NUREG/IA-0467 

International 

 Agreement Report

RELAP5 Analysis of Mitigation Strategy for 

Extended Blackout Power Condition in 

PWR

 

 

Prepared by: 

Andrej Prošek, Andrija Volkanovski

Jožef Stefan Institute 

Jamova cesta 39 

Sl-1000 Ljubljana, Slovenia

K. Tien, NRC Project Manager 

 

Division of Systems Analysis

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

Manuscript Completed: November 2015

Date Published: May 2016

Prepared as part of 

The Agreement on Research Participation and Technical Exchange

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

Published by

U.S. Nuclear Regulatory Commission

NUREG/IA-0467 

International 

 Agreement Report

RELAP5 Analysis of Mitigation Strategy for 

Extended Blackout Power Condition in 

PWR

 

 

Prepared by: 

Andrej Prošek, Andrija Volkanovski

Jožef Stefan Institute 

Jamova cesta 39 

Sl-1000 Ljubljana, Slovenia

K. Tien, NRC Project Manager 

 

Division of Systems Analysis

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

Manuscript Completed: November 2015

Date Published: May 2016

Prepared as part of 

The Agreement on Research Participation and Technical Exchange

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

Published by

U.S. Nuclear Regulatory Commission

iii

0BABSTRACT

A set of measures have been proposed and currently implemented in response of the accident 

at the Fukushima Daiichi nuclear power plant. Those measures include diverse and flexible 

mitigation strategies that increase the defense-in-depth for beyond-design-basis scenarios. 

Mitigation strategies are based on the utilization of the portable equipment to provide power and 

water to the nuclear power plants in order to maintain or restore key safety functions. The 

verification of the proposed measures with the plant specific safety analyses is endorsed in the 

mitigation strategies. The purpose of the study was to investigate the application of the 

deterministic safety analysis for mitigation strategy of the extended station blackout (SBO). A 

methodology for the assessment of flowrates for steam generator makeup using portable pump 

is proposed. The aim is to fill steam generator without available information on level in such a 

way that makeup is sufficient and that at the same time the steam generators are not overfilled. 

The RELAP5/MOD3.3 computer code and input model of a two-loop pressurized water reactor 

is used for analyses, assuming different injection start times, flowrates and reactor coolant 

system losses. The calculated results show effectiveness of the proposed extended SBO 

mitigation strategy. The applicability of the developed method on operational power plant is 

validated.

v

CONTENTS --

Page

0BABSTRACT...............................................................................................................................iii

CONTENTS............................................................................................................................... v

1BFIGURES..................................................................................................................................vii

2BTABLES ....................................................................................................................................ix

3BEXECUTIVE SUMMARY ...........................................................................................................xi

4BACKNOWLEDGMENTS..........................................................................................................xiii

5BABBREVIATIONS ....................................................................................................................xv

1. 6BINTRODUCTION ................................................................................................................ 1

2. 7BMETHOD FOR ASSESSMENT OF STEAM GENERATOR MAKEUP FLOW.................... 3

3. 8BRELAP5 INPUT MODEL AND SCENARIOS DESCRIPTION ............................................ 5

3.1 12BRELAP5 Input Model................................................................................................... 5

3.2 13BMain Assumptions for Extended SBO.......................................................................... 6

3.3 14BScenarios Description.................................................................................................. 7

4. 9BRESULTS........................................................................................................................... 9

4.1 15BBase Calculations........................................................................................................ 9

4.2 16BAssessment of Necessary Flowrates..........................................................................12

4.3 17BVerification Calculations .............................................................................................13

5. 10BCONCLUSIONS................................................................................................................21

6. 1BREFERENCES..................................................................................................................23

vii

1BFIGURES

Page

Figure 1 Method for Necessary Makeup Flow Into SGs During Extended SBO ..................... 3

Figure 2 RELAP5 Krško NPP Hydraulic Components View ................................................... 5

Figure 3 Influence of SG Injection Delay on Average Rod Cladding Temperature for 

SL_LOSS Scenario Type: (a) Case C1, (b) Case C2 ............................................... 9

Figure 4 Influence of SG Injection Delay on RCS Pressure for SL_LOSS Scenario Type:

(a) Case C1, (b) Case C2 .......................................................................................10

Figure 5 Comparison of Scenario Types for Average Rod Cladding Temperature: (a) Case 

