NUREG/IA-0424 

International 

 Agreement Report

RELAP5 Extended Station Blackout Analyses 

 

 

 

Prepared by:

A. Prošek, B. Mavko

 

 

 Jožef Stefan Institute

 Jamova cesta 39

SI-1000 Ljubljana, Slovenia

 

 

A. Calvo, NRC Project Manager

 

 Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

Manuscript Completed: January 2013

Date Published: March 2013

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-0424

International 

 Agreement Report

RELAP5 Extended Station Blackout Analyses

Prepared by:

A. Prošek, B. Mavko

 

 Jožef Stefan Institute

 Jamova cesta 39

SI-1000 Ljubljana, Slovenia

 

A. Calvo, NRC Project Manager

 Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

Manuscript Completed: January 2013

Date Published: March 2013

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

Following the accident at the nuclear power plant Fukushima in Japan the stress test were 

performed in European countries. Considering the stress tests specifications it was necessary to 

evaluate the consequences of loss of safety functions from any initiating event (earthquake or 

flooding) for loss of electrical power, including Station Blackout (SBO), loss of the ultimate heat 

sink or combination of both. In this report long term station blackout sequences for Krško twoloop pressurized water reactor with loss of normal or all secondary side heat sinks were 

performed. For calculations the latest RELAP5/MOD3.3 Patch 04 computer code was used. The 

verified standard RELAP5/MOD3.3 input model delivered by Krško nuclear power plant was 

used.

SBO scenario involves a loss of offsite power, failure of the redundant emergency diesel 

generators, failure of alternate current power restoration and the eventual degradation of the 

reactor coolant pump seals resulting in a long term loss of coolant. In the study different reactor 

coolant pump seal leaks were studied due to SBO. Besides, scenarios were performed for 

different primary side depressurizations performed by operator through the secondary side 

power operated relief valves, providing that turbine driven auxiliary feedwater pump is available. 

Finally, the effect of having some injection into the reactor coolant system was also evaluated. It 

can be concluded that calculated results obtained by RELAP5 give good indication about time 

available before core degradation started. The results suggest that RELAP5 can be used for 

extended SBO studies until core damage started. It is especially useful in studying maintaining 

core cooling function and time available before core uncovers as part of severe accident 

management. The benefit of using RELAP5 is in the fact that best estimate system codes are 

more accurate than severe accident codes in phases before core degradation started.

v

CONTENTS

Page

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

FIGURES....................................................................................................................................vii

TABLES.......................................................................................................................................ix

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

ACKNOWLEDGEMENTS .........................................................................................................xiii

ABBREVIATIONS ......................................................................................................................xv

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

2. PLANT DESCRIPTION...........................................................................................................3

3. RELAP5 INPUT MODEL DESCRIPTION...............................................................................5

4. DESCRIPTION OF SCENARIOS ANALYZED.......................................................................9

5. RESULTS .............................................................................................................................11

5.1 Dependence on depressurization of selected SBO scenarios .....................................11

5.1.1 Scenario S0-21....................................................................................................11

5.1.2 Scenario S0-50....................................................................................................15

5.1.3 Scenario S0-75....................................................................................................18

5.1.4 Scenario S0-100..................................................................................................22

5.1.5 Scenario S0-150..................................................................................................25

5.1.6 Scenario S0-150p................................................................................................29

5.1.7 Scenario S0-300..................................................................................................32

5.1.8 Scenario S1-21....................................................................................................36

5.1.9 Scenario S1-21v1................................................................................................39

5.1.10 Scenario S1-21v2................................................................................................43

5.1.11 Scenario S1-21p..................................................................................................46

5.2 Dependence on RCPs seal leak of SBO scenarios with TD AFW available................50

5.2.1 S0 scenarios depressurized to 1.33 MPa (12.5 kp/cm2 gauge) ..........................50

5.2.2 S0 scenarios depressurized to 1.57 MPa (15 kp/cm2 gauge) .............................53

5.2.3 S0 scenarios depressurized to 2.06 MPa (20 kp/cm2 gauge) .............................56

5.2.4 S0 scenarios depressurized to 2.55 MPa (25 kp/cm2 gauge) .............................59

5.3 Dependence on equipment operation of SBO scenario with TD AFW lost 4 hours after 

transient start ...............................................................................................................62

5.3.1 S1 scenarios depressurized to 1.33 MPa (12.5 kp/cm2 gauge) ..........................62

5.3.2 S1 scenarios depressurized to 1.57 MPa (15 kp/cm2 gauge) .............................65
vi

5.3.3 S1 scenarios depressurized to 2.06 MPa (20 kp/cm2 gauge) .............................67

5.3.4 S1 scenarios depressurized to 2.55 MPa (25 kp/cm2 gauge) .............................70

6. RUN STATISTICS ................................................................................................................73

7. CONCLUSIONS....................................................................................................................75

8. REFERENCES......................................................................................................................77
vii

FIGURES

Page

Figure 1 Krško NPP base nodalization scheme – SNAP hydraulics component view .............6

Figure 2 Krško NPP modified nodalization scheme – SNAP hydraulics component view........7

Figure 3 RCS pressure – scenario S0-21...............................................................................12

Figure 4 Core exit temperature – scenario S0-21...................................................................12

Figure 5 Core collapsed liquid level – scenario S0-21............................................................13

Figure 6 Average fuel cladding temperature – scenario S0-21 ..............................................13

Figure 7 Total mass discharged from RCS – scenario S0-21 ................................................14

Figure 8 Mass injected into RCS – scenario S0-21................................................................14

Figure 9 RCS pressure – scenario S0-50...............................................................................15

Figure 10 Core exit temperature – scenario S0-50...................................................................16

Figure 11 Core collapsed liquid level – scenario S0-50............................................................16

Figure 12 Average fuel cladding temperature – scenario S0-50 ..............................................17

Figure 13 Total mass discharged from RCS – scenario S0-50 ................................................17

Figure 14 Mass injected into RCS – scenario S0-50................................................................18

Figure 15 RCS pressure – scenario S0-75...............................................................................19

Figure 16 Core exit temperature – scenario S0-75...................................................................19

Figure 17 Core collapsed liquid level – scenario S0-75............................................................20

Figure 18 Average fuel cladding temperature – scenario S0-75 ..............................................20

Figure 19 Total mass discharged from RCS – scenario S0-75 ................................................21

Figure 20 Mass injected into RCS – scenario S0-75................................................................21

Figure 21 RCS pressure – scenario S0-100.............................................................................22

Figure 22 Core exit temperature – scenario S0-100.................................................................23

Figure 23 Core collapsed liquid level – scenario S0-100..........................................................23

Figure 24 Average fuel cladding temperature – scenario S0-100 ............................................24

Figure 25 Total mass discharged from RCS – scenario S0-100 ..............................................24

Figure 26 Mass injected into RCS – scenario S0-100..............................................................25

Figure 27 RCS pressure – scenario S0-150.............................................................................26

Figure 28 Core exit temperature – scenario S0-150.................................................................26

Figure 29 Core collapsed liquid level – scenario S0-150..........................................................27

Figure 30 Average fuel cladding temperature – scenario S0-150 ............................................27

Figure 31 Total mass discharged from RCS – scenario S0-150 ..............................................28

Figure 32 Mass injected into RCS – scenario S0-150..............................................................28

Figure 33 RCS pressure – scenario S0-150p...........................................................................29

Figure 34 Core exit temperature – scenario S0-150p...............................................................30

Figure 35 Core collapsed liquid level – scenario S0-150p........................................................30

Figure 36 Average fuel cladding temperature – scenario S0-150p ..........................................31

Figure 37 Total mass discharged from RCS – scenario S0-150p ............................................31

Figure 38 Mass injected into RCS – scenario S0-150p............................................................32

Figure 39 RCS pressure – scenario S0-300.............................................................................33

Figure 40 Core exit temperature – scenario S0-300.................................................................33

Figure 41 Core collapsed liquid level – scenario S0-300..........................................................34

Figure 42 Average fuel cladding temperature – scenario S0-300 ............................................34

Figure 43 Total mass discharged from RCS – scenario S0-300 ..............................................35

Figure 44 Mass injected into RCS – scenario S0-300..............................................................35

Figure 45 RCS pressure – scenario S1-21...............................................................................36

Figure 46 Core exit temperature – scenario S1-21...................................................................37

Figure 47 Core collapsed liquid level – scenario S1-21............................................................37
viii

Figure 48 Average fuel cladding temperature – scenario S1-21 ..............................................38

Figure 49 Total mass discharged from RCS – scenario S1-21 ................................................38

Figure 50 Mass injected into RCS – scenario S1-21................................................................39

Figure 51 RCS pressure – scenario S1-21v1...........................................................................40

Figure 52 Core exit temperature – scenario S1-21v1...............................................................40

Figure 53 Core collapsed liquid level – scenario S1-21v1........................................................41

Figure 54 Average fuel cladding temperature – scenario S1-21v1...........................................41

Figure 55 Total mass discharged from RCS – scenario S1-21v1.............................................42

Figure 56 Mass injected into RCS – scenario S1-21v1 ............................................................42

Figure 57 RCS pressure – scenario S1-21v2...........................................................................43

Figure 58 Core exit temperature – scenario S1-21v2...............................................................44

Figure 59 Core collapsed liquid level – scenario S1-21v2........................................................44

Figure 60 Average fuel cladding temperature – scenario S1-21v2...........................................45

Figure 61 Total mass discharged from RCS – scenario S1-21v2.............................................45

Figure 62 Mass injected into RCS – scenario S1-21v2 ............................................................46

Figure 63 RCS pressure – scenario S1-21p.............................................................................47

Figure 64 Core exit temperature – scenario S1-21p.................................................................47

Figure 65 Core level – scenario S1-21p ...................................................................................48

Figure 66 Average fuel cladding temperature – scenario S1-21p ............................................48

Figure 67 Total mass discharged from RCS – scenario S1-21p ..............................................49

Figure 68 Mass injected into RCS – scenario S1-21p..............................................................49

Figure 69 RCS pressure – S0 scenarios (1.33 MPa) ...............................................................50

Figure 70 SG no.1 pressure – S0 scenarios (1.33 MPa)..........................................................51

Figure 71 Core collapsed liquid level – S0 scenarios (1.33 MPa) ............................................51

Figure 72 Average fuel cladding temperature – S0 scenarios (1.33 MPa)...............................52

Figure 73 RCS mass inventory – S0 scenarios (1.33 MPa) .....................................................52

Figure 74 RCS pressure – S0 scenarios (1.57 MPa) ...............................................................53

Figure 75 SG no.1 pressure – S0 scenarios (1.57 MPa)..........................................................54

