NUREG/IA-0225

 International 

Agreement Report

Analyzing Operator Actions 

During Loss of AC Power 

Accident with Subsequent

Loss of Secondary Heat Sink 

Prepared by:

I. Parzer1, B. Krajnc2, B. Mavko1

1

Jožef Stefan Institute

Jamova cesta 39

SI-1000, Ljubljana, slovenia

2NPP Krško

Vrbina 12 

S1-8270 Krško, Slovenia

A. Calvo, NRC Project Manager 

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

April 2010 

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-0225

Analyzing Operator Actions During Loss of 

AC Power Accident with Subsequent Loss 

of Secondary Heat Sink 

Manuscript Completed: December 2004 

Date Published: April 2010 

Prepared by: Iztok Parzer1

, Božidar Krajnc2, Borut Mavko1

Jožef Stefan Institute 

Jamova cesta 39 

SI-1000 Ljubljana, Slovenia 

NPP Krško 

Vrbina 12 

SI-8270 Krško, Slovenia 

Antony Calvo, NRC Project Manager 

Prepared for:

 

Office of Nuclear Regulatory Research 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-0001

iii 

ABSTRACT

The thermo-hydraulic analysis of plant response on total loss of AC power is very demanding 

and challenging job due to a number of phenomena included. Such an analysis becomes even 

more complicated and interesting if we include also the assumption of total loss of secondary 

heat sink. 

In this paper we are presenting the NPP Krško specific analysis of complete loss of AC power 

with subsequent total loss of secondary heat sink and influence of specific operator actions. The 

aim of this analysis is to verify if emergency operating (EOP) or severe accident management 

guidelines (SAMG) procedures should be changed and if design change on pressurizer 

pressure relief valves (PORV) should be implemented to be able to cope with this kind of 

accidents better. 

The analyses were performed with three different state of the art codes used at NPP Krško and 

IJS: RELAP5/MOD3.3, ANTHEM and MAAP4. The last two codes are used in the NPP Krško 

plant specific full scope simulator, one for the simulation of the design bases transients and 

accidents and the second for simulation of the severe accidents. 

This type of analyses has been done also for the simulator validation, performed during vendor 

and site acceptance testing. 

v 

CONTENTS

Page

Abstract ........................................................................................................................... iii

Acknowledgement ......................................................................................................... viii

Abbreviations .................................................................................................................. ix

1. Introduction ............................................................................................................. 1-1

2. Plant description ..................................................................................................... 2-1

2.1 Containment ................................................................................................................... 2-1

2.2 Turbine Building ............................................................................................................. 2-1

2.3 Auxiliary Building ............................................................................................................ 2-1

2.4 Intake Structure .............................................................................................................. 2-1

2.5 Fuel Handling Building ................................................................................................... 2-1

2.6 Nuclear Steam Supply System ....................................................................................... 2-2

2.7 Reactor Core .................................................................................................................. 2-2

2.8 Reactor Coolant System ................................................................................................ 2-2

2.9 Auxiliary Systems ........................................................................................................... 2-3

2.10 Engineered Safety Features Systems .................................................................... 2-3

3. Accident progression and operator actions ............................................................. 3-1

4. Input model description ........................................................................................... 4-1

4.1 Hydrodynamic component description ........................................................................... 4-1

4.2 Regulation and protection logic ...................................................................................... 4-4

5. Results .................................................................................................................... 5-1

5.1 Base analyses ................................................................................................................ 5-1

5.2 Variation analyses ........................................................................................................ 5-10

6. RELAP5/MOD3.3 run statistics ............................................................................... 6-1

7. Conclusions ............................................................................................................ 7-1

8. References.............................................................................................................. 8-1
vi 

Figures

Page

1. NPP Krško nodalization scheme ........................................................................................... 4-2

2. Primary system pressure - RELAP5/MOD3.3 ....................................................................... 5-4

3. Primary system pressure - SIM- ANTHEM ............................................................................ 5-4

4. Primary system pressure - MAAP4 ....................................................................................... 5-5

5. Total primary leak - RELAP5/MOD3.3 .................................................................................. 5-5

6. SG 1 valves flow - RELAP5/MOD3.3 .................................................................................... 5-6

7. SG 1 WR level - RELAP5/MOD3.3 ....................................................................................... 5-6

8. Accumulator no.1 level - RELAP5/MOD3.3 ........................................................................... 5-7

9. PRZ level - RELAP5/MOD3.3 ............................................................................................... 5-7

10. Primary mass - RELAP5/MOD3.3 ........................................................................................ 5-8

11. Core level - RELAP5/MOD3.3 .............................................................................................. 5-8

12. Rod cladding temperature - RELAP5/MOD3.3 .................................................................... 5-9

13. Core exit temperature - RELAP5/MOD3.3 ........................................................................... 5-9

14. Primary system pressure without and with depressurization using PRZ relief valves – 

MAAP4 ..................................................................................................................................... 5-11

15. Primary system pressure without and with depressurization using PRZ relief valves – 

RELAP5/MOD3.3 ..................................................................................................................... 5-11

16. Consumed CPU time for all base and variation analyses cases– RELAP5/MOD3.3 .......... 6-1

17. Mass error for 3 base analyses cases– RELAP5/MOD3.3 .................................................. 6-3

18. Time step for 3 base analyses cases– RELAP5/MOD3.3 .................................................... 6-3

19. Courant Δt for 3 base analyses cases– RELAP5/MOD3.3 .................................................. 6-4

20. Mass error for 3 variation analyses cases– RELAP5/MOD3.3 ............................................ 6-4

21. Time step for 3 variation analyses cases– RELAP5/MOD3.3 .............................................. 6-5

22. Courant Δt for 3 variation analyses cases– RELAP5/MOD3.3 ............................................ 6-5
vii 

Tables

Page

1. Loss of all AC power - assumptions common for all analyses .............................................. 5-2

2. Comparison of available time - time until the core heatup - degradations starts................... 5-2

3. Loss of all AC power - assumptions for MAAP analyses .................................................... 5-10

4. Run time statistics ................................................................................................................. 6-1
viii 

ACKNOWLEDGEMENT 

The RELAP5/MOD3.3 NPP Krško base input model, nodalization diagram and plant full scope 

simulator data are courtesy of Krško NPP. 
ix 

ABBREVIATIONS 

AC Alternating Current 

AFW Auxiliary Feedwater 

CC Component Cooling 

CET Core Exit Temperature 

CFR Code of Federal Regulations 

CVCS Chemical and Volume Control System 

ECCS Emergency Core Cooling system 

EOP Emergency Operating Procedures 

HPIS High Pressure Injection System 

IJS Institut “Jožef Stefan” 

LPIS Low Pressure Injection System 

MFW Main Feedwater 

NPP Nuclear Power Plant 

NSSS Nuclear Steam Supply System 

PORV Power Operated Relief Valve 

PRZ Pressurizer 

PWR Pressurized Water Reactor 

RCS Reactor Coolant System 

RCP Reactor Coolant Pump 

RHRS Residual Heat Removal System 

SAMG Severe Accident Management Guidelines 

SG Steam Generator 

TSC Technical Support Center

1-1 

1. INTRODUCTION

Station blackout is the accident where it is assumed that all plant AC power sources are lost. 