C1 With 4 H Delay, (b) Case C2 With 2 H Delay.....................................................11

Figure 6 Comparison of Scenario Types for Average Rod Cladding Temperature: (a) Case 

C1 with 5 h Delay, (b) Case C2 with 3 h Delay........................................................12

Figure 7 Main Parameters for SLD_LOSS Verification Calculations: (a) RCS Pressure, 

(b) Average Rod Cladding Temperature, (c) SG No. 1 WR Level,

(d) SG No. 2 WR Level ...........................................................................................15

Figure 8 Main Parameters for NSLD_LOSS Verification Calculations: (a) RCS Pressure, 

(b) Average Rod Cladding Temperature, (c) SG No. 1 WR Level,

(d) SG No. 2 WR Level. ..........................................................................................16

Figure 9 Main Parameters for S_LOSS Verification Calculations: (a) RCS Pressure, 

(b) Core Collapsed Liquid Level, (c) SG No.1 WR Level, 

(d) SG No.2 WR Level. ...........................................................................................17

Figure 10 Main Parameters for N_LOSS Verification Calculations: (a) RCS Pressure, 

(b) Core Collapsed Liquid Level, (c) SG No. 1 WR Level,

(d) SG No. 2 WR Level. ..........................................................................................18

Figure 11 Main Parameters for NO_LOSS Verification Calculations: (a) RCS Pressure,

(b) Core Collapsed Liquid Level, (c) SG No. 1 WR Level,

(d) SG No. 2 WR Level. ..........................................................................................19

ix

2BTABLES

Page

Table 1 Types of RCS Coolant Loss Scenarios .................................................................... 7

Table 2 Minimum Necessary Constant Flows within the 24 h After SBO Start .....................13

xi

3BEXECUTIVE SUMMARY

The events at the Fukushima Daiichi nuclear power plant and stress tests showed that the loss 

of electrical power (LOOP) followed by station blackout event (SBO) and loss of the ultimate 

heat sink (UHS) can have large impact on the safety of the nuclear power plant (NPP). The 

SBO event when power from all emergency power sources, including batteries, is lost is named 

extended SBO and leads ultimately to core heatup and core damage.

The strategies proposed for coping with these events include utilization of portable equipment, 

permanent equipment or combinations of portable and permanent equipment. The Diverse and 

Flexible Coping Strategies (FLEX) are focused on maintaining or restoring key plant safety 

functions. The FLEX strategies propose development and application of thermal hydraulic 

analyses to support plant specific decision-making.

In this study utilization of the pump, either turbine driven auxiliary feedwater or portable pump, 

for mitigation of the extended SBO event and prevention of core damage in pressurized water 

reactor (PWR) is investigated. It presents a follow-up study to European Union stress tests. 

Methodology is developed for assessment of the required pump flowrates within analyzed time 

period. The assessed flow rates should prevent core damage without overfilling the steam 

generators in the analyzed scenarios.

The RELAP/MOD3.3 Patch 04 input model of NPP with two-loop PWR is used in this study. The 

major difference between boiling water reactor and PWR is PWR has coolant under pressure in 

its primary cooling/heat transfer circuit, and generates steam in a secondary circuit while boiling 

water reactor (BWR) makes steam in the primary circuit above the reactor core. This PWR 

feature gives possibility to cool the reactor through the secondary side also during extended 

SBO providing that makeup water for secondary side is established.

Different scenario types are investigated including plant design improvements which may be 

done to better cope with extended SBO. Three different reactor coolant system (RCS) coolant 

loss pathways, with corresponding leakage rate, can be expected in the PWR plant during the 

extended SBO: normal system leakage, reactor coolant pump seal leakage and loss of RCS 

coolant through letdown relief valve unless automatically isolated or until isolation is 

procedurally directed.

The required injection rates to the steam generators in the first 24 h and from 24 h to 72 h are

calculated from the cumulative water mass injected by the turbine driven auxiliary feedwater 

pump in the analyzed scenarios, when desired normal level is maintained automatically. It 

means that flow has to be adjusted just two times in 72 h. After determining the required 

flowrates the verification analyses using portable pump with required flows are performed.