Figure 76 Core collapsed liquid level – S0 scenarios (1.57 MPa) ............................................54

Figure 77 Average fuel cladding temperature – S0 scenarios (1.57 MPa)...............................55

Figure 78 RCS mass inventory – S0 scenarios (1.57 MPa) .....................................................55

Figure 79 RCS pressure – S0 scenarios (2.06 MPa) ...............................................................56

Figure 80 SG no.1 pressure – S0 scenarios (2.06 MPa)..........................................................57

Figure 81 Core collapsed liquid level – S0 scenarios (2.06 MPa) ............................................57

Figure 82 Average fuel cladding temperature – S0 scenarios (2.06 MPa)...............................58

Figure 83 RCS mass inventory – S0 scenarios (2.06 MPa) .....................................................58

Figure 84 RCS pressure – S0 scenarios (2.55 MPa) ...............................................................59

Figure 85 SG no.1 pressure – S0 scenarios (2.55 MPa)..........................................................60

Figure 86 Core collapsed liquid level – S0 scenarios (2.55 MPa) ............................................60

Figure 87 Average fuel cladding temperature – S0 scenarios (2.55 MPa)...............................61

Figure 88 RCS mass inventory – S0 scenarios (2.55 MPa) .....................................................61

Figure 89 RCS pressure – S1 scenarios (1.33 MPa) ...............................................................63

Figure 90 RCS mass inventory – S1 scenarios (1.33 MPa) .....................................................63

Figure 91 Average fuel cladding temperature – S1 scenarios (1.33 MPa)...............................64

Figure 92 Total mass discharged from RCS – S1 scenarios (1.33 MPa).................................64

Figure 93 RCS pressure – S1 scenarios (1.57 MPa) ...............................................................65

Figure 94 RCS mass inventory – S1 scenarios (1.57 MPa) .....................................................66

Figure 95 Average fuel cladding temperature – S1 scenarios (1.57 MPa)...............................66

Figure 96 Total mass discharged from RCS – S1 scenarios (1.57 MPa).................................67

Figure 97 RCS pressure – S1 scenarios (2.06 MPa) ...............................................................68

Figure 98 RCS mass inventory – S1 scenarios (2.06 MPa) .....................................................68
ix

Figure 99 Average fuel cladding temperature – S1 scenarios (2.06 MPa)...............................69

Figure 100 Total mass discharged from RCS – S1 scenarios (2.06 MPa).................................69

Figure 101 RCS pressure – S1 scenarios (2.55 MPa) ...............................................................70

Figure 102 RCS mass inventory – S1 scenarios (2.55 MPa) .....................................................71

Figure 103 Average fuel cladding temperature – S1 scenarios (2.55 MPa)...............................71

Figure 104 Total mass discharged from RCS – S1 scenarios (2.55 MPa).................................72

TABLES

Page

Table 1 Set of scenarios analyzed for each of four depressurization cases .........................10

Table 2 Run statistics ............................................................................................................73

xi

EXECUTIVE SUMMARY

Following the accident at the nuclear power plant Fukushima in Japan the stress test were 

performed in European countries. Considering the stress tests specifications it was necessary to 

evaluate the consequences of loss of safety functions from any initiating event (earthquake or 

flooding) for loss of electrical power, including Station Blackout (SBO), loss of the ultimate heat 

sink or combination of both. In this report long term station blackout sequences for Krško twoloop pressurized water reactor with loss of normal or all secondary side heat sinks were 

performed. For calculations the latest RELAP5/MOD3.3 Patch 04 computer code was used. The 

verified standard RELAP5/MOD3.3 input model delivered by Krško nuclear power plant was 

used.

SBO scenario involves a loss of offsite power, failure of the redundant emergency diesel 

generators, failure of alternate current (AC) power restoration and the eventual degradation of 

the reactor coolant pump seals resulting in a long term loss of coolant. It is assumed that AC 

power exists only on the AC buses powered by inverters connected to the station batteries. 

Loss of all AC power results in unavailability of all normal electrical equipment and most of the 

safety electrical equipment. The only possible corrective actions are reactor trip and residual 

heat removal using steam generator safety and relief valves and turbine (steam) driven auxiliary 

feedwater pump.

In the study six different seal leaks per reactor coolant pump were studied due to SBO, ranging 

from 9.4 cm2 to 148.8 cm2

. These breaks represent leaks from 1.32 l/s (21 gpm) to 18.93 l/s 

(300 gpm) at nominal conditions. The same sizes of breaks on both reactor coolant pumps were 

assumed. Besides, scenarios were performed for four different primary side depressurizations 

performed by operator through the secondary side power operated relief valves, providing that 

turbine driven auxiliary feedwater is available. Finally, the effect of having some injection into 

the reactor coolant system was also evaluated (e.g. if positive displacement charging pump 

would be powered from mobile diesel generator). It can be concluded that calculated results 

obtained by RELAP5 give good indication about time available before core degradation started. 

The results suggest that best estimate system codes like RELAP5 can be used for extended 

SBO studies until core damage started. It is especially useful in studying maintaining core 

cooling function and time available before core uncovers as part of severe accident 

management. The benefit of using RELAP5 is in the fact that best estimate system codes are 

more accurate than severe accident codes in phases before core degradation started.

xiii

ACKNOWLEDGEMENTS

The authors acknowledge the financial support from Krško Nuclear Power Plant and Slovenian 

Nuclear Safety Administration within CAMP program (project no. POG-3473) and from the state 

budget by the Slovenian Research Agency (program no. P2-0026).

xv

ABBREVIATIONS

ACC accumulator

AFW auxiliary feedwater

CPU central processing unit

CVCS chemical and volume control system

ECCS emergency core cooling system

HPSI high-pressure safety injection

LD letdown

LOCA loss of coolant accident

LPSI low-pressure safety injection

MAAP Modular Accident Analysis Program

MD motor driven

MFW main feedwater

MSIV main steam isolation valve

NEK Krško nuclear power plant (in Slovene: Nuklearna elektrarna Krško)

NPP nuclear power plant

PORV power operated relief valve

PRZ pressurizer

PWR pressurized water reactor

RCP reactor coolant pump

RCS reactor coolant system

RPV reactor protection vessel

RELAP Reactor Excursion and Leak Analysis Program

SG steam generator

SL surge line

TD turbine driven

USAR updated safety analysis report

1

1. INTRODUCTION

Following the accident at the nuclear power plant Fukushima in Japan the “stress tests” were 

performed in European countries (Ref. 1). Considering the stress tests specifications it was 

necessary to evaluate the consequences of loss of safety functions from any initiating event 

(earthquake or flooding) for loss of electrical power, including Station Blackout (SBO), loss of 

the ultimate heat sink or combination of both. SBO scenario involves a loss of offsite power, 

failure of the redundant emergency diesel generators, failure of alternate current (AC) power 

restoration and the eventual degradation of the reactor coolant pump seals resulting in a long 

term loss of coolant. It is assumed that AC power exists only on the AC buses powered by 

inverters connected to the station batteries. Loss of all AC power results in unavailability of all 

normal electrical equipment and most of the safety electrical equipment. The only possible 

corrective actions are reactor trip and residual heat removal using steam generator safety 

and relief valves and turbine (steam) driven auxiliary feedwater pump.

“Stress tests” had to be performed also for Krško nuclear power plant (NPP), which is a twoloop pressurized water reactor. Normally such long scenarios are simulated with severe 

accident codes. For example, in the case of Krško NPP the Modular Accident Analysis 

Program (MAAP) Version 4.0.5 was used to analyze long term Station Blackout (SBO) 

accident sequences in 2011 (Ref. 2). The study presented analyses performed by MAAP in 

which the operator action was used to rapidly depressurize the secondary side to 2.1 MPa 

(20 kp/cm2 gauge) and then maintain this pressure. Secondary side depressurization leads to 

primary side cooldown and depressurization. The calculations performed by MAAP were 

performed for different break sizes of reactor coolant pump (RCP) seals. Namely, following 

the loss of all AC power the RCP seals would lose their cooling support systems (the RCP 

seal injection flow and component cooling water to the RCP thermal barrier heat exchanger 

would be unavailable) and would undergo a severe thermal transient. The MAAP long term 

SBO accident sequences were analyzed with the focus on the containment response after 

the core damage.

On the opposite, the focus in this study is to evaluate the plant response before the core 

degradation. Please note that base case MAAP scenario demonstrated that with turbine 

driven auxiliary feedwater (TD AFW) pump available and by operator depressurization of 

primary system to 2.1 MPa the core damage can be prevented for the first seven days. Same 

scenario could be performed by RELAP5, which is best estimate system code for design 

basis transient and accidents, while the MAAP code is intended for simulations of severe 

accidents. It is expected that the calculations by RELAP5 could therefore be performed with 

smaller uncertainties than those with MAAP code. Performing RELAP5 calculations direct 

code to code comparison could be done.

For example, in the study (Ref. 3) describing the comparison of the SBO results obtained by 

MAAP4 and CENTS computer codes indicates that: (1) the overall trends of key parameters 

are similar, and (2) there are differences in the timing of significant occurrences (e.g., SG 

dryout, core uncovery). Nevertheless it is stated that although the timings and durations of 

key occurrences and actuations vary, MAAP4 predictions of core uncovery tend to be 

conservatively biased. They concluded that the simplified single phase natural circulation 

model utilized by MAAP4 drives differences in the thermal hydraulic response of the SGs as 

well as the RCS. They also pointed out that older generations of MAAP4, such as MAAP 

4.0.5, have been known to skew the RCS pressure responses for feed and bleed and SBO 
2

transients. MAAP versions 4.0.6 and beyond include a number of enhancements that yield 

more consistent results for the pressure traces of the SBO and the feed and bleed transients. 

It is recommended that the user be cautious in the selection of the code version that is 

employed for SBO event analysis. Further it was stated (Ref. Ref. 3) that a new version of the 

MAAP code, MAAP5, incorporates a momentum equation to model the primary side natural 

circulation flowrate and a more detailed SG model to more accurately predict secondary side 

behavior. These code modifications are expected to minimize the impact of the uncertainties 

seen with the MAAP4 version of the code. The study for Krško NPP (Ref. 2) was performed 

by MAAP 4.0.5.

In RELAP5 study six different reactor coolant pump seal leaks were studied due to SBO, 

ranging from 9.4 cm2 to 148.8 cm2

. These breaks represent leak from 1.32 l/s (21 gpm) to 

18.93 l/s (300 gpm) at nominal conditions. Besides, scenarios were performed for four 

different primary side depressurizations performed by operator through the secondary side 

power operated relief valves, providing that turbine driven auxiliary feedwater is available. 

