. After Fukushima-Daiichi in the Europe the design extension conditions

were introduced as preferred method for giving due consideration to the complex sequences and severe 

accidents without including them in the design basis conditions. In the study, the analysis was performed to

see if the plant design can prevent spectrum of loss of coolant accidents together with the complete loss of

one emergency core cooling function (e.g. high pressure injection or low pressure injection). The analyzed 

break spectrum ranged from 1.27 cm to 30.48 cm. For calculations the latest RELAP5/MOD3.3 Patch 5 has

been used and the RELAP5 input model of Krško nuclear power plant. 

Accident Management (AM)

Anticipated Transient without Scram (ATWS)

Beyond Design Basis Accidents (BDBA)

Design Extension Conditions (DEC)

Station Blackout (SBO)

Ultimate Operating Procedure (UOP)

Western European Nuclear Regulators Association (WENRA)

January 2020

Technical




NUREG/IA-0516 LOCAs With Loss of One Active Emergency Cooling System Simulated by RELAP5 January 2020