That means that all the offsite AC power sources are unavailable and at the same time it is also 

assumed all AC sources on the site are lost. Within this paper we are going to discuss 

concurrent total loss of all AC and loss of secondary heat sink (no SG feed available). One 

should know that this type of the accident is beyond the design bases of all the nuclear power 

plants currently operating around the word. If AC power is not recovered within certain time this 

scenario will lead to the severe accident sequence of events when the core damage and 

degradation becomes unavoidable. 

This type of the accidents would also result in loss of coolant accident due to the fact that the 

operators would not be able to isolate the letdown and reactor coolant pump seal leakoff lines. 

This is true in case of NPP Krško and many other plants of same or similar design, since 

isolation valves on those lines are motor driven valves. As a consequence of loss of seal 

injection and the fact that the thermal barrier cooling is also lost, reactor coolant pumps (RCP) 

seals are exposed to high temperature of the primary coolant. This will cause seals degradation 

and seal leakoff flow will develop to a break flow. This is one of the well-known safety issues, 

which are known as the unresolved safety issues. This one was recognized as no.25 [1]. As a 

response on this issue the industry initiated a large research program, which resulted in certain 

seal improvements. New seals have higher temperature resistance so it is predicted that instead 

after approximately 30 minutes they would fail after approximately 2 hours. 

If the AC power is not recovered before the core is uncovered and significantly overheated, core 

melt and consequently primary system failure from high pressure could not be avoided. As soon 

as the core exit temperature is higher than 650  ̊C for more than 30 minutes it is considered that 

accident becomes a severe accident. For this type of the accidents nuclear industry, including 

NPP Krško, have developed so called Severe Accident Management Guidelines (SAMG), 

where the main objectives are to maintain fission products barriers intact (primary system and 

containment). If this is not possible or in case that release of fission product is in progress then 

these procedures (SAMG) will suggest actions to minimize the fission products release to the 

environment, with the purpose to protect the public health. 

2-1 

2. PLANT DESCRIPTION 

NPP Krško is a Westinghouse 2-loop PWR plant, in commercial operation since 1983. 

2.1 Containment

The NPP Krško Nuclear Power Plant utilizes a cylindrical steel shell with a hemispherical dome 

and ellipsoidal bottom designed to accommodate normal operating loads, functional loads 

resulting from a loss-of-coolant accident, and the most severe loading predicted for seismic 

activity. A concrete shield building surrounds the steel shell to provide biological shielding for 

both normal and accident conditions and to provide collection and holdup for leakage from the 

containment vessel. Inside the containment structure, the reactor and other NSSS components 

are shielded with concrete. In addition to a containment spray system, a containment 

recirculation and cooling system is provided to remove post-accident heat. 

2.2 Turbine Building

The turbine building contains the turbine generator and all the power conversion related 

accessories. The building is of closed construction. The building does not contain any safety 

related equipment and is designed in accordance with local and national building codes. 

2.3 Auxiliary Building

The auxiliary building structures are of reinforced concrete design with shear walls and beam 

and slab floor systems. The portion of the auxiliary building that is below grade elevation is 

suitably protected with a waterproofing membrane to prevent the intrusion of groundwater. In 

addition, redundant safety equipment below grade is located in separate compartments to 

preclude simultaneous flooding due to a fluid-system rupture. 

2.4 Intake Structure

The intake structure consists of two separate substructures: a non-safety category structure 

containing the main condenser circulating pumps and related equipment and a safety category 

structure containing the service water pumps and the related equipment. For cooling water 

intake, a dam has been built across the Sava River with the pumping station, and water intake 

and discharge structures. Two batteries of cooling cells are included for combined cooling in the 

event of low river flow rates. 

2.5 Fuel Handling Building

The fuel handling building is an integral part of the auxiliary building and is a reinforced concrete 

structure that utilizes shear walls and beam and slab floor systems. The spent fuel pool within 

the fuel handling building is lined with stainless steel to prevent leakage of water. 
2-2 

2.6 Nuclear Steam Supply System

The power rating of the NPP Krško nuclear steam supply system (NSSS) is 2000 MWt, 

composed of 1994 MWt core power output plus 6 MWt of reactor coolant pump heat input. The 

NSSS consists of a pressurized water reactor, reactor coolant system (RCS) and associated 

auxiliary fluid systems. The RCS is arranged as two closed reactor coolant loops connected in 

parallel to the reactor vessel, each containing a reactor coolant pump and a steam generator. 

An electrically heated pressurizer is connected to one of the loops. 

2.7 Reactor Core

The reactor core is composed of 121 fuel assemblies. Square spacer grid assemblies and the 

upper and lower end fitting assemblies support the fuel rods in fuel assemblies. Each fuel 

assembly is composed of 16 x 16 rods; of these only 235 places are used by fuel rods; of the 21 

remaining places, 20 places which are evenly and symmetrically distributed across the cross 

section of the assembly, are provided with thimble tubes which may be reserved for control 

rods, and one control instrumentation tube for incore thimble. 

Of all fuel assemblies in the core, 33 are equipped with control rod clusters. The core is of the 

multi-region type. All fuel assemblies are mechanically identical, although the fuel enrichment is 

not the same in all assemblies. Fuel assemblies with the highest enrichments were placed in 

the core periphery, and the two groups of lower enrichment fuel were arranged in a selected 

pattern in the central region. Core design strategy depends on plant operation strategy (length 

of cycle) and improvements in fuel design. 

2.8 Reactor Coolant System

The RCS consists of two reactor coolant loops connected in parallel to the reactor vessel, each 

loop containing a reactor coolant pump and a steam generator. The coolant loops are filled with 

high pressure water driven by the reactor coolant pumps. The water circulates through the 

reactor core to remove the heat from the fuel assemblies generated by the nuclear chain 

reaction. The heated water exits from the reactor vessel and passes via the coolant loop piping 

to the steam generator. Here it gives up its heat to the feedwater to generate steam for the 

turbine generator. 

The reactor coolant pumps, one per coolant loop, are Westinghouse vertical, single-stage, 

centrifugal pumps of the shaft-seal type. The power supply system to the pumps is designed so 

that adequate coolant flow is maintained to cool the reactor core under all conceivable 

circumstances. The pump capacity is about 17,000 t/h. All pump parts in contact with the 

coolant are made of austenitic steel or stainless steel covered. 