One of the aims of the stress tests was also to indicate time before water level reaches the 

top of the core, and time before fuel degradation (fast cladding oxidation with hydrogen 

production). Therefore the purpose of our study was also to estimate the effect of 

depressurization on the time before water level reaches the top of the core, and time before 

time before fuel degradation. For example, in Ref. 4 it is stated that at temperatures above 

1200 °C the rapid oxidation of Zircaloy and of stainless steel by steam is present.

Finally, it was investigated the effect of having some small injection into the reactor coolant 

system by having positive displacement charging pump powered from mobile diesel 

generator. 

The organization of the report is as follows. In the Section 2 the plant analyzed is briefly 

described, while RELAP5 input model is described in Section 3. The description of scenarios 

analyzed is given in Section 4. The analysis results for the selected scenarios are described 

in Section 5. Finally, the conclusions are given in Section 6.
3

2. PLANT DESCRIPTION

Krško NPP is a Westinghouse two-loop pressurized-water reactor (PWR) plant with a large dry 

containment. The plant has been in commercial operation since 1983. After modernization in 

2000, the plant’s fuel cycle was gradually prolonged from 12 (cycle 17) to 18 months (cycle 21).

The power rating of the Krško NPP nuclear steam supply system is 2,000 megawatt thermal 

(MWt) (1,882 MWt before the plant modernization and power uprate), comprising 1,994 MWt 

(1,876 MWT before the plant modernization and power uprate) of core power output plus 6 MWt 

of reactor coolant pumps (RCPs) heat input. The reactor coolant system (RCS) is arranged as 

two closed reactor coolant loops connected in parallel to the reactor vessel, each containing an 

RCP and a steam generator (SG). An electrically heated pressurizer is connected to one of the 

loops.

The reactor core is composed of 121 fuel assemblies. The RCPs, one per coolant loop, are 

Westinghouse vertical, single-stage, centrifugal pumps of the shaft-seal type. The SGs, one per 

loop, are vertical U-tube, Siemens-Framatome type SG 72 W/D4-2 units, installed during the 

plant modernization in 2000.

For more detailed description of the plant the reader is referred to Ref. 5.

5

3. RELAP5 INPUT MODEL DESCRIPTION

To perform the analysis, Krško NPP has provided the base RELAP5 input model, so called 

“Master input deck”, which have been used for several analyses, including reference 

calculations for Krško full scope simulator verification (Refs. 6, 7, 8). The analysis was 

performed for uprated conditions (2000 MWt) with new steam generators (SGs) and Cycle 23 

settings, corresponding to the expected plant state after outage and refueling in October 2007.

The base model consists of 469 control volumes, 497 junctions and 378 heat structures with 

2107 radial mesh points and is shown in Figure 1. The analyses were performed with direct 

injection of TD AFW into steam generators (AFW piping was removed from the RELAP5 model)

as it was shown that the influence on results is negligible, while calculation performs about ten 

times faster. The model without TD AFW piping consisted of 432 control volumes, 459 junctions

while the number of heat structures remained unchanged as can be seen from Figure 2. 

Modeling of the primary side without the reactor vessel and both loops includes 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 RCP seal flow. The reactor vessel 

(RPV) consists of the lower downcomer, lower head, lower plenum, core inlet, reactor core, core 

baffle bypass, core outlet, upper plenum, upper head, upper downcomer, and guide tubes. The 

primary loop is represented by the hot leg, primary side of the steam generator (SG), 

intermediate leg with cold leg loop seal, and cold leg, separately for loop 1 and loop 2. Loops 

are symmetrical except for the pressurizer surge line and the chemical and volume control 

system connections layout. The primary side of the SG consists of the inlet and outlet plenum, 

tubesheet, and the U-tube bundle represented by a single pipe. Emergency core cooling system 

(ECCS) piping includes high-pressure safety injection (HPSI) pumps, accumulators (ACCs), and 

low-pressure safety injection (LPSI) pumps.

The secondary side consists of the SG secondary side (riser, separator and separator pool, 

downcomer, steam dome), main steamline, main steam isolation valves (MSIVs), SG relief and 

safety valves, and main feedwater (MFW) piping. Auxiliary feedwater (AFW) piping was 

removed and only TD AFW injects above the SG riser. The main steam no. 1 has same 

volumes as main steam no. 2, but the geometry data differ depending on pipeline. Turbine valve 

is modeled by the corresponding logic, while turbine is represented by time dependent volume. 

MFW and AFW pumps are modeled as time dependent junctions, pumping water from time 

dependent volumes, representing the condensate storage tank.

In order to accurately represent the Krško NPP behavior, a considerable number of control 

variables and general tables are part of the model. They represent protection, monitoring and 

simplified control systems used only during steady state initialization, as well as the following 

main plant control systems: (a) rod control system, (b) PRZ pressure control system, (c) PRZ 

level control system, (d) SG level control system, and (e) steam dump. It must be noted that rod 

control system has been modeled for point kinetics. The reactor protection system was based 

on trip logic. It includes reactor trip signal, safety injection signal, turbine trip signal, steam line 

isolation signal, MFW isolation signal, and AFW start signal.

For further details of the above mentioned plant systems and components, plant signals and 

control systems schemes the reader can refer to Reference 5.
6

Figure 1 Krško NPP base nodalization scheme – SNAP hydraulics component view
7

Figure 2 Krško NPP modified nodalization scheme – SNAP hydraulics component view

9

4. DESCRIPTION OF SCENARIOS ANALYZED

In this study the RELAP5 analyses were performed until significant reactor heatup occurred (up 

to 1500 K). The set of scenarios analyzed by RELAP5/MOD3.3 Patch 04 is shown in Table 1. 

Each set of scenarios was analyzed at for different depressurization pressures (fast 

depressurization to the specified SG pressure and maintaining the specified SG pressure): 

2.55 MPa (25 kp/cm2 gauge), 2.06 MPa (20 kp/cm2 gauge), 1.57 MPa (15 kp/cm2 gauge) and 

1.33 MPa (12.5 kp/cm2 gauge).

The scenarios (in majority) and the initial and boundary conditions used were based on Ref. 2. 

To better understand the transient progression, the assumptions used in the RELAP5 scenarios 

are listed below.

At time 0 s the following trips were actuated:

• reactor trip,

• turbine trip,

• MFW1 and MFW2 isolation,

• SI signal generation,

• RCP 1 and RCP 2 trip,

• MSIV1 and MSIV2 isolation.

The following safety systems were assumed unavailable at time 0 s:

• HPSI pump 1 and 2 unavailable,

• LPSI pump 1and 2 unavailable,

• AFW MD pump 1 and 2 unavailable.

Other systems unavailable or disabled at time 0 s were:

• pressurizer proportional and backup heaters disabled,

• pressurizer spray disabled,

• CVCS (charging and letdown) flow not available,

• condenser (steam dump) unavailable.

The following assumptions were also used:

• opening of letdown relief valve 8120 to pressurizer relief tank, if RCS pressure greater 

than 4.23 MPa,

• TD AFW available if SG pressure greater than 0.79 MPa,

• TD AFW control valves available.

Also it was assumed, when TD AFW was available, that condensate storage tanks are refilled, 

thus providing unlimited source of water.

The following operator actions were modeled:

• control of the TD AFW flow to maintain SG NR level around 60%,

• fast depressurization of the SGs to selected pressure (2.55 MPa, 2.06 MPa, 1.57 MPa or 

1.33 MPa) by opening the SG PORVs,

• maintaining SGs pressure at selected pressure (2.55 MPa, 2.06 MPa, 1.57 MPa or 1.33 

MPa) by a manual control of SG PORVs.
10

Besides main assumptions listed above, for each scenario additional assumptions were used as 

shown in Table 1. Two different groups of scenarios were simulated: S0 scenarios with TD AFW 

pump assumed available all the time and S1 scenarios with TD AFW pump assumed available 

first four hours. The calculations were performed up to 604800 s (7 days). Turbine Driven AF 

pump does not require electric power and can operate if SG pressure is above 0.79 MPa (7 

kp/cm2 gauge) so it can provide AFW injection. Namely, the AFW control valves are airoperated and provided with a 4-hour supply of nitrogen gas to control the TD AFW pump and 

the power-operated relief valves for releasing steam from SGs. When taking into account the 

assumption that the AFW regulator valves are operable (nitrogen or alternative compressed air 

supply is assumed available) and that condensate storage tanks can be refilled, the secondary 

side heat sink is available towards the whole transient. For the transient analysis duration seven 

days the SG pressure is above 0.79 MPa.

Different seal leaks per reactor coolant pump were assumed (1.32 l/s, 3.15 l/s, 4.73 l/s, 6.31 l/s, 

9.46 l/s and 18.93 l/s) for scenarios with TD AFW pump assumed available all the time (S0 

scenarios). For convenience the break flows are specified also in gallons per minute (gpm). 

Volumetric break flow was first converted to mass flow (density was considered 753.5 kg/m3

).

The mass flows were modeled by equivalent break area giving specified mass flow at nominal 

pressure and temperature conditions (15.51 MPa, 578 K). If RCS pressure was greater than 

4.23 MPa, also the letdown (LD) leak was considered (5.68 l/s).

Besides simulating scenarios with TD AFW available all the time also scenarios with TD AFW 

available (S1 scenarios) first four hours were performed for the case with 1.32 l/s volumetric 

break flow. Besides base case (S1-21) also cases with primary side depressurization were 

analyzed using one (S1-21v1) and two PRZ PORVs (S1-21v2). In the last case (S1-21p) 

injection by PDP charging pump was assumed.

Table 1 Set of scenarios analyzed for each of four depressurization cases

Scenario Seal Leak

l/s (gpm)

LD Leak

l/s (gpm)

TD AFW

Pump PRZ PORV PDP

S0-21 1.32 (21) 90 ON No No

S0-50 3.15 (50) 90 ON No No

S0-75 4.73 (75) 90 ON No No

S0-100 6.31 (100) 90 ON No No

S0-150 9.46 (150) 90 ON No No

S0-150p 9.46 (150) 90 ON No Yes

S0-300 18.93 (300) 90 ON No No

S1-21 1.32 (21) 90 OFF at 4hr No No

S1-21v1 1.32 (21) 90 OFF at 4hr Yes

(one PORV) No

S1-21v2 1.32 (21) 90 OFF at 4hr Yes

(two PORV) No

S1-21p 1.32 (21) 90 OFF at 4hr No Yes
11

5. RESULTS

In total 44 calculations were performed for eleven scenarios at four different SG 

depressurization pressure setpoints. The calculations were performed up to 604800 s (7 days) 

or heatup of the core (calculations were aborted due to high clad temperature or due to reactor 

kinetics error), whatever occurred first. First set of plots shows the dependence of each 

calculated scenario on the depressurization of SGs to selected values and maintaining that 

pressure. Second set of plots shows the RCP seal leak dependence of S0 scenarios with TD 

AFW pump available all the time and the third set shows S1 scenarios with TD AFW available 

four hours after SBO event start, in which the equipment used was varied.