The effective prevention strategy of core heatup during the extended blackout condition can be 

developed with the utilization of the presented method. The method can be extended from the 

analyzed 72 h and can be modified for different type of pumps and their characteristics.

xiii

4BACKNOWLEDGMENTS

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

Nuclear Safety Administration within CAMP program (project no. POG-U3-KE-R4/104/12 – also 

NEK no.: 3120118) and from the state budget by the Slovenian Research Agency (program 

no. P2-0026).

xv

5BABBREVIATIONS

AFW auxiliary feedwater

ASCII American Standard Code for Information Interchange

BWR boiling water reactor

CVCS chemical and volume control system

ECCS emergency core cooling system

EDG emergency diesel generator

FLEX Diverse and Flexible Coping Strategies

HPSI high pressure safety injection

LOOP loss of electrical power

LPSI low pressure safety injection

MFW main feedwater

MSIV main steam isolation valve

NPP nuclear power plant

PORV power operated relief valve

PRZ pressurizer

PWR Pressurized Water Reactor

RCP reactor coolant pump

RCS reactor coolant system

RELAP Reactor Excursion and Leak Analysis Program

RPV reactor pressure vessel

SBO station blackout

SG steam generator

SL surge line

SNAP Symbolic Nuclear Analysis Package

TD-AFW turbine driven auxiliary feedwater

UHS ultimate heat sink

WR wide range

1

1. 6BINTRODUCTION

The events at the Fukushima Daiichi nuclear power plant (Ref. 1) and stress tests (Refs. 2 and 

3) showed that the loss of electrical power (LOOP) followed by station blackout event (SBO) 

and loss of the ultimate heat sink (UHS) can have large impact on the safety of the nuclear 

power plant (NPP). The SBO event when power from all emergency power sources, including 

batteries, is lost is named extended SBO and leads ultimately to core heatup and core damage 

(Ref. 4).

The strategies proposed for coping with these events include utilization of portable equipment, 

permanent equipment or combinations of portable and permanent equipment. The Diverse and 

Flexible Coping Strategies (FLEX) (Ref. 5) are focused on maintaining or restoring key plant 

safety functions. The FLEX strategies propose development and application of thermal hydraulic 

analyses to support plant specific decision-making (Ref. 5).

For the LOOP, SBO and the loss of the UHS scenarios, maintaining water injection to reactors 

and/or steam generators (SGs) offers an ultimate means of cooling the core. Different systems 

for performing this function have been identified in the stress test reports, including electric 

power-independent turbine driven pumps, arrangements for gravity feed, dedicated diesel 

driven pumps and pre-installed connections for external feed, such as from the on-site fire 

trucks (Ref. 3). The availability of the pre-arranged connections and drills for actual 

establishment of the connection for feeding of steam generators is identified as prerequisite for 

the utilization of the fire trucks.

In this study the utilization of the pump, either turbine driven auxiliary feedwater or portable 

pump, for mitigation of the extended SBO event and prevention of core damage in pressurized 

water reactor (PWR) is investigated. It presents a follow-up study to stress tests. Methodology is 

developed for assessment of the necessary pump flowrates within analyzed time period. The 

assessed flow rates should prevent core damage without overfilling the steam generators in the 

analyzed scenarios, therefore verification analyses are performed. The description of the 

deterministic safety analysis input model, the methodology used for the assessment of the 

necessary injection flow of the pump to SGs and developed case scenarios is given in Section 

2. Obtained results from the deterministic safety analyses are given in Section 3 both for base

calculations to assess the necessary flow and the verification calculations using the assessed 

pump constant flows.

3

2. 7BMETHOD FOR ASSESSMENT OF STEAM GENERATOR MAKEUP 

FLOW

The operation of pump, utilized for SG injection during extended SBO, is assumed to be divided 

into several intervals. The flow rate of the pump is constant in each interval. The pump can operate 

in continuous or discrete mode. The methodology is applicable for both cases. The operation with 

constant flow and long time intervals minimizes the required number of flow changes and 

operators errors during required manipulations. The method for assessing the necessary makeup 

flow into SG during extended SBO is shown in Figure 1.