The steam generators, one per loop, are vertical U-tube units, recently installed SiemensFramatome steam generators type SG 72 W/D4-2, which replaced highly degraded 

Westinghouse D-4 steam generators with preheater. Internal moisture separation equipment 

reduces the moisture content of steam to 0.1 % or less. 

The reactor coolant piping and all of the pressure-containing and heat transfer surfaces in 
2-3 

contact with reactor water are stainless steel except the steam generator tubes and fuel tubes, 

which are Inconel and zircaloy, respectively. Reactor core internals, including control rod drive 

shafts, are primarily stainless steel. 

An electrically heated pressurizer connected to one reactor coolant loop maintains RCS 

pressure during normal operation, limits pressure variations during plant load transients, and 

keeps system pressure within design limits during abnormal conditions. 

2.9 Auxiliary Systems

Auxiliary systems components are provided to charge the Reactor Coolant System and to add 

makeup water, purify reactor coolant water, provide chemicals for corrosion inhibition and 

reactivity control (CVCS – Chemical and Volume Control System), cool system components 

(CC – Component Cooling), remove residual heat when the reactor is shut down (RHRS - 

Residual Heat Removal System), change fuel and cool the spent fuel storage pool, sample the 

reactor coolant water, provide for safety emergency injection, and vent and drain the Reactor 

Coolant System. 

2.10 Engineered Safety Features Systems

Engineered safety features are provided to prevent accident propagation or to limit the 

consequences of postulated accidents, which might otherwise lead to damage of the system 

and release of fission products. The principal criteria is that under the conditions of a 

hypothetical loss of coolant accident, the system can, even when operating with partial 

effectiveness, maintain the integrity of the containment, and limit the potential offsite radiation 

dose to less than the values of applicable US Federal Regulations (10 CFR Part 100). A number 

of engineered safety features are included in this plant, as follows: 

$ Containment Spray System 

$ Hydrogen Control System 

$ Emergency Core Cooling System 

$ Component Cooling Water System 

$ Essential Service Water System 

$ Auxiliary Feedwater System 

3-1 

3. ACCIDENT PROGRESSION AND OPERATOR ACTIONS

Current emergency operating procedures cover design bases accidents and also a part of the 

beyond design bases accidents. One should know that this type of the procedures have been 

enhanced and completely revised after Three Mile Island accident in beginning of eighties of the 

past century. Emergency operating procedures have been developed with the goal to cover 

accidents with initiating frequency higher than 1E-6 per year. They have been developed on a 

bases that the operators do not need to know what kind of accident is in progress, that's why 

they are also called as "symptom based" and not "event based" as before. Among these 

procedures one covers the case in which all AC power is lost. Within this procedure operators 

are instructed to try to restore the offsite or onsite AC power source, to power at least one train 

of emergency core cooling systems. In the meantime the operators are instructed to isolate 

paths through which primary coolant is lost - mainly letdown and reactor coolant pump seal 

leakoff lines. Due to high primary coolant temperature and no thermal barrier cooling, reactor 

coolant pump seal degradation is expected due to high temperature seal leakoff flow. As a 

consequence a relatively small leak flow will be increased to a break flow with the upper bound 

value of 25 gpm [2]. If the secondary heat sink is available the operators are instructed to 

cooldown and depressurize the primary system with secondary heat sink - steam generators. 

There are the following main benefits of this action. The first is to decrease the leakage - break 

flow of the primary coolant due to lower primary pressure and the second is that if the primary 

pressure will decrease low enough so the passive injection of additional borated water from the 

accumulators can be expected. By this we would gain some cold borated water and increase 

the mass of the primary coolant. As a consequence operators are gaining some additional time 

for AC power restoration. 

If in a case of loss of all AC power we also assume the failure of the turbine driven auxiliary 

feedwater pump, this will cause also the loss of secondary heat sink. Per the current EOP 

procedures [3] operators are not instructed to initiate primary system cooldown by opening of 

the steam generators pressure control valve(s). They are instructed to loop in the procedure, 

trying to restore secondary heat sink and AC power. Due to that the primary temperature and 

pressure will, after all the secondary side steam generator water will boil off, start to increase 

(see Figure 1, 2 or 3 case with 0 SG and 0 PRZ PORVs). Available analyses [2], [4], [5], [6] 

indicate that the break flow through the letdown line and reactor coolant seals will not be large 

enough to bring the pressure below the accumulator injection pressure. Above-mentioned seal 

improvements support such a conclusion even further. 

Since the pressure and the temperature will increase in the primary system and the fact that 

pressurizer pressure relief control valves are not operable due to the fact that there is no 

instrument air available, primary safety valves will control the pressure in the reactor coolant 

system. Consequently the break flow rate will significantly increase and speed up the coolant 

depletion. This will cause the core uncovery and heatup of the core up and above the 650 o

C. At 

this point the operators are instructed to leave emergency operating procedures and enter the 

severe accident management guidelines. Within these procedures [7] technical support center 

(TSC), which is responsible for making the decisions and instructions for the operators, will not 

be able to provide the instructions how to decrease the primary system pressure. For further 

development of this type of the accident it is beneficial that the primary system fails at low 
3-2 

internal pressure (reactor vessel failure or hot leg creep failure) [5], [7]. 

Even with use of SAMG procedures TSC and the operators will not be able to prevent further 

core degradation without any AC power. Eventually core will melt and primary system will fail 

from high internal pressure (see Figure 3 and 4). Operators and plant technical support team 

will then focus on trying to prevent containment failure and to minimize fission product release 

to the environment. 

As can bee seen from the above discussion it is essential to try to meet the following objectives 

during such an accident: 

$ prolong as much as possible the time until the core will be overheated, melted and the 

primary system would fail - to gain additional time for AC power restoration and 

$ decrease the primary pressure to prevent high pressure reactor vessel failure, which is 

causing potential threat to the containment integrity due to so call direct containment 

heating phenomena [4]. Small very hot particles could cause very rapid containment 

pressure increase and its failure. In case that the containment reactor vessel would fail 

at low pressure, core debris would be relocated into the reactor cavity and no immediate 

threat to the containment integrity is expected [5]. 

To get appreciation of certain key operator actions and to see if it is feasible also to consider 

any potential design change in the future, we decided to perform further analyses of this 

transient. 
4-1 

4. INPUT MODEL DESCRIPTION 

To perform this analysis, NPP Krško has provided the base input model, so called “Master input 

deck”. The scheme of the NPP Krško nodalization for the RELAP5/MOD3.3 [8] code is 

presented in Figure 1. A full two-loop plant model was developed, including the new SiemensFramatome replacement steam generators (RSG) type SG 72 W/D4-2. 

The model consists of 469 volumes, connected with 497 junctions. Plant structure is 

represented by 376 heat structures with 2101 mesh points, while the reactor protection and 

regulation systems, safety systems operational logic and plant instrumentation is represented by 

401 logical conditions (trips) and 575 control variables. 