5.1 Dependence on depressurization of selected SBO scenarios

For each scenario, the following six variables are shown: RCS pressure, core exit temperature, 

core collapsed liquid level, average fuel cladding temperature, total mass discharged from RCS 

(through letdown isolation valve when RCS pressure greater than 4.23 MPa, both RCP seal 

leaks and PRZ PORVs – only scenarios S1-21p1 and S1-21p2) and mass injected into RCS 

(accumulators (opening below 4.96 MPa) and PDP pump in cases S0150p and S1-21p).

5.1.1 Scenario S0-21

The results for scenario S0-21 are shown in Figures 3 through 8. It can be seen that primary 

pressure (Figure 3) follows the secondary side depressurization. At the end of transient analysis 

the pressure start to drop below the depressurization pressure setpoint, what means that 

cooling through the break and by steam assumed to be consumed by TD AFW pump is 

sufficient. Core exit temperature shown in Figure 4 has similar trend as primary pressure. Figure 

5 shows the core collapsed liquid level. It can be seen, that initially the level dropped due to 

stopped injection from accumulators and later remains around 80% (please note that this is 

collapsed liquid level, denoting voids in the core without real core uncovery). Figure 6 shows the 

average fuel cladding temperature, which slowly decrease during transient. The mass 

discharged from the primary system is shown in Figure 7. Initially more mass is discharged in 

the cases with larger depressurization (the larger injection from the accumulators the larger 

break flow). From Figure 8 it can be seen that the accumulators were not completely emptied

(further depressurization would be needed to enable discharge of all 72 tons of water from both 

accumulators). The results showed that one TD AFW pump is sufficient to cool the primary 

system. Also it can be concluded, that depressurization is beneficial, especially below 

4.23 MPa, by eliminating letdown break flow and enabling accumulator injection. Nevertheless, 

after one week the plant is in similar state for all selected depressurizations.
12

0

2

4

6

8

10

12

14

16

0 100000 200000 300000 400000 500000 600000

Pressure (MPa)

Time (s)

12.5 kp/cm2

15 kp/cm2

20 kp/cm2

25 kp/cm2

Figure 3 RCS pressure – scenario S0-21

400

440

480

520

560

600

0 100000 200000 300000 400000 500000 600000

Temperature (K)

Time (s)

12.5 kp/cm2

15 kp/cm2

20 kp/cm2

25 kp/cm2

Figure 4 Core exit temperature – scenario S0-21
13

0

20

40

60

80

100

120

0 100000 200000 300000 400000 500000 600000

Level (%)

Time (s)

12.5 kp/cm2

15 kp/cm2

20 kp/cm2

25 kp/cm2

Figure 5 Core collapsed liquid level – scenario S0-21

400

450

500

550

600

650

700

0 100000 200000 300000 400000 500000 600000

Temperature (K)

Time (s)

12.5 kp/cm2

15 kp/cm2

20 kp/cm2

25 kp/cm2

Figure 6 Average fuel cladding temperature – scenario S0-21
14

0

20

40

60

80

100

0 100000 200000 300000 400000 500000 600000

Mass (ton)

Time (s)

12.5 kp/cm2

15 kp/cm2

20 kp/cm2

25 kp/cm2

Figure 7 Total mass discharged from RCS – scenario S0-21

0

20

40

60

80

0 100000 200000 300000 400000 500000 600000

Mass (ton)

Time (s)

12.5 kp/cm2

15 kp/cm2

20 kp/cm2

25 kp/cm2

Figure 8 Mass injected into RCS – scenario S0-21
15

5.1.2 Scenario S0-50

The results for scenario S0-50 are shown in Figures 9 through 14. It can be seen that primary 

pressure (Figure 9) follows the secondary side depressurization. As break in scenario S0-50 is 

larger than in scenario S0-21, the importance of RCS depressurization increases. Core exit 

temperature (Figure 10) shows that only the case with largest depressurization does not lead to 

core heatup in the first seven days. Figure 11 shows the core uncovery for cases with lower 

depressurization. It can be seen, that initially the level dropped due to stopped injection from 

accumulators and then remains around 80% until core uncovery. Figure 12 shows the average 

fuel cladding temperature. At larger depressurization initially more mass is discharged from the 

primary system as shown in Figure 13 because also more mass is injected from the 

accumulators (Figure 14). Later larger depressurization means smaller discharge of RCS 

inventory.

0

2

4

6

8

10

12

14

16

0 100000 200000 300000 400000 500000 600000

Pressure (MPa)

Time (s)

12.5 kp/cm2

15 kp/cm2

20 kp/cm2

25 kp/cm2

Figure 9 RCS pressure – scenario S0-50
16

400

600

800

1000

1200

1400

0 100000 200000 300000 400000 500000 600000

Temperature (K)

Time (s)

12.5 kp/cm2

15 kp/cm2

20 kp/cm2

25 kp/cm2

Figure 10 Core exit temperature – scenario S0-50

0

20

40

60

80

100

120

0 100000 200000 300000 400000 500000 600000

Level (%)

Time (s)

12.5 kp/cm2

15 kp/cm2

20 kp/cm2

25 kp/cm2

Figure 11 Core collapsed liquid level – scenario S0-50
17

400

600

800

1000

1200

1400

1600

0 100000 200000 300000 400000 500000 600000

Temperature (K)

Time (s)

12.5 kp/cm2

15 kp/cm2

20 kp/cm2

25 kp/cm2

Figure 12 Average fuel cladding temperature – scenario S0-50

0

40

80

120

160

0 100000 200000 300000 400000 500000 600000

Mass (ton)

Time (s)

12.5 kp/cm2

15 kp/cm2

20 kp/cm2

25 kp/cm2

Figure 13 Total mass discharged from RCS – scenario S0-50
18

0

10

20

30

40

50

60

70

0 100000 200000 300000 400000 500000 600000

Mass (ton)

Time (s)

12.5 kp/cm2

15 kp/cm2

20 kp/cm2

25 kp/cm2

Figure 14 Mass injected into RCS – scenario S0-50

5.1.3 Scenario S0-75

The results for scenario S0-75 are shown in Figures 15 through 20. It can be seen that primary 

pressure (Figure 15) follows the secondary side depressurization. Core exit temperature (Figure 

16) shows that core heatup occurred for all cases in the first five days. The core collapsed liquid 

level as a function of depressurization is shown in Figure 17. It can be seen, that initially the 

level dropped due to stopped injection from accumulators and then remains around 80% until 

core uncovery. Figure 18 shows the average fuel cladding temperature. Again, at larger

depressurization initially more mass is discharged from the primary system than at lower 

depressurization (see Figure 19) because more mass is injected from the accumulators (see 

Figure 20). Later larger depressurization means smaller discharge of RCS inventory. With high 

depressurization (case 12.5 kp/cm2) the RCS could be sufficiently cooled two days more than 

by low depressurization case (25 kp/cm2). Nevertheless, with transient progression the core 

heatup happened in all cases before five days.
19

0

2

4

6

8

10

12

14

16

0 100000 200000 300000 400000

Pressure (MPa)

Time (s)

12.5 kp/cm2

15 kp/cm2

20 kp/cm2

25 kp/cm2

Figure 15 RCS pressure – scenario S0-75

400

500

600

700

800

900

1000

0 100000 200000 300000 400000

Temperature (K)

Time (s)

12.5 kp/cm2

15 kp/cm2

20 kp/cm2

25 kp/cm2

Figure 16 Core exit temperature – scenario S0-75
20

0

20

40

60

80

100

120

0 100000 200000 300000 400000

Level (%)

Time (s)

12.5 kp/cm2

15 kp/cm2

20 kp/cm2

25 kp/cm2

Figure 17 Core collapsed liquid level – scenario S0-75

400

600

800

1000

1200

1400

1600

0 100000 200000 300000 400000

Temperature (K)

Time (s)

12.5 kp/cm2

15 kp/cm2

20 kp/cm2

25 kp/cm2

Figure 18 Average fuel cladding temperature – scenario S0-75
21

0

20

40

60

80

100

120

140

160

180

0 100000 200000 300000 400000

Mass (kg)

Time (s)

12.5 kp/cm2

15 kp/cm2

20 kp/cm2

25 kp/cm2

Figure 19 Total mass discharged from RCS – scenario S0-75

0

10

20

30

40

50

60

70

0 100000 200000 300000 400000

Mass (ton)

Time (s)

12.5 kp/cm2

15 kp/cm2

20 kp/cm2

25 kp/cm2

Figure 20 Mass injected into RCS – scenario S0-75
22

5.1.4 Scenario S0-100

The results for scenario S0-100 are shown in Figures 21 through 26. It can be seen that primary 

pressure (Figure 21) follows the secondary side depressurization. Core exit temperature (Figure 

22) shows that core heatup occurred for all cases again faster than in the previous case. The 

core collapsed liquid level as a function of depressurization is shown in Figure 23. It can be 

seen, that initially the level dropped due to stopped injection from accumulators and then 

remains around 80% until core uncovery. Figure 24 shows the average fuel cladding 

temperature. Again, at larger depressurization initially more mass is discharged from the 

primary system than at lower depressurization (see Figure 25) because more mass is injected 

from the accumulators (see Figure 26). Later larger depressurization means smaller discharge 

of RCS inventory. With high depressurization (case 12.5 kp/cm2) the RCS could be sufficiently 

cooled one and half day longer than by low depressurization case (25 kp/cm2). Nevertheless, 

with transient progression the core heatup happened in all cases before four days.