Figure 1 Method for Necessary Makeup Flow Into SGs During Extended SBO

It consists of three steps. In the Step 1 the base case calculation is performed for selected SBO 

scenario, but it is assumed that SG levels are maintained at normal level using artificial level 

control system. In this way, the required water mass, which has to be injected in the steam 

generator to restore and maintain the desired water level, is obtained. The analysis and 

assessment of the required water mass is done for time a window equal to the operational time 

of the pump during extended SBO. The scenarios that in the analysed time window result in the 

core heatup and damage are not analyzed further. If core heatup is prevented in Step 1 then the 

mass flow rate of the portable pump has to be set so, that the integral water mass injected in the 

steam generator at a given time ideally corresponds to the mass, determined in Step 1, but in any 

case remains in the operable range of the steam generator (between 8% and 96% for selected 

PWR). In the simplest case the mass flow rate of the portable pump is not adjusted regularly, but 

is set constant for a longer time period. Care should be taken that the mass flow rate is high 

enough that in the initial period, when the residual power is higher, the steam generator is not 

emptied. It may turn out that for longer time periods this may not be achieved with a constant 

mass flow rate and that more adjustments of the mass flow rate are required. Finally, in Step 3 

the RELAP5 verification calculations are performed with the pump flows set to the necessary flow 

rates assessed in the Step 2. These calculations verify if calculated water injection in the SGs, for 

given scenario, prevents core heat up and SGs overfill.

Step 3: Perform verification calculation

The verification calculations are performed with the makeup flows 

set to the makeup flowrates assessed in the Step 2. 

Step 2: Assess the necessary makeup flowrate

The portable pump flow is set so, that the integral water mass 

injected in the steam generator at a given time ideally 

corresponds to the mass, determined in Step 1. 

Step 1: Perform base case calculation

The water inventory needed for feeding steam generators is 

determined as integral of makeup flow to steam generator when 

level is maintained at normal level.

Continue, if core heatup

is prevented in Step 1 

5

3. 8BRELAP5 INPUT MODEL AND SCENARIOS DESCRIPTION

3.1 12BRELAP5 Input Model

To perform the analyses, the base RELAP5 input model of Krško NPP has been used. Krško 

NPP is a two loop PWR, Westinghouse type, with reactor power uprated to 1994 MW. The input 

model has been validated by plant transients (e.g. Ref. 6). It has been used for several safety 

analyses including reference calculations for Krško full scope simulator verification (Refs. 7 and 

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

2107 radial mesh points. When imported ASCI file into SNAP, the following hydraulic 

components view has been generated semi automatically (hydraulic components with 

connections generated automatically, annotations and layout manually). In terms of SNAP this 

gives 304 hydraulic components and 108 heat structures. Hydraulic components in SNAP 

consist of both volumes and junctions, where pipe with more volumes is counted as one 

component. Each heat structure in SNAP connected to pipe is counted as one component in 

SNAP and not as many heat structures as pipe volumes like counted in RELAP5 output file. 

This explains the difference in numbers of heat structures in Figure 2 and that reported in 

RELAP5 output file.

Figure 2 RELAP5 Krško NPP Hydraulic Components View

Modeling of the primary side includes the reactor pressure vessel (RPV), both loops (LOOP 1 
6

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

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

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

coolant pump (RCP) seal flow. 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, main steam line, main 

steam isolation valves (MSIVs), SG relief and safety valves, and main feedwater (MFW) piping. 

The turbine valve is modeled by the corresponding logic. The turbine is represented by time 

dependent volume. The MFW and AFW (auxiliary feedwater) pumps are modeled as time 

dependent junctions.

3.2 13BMain Assumptions for Extended SBO

The main assumptions considered in the input model for base calculations of extended SBO are 

the following:

A1) The reactor automatically trips and all rods are inserted following the LOOP;

A2) The systems that are available for cooling of the reactor coolant system (RCS) during 

LOOP and SBO are assumed unavailable in this study. In depressurization scenarios 

the steam generators pressure indication is assumed to be available.

A3) The cooling on the primary side is through the pressurizer safety valves (SV) and/or 

leaks causing RCS inventory loss.

A4) The integrity of the RCS is maintained with the exception of assumed loss of coolant in 

the analysed scenario.

A5) Connection point to the outlet of the TD-AFW is assumed to be available for portable 

pump and included in the model;

A6) Availability of water for operation of the water pump is assumed in the model.

A7) Pressurizer and steam generator safety valves are assumed available.

A8) The nitrogen gas required for the operation of the steam generator power operated 

relief valves (SG PORVs) is assumed available in the depressurization scenarios. The 

alternative compressed air supply from the portable diesel compressor is providing 

required gas.

A9) The criterion used as indication for the steam generator overfill is 96% of wide range 

(WR) steam generator level, while for the loss of heat sink is 8%.

A10) The criterion used for core heatup start is significant core uncovery causing core 

heatup.