4.1 Hydrodynamic component description

Components numbered from 101 to 165 represent reactor vessel in the following manner: 

171, 173 and 175 - lower downcomer 

101 and 103 - lower head 

105 - lower plenum 

107 - core inlet 

111 - reactor core 

115 - core baffle bypass 

121 - core outlet 

125, 131 and 141 - upper plenum 

151 and 153 - upper head 

165 - upper downcomer 

113 and 145 - guide tubes 

Components numbered 51, 53 and 55 represent the pressurizer surge line and volumes 61, 63, 

65, 67 and 69 represent the pressurizer vessel. Pressurizer spray lines (80, 81 and 84) are 

connected to the top of the pressurizer vessel and include the spray valves numbered 82 and 

83. Valves numbered 28 and 32 represent the two pressurizer PORVs and valves numbered 14 

and 22 represent pressurizer safety valves. 

Primary piping is represented by the following components: 

201, 203, 205, 207, 209 and 211 - hot leg no.1 

251, 253, 255, 257 and 259 - intermediate leg no.1 with cold leg no.1 loop seal 

265, 271, 273, 275, 277 and 279 - cold leg no.1 with the primary coolant pump no.1 

301, 303, 305, 307, 309 and 311 - hot leg no.2 

351, 353, 355, 357 and 359 - intermediate leg no.2 with cold leg no.2 loop seal 

365, 371, 373, 375, 377 and 379 - cold leg no.2 with the primary coolant pump no.2 

Loops are symmetrical except for the pressurizer surge line and CVCS connections layout. 
151

171

121

111 0605

04030201

115

01

020304

05

06

07

08

175

107

105

103

101

69

656361

201 203 205 207

51

209

53

01020304 05 060708

11

10

09

55

257

01 02 03 04

259

265 271 273 275 277 279

313

379 377 375 373 371 365

359

357

353

355

01 02 03 04

551

553555

557

559

561

301 303 305 307 309 311

211

253

255

351

906

905

11

10

09

08

0706

05

04

03

02

01

67

901

902

165

131

171

173

0201

719

718

717

715

713 711

801

811 813

815

817

819

818

02 01

02

01

992

991

993

701

Base deck ID:

branchpipetime dependent volume

pumpaccumulatorjunction orientation

crossflow junctiontime dependent junctionvalve

601

603

605

05

04

03

02

01

604

598

271 371

01 02

04

03

05

02

03 04 05

07

06 80 81

0201

03 04

84

82 83

90

91

96

95

88

93

86

92

912 911

913 914

171

727725723

722

721

720

705 805

820

745

743

746 748

749747

872

882

871 881 861

760 860

857

853

851

757

753751

731

732

740

741

831

832

727

582

584

586

588

592

594

951 953

01

08

01

02

02

01

01

0101

01

02

01

01

01

02

01 02

01

141

0302

01

131125

02

02

01

02

01

01 02

03

01 02

01 01 01 01 01 01 01

01

01

02 01

01

02 01

02

01 01 01 01

01

703

01 01

01

01

01

01

01

01 02

01 01

750

01

01

01

803

01 01

01

01

01

850

01

01

01

02

01

02

01 01

375 377

01 01

01

01

01

01

01 01 01 01 01 01

01

01

01

01

02

0101

02 01

02

01 01 01 01

xxx junction described at branch xxx

915

804 704

612

611

916

03

145

04

03

02

01

131

05

ACC 1 ACC 2

RCP 2 RCP 1

SG 2

RPV

PRZ

charging

letdown

451

453

455

457

459

461

217

215 245

243

219 241

933

223

225

07

06

05

04

03

02

01

227

237

08

429

427

423

415

010203040506

411

213

251

0201

413

484

486

488492494

951 953

482

498

03 04 05

01

01

01

01

02

02

01

01

01

0101

01

02

03

01

01

01

02

01

01 01

01

01

01

01

601

SG 1

05

04

030201

417

421

425

419

02 934

443

447

445

01

01

02

01

01

01

317

315345

343

319 341

935

323

325

07

06

0504

03

02

01

327

337

08

529

527

523

515

511

513

01

01

01

02

02

01

01

01

01

01

01

02

03

01

01

01

02

01

517

521

525

519

02 936

543 545

547

01

01

02

01

01

01

113

01

01

nekCClrv

458 558

01

02

03 04

013

014

01

02

03 04

021

022 028

027

032

01

02 03 04

025 011

01

07 06 05 04

03

02

01

407 409

930

931

01

07 060504

0302

01

507

233

235

01

0102030405060708

333

335

01

010203040506

07

08

121110

09

0807

121110

09

080706

05

04030201

121110090807060504030201

091011

121314

06

153

01

02

01

07

08

09

10

09

08

0706

01

0203040506

07

08

09

10

0908

07

06

05

04030201

509

475

0706

0504030201

406

474

473

0504030201

472

471500

930

931

575

0706

0504030201

506

574

573

04

03

0201

572

571500

939

938

01

02 03

671

672 01 02 03 673 435

941

940

01

02

681

683

01

02

685

686

695

02 01 03

696

687 02

01

697

02

01

943

942

01

02 03

675

676

677

01 02 03 04

535

01 02 03 04 05 06

07 08

10

09

437

11

01

02 03 04 05

08

07

06

537

01

02

03

0101

01 01

01

02

945

944

947

946

949

948

978

976

974 972

443

543

437 537

MD AFWpump 01

TD AFWpumpMD AFWpump 02

HPIS

LPIS

MFWheader

MFWheader

674

688

678

698

01

607

606

Figure 1: NPP Krško nodalization scheme 

4-2 
4-3 

ECCS piping nodalization and connections are represented by hydrodynamic components 

numbered from 701 to 882. The hydrodynamic components representing HPIS pumps are timedependent junctions 703 and 803, while time-dependent junctions 750 and 850 represent LPIS 

pumps. Accumulators are numbered 701 and 801 their lineup provides cold leg injection only. 

ECCS connects to both cold legs (junctions 719-01 and 819-01). Direct vessel ECCS injection 

through junction no. 746 and 748 opens simultaneously at the SI signal generation. 