0

2

4

6

8

10

12

14

16

0 60000 120000 180000 240000 300000

Pressure (MPa)

Time (s)

12.5 kp/cm2

15 kp/cm2

20 kp/cm2

25 kp/cm2

Figure 21 RCS pressure – scenario S0-100
23

400

600

800

1000

1200

1400

0 60000 120000 180000 240000 300000

Temperature (K)

Time (s)

12.5 kp/cm2

15 kp/cm2

20 kp/cm2

25 kp/cm2

Figure 22 Core exit temperature – scenario S0-100

0

20

40

60

80

100

120

0 60000 120000 180000 240000 300000

Level (%)

Time (s)

12.5 kp/cm2

15 kp/cm2

20 kp/cm2

25 kp/cm2

Figure 23 Core collapsed liquid level – scenario S0-100
24

400

600

800

1000

1200

1400

1600

0 60000 120000 180000 240000 300000

Temperature (K)

Time (s)

12.5 kp/cm2

15 kp/cm2

20 kp/cm2

25 kp/cm2

Figure 24 Average fuel cladding temperature – scenario S0-100

0

20

40

60

80

100

120

140

160

0 60000 120000 180000 240000 300000

Mass (ton)

Time (s)

12.5 kp/cm2

15 kp/cm2

20 kp/cm2

25 kp/cm2

Figure 25 Total mass discharged from RCS – scenario S0-100
25

0

10

20

30

40

50

60

70

0 60000 120000 180000 240000 300000

Mass (ton)

Time (s)

12.5 kp/cm2

15 kp/cm2

20 kp/cm2

25 kp/cm2

Figure 26 Mass injected into RCS – scenario S0-100

5.1.5 Scenario S0-150

The results for scenario S0-150 are shown in Figures 27 through 32. It can be seen that primary 

pressure (Figure 27) follows the secondary side depressurization. Core exit temperature (Figure 

28) shows that core heatup occurred for all cases again faster than in the previous case. The 

core collapsed liquid level as a function of depressurization is shown in Figure 29. It can be 

seen, that initially the level dropped due to stopped injection from accumulators and then 

remains around 80% until core uncovery. Figure 30 shows the average fuel cladding 

temperature. Again, at larger depressurization initially more mass is discharged from the 

primary system than at lower depressurization (see Figure 31) because more mass is injected 

from the accumulators (see Figure 32). Later larger depressurization means smaller discharge 

of RCS inventory. With high depressurization (case 12.5 kp/cm2) the RCS could be sufficiently 

cooled one day more than by low depressurization case (25 kp/cm2). Nevertheless, with 

transient progression the core heatup happened in all cases before two and half days.
26

0

2

4

6

8

10

12

14

16

0 50000 100000 150000 200000

Pressure (MPa)

Time (s)

12.5 kp/cm2

15 kp/cm2

20 kp/cm2

25 kp/cm2

Figure 27 RCS pressure – scenario S0-150

400

600

800

1000

1200

1400

0 50000 100000 150000 200000

Temperature (K)

Time (s)

12.5 kp/cm2

15 kp/cm2

20 kp/cm2

25 kp/cm2

Figure 28 Core exit temperature – scenario S0-150
27

0

20

40

60

80

100

120

0 50000 100000 150000 200000

Level (%)

Time (s)

12.5 kp/cm2

15 kp/cm2

20 kp/cm2

25 kp/cm2

Figure 29 Core collapsed liquid level – scenario S0-150

400

600

800

1000

1200

1400

1600

0 50000 100000 150000 200000

Temperature (K)

Time (s)

12.5 kp/cm2

15 kp/cm2

20 kp/cm2

25 kp/cm2

Figure 30 Average fuel cladding temperature – scenario S0-150
28

0

20

40

60

80

100

120

140

160

180

0 50000 100000 150000 200000

Mass (ton)

Time (s)

12.5 kp/cm2

15 kp/cm2

20 kp/cm2

25 kp/cm2

Figure 31 Total mass discharged from RCS – scenario S0-150

0

10

20

30

40

50

60

70

0 50000 100000 150000 200000

Mass (ton)

Time (s)

12.5 kp/cm2

15 kp/cm2

20 kp/cm2

25 kp/cm2

Figure 32 Mass injected into RCS – scenario S0-150
29

5.1.6 Scenario S0-150p

The results for scenario S0-150p are shown in Figures 33 through 38. The only difference 

comparing to S0-150 scenario is, that PDP charging pump with the capacity to inject 2.2 kg/s 

into RCS is used after 4 hours. Namely, it is not sufficient just to cool the primary system, as 

core heatup resulted from RCS inventory depletion. Therefore RCS inventory injection is 

needed. It can be seen that RCS pressure (Figure 33) follows the secondary side 

depressurization in the first part only. Later cooling through the breaks and steam consumption 

by TD AFW pumps is sufficient provided that RCS inventory makeup is provided. Core exit 

temperature (Figure 34) shows that core heatup does not occur. The core collapsed liquid level

as a function of depressurization is shown in Figure 35. It can be seen, that initially the level 

dropped due to stopped injection from accumulators and then remains around 90% until core 

uncovery. Figure 36 shows the average fuel cladding temperature, which is decreasing. The 

mass discharged from RCS (see Figure 37) and injected mass (see Figure 38) is not dependent 

on the depressurization in the second part of transient. The injected mass to and discharged 

mass from RCS are practically balanced. The RCS system is efficiently cooled through breaks 

besides secondary side cooling. The selected case clearly showed that RCS injection is also 

very important for preventing core uncovery; especially in the cases with larger breaks this is the 

only way to prevent core heatup in the first seven days.

0

2

4

6

8

10

12

14

16

0 100000 200000 300000 400000 500000 600000

Pressure (MPa)

Time (s)

12.5 kp/cm2

15 kp/cm2

20 kp/cm2

25 kp/cm2

Figure 33 RCS pressure – scenario S0-150p
30

400

450

500

550

600

650

700

0 100000 200000 300000 400000 500000 600000

Temperature (K)

Time (s)

12.5 kp/cm2

15 kp/cm2

20 kp/cm2

25 kp/cm2

Figure 34 Core exit temperature – scenario S0-150p

0

20

40

60

80

100

120

0 100000 200000 300000 400000 500000 600000

Level (%)

Time (s)

12.5 kp/cm2

15 kp/cm2

20 kp/cm2

25 kp/cm2

Figure 35 Core collapsed liquid level – scenario S0-150p
31

0

100

200

300

400

500

600

700

0 100000 200000 300000 400000 500000 600000

Temperature (K)

Time (s)

12.5 kp/cm2

15 kp/cm2

20 kp/cm2

25 kp/cm2

Figure 36 Average fuel cladding temperature – scenario S0-150p

0

200

400

600

800

1000

1200

1400

1600

0 100000 200000 300000 400000 500000 600000

Mass (ton)

Time (s)

12.5 kp/cm2

15 kp/cm2

20 kp/cm2

25 kp/cm2

Figure 37 Total mass discharged from RCS – scenario S0-150p
32

0

200

400

600

800

1000

1200

1400

0 100000 200000 300000 400000 500000 600000

Mass (ton)

Time (s)

12.5 kp/cm2

15 kp/cm2

20 kp/cm2

25 kp/cm2

Figure 38 Mass injected into RCS – scenario S0-150p

5.1.7 Scenario S0-300

The results for scenario S0-300 are shown in Figures 39 through 44. It can be seen that RCS

pressure (Figure 39) follows the secondary side depressurization. Core exit temperature (Figure 

40) shows that core heatup occurred for all cases again faster than in the S0-150 case. The 

core collapsed liquid level as a function of depressurization is shown in Figure 41. It can be 

seen, that initially the level dropped due to stopped injection from accumulators and then 

remains around 80% until core uncovery. Figure 42 shows the average fuel cladding 

temperature. Again, at larger depressurization initially more mass is discharged from the 

primary system than at lower depressurization (see Figure 43) because more mass is injected 

from the accumulators (see Figure 44). Later larger depressurization means smaller discharge 

of RCS inventory. With high depressurization (case 12.5 kp/cm2) the RCS could be sufficiently 

cooled half day more than by low depressurization case (25 kp/cm2). Nevertheless, with 

transient progression the core heatup happened in all cases in one day. The only alternative to 

avoid core uncovery is to provide RCS inventory by PDP charging pump or other means.
33

0

2

4

6

8

10

12

14

16

18

0 20000 40000 60000 80000 100000

Pressure (MPa)

Time (s)

12.5 kp/cm2

15 kp/cm2

20 kp/cm2

25 kp/cm2

Figure 39 RCS pressure – scenario S0-300

400

600

800

1000

1200

1400

0 20000 40000 60000 80000 100000

Temperature (K)

Time (s)

12.5 kp/cm2

15 kp/cm2

20 kp/cm2

25 kp/cm2

Figure 40 Core exit temperature – scenario S0-300
34

0

20

40

60

80

100

120

0 20000 40000 60000 80000 100000

Level (%)

Time (s)

12.5 kp/cm2

15 kp/cm2

20 kp/cm2

25 kp/cm2

Figure 41 Core collapsed liquid level – scenario S0-300

0

500

1000

1500

2000

2500

0 20000 40000 60000 80000 100000

Temperature (K)

Time (s)

12.5 kp/cm2

15 kp/cm2

20 kp/cm2

25 kp/cm2

Figure 42 Average fuel cladding temperature – scenario S0-300
35

0

20

40

60

80

100

120

140

160

180

0 20000 40000 60000 80000 100000

Mass (ton)

Time (s)

12.5 kp/cm2

15 kp/cm2

20 kp/cm2

25 kp/cm2

Figure 43 Total mass discharged from RCS – scenario S0-300

0

10

20

30

40

50

60

70

0 20000 40000 60000 80000 100000

Mass (ton)

Time (s)

12.5 kp/cm2

15 kp/cm2

20 kp/cm2

25 kp/cm2

Figure 44 Mass injected into RCS – scenario S0-300
36

5.1.8 Scenario S1-21

The results for scenario S1-21 are shown in Figures 45 through 50. It can be seen that primary 

pressure (Figure 45) follows the secondary pressures until SGs as heat sink are lost. The steam 

generator pressures are maintained by operator until 14400 s. Later the TD AFW pump was

assumed not to be available. Core exit temperature shown in Figure 46 increased after core 

heatup. Only small heatup is shown because calculations were stopped due to reactor kinetics 

time step reduced below minimum value. No attempt was made to restart calculations as from 

Figure 47 showing the core collapsed liquid level the core uncovery is evident at the time of 

calculation abortion and restarts were not very much successful when such error occured. It can 

also be seen, that after decreasing RCS inventory also the core collapsed liquid levels 

decrease. However, significant core uncovery happens after 40000 s for all depressurization 

cases. Figure 48 shows the average fuel cladding temperature, which starts to increase after 

the core is significantly uncovered. Figure 49 shows mass discharged from RCS, which after 4 

hours increases again due to the lost heat sink. Finally, Figure 50 shows the mass injected by 

accumulators. After RCS system repressurization further injection by accumulators was 

prevented. The scenario clearly showed that after losing the heat sink after 4 hours on 

secondary side the plant can survive additional 8 to 10 hours (depending on depressurization) 

as some cooling is provided by RCP seal leaks.