In addition to above assumptions the following two assumptions were used for verification 

calculations of extended SBO:

A11) Loss of all electric power in the plant, including batteries, during SBO condition is 

assumed. This plant condition is referred to in this paper as extended SBO. This results 

in loss of all instrumentation and control in the plant.

A12) The pump, TD-AFW or portable, is available for the whole analysed period for injection 

of water in the steam generators. The portable pump flow measurement and regulation 

is also assumed to be available in the study. The TD-AFW pump speed is manually 

controlled. Hand wheels are provided for local manual operation of TD-AFW control 

valves. The TD-AFW pump flowrate local indicators, not relying on electric power, are 

also available locally.

In base calculations normally operable TD-AFW is assumed, with all instrumentation and control. 

The regulation of the TD-AFW is set to restore and maintain narrow range level at 69% (plant 

normal level, at normal power condition this means 77% wide range level). Using such 

assumptions the integrated mass injected to steam generators is obtained, which satisfy A9
7

criterion than SGs are not overfilled. Namely, in SBO event the permanent steam generator level 

indication is not available and when developing mitigation strategies it is helpful to know how 

much of mass need to be injected to SGs. The SGs are overfilled if feed flow to the steam 

generators is larger than required. The SG overfill can result in different deteriorating effects (Ref. 

11). The potential effects of steam generator overfill are hydraulic forces, excessive dead weight 

loads, failure of valves to reseat, loss of emergency (auxiliary) feedwater pump turbine, steam 

generator tube rupture and acceleration of accumulated water.

3.3 14BScenarios Description

The method for assessment of necessary makeup flowrate to SGs is demonstrated on the input 

model of the two-loop PWR design described in Section 3.1. Different scenario types are 

investigated including plant design improvements which may be done to better cope with 

extended SBO. Three different RCS coolant loss pathways, with corresponding leakage rate, can 

be expected in the PWR plant during the extended SBO:

1. Normal system leakage. It is assumed that this kind of leakage is equal to plant technical 

specifications identified leakage (0.63 l/s at nominal RCS conditions).

2. RCPs seal leakage. During the extended SBO the cooling of the seals is not available 

resulting in their heat up and loss of integrity. The seal leak rate of 1.32 l/s per RCP (at 

nominal RCS conditions) due to loss of seal cooling and RCP pump stop is assumed as 

representative for the plants using a high temperature O-ring RCP seal package, as in 

reference NPP (Ref. 9). The seal leakage can be practically prevented (negligible loss of 

the order of 0.06 l/s) with the installation of special passive RCP thermal shutdown seals 

(Ref. 10)

3. Loss of RCS coolant through letdown relief valve unless automatically isolated or until 

isolation is procedurally directed. During extended SBO and nominal RCS pressure the 

letdown loss of 5.68 l/s is expected. Some NPPs have option of manual isolation of the 

letdown. The depressurization of the primary side through the secondary side results in 

decrease of the RCS pressure. Decreased RCS pressure stops letdown loss of coolant 

and allows RCS coolant makeup from accumulators. In case of heat sink loss, the 

decrease of the RCS pressure also delays opening of the pressurizer safety valves and 

loss of the RCS coolant over those valves.

The pressurizer safety valves open when RCS increases to their set points. The RCS pressure 

depends on the plant parameters and is specific for each analysis. Therefore the loss of coolant 

over pressurizer safety valves is not included in Table 1.

Six types of RCS coolant loss scenarios, given in Table 1, are developed and analysed: the 

NO_LOSS (no RCS loss), N_LOSS with normal system leakage, the S_LOSS with RCP seal 

loss starting one hour after the start of extended SBO, the SL_LOSS with RCP seal loss and 

loss of coolant through the letdown relief valve when RCS pressure is greater than 4.24 MPa, 

SLD_LOSS with RCP seal and letdown loss (if RCS pressure is greater than 4.24 MPa) and 

depressurization of the primary side through the secondary side to 1.57 MPa, started 30 

minutes after SBO occurrence, and NSLD (SLD case with additionally assumed system 

leakage).

Table 1 Types of RCS Coolant Loss Scenarios

Scenario type Normal system 

leakage

Seal loss Letdown loss Depressurization

NO_LOSS no no no no

N_LOSS yes no no no
8

S_LOSS no yes no no

SL_LOSS no yes yes no

SLD_LOSS no yes yes yes

NSLD_LOSS yes yes yes yes