Primary side of the SG is represented by inlet and outlet plenum, among which a single pipe is 

representing the U-tube bundle: 

215, 217 and 219 - - SG 1 inlet plenum (hot side) and tube sheet inlet 

223, 225, 227, 233, 235, and 237 - SG 1 U-tubes 

241, 243 and 245 - SG 1 tube sheet outlet and outlet plenum (cold side) 

315, 317 and 319 - - SG 2 inlet plenum (hot side) and tube sheet inlet 

323, 325, 327, 333, 335, and 337 - SG 2 U-tubes 

341, 343 and 345 - SG 2 tube sheet outlet and outlet plenum (cold side) 

The parts of the SG secondary side are represented by the following hydrodynamic 

components: 

415, 417 and 419 - SG 1 riser 

421 and 427 - SG 1 separator and separator pool 

411 and 413 - SG 1 downcomer 

423, 425 and 429 - SG 1 steam dome 

515, 517 and 519 - SG 2 riser 

521 and 527 - SG 2 separator and separator pool 

511 and 513 - SG 2 downcomer 

523, 525 and 529 - SG 2 steam dome 

Main steamlines are represented by volumes 451, 453, 455, 457, 459 and 461 (SG 1) and 551, 

553, 555, 557, 559 and 561 (SG 1), divided by main steam isolation valves (458 and 558). SG 

relief (482 and 582) and safety valves (484, 486, 488, 492, 494 and 584, 586, 588, 592, 594) 

are situated upstream the isolation valves. Turbine valve (604) and steam dump (611) flow is 

regulated by corresponding logic. 

Main feedwater (MFW) piping is represented by volumes 471, 473, 475, 407, 409 (SG 1) and 

471, 573, 575, 507, 509 (SG 2), branching from main feedwater header (500). 

Auxiliary feedwater (AFW) is injecting above the SG riser (via volumes 437, 443, 445 and 447) 

and its piping is represented by volumes 671, 673 (motor driven AFW 1), 675, 677 (AFW 2), and 

681, 683, 685, 687, 695, 697 (turbine driven AFW). 
4-4 

4.2 Regulation and protection logic

In order to accurately represent the NEK 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 main plant control 

systems: 

$ rod control system, 

$ PRZ pressure control system, 

$ PRZ level control system, 

$ SG level control system and 

$ steam dump. 

It must be noted that rod control system has been modeled for point kinetics. Present model can 

be used for transient analysis with two options: 

$ with constant or predefined core power transient as function of time (including decay 

power calculation) or 

$ with rod control system in auto or manual mode. 

The following plant protection systems are defined using trip logic: 

$ reactor trip, 

$ SI signal, 

$ turbine trip, 

$ steamline isolation, 

$ main feedwater isolation and 

$ auxiliary feedwater start. 
5-1 

5. RESULTS 

For the reason stated above the analyses with the state of the art tools were performed to check 

below listed potential procedure and/or design change: 

$ Potential emergency operating procedure change by which the operators would be 

instructed to start secondary side depressurization even in the case if there is no feed 

into SG. 

$ If the potential design change which will assure pressurizer PORV operability under total 

loss of all AC power would increase available time before the core heatup starts (to open 

the primary system bleed path with the purpose to get feed from the accumulators) and 

would depressurize primary system enough after core degradation and relocation. 

Since this accident is very complicated and the operator actions are very important the following 

tools have been used to perform required analyses: 

$ RELAP5/MOD3.3 (for the analyses of available time before the core degradations 

starts), 

$ ANTHEM (for the analyses of available time before the core degradations starts) and 

$ MAAP4 (for the analyses of available time before the core degradations starts and 

analysis for potential primary system depressurization at the point of entrance to the 

SAMG procedures and no operator actions for primary system cooldown and 

depressurization was performed before). 

For above mentioned, the analyses of the accident up to the point of fuel heatup due to low 

coolant level in the core was first performed with the RELAP5/MOD3.3 code [7], [9] and with the 

NPP Krško plant engineering simulator using ANTHEM. 

ANTHEM [10], [11] is the two phase 5 equation (plus conservation equation for noncondensable mass), drift flux code build in the simulator for the simulation of the nuclear steam 

supply systems. With this plant engineering simulator we have capability to simulate all the 

design bases and beyond design bases accidents. Due to that for the nuclear steam supply 

systems and containment simulation we use ANTHEM for simulation of all the accidents that are 

covered by the emergency operating procedures and MAAP4 code if we want to simulate 

severe accidents including core degradation, core relocation, reactor vessel failure, molten core 

concrete interaction and containment failure. 

5.1 Base analyses

For the first set of the analyses the assumed operator actions and sequence of main events is 

seen from Table 1 and Table 2. 

All RELAP5/MOD3.3 runs include 1000 s of initial steady-state, while ANTHEM and MAAP4 

curves start from the occurrence of Loss of all AC power. 
5-2 

Table 1: Loss of all AC power - assumptions common for all analyses 

Assumed action - Event Time (s) 

(All analyses) 

Number of PRZ PORVs --> 0 1 2 

Loss of all AC/Reactor trip/loss of all SG feed 0.0 0.0 0.0 

Assumed letdown isolation and start of RCP seal 

degradation 300 300 300 

Assumed complete RCP seal degradation 2100 2100 2100 

Assumed start of the first SG depressurization1 N/A 300 300 

Assumed first PRZ PORV opening2 N/A Note 2 N/A 

Assumed second PRZ PORV opening2 N/A Note 2 Note 2 

Table 2: Comparison of available time - time until the core heatup - degradations starts 

Assumed action 

- Event 

Time (s) 

RELAP5/MOD3.3 E-SIM-ANTEM E-SIM-MAAP4 

PRZ PORVs -> 0 1 2 0 1 2 0 1 2 

Time before 

core heatup3 9600 7040 12100 8500 6110 12270

RCS/Reactor 

vessel failure 10400 9190 15200

On Figure 2 primary pressure behavior for all the cases analyzed by RELAP5/MOD3.3 is 

shown. On Figure 3 the same parameter is shown calculated by plant engineering simulator - 

ANTHEM and on Figure 4 with MAAP4 (with the maximum seal break -300 gpm). MAAP4 code 

is built in the plant engineering simulator for the simulation of the severe accidents. 

As it can be seen from the figures all the three codes are providing very similar prediction. The 

differences seen are mainly due to the fact that in RELAP5 analyses for all the breaks we 

defined the fix boundary conditions, while on simulator the break flow depends on the variable 

conditions in the system to which the break flow is directed. One can also observe that the plant 

engineering simulator is predicting higher rate of primary system pressurization after plant 

cooldown by the secondary side is terminated (no secondary heat sink). The reason for that is 

slightly higher decay power calculated by the core model and as already mentioned slightly 

different boundary conditions. It has to be pointed out that for the MAAP4 analyses we doubled 

the size of the assumed break on the RCP seals, since this assumption is conservative. 

Based on Table 2 and Figure 2, Figure 3 and Figure 4 it can be concluded that in the case of 

loss of all AC and secondary heat sink the operators should try to depressurize the primary 

system as soon the steam generators are empty. 

 

1 Second SG depressurization was assumed to be started by the operator when in the first SG wide range level drops 

below 8% 

2 Pressurizer PORV opening depends on second SG wide range Level. It is assumed that operators would open one 

or both pressurizer PORVs after the second SG water on the secondary side is almost depleted (level < 4%) 

3 Time until the fuel rod temperature is below 850 o

C (RELAP) and core exit temperature below 750 oC 
5-3 

For better insight into the analyzed scenarios, additional parameters are shown in Figure 5 to 

Figure 13. 