0

2

4

6

8

10

12

14

16

18

0 10000 20000 30000 40000 50000

Pressure (MPa)

Time (s)

12.5 kp/cm2

15 kp/cm2

20 kp/cm2

25 kp/cm2

Figure 45 RCS pressure – scenario S1-21
37

400

450

500

550

600

650

700

0 10000 20000 30000 40000 50000

Temperature (K)

Time (s)

12.5 kp/cm2

15 kp/cm2

20 kp/cm2

25 kp/cm2

Figure 46 Core exit temperature – scenario S1-21

0

20

40

60

80

100

120

0 10000 20000 30000 40000 50000

Level (%)

Time (s)

12.5 kp/cm2

15 kp/cm2

20 kp/cm2

25 kp/cm2

Figure 47 Core collapsed liquid level – scenario S1-21
38

400

500

600

700

800

900

0 10000 20000 30000 40000 50000

Temperature (K)

Time (s)

12.5 kp/cm2

15 kp/cm2

20 kp/cm2

25 kp/cm2

Figure 48 Average fuel cladding temperature – scenario S1-21

0

20

40

60

80

100

120

0 10000 20000 30000 40000 50000

Mass (ton)

Time (s)

12.5 kp/cm2

15 kp/cm2

20 kp/cm2

25 kp/cm2

Figure 49 Total mass discharged from RCS – scenario S1-21
39

0

10

20

30

40

0 10000 20000 30000 40000 50000

Mass (ton)

Time (s)

12.5 kp/cm2

15 kp/cm2

20 kp/cm2

25 kp/cm2

Figure 50 Mass injected into RCS – scenario S1-21

5.1.9 Scenario S1-21v1

The results for scenario S1-21v1, which differ from S1-21 scenario in that one pressurizer 

PORV is used arbitrarily 25000 s after transient start, are shown in Figures 51 through 56. It can 

be seen that RCS pressure (Figure 51) follows the secondary pressures until SGs as heat sink 

are lost. The steam generator pressures are maintained by operator until 14400 s. Later the TD 

AFW pump was assumed to be lost and SG PORVs were assumed not available. When 

setpoint was reached, the SG safety valves opened and provide cooling until SGs dry out. At 

that time RCS pressure increases again. PRZ PORV was used to limit the RCS pressure. Core 

exit temperature shown in Figure 52 increased after PRZ PORV opening causing core uncovery 

(Figure 53). From Figure 53 it can be seen, that due to the RCP leaks and lost TD AFW the core

collapsed liquid levels decrease. However, significant core uncovery happens after PRZ PORV 

opening for all depressurization cases. Figure 54 shows the average fuel cladding temperature, 

which starts to increase when the core is significantly uncovered. Figure 55 shows mass release 

from RCS, which after 4 hours increases again due to lost heat sink. When PRZ PORV was 

opened, further large mass discharge happened. Finally, Figure 56 shows the mass injected by 

accumulators. After RCS system repressurization further injection by accumulators was 

prevented. The scenario clearly showed that after losing the heat sink after 4 hours on 

secondary side the plant heatup happened around one hour (independent on depressurization)

after PRZ PORV is used, causing much discharge of RCS inventory.
40

0

2

4

6

8

10

12

14

16

0 5000 10000 15000 20000 25000 30000

Pressure (MPa)

Time (s)

12.5 kp/cm2

15 kp/cm2

20 kp/cm2

25 kp/cm2

Figure 51 RCS pressure – scenario S1-21v1

400

450

500

550

600

650

700

750

800

850

900

0 5000 10000 15000 20000 25000 30000

Temperature (K)

Time (s)

12.5 kp/cm2

15 kp/cm2

20 kp/cm2

25 kp/cm2

Figure 52 Core exit temperature – scenario S1-21v1
41

0

20

40

60

80

100

120

0 5000 10000 15000 20000 25000 30000

Level (%)

Time (s)

12.5 kp/cm2

15 kp/cm2

20 kp/cm2

25 kp/cm2

Figure 53 Core collapsed liquid level – scenario S1-21v1

400

600

800

1000

1200

1400

1600

0 5000 10000 15000 20000 25000 30000

Temperature (K)

Time (s)

12.5 kp/cm2

15 kp/cm2

20 kp/cm2

25 kp/cm2

Figure 54 Average fuel cladding temperature – scenario S1-21v1
42

0

20

40

60

80

100

120

140

0 5000 10000 15000 20000 25000 30000

Mass (ton)

Time (s)

12.5 kp/cm2

15 kp/cm2

20 kp/cm2

25 kp/cm2

Figure 55 Total mass discharged from RCS – scenario S1-21v1

0

5

10

15

20

25

30

35

40

0 5000 10000 15000 20000 25000 30000

Mass (ton)

Time (s)

12.5 kp/cm2

15 kp/cm2

20 kp/cm2

25 kp/cm2

Figure 56 Mass injected into RCS – scenario S1-21v1
43

5.1.10 Scenario S1-21v2

The results for scenario S1-21v2, which differ from S1-21v1 scenario in that two pressurizer 

PORV are used instead of one in 25000 s after transient start, are shown in Figures 57 through 

62. It can be seen that RCS pressure (Figure 57) follows the secondary pressures until SGs as 

heat sink are lost. The steam generator pressures are maintained by operator until 14400 s. 

Later the TD AFW pump was assumed to be lost and SG PORVs were assumed not available.

When setpoint was reached, the SG safety valves opened and provide cooling until SGs dry 

out. At that time RCS pressure increases again. PRZ PORVs were used to limit the RCS 

pressure. Core exit temperature shown in Figure 58 increased after PRZ PORVs opening 

causing core uncovery (Figure 59). From Figure 59 it can be seen, that due to the RCP leaks 

and lost TD AFW the core collapsed liquid levels start to decrease. However, significant core 

uncovery happens after PRZ PORVs opening for all depressurization cases. Figure 60 shows 

the average fuel cladding temperature, which starts to increase when the core is significantly 

uncovered soon after PRZ PORVs opening. Figure 61 shows mass discharge from RCS, which 

after 4 hours increases again due to lost heat sink. When PRZ PORVs were opened, further 

large mass discharge happened. Finally, Figure 62 shows the mass injected by accumulators. 

After RCS system repressurization further injection by accumulators was prevented. The 

scenario clearly showed that after losing the heat sink after 4 hours on secondary side the plant 

heatup happened around one hour (independent on depressurization) after PRZ PORVs are 

used, causing much release of RCS inventory.

0

2

4

6

8

10

12

14

16

0 5000 10000 15000 20000 25000 30000

Pressure (MPa)

Time (s)

12.5 kp/cm2

15 kp/cm2

20 kp/cm2

25 kp/cm2

Figure 57 RCS pressure – scenario S1-21v2
44

400

500

600

700

800

900

1000

0 5000 10000 15000 20000 25000 30000

Temperature (K)

Time (s)

12.5 kp/cm2

15 kp/cm2

20 kp/cm2

25 kp/cm2

Figure 58 Core exit temperature – scenario S1-21v2

0

20

40

60

80

100

120

0 5000 10000 15000 20000 25000 30000

Level (%)

Time (s)

12.5 kp/cm2

15 kp/cm2

20 kp/cm2

25 kp/cm2

Figure 59 Core collapsed liquid level – scenario S1-21v2
45

400

600

800

1000

1200

1400

1600

0 5000 10000 15000 20000 25000 30000

Temperature (K)

Time (s)

12.5 kp/cm2

15 kp/cm2

20 kp/cm2

25 kp/cm2

Figure 60 Average fuel cladding temperature – scenario S1-21v2

0

20

40

60

80

100

120

140

160

0 5000 10000 15000 20000 25000 30000

Mass (ton)

Time (s)

12.5 kp/cm2

15 kp/cm2

20 kp/cm2

25 kp/cm2

Figure 61 Total mass discharged from RCS – scenario S1-21v2
46

0

5

10

15

20

25

30

35

40

0 5000 10000 15000 20000 25000 30000

Mass (ton)

Time (s)

12.5 kp/cm2

15 kp/cm2

20 kp/cm2

25 kp/cm2

Figure 62 Mass injected into RCS – scenario S1-21v2

5.1.11 Scenario S1-21p

The results for scenario S1-21p, which differ from S1-21 scenario in that PDP charging pump is 

used 4 hours after transient start, when TD AFW is lost, are shown in Figures 63 through 68.

It can be seen that RCS pressure (Figure 63) follows the secondary pressures until SGs as heat 

sink are lost. The steam generator pressures are maintained by operator until 14400 s. Later the 

TD AFW pump was assumed to be lost and SG PORVs were assumed not available. Core exit 

temperature shown in Figure 64 increased after core heatup. Figure 65 shows the core 

collapsed liquid level. It can be seen, that after decreasing RCS inventory also the core 

collapsed liquid levels decrease. However, significant core uncovery happens after 50000 s for 

all depressurization cases due to primary pressure increase causing automatic PRZ relief valve

opening. Figure 66 shows the average fuel cladding temperature, which starts to increase after 

the core is significantly uncovered as result of PRZ relief valve opening.

Figure 67 shows mass release from RCS, which after 4 hours increases again due to lost heat 

sink. When PRZ relief valve was opened, further large mass discharge happened. Finally, 

Figure 68 shows the mass injected by accumulators and later by PDP charging pump. After 

RCS system repressurization further injection by accumulators was prevented. The scenario 

clearly showed that after losing the heat sink after 4 hours on secondary side the plant can 

survive at least 10 hours (depending on depressurization) as some cooling is provided also by 

RCP seal leaks. Comparing to base case the operation of PDP prolongs the time with core 

uncovery for few hours. This example clearly showed that TD AFW operation is needed to 

provide long term cooling, if RCS makeup is too small as in our case.
47

0

2

4

6

8

10

12

14

16

18

0 10000 20000 30000 40000 50000 60000

Pressure (MPa)

Time (s)

12.5 kp/cm2

15 kp/cm2

20 kp/cm2

25 kp/cm2

Figure 63 RCS pressure – scenario S1-21p

400

450

500

550

600

650

700

0 10000 20000 30000 40000 50000 60000

Temperature (K)

Time (s)

12.5 kp/cm2

15 kp/cm2

20 kp/cm2

25 kp/cm2

Figure 64 Core exit temperature – scenario S1-21p
48

0

20

40

60

80

100

120

0 10000 20000 30000 40000 50000 60000

Level (%)

Time (s)

12.5 kp/cm2

15 kp/cm2

20 kp/cm2

25 kp/cm2

Figure 65 Core level – scenario S1-21p

0

100

200

300

400

500

600

700

0 10000 20000 30000 40000 50000 60000

Temperature (K)

Time (s)

12.5 kp/cm2

15 kp/cm2

20 kp/cm2

25 kp/cm2

Figure 66 Average fuel cladding temperature – scenario S1-21p
49

0

50

100

150

200

250

0 10000 20000 30000 40000 50000 60000

Mass (ton)

Time (s)

12.5 kp/cm2

15 kp/cm2

20 kp/cm2

25 kp/cm2

Figure 67 Total mass discharged from RCS – scenario S1-21p

0

20

40

60

80

100

120

140

0 10000 20000 30000 40000 50000 60000

Mass (ton)

Time (s)

12.5 kp/cm2

15 kp/cm2

20 kp/cm2

25 kp/cm2

Figure 68 Mass injected into RCS – scenario S1-21p
50

5.2 Dependence on RCPs seal leak of SBO scenarios with TD AFW available

For each set of S0 scenarios (consisting of six different leaks) on depressurization pressure, the 

following five variables are shown: RCS pressure, SG no. 1 pressure, core collapsed liquid 

level, average fuel cladding temperature, and RCS inventory. As the variables have already 

been plotted for all cases, Figures 69 through 88 will not be described in detail. Rather, some

remarks will be made. Please note, that SG no. 1 pressure and RCS mass inventory have not 

been plotted in Section 5.1.

5.2.1 S0 scenarios depressurized to 1.33 MPa (12.5 kp/cm2 gauge)

The results of S0 scenarios, in which the operator depressurizes and maintains SG pressure at 

1.33 MPa are shown in Figures 69 through 73. The larger the break is, the shorter the time 

available before core uncovery and core heatup is.

0

2

4

6

8

10

12

14

16

0 100000 200000 300000 400000 500000 600000

Pressure (MPa)

Time (s)

21 gpm

50 gpm

75 gpm

100 gpm

150 gpm

300 gpm

Figure 69 RCS pressure – S0 scenarios (1.33 MPa)
51

0

1

2

3

4

5

6

7

8

9

0 100000 200000 300000 400000 500000 600000

Pressure (MPa)

Time (s)

21 gpm

50 gpm

75 gpm

100 gpm

150 gpm

300 gpm

Figure 70 SG no.1 pressure – S0 scenarios (1.33 MPa)

0

20

40

60

80

100

120

0 100000 200000 300000 400000 500000 600000

Level (%)

Time (s)

21 gpm

50 gpm

75 gpm

100 gpm

150 gpm

300 gpm

Figure 71 Core collapsed liquid level – S0 scenarios (1.33 MPa)
52

400

600

800

1000

1200

1400

1600

1800

2000

0 100000 200000 300000 400000 500000 600000

Temperature (K)

Time (s)

21 gpm

50 gpm

75 gpm

100 gpm

150 gpm

300 gpm

Figure 72 Average fuel cladding temperature – S0 scenarios (1.33 MPa)

0

20

40

60

80

100

120

140

160

0 100000 200000 300000 400000 500000 600000

Mass (kg)

Time (s)

21 gpm

50 gpm

75 gpm

100 gpm

150 gpm

300 gpm

Figure 73 RCS mass inventory – S0 scenarios (1.33 MPa)
53

5.2.2 S0 scenarios depressurized to 1.57 MPa (15 kp/cm2 gauge)

The results scenarios S0, in which the operator depressurizes and maintains SG pressure at 

1.57 MPa are shown in Figures 74 through 78. The larger the break is, the shorter the time 

available before core uncovery and core heatup is.

0

2

4

6

8

10

12

14

16

0 100000 200000 300000 400000 500000 600000

Pressure (MPa)

Time (s)

21 gpm

50 gpm

75 gpm

100 gpm

150 gpm

300 gpm

Figure 74 RCS pressure – S0 scenarios (1.57 MPa)
54

0

1

2

3

4

5

6

7

8

9

0 100000 200000 300000 400000 500000 600000

Pressure (MPa)

Time (s)

21 gpm

50 gpm

75 gpm

100 gpm

150 gpm

300 gpm

Figure 75 SG no.1 pressure – S0 scenarios (1.57 MPa)

0

20

40

60

80

100

120

0 100000 200000 300000 400000 500000 600000

Level (%)

Time (s)

21 gpm

50 gpm

75 gpm

100 gpm

150 gpm

300 gpm

Figure 76 Core collapsed liquid level – S0 scenarios (1.57 MPa)
55

400

600

800

1000

1200

1400

1600

1800

0 100000 200000 300000 400000 500000 600000

Temperature (K)

Time (s)

21 gpm

50 gpm

75 gpm

100 gpm

150 gpm

300 gpm

Figure 77 Average fuel cladding temperature – S0 scenarios (1.57 MPa)

0

20

40

60

80

100

120

140

160

0 100000 200000 300000 400000 500000 600000

Mass (ton)

Time (s)

21 gpm

50 gpm

75 gpm

100 gpm

150 gpm

300 gpm

Figure 78 RCS mass inventory – S0 scenarios (1.57 MPa)
56

5.2.3 S0 scenarios depressurized to 2.06 MPa (20 kp/cm2 gauge)

The results scenarios S0, in which the operator depressurizes and maintains SG pressure at 

2.06 MPa are shown in Figures 79 through 83. The larger the break is, the shorter the time 

available before core uncovery and core heatup is.

0

2

4

6

8

10

12

14

16

0 100000 200000 300000 400000 500000 600000

Pressure (MPa)

Time (s)

21 gpm

50 gpm

75 gpm

100 gpm

150 gpm

300 gpm

Figure 79 RCS pressure – S0 scenarios (2.06 MPa)
57

0

1

2

3

4

5

6

7

8

9

0 100000 200000 300000 400000 500000 600000

Pressure (MPa)

Time (s)

21 gpm

50 gpm

75 gpm

100 gpm

150 gpm

300 gpm

Figure 80 SG no.1 pressure – S0 scenarios (2.06 MPa)

0

20

40

60

80

100

120

0 100000 200000 300000 400000 500000 600000

Level (%)

Time (s)

21 gpm

50 gpm

75 gpm

100 gpm

150 gpm

300 gpm

Figure 81 Core collapsed liquid level – S0 scenarios (2.06 MPa)
58

400

600

800

1000

1200

1400

1600

0 100000 200000 300000 400000 500000 600000

Temperature (K)

Time (s)

21 gpm

50 gpm

75 gpm

100 gpm

150 gpm

300 gpm

Figure 82 Average fuel cladding temperature – S0 scenarios (2.06 MPa)

0

20000

40000

60000

80000

100000

120000

140000

160000

0 100000 200000 300000 400000 500000 600000

Mass (ton)

Time (s)

21 gpm

50 gpm

75 gpm

100 gpm

150 gpm

300 gpm

Figure 83 RCS mass inventory – S0 scenarios (2.06 MPa)
59

5.2.4 S0 scenarios depressurized to 2.55 MPa (25 kp/cm2 gauge)

The results scenarios S0, in which the operator depressurizes and maintains SG pressure at 

2.55 MPa are shown in Figures 84 through 88. The larger the break is, the shorter the time 

available before core uncovery and core heatup is.

0

2

4

6

8

10

12

14

16

0 100000 200000 300000 400000 500000 600000

Pressure (MPa)

Time (s)

21 gpm

50 gpm

75 gpm

100 gpm

150 gpm

300 gpm

Figure 84 RCS pressure – S0 scenarios (2.55 MPa)
60

0

1

2

3

4

5

6

7

8

9

0 100000 200000 300000 400000 500000 600000

Pressure (MPa)

Time (s)

21 gpm

50 gpm

75 gpm

100 gpm

150 gpm

300 gpm

Figure 85 SG no.1 pressure – S0 scenarios (2.55 MPa)

0

20

40

60

80

100

120

0 100000 200000 300000 400000 500000 600000

Level (%)

Time (s)

21 gpm

50 gpm

75 gpm

100 gpm

150 gpm

300 gpm

Figure 86 Core collapsed liquid level – S0 scenarios (2.55 MPa)
61

400

600

800

1000

1200

1400

1600

0 100000 200000 300000 400000 500000 600000

Temperature (K)

Time (s)

21 gpm

50 gpm

75 gpm

100 gpm

150 gpm

300 gpm

Figure 87 Average fuel cladding temperature – S0 scenarios (2.55 MPa)

0

20

40

60

80

100

120

140

160

0 100000 200000 300000 400000 500000 600000

Mass (ton)

Time (s)

21 gpm

50 gpm

75 gpm

100 gpm

150 gpm

300 gpm

Figure 88 RCS mass inventory – S0 scenarios (2.55 MPa)
62

5.3 Dependence on equipment operation of SBO scenario with TD AFW lost 4 

hours after transient start

For each set of S1 scenarios (consisting of four cases) on equipment operation, when TD AFW 

was lost after four hours into transient. The following four variables are shown: RCS pressure, 

RCS inventory, average fuel cladding temperature and total mass discharged from RCS 

(through letdown isolation valve when RCS pressure greater than 4.23 MPa, both RCP seal 

leaks and PRZ PORVs – only scenarios S1-21p1 and S1-21p2). As the variables have already 

been plotted in Section 5.1 for all cases except RCS mass inventory (it provides new 

information), the Figures 89 through 104 will not be described in detail. Rather, some remarks 

will be made.

It should be also noted that similar studies making depressurization by PRZ PORVs as in this 

section have been performed before Fukushima accident for scenarios in which also all heat 

sinks were lost besides loss of all AC power (see Ref. 5).

5.3.1 S1 scenarios depressurized to 1.33 MPa (12.5 kp/cm2 gauge)

The results for scenarios S1, in which the operator depressurizes and maintains SG pressure at 

1.33 MPa are shown in Figures 69 through 73. It is clearly shown that use of PRZ PORVs 

causes faster heatup when not using them. The results also emphasized the need for a heat 

sink (TD AFW) since it is essential to prevent core damage. Also, PDP charging pump prolongs 

the time when core uncovery and heatup occurred.

Regarding use of PRZ PORV the conclusions derived in Ref. 5 should be considered: “However 

it can be concluded that if the operators would be able to open both pressurizer relief valves 

after the core heatup starts, this would have positive effect on further progression of the severe 

accident. As it can be concluded from the presented analyses by performing this action within 

SAMG procedures, primary pressure will be at the time of the primary system vessel failure 

significantly lower than in the case that there will be no operator actions for primary system 

depressurization. If in such situation operators would be able to open only one pressurizer 

PORV would this be beneficial for later accident progression.”
63

0

2

4

6

8

10

12

14

16

18

0 10000 20000 30000 40000 50000 60000

Pressure (MPa)

Time (s)

base

one PRZ PORV

two PRZ PORV

PDP pump

Figure 89 RCS pressure – S1 scenarios (1.33 MPa)

0

20

40

60

80

100

120

140

160

0 10000 20000 30000 40000 50000 60000

Mass (ton)

Time (s)

base

one PRZ PORV

two PRZ PORV

PDP pump

Figure 90 RCS mass inventory – S1 scenarios (1.33 MPa)
64

400

600

800

1000

1200

1400

1600

1800

0 10000 20000 30000 40000 50000 60000

Temperature (K)

Time (s)

base

one PRZ PORV

two PRZ PORV

PDP pump

Figure 91 Average fuel cladding temperature – S1 scenarios (1.33 MPa)

0

50

100

150

200

250

0 10000 20000 30000 40000 50000 60000

Mass (ton)

Time (s)

base

one PRZ PORV

two PRZ PORV

PDP pump

Figure 92 Total mass discharged from RCS – S1 scenarios (1.33 MPa)
65

5.3.2 S1 scenarios depressurized to 1.57 MPa (15 kp/cm2 gauge)

The results for scenarios S1, in which the operator depressurizes and maintains SG pressure at

1.57 MPa are shown in Figures 93 through 96. It is clearly shown that use of PRZ PORVs 

causes faster heatup when not using them. The results also emphasized the need for a heat 

sink (TD AFW) since it is essential to prevent core damage. Also, PDP charging pump prolongs 

the time when core uncovery and heatup occurred.

0

2

4

6

8

10

12

14

16

18

0 10000 20000 30000 40000 50000 60000

Pressure (MPa)

Time (s)

base

one PRZ PORV

two PRZ PORV

PDP pump

Figure 93 RCS pressure – S1 scenarios (1.57 MPa)
66

0

20

40

60

80

100

120

140

160

0 10000 20000 30000 40000 50000 60000

Mass (ton)

Time (s)

base

one PRZ PORV

two PRZ PORV

PDP pump

Figure 94 RCS mass inventory – S1 scenarios (1.57 MPa)

400

600

800

1000

1200

1400

1600

1800

0 10000 20000 30000 40000 50000 60000

Temperature (K)

Time (s)

base

one PRZ PORV

two PRZ PORV

PDP pump

Figure 95 Average fuel cladding temperature – S1 scenarios (1.57 MPa)
67

0

50

100

150

200

250

0 10000 20000 30000 40000 50000 60000

Mass (ton)

Time (s)

base

one PRZ PORV

two PRZ PORV

PDP pump

Figure 96 Total mass discharged from RCS – S1 scenarios (1.57 MPa)

5.3.3 S1 scenarios depressurized to 2.06 MPa (20 kp/cm2 gauge)

The results for scenarios S1, in which the operator depressurizes and maintains SG pressure at 

1.57 MPa are shown in Figures 97 through 100. It is clearly shown that use of PRZ PORVs 

causes faster heatup when not using them. The results also emphasized the need for a heat 

sink (TD AFW) since it is essential to prevent core damage. Also, PDP charging pump prolongs 

the time when core uncovery and heatup occurred.
68

0

2

4

6

8

10

12

14

16

18

0 10000 20000 30000 40000 50000 60000

Pressure (MPa)

Time (s)

base

one PRZ PORV

two PRZ PORV

PDP pump

Figure 97 RCS pressure – S1 scenarios (2.06 MPa)

0

20

40

60

80

100

120

140

160

0 10000 20000 30000 40000 50000 60000

Mass (ton)

Time (s)

base

one PRZ PORV

two PRZ PORV

PDP pump

Figure 98 RCS mass inventory – S1 scenarios (2.06 MPa)
69

400

600

800

1000

1200

1400

1600

1800

0 10000 20000 30000 40000 50000 60000

Temperature (K)

Time (s)

base

one PRZ PORV

two PRZ PORV

PDP pump

Figure 99 Average fuel cladding temperature – S1 scenarios (2.06 MPa)

0

50

100

150

200

0 10000 20000 30000 40000 50000 60000

Mass (ton)

Time (s)

base

one PRZ PORV

two PRZ PORV

PDP pump

Figure 100 Total mass discharged from RCS – S1 scenarios (2.06 MPa)
70

5.3.4 S1 scenarios depressurized to 2.55 MPa (25 kp/cm2 gauge)

The results for scenarios S1, in which the operator depressurizes and maintains SG pressure at 

2.55 MPa are shown in Figures 101 through 104. It is clearly shown that use of PRZ PORVs 

causes faster heatup when not using them. The results also emphasized the need for a heat 

sink (TD AFW) since it is essential to prevent core damage. Also, PDP charging pump prolongs 

the time when core uncovery and heatup occurred.

0

2

4

6

8

10

12

14

16

18

0 10000 20000 30000 40000 50000 60000

Pressure (MPa)

Time (s)

base

one PRZ PORV

two PRZ PORV

PDP pump

Figure 101 RCS pressure – S1 scenarios (2.55 MPa)
71

0

20

40

60

80

100

120

140

160

0 10000 20000 30000 40000 50000 60000

Mass (ton)

Time (s)

base

one PRZ PORV

two PRZ PORV

PDP pump

Figure 102 RCS mass inventory – S1 scenarios (2.55 MPa)

400

600

800

1000

1200

1400

1600

1800

0 10000 20000 30000 40000 50000 60000

Temperature (K)

Time (s)

base

one PRZ PORV

two PRZ PORV

PDP pump

Figure 103 Average fuel cladding temperature – S1 scenarios (2.55 MPa)
72

0

20

40

60

80

100

120

140

160

180

200

0 10000 20000 30000 40000 50000 60000

Mass (ton)

Time (s)

base

one PRZ PORV

two PRZ PORV

PDP pump

Figure 104 Total mass discharged from RCS – S1 scenarios (2.55 MPa)
73

6. RUN STATISTICS

The calculations with the RELAP5/MOD3.3 Patch 04 computer code (linux version relap5-33iylinux-ifc-opt.x) were performed on cluster Krn with 50 nodes and 600 processor cores. Each 

node has two Intel Xeon 5670 @ 2.93 GHz processor, each having 6 cores and 6 threads. The 

operating system is SUSE Linux Enterprise Server 11 (x86_64) - service pack 1.

Table 2 shows the run statistics for base calculation (scenario S0-21, depressurization 

2.55 MPa). For other calculations the statistic is similar for the same length of run. If runs are 

shorter, the CPU time is smaller accordingly. For all calculations, the number of volumes was 

432. The calculations run five times faster than real time. Steady-state calculations for all runs 

lasted 1,000 seconds and required 231.2 seconds of central processing unit (CPU) and 25,948

steps.

Table 2 Run statistics

Transient Time

(s)

CPU Time

(s)

CPU/Transient 

Time

Number of Time 

Steps

604800 119206 0.197 15255188

75

7. CONCLUSIONS

In this report long term station blackout sequences for Krško two-loop pressurized water reactor 

with loss of normal or all secondary side heat sinks were studied. For calculations the latest 

RELAP5/MOD3.3 Patch 04 computer code was used. The verified standard RELAP5/MOD3.3 

input model from 2008 (cycle 23) was delivered by Krško nuclear power plant.

SBO scenarios were analyzed for different RCP leak seals. Besides, scenarios were performed 

for different primary side depressurizations performed by operator through the secondary side 

power operated relief valves, providing that turbine driven auxiliary feedwater pump is available. 

The results showed that when RCP seal leaks are small the core uncovery could be prevented 

in the first seven days by using TD AFW pump and manually depressurizing the RCS through 

SGs depressurization. When RCP seal leaks are larger, small capacity RCS makeup is needed 

in addition to TD AFW pump to prevent core uncovery and core heatup. It was also shown that 

with TD AFW not available after 4 hours after around 10 hours the core will start to heatup.

The results clearly showed that alternative RCS makeup (when emergency core cooling system

is lost) is also very important for preventing core uncovery; especially in the cases with larger 

breaks this is the only way to prevent core heatup in the first seven days. It can be concluded, 

provided that TD AFW and some RCS injection of the order of 2 kg/s mass flowrate are 

available, RCP seal leaks are of no concern. If only TD AFW is available, it is very important to 

limit the RCP seal leaks. One of the ways to limit the RCP seal leaks is manual depressurization 

strategies, which is therefore very important in the absence of RCS makeup. Primary side 

depressurization is of very limited use in preventing core heatup. However, would have positive 

effect on further progression of the severe accident according to study in Ref. 5.

Finally, the results suggest that RELAP5 can be used for extended SBO studies until core 

damage started. It is especially useful in studying maintaining core cooling function and time 

available before core uncovers as part of severe accident management. The benefit of using 

RELAP5 is in the fact that best estimate system codes are more accurate than severe accident 

codes in phases before core degradation started.

77

8. REFERENCES

1. WENRA, »Stress tests« specifications, Proposal by the WENRA Task force, 21 April 

2011.

2. B. Krajnc, B. Glaser, R. Jalovec, S. Špalj, “MAAP Station Blackout Analyses as a 

Support for the NPP Krško STORE (Safety Terms of Reference) Actions”, Proc. of 

International Conference Nuclear Energy for New Europe NENE-2012, Bovec, Slovenia, 

September 12-15, 2011.

3. N. R. LaBarge, B. R. Baron, R. E. Schneider, M. C. Jacob, “Comparison of thermal 

hydraulic simulations of beyond design basis events using the MAAP4 and CENTS 

computer codes”, Proceedings of the 17th International Conference on Nuclear 

Engineering ICONE17, Brussels, Belgium, July 12-16, 2009.

4. P. Hofmann, “Current knowledge on core degradation phenomena, a review”, Journal of 

Nuclear Materials, Vol. 270, Issues 1-2, 1 April 1999, pp. 194-211.

5. I. Parzer, B. Krajnc, B. Mavko, “Analyzing Operator Actions During Loss of AC Power 

Accident with Subsequent Loss of Secondary Heat Sink”, NUREG/IA-0225, April 2010.

6. A. Prošek, I. Parzer, and B. Krajnc, “Simulation of hypothetical small-break loss-ofcoolant accident in modernized nuclear power plant”, Electrotechnical Review, Vol. 71, 

No. 4: 2004.

7. I. Parzer, B. Mavko, and B. Krajnc, “Simulation of a hypothetical loss-of-feedwater 

accident in a modernized nuclear power plant”, Journal of Mechanical Engineering, Vol. 

49, No. 9: 2003.

8. I. Parzer, “Break model comparison in different RELAP5 versions”, Proc. of International 

Conference Nuclear Energy for New Europe 2003. Nuclear Society of Slovenia (NSS), 

Portorož, September 8-11, 2003. Nuclear Society of Slovenia: Ljubljana, Slovenia. 2003.




UNITED STATES 

NUCLEAR REGULATORY COMMISSION

WASHINGTON, DC 20555-0001

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OFFICIAL BUSINESS
NUREG/IA-0424 RELAP5 Extended Station Blackout Analyses