From total primary leak flow (seal leak + PRZ valves) curve, shown in Figure 5, one can 

recognize period of increasing RCP leak flow (300 s – 2100s). This is followed by a certain 

period of quasi steady leak flow. At the end of this period secondary heat sink was lost due to 

SG valves opening (Figure 6) and consequent depletion of the secondary inventory (Figure 7). 

After the secondary heat sink was completely lost, primary pressure increased to the point, 

when PRZ safety valves started to open. In the case labeled “0 PRZ PORVs”, only PRZ safety 

valves were opening/closing on their set/reset pressure setpoints, while in the other two cases, 

labeled “1 PRZ PORV” and “2 PRZ PORVs”, PRZ PORV(s) were opened by the operator to 

mitigate the consequences of the station blackout sequence of events. After the valves opening 

primary pressure was successfully decreased for a certain period of time, so RCP leak flow was 

decreased. Some spikes can be observed in the cases “1 PRZ PORV” and “2 PRZ PORVs”. 

These origin from accumulator injection (Figure 8). 

PRZ gradually emptied (Figure 9) after the PRZ valves opening and was not restored due to 

continuous primary coolant leakage. Primary inventory was also gradually lost (Figure 10), 

which soon caused core uncovery (Figure 11) and unavoidably led to core heatup (Figure 12 

and Figure 13). 

By opening this additional break in the primary system intentionally and in fact starting the bleed 

procedure they would be able to get feed from the accumulators which will have two long term 

effects. Time between beginning of the accident and core heatup will increase giving more time 

to the operators for the AC power restoration. This would only be true in the case when both 

pressurizer PORVs could be opened. If this is not the case then it is better not to perform these 

actions, since we only increase the coolant depletion rate but we are not able to get passive 

injection from the accumulators. 

The second important effect, which is seen from the Figure 2 and Figure 3 as well as from the 

Table 1, is the primary system depressurization, which can be achieved by two pressurizer 

PORVs. This is one of the important objectives when we entered the area of severe accidents. 
PRZ pressure (p 069010000)

0

20

40

60

80

100

120

140

160

180

200

0 5000 10000 15000

Time [s]

Pressure [kp/cm2]

0 PRZ&SG PORVs

1 PRZ PORV

2 PRZ PORVs

Figure 2: Primary system pressure - RELAP5/MOD3.3 

PRZ pressure

0

20

40

60

80

100

120

140

160

180

200

0 5000 10000 15000

Time [s]

Pressure [kp/cm2]

0 PRZ&SG PORVs

1 PRZ PORV

2 PRZ PORVs

Figure 3: Primary system pressure - SIM- ANTHEM 

5-4 
PRZ pressure

0

20

40

60

80

100

120

140

160

180

200

0 5000 10000 15000

Time [s]

Pressure [kp/cm2]

0 PRZ&SG PORVs

1 PRZ PORV

2 PRZ PORVs

Figure 4: Primary system pressure - MAAP4 

Total primary leak (RCP seal + PRZ valves) flow 

(cntrlvar 761)

0

20

40

60

80

100

120

0 5000 10000 15000

Time [s]

Mass flow [kg/s]

0 PRZ&SG PORVs

1 PRZ PORV

2 PRZ PORVs

Figure 5: Total primary leak - RELAP5/MOD3.3 

5-5 
SG 1 valves integral flow (cntrlvar 847)

0

10000

20000

30000

40000

50000

60000

0 5000 10000 15000

Time [s]

Mass [kg]

0 PRZ&SG PORVs

1 PRZ PORV

2 PRZ PORVs

Figure 6: SG 1 valves flow - RELAP5/MOD3.3 

SG 1 level (cntrlvar 40)

0

10

20

30

40

50

60

70

80

90

100

0 5000 10000 15000

Time [s]

Level [%]

0 PRZ&SG PORVs

1 PRZ PORV

2 PRZ PORVs

Figure 7: SG 1 WR level - RELAP5/MOD3.3 

5-6 
Accum. no.1 level (cntrlvar 623)

0

10

20

30

40

50

60

70

80

90

100

0 5000 10000 15000

Time [s]

Level [%]

0 PRZ&SG PORVs

1 PRZ PORV

2 PRZ PORVs

Figure 8: Accumulator no.1 level - RELAP5/MOD3.3 

PRZ level (cntrlvar 4)

-20

0

20

40

60

80

100

120

140

0 5000 10000 15000

Time [s]

Level [%]

0 PRZ&SG PORVs

1 PRZ PORV

2 PRZ PORVs

Figure 9: PRZ level - RELAP5/MOD3.3 

5-7 
Primary mass (cntrlvar 191)

0

20000

40000

60000

80000

100000

120000

140000

0 5000 10000 15000

Time [s]

Mass [kg]

0 PRZ&SG PORVs

1 PRZ PORV

2 PRZ PORVs

Figure 10: Primary mass - RELAP5/MOD3.3 

Core collapsed level (cntrlvar 5)

0

20

40

60

80

100

120

0 5000 10000 15000

Time [s]

Level [%]

0 PRZ&SG PORVs

1 PRZ PORV

2 PRZ PORVs

Figure 11: Core level - RELAP5/MOD3.3 

5-8 
Rod cladding temperature (httemp 111900816)

0

500

1000

1500

2000

0 5000 10000 15000

Time [s]

Temperature [oC]

0 PRZ&SG PORVs

1 PRZ PORV

2 PRZ PORVs

Figure 12: Rod cladding temperature - RELAP5/MOD3.3 

Core exit temperature (cntrlvar 609)

0

200

400

600

800

1000

1200

0 5000 10000 15000

Time [s]

Temperature [oC]

0 PRZ&SG PORVs

1 PRZ PORV

2 PRZ PORVs

Figure 13: Core exit temperature - RELAP5/MOD3.3 

5-9 
5-10 

5.2 Variation analyses

Second set of analyses performed only with the plant engineering simulator with MAAP4 code 

was done with the objective to verify if the primary system can be depressurized if we do not 

assume operator actions for SG depressurization. Here the objective was to check if it is 

feasible to expect that the operators will be able to depressurize primary system below the 

pressure for high-pressure debris ejection. For NPP Krško that would mean that at the time of 

the primary system failure internal pressure would lower than 22 kp/cm2

. For this analyses it 

was assumed that we would not change any of the steps in the emergency operating 

procedures, but rather in the SAMG procedures. It was assumed in the analyses that the 

operators would deliberately open both PRZ PORVs as soon as the core would be overheated - 

core exit temperature would be higher than 750 o

C. During such an accident, assuming station 

blackout, loss of secondary heat sink, no injection and seal failure, core damage can not be 

avoided any more. Then the primary objective is to decrease the primary system pressure. 

Assumed operator actions and sequence of main events listed in Table 3 

As it can be seen from the Figure 14, if the operators would be able to open both pressurizer 

PORVs, primary pressure would drop below 22 kp/cm2

. It is then expected that primary system 

would fail from low internal pressure. Even if they would open only one PORV this would not 

prevent the vessel failure however will fail from significantly lower pressure. 

The same sorts of analyses have been performed by RELAP5/MOD3.3, to compare with 

MAAP4 results, of course only up to the point of significant core overheat. The only difference 

from the assumptions given in Table 3 was, that the second PRZ PORV opening (one or both 

PORVs) was performed when rod cladding temperature in core node no.8 exceeded 850 o

C 

(instead core exit temperature > 750 o

C). 

Pressure development for the base case and the two variation cases is shown in Figure 15. It 

can be observed that even quantitatively both RELAP5/MOD3.3 and MAAP4 predicted the 

course of the transient very similarly. 

Table 3: Loss of all AC power - assumptions for MAAP analyses 

Assumed action – Event Time (s) 

(All analyses) 

Number of PRZ PORVs 2 

Loss of all AC/Reactor trip/loss of all SG feed 0.0 

Assumed letdown isolation and start of RCP 

seal degradation 300 

Assumed complete RCP seal degradation 2100 

Assumed start of first SG depressurization N/A 

Assumed first PRZ PORV opening N/A 

Assumed second PRZ PORV opening Core exit temperature > 750 DEGC 
PRZ pressure

0

20

40

60

80

100

120

140

160

180

200

0 5000 10000 15000 20000

Time [s]

Pressure [kp/cm2]

0 PRZ&SG

PORVs

2 PRZ PORVs at

CET 750 DEGC

1 PORV at CET

750 DEGC

Figure 14: Primary system pressure without and with depressurization using PRZ relief valves – 

MAAP4 

PRZ pressure (p 069010000)

0

20

40

60

80

100

120

140

160

180

200

0 5000 10000 15000 20000

Time [s]

Pressure [kp/cm2]

0 PRZ&SG PORVs

2 PRZ PORVs at clad T 850oC

1 PRZ PORV at clad T 850oC

Figure 15: Primary system pressure without and with depressurization using PRZ relief 

valves – RELAP5/MOD3.3 

5-11 

6. RELAP5/MOD3.3 RUN STATISTICS 

MOD3.3 Calculations were performed on SUN FIRE V880 server with 4 UltraSPARC III 750 

MHz processors, with 16 GB main RAM, running under SOLARIS 9 operating System. 

Table 4 shows run-time statistics for all the analyzed cases. 

Table 4: Run time statistics 

Analyzed case Computer CPU

time [s]

Total number of

time steps (NT)

Total number of

volumes (N)

Grind time

CPU/(NT*N)

all cases - 1000 s of steady state 1669.90 26172 469 1.36E-04

0 PRZ&SG PORVs 7293.84 98268 469 1.58E-04

1 PRZ PORV 45931.01 688048 469 1.42E-04

2 PRZ PORVs 55307.01 850588 469 1.39E-04

1 PRZ PORV at clad T 850o

C 11408.40 169260 469 1.44E-04

2 PRZ PORVs at clad T 850o

C 42654.50 659750 469 1.38E-04

Consumed CPU time for all base and variation analysis cases is shown in Figure 16. 

Mass error, time step and Courant Δt for the 3 base analyses cases and the 3 variation 

analyses cases are shown in Figure 17 to Figure 19 and Figure 20 to Figure 22, respectively. 

Consumed CPU time

0

10000

20000

30000

40000

50000

60000

0 5000 10000 15000 20000

Time [s]

CPU time [s]

0 PRZ&SG PORVs

1 PRZ PORV

2 PRZ PORVs

1 PRZ PORV at clad T 850oC

2 PRZ PORVs at clad T 850oC

Figure 16: Consumed CPU time for all base and variation analyses cases– 

RELAP5/MOD3.3 

6-1 
6-2 

Next is the excerpt from base case simulation with 0 PRZ&SG PORVs (time step statistics): 

0 advancement total between edits|  attempted: 26302 26302 | min.dt= 0. sec last dt= 3.823475E-02 sec emass=  repeated: 0 0 | max.dt= 0. sec crnt.dt= 3.823475E-02 sec tmass=  successful: 26302 26302 | avg.dt= 0. sec merr.est= 2.385202E-10 em/tm=  requested: 26304 26304 | req.dt= 0.100000 sec cpu= 2.15907 sec time=  attempted: 28550 2248 | min.dt= 3.293432E-02 sec last dt= 5.839124E-02 sec emass=  repeated: 0 0 | max.dt= 0.100000 sec crnt.dt= 0.220598 sec tmass=  successful: 28550 2248 | avg.dt= 8.897246E-02 sec merr.est= 3.884262E-07 em/tm=  requested: 28552 2248 | req.dt= 0.100000 sec cpu= 179.494 sec time=  attempted: 36886 8336 | min.dt= 6.002535E-02 sec last dt= 6.672375E-02 sec emass=  repeated: 0 0 | max.dt= 0.100000 sec crnt.dt= 0.414794 sec tmass=  successful: 36886 8336 | avg.dt= 9.596929E-02 sec merr.est= 1.421074E-07 em/tm=  requested: 36888 8336 | req.dt= 0.100000 sec cpu= 844.417 sec time=  attempted: 45161 8275 | min.dt= 5.408423E-02 sec last dt= 5.408423E-02 sec emass=  repeated: 0 0 | max.dt= 0.100000 sec crnt.dt= 0.333887 sec tmass=  successful: 45161 8275 | avg.dt= 8.459215E-02 sec merr.est= 9.242832E-08 em/tm=  requested: 45163 8275 | req.dt= 0.100000 sec cpu= 1499.03 sec time=  attempted: 46394 1233 | min.dt= 5.949265E-02 sec last dt= 0.216676 sec emass=  repeated: 0 0 | max.dt= 0.250000 sec crnt.dt= 0.288100 sec tmass=  successful: 46394 1233 | avg.dt= 0.243309 sec merr.est= 1.825119E-06 em/tm=  requested: 46396 1233 | req.dt= 0.250000 sec cpu= 1597.83 sec time=  attempted: 65445 19051 | min.dt= 1.984276E-03 sec last dt= 0.132841 sec emass=  repeated: 730 730 | max.dt= 0.250000 sec crnt.dt= 0.328195 sec tmass=  successful: 64715 18321 | avg.dt= 0.163747 sec merr.est= 2.378124E-07 em/tm=  requested: 64717 18321 | req.dt= 0.250000 sec cpu= 2956.55 sec time=  attempted: 69494 4049 | min.dt= 0.125000 sec last dt= 0.191908 sec emass=  repeated: 732 2 | max.dt= 0.250000 sec crnt.dt= 0.487488 sec tmass=  successful: 68762 4047 | avg.dt= 0.247097 sec merr.est= 2.395330E-06 em/tm=  requested: 68764 4047 | req.dt= 0.250000 sec cpu= 3260.92 sec time=  attempted: 113784 44290 | min.dt= 5.171839E-03 sec last dt= 6.922516E-02 sec emass=  repeated: 739 7 | max.dt= 0.100000 sec crnt.dt= 0.224675 sec tmass=  successful: 113045 44283 | avg.dt= 4.064765E-02 sec merr.est= 3.580817E-07 em/tm=  requested: 113047 44283 | req.dt= 0.100000 sec cpu= 6474.08 sec time=  attempted: 124566 10782 | min.dt= 7.079435E-03 sec last dt= 0.200000 sec emass=  repeated: 740 1 | max.dt= 0.200000 sec crnt.dt= 0.240902 sec tmass=  successful: 123826 10781 | avg.dt= 0.179975 sec merr.est= 2.737707E-05 em/tm=  requested: 123827 10780 | req.dt= 0.200000 sec cpu= 7295.71 sec time= 
Mass error (emass 0)

0

50

100

150

200

250

0 5000 10000 15000

Time [s]

Mass [kg]

0 PRZ&SG PORVs

1 PRZ PORV

2 PRZ PORVs

Figure 17: Mass error for 3 base analyses cases– RELAP5/MOD3.3 

Time step (dt 0)

0

0.05

0.1

0.15

0.2

0.25

0.3

0 5000 10000 15000

Time [s]

Time [s]

0 PRZ&SG PORVs

1 PRZ PORV

2 PRZ PORVs

Figure 18: Time step for 3 base analyses cases– RELAP5/MOD3.3 

6-3 
Courant time step (dtcrnt 0)

0

0.1

0.2

0.3

0.4

0.5

0.6

0.7

0.8

0.9

0 5000 10000 15000

Time [s]

Time [s]

0 PRZ&SG PORVs

1 PRZ PORV

2 PRZ PORVs

Figure 19: Courant Δt for 3 base analyses cases– RELAP5/MOD3.3 

Mass error (emass 0)

0

50

100

150

200

250

0 5000 10000 15000 20000

Time [s]

Mass [kg]

0 PRZ&SG PORVs

2 PRZ PORVs at clad T 850oC

1 PRZ PORV at clad T 850oC

Figure 20: Mass error for 3 variation analyses cases– RELAP5/MOD3.3 

6-4 
Time step (dt 0)

0

0.05

0.1

0.15

0.2

0.25

0.3

0 5000 10000 15000 20000

Time [s]

Time [s]

0 PRZ&SG PORVs

2 PRZ PORVs at clad T 850oC

1 PRZ PORV at clad T 850oC

Figure 21: Time step for 3 variation analyses cases– RELAP5/MOD3.3 

Courant time step (dtcrnt 0)

0

0.1

0.2

0.3

0.4

0.5

0.6

0.7

0.8

0.9

0 5000 10000 15000 20000

Time [s]

Time [s]

0 PRZ&SG PORVs

2 PRZ PORVs at clad T 850oC

1 PRZ PORV at clad T 850oC

Figure 22: Courant Δt for 3 variation analyses cases– RELAP5/MOD3.3 

6-5 

7-1 

7. CONCLUSIONS

Loss of all AC power is one of the most challenging plant transients. If the power is not restored 

within certain time, and if the capability to feed the steam generators is also lost this accident 

will progress to a severe accident. The aim of the presented analysis was to verify if the 

emergency operating or SAMG procedures should be changed and if the design change on 

pressurizer PORV should be implemented to be able to cope with this kind of accidents better. 

The analyses were performed with three different state of the art codes used at NPP Krško and 

IJS: RELAP5/MOD3.3, ANTHEM and MAAP4. It can be concluded that no change to the 

existing emergency operating procedures can be recommended. 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. 

Comparing RELAP5/MOD3.3, ANTHEM and MAAP4, it can be concluded that all the three 

codes predicted very similar transient course in all the analyzed scenarios. 

It has to be pointed out that before making any specific conclusions related to design change 

and SAMG procedure change, further analyses including cost benefit analyses shall be 

performed, since we are dealing with very low probability events. 

8-1 

8. REFERENCES

1. NUREG-0933, "A Prioritization of Generic Safety Issues", Rev.3, U.S. Nuclear Regulatory 

Commission, July 1991 

2. WENX 91/04 Station Blackout analyses for NPP Krško, Westinghouse, Brussels, March 

1991 

3. NPP Krško Emergency Operating Procedures, Rev.10, February 2001 

4. US NRC, NUREG 1150, Severe Accident Risk, An assessment for Five U.S. Nuclear Power 

Plants, Final Report, January 1991 

5. EPRI - FAI 91/19 "Severe Accident Management Guidelines Technical bases Report", 

September 1991 

6. Probabilistic Safety Assessment of Nuclear Power Plant Krško, SUBMITAL REPORT, 

Volume 1 and 2, August 1995 

7. NPP Krško Severe Accident Management Guidelines - SAMG, Rev.2, April 2002 

8. Information Systems Laboratories, Inc., Nuclear Safety Analysis Division: RELAP5/MOD3.3 

Code Manual, Vol.#1: Code Structure, System Models, and Solution Methods, Vol.#2: 

Users’ Guide and Input Requirements, Vol.#3: Developmental Assessment Problems, 

Vol.#4: Models and Correlations, Vol.#5: Users’ Guidelines, Vol.#7: Summaries and 

Reviews of Independent Code Assessment Reports, Vol.#8: Programmers Manual 

NUREG/CR5535 Rev.1, Rockville, Maryland, Idaho Falls, Idaho, USA, December 2001, 

Vol.#6: A. S. Shieh, V. H. Ransom, R. Krishnamurthy: RELAP5/MOD3 Code Manual - 

Validation of Numerical Techniques in RELAP5/MOD3.0, October 1994 [Issued with minor 

revisions December 2001] 

9. Krajnc, B. and Parzer, I.: Analyzing Operator Actions To Gain Time In Loss of AC Power 

with Subsequent Loss of Secondary Heat Sink Accident, 4th International Conference on 

Nuclear Option in Countries with Small and Medium Electricity Grids Dubrovnik, Croatia, 

June 16-20, 2002. 

10. Boire, R and Salim, J. "ANTHEM: Advanced Thermal Hydraulic Model for Power Plant 

Simulation", CSNI Specialist Meeting on Simulators and Plant Analyzers, Technical 

Research Center of Finland, Espoo 1994 

11. Salim, G., Vivier, P., Filiatrault, P. and Boire, R., "ANTHEMTM NSSS Model Validation", 

Proceedings of the 2000 Western Multconference, Society of Computer Simulation, San 

Diego, California, January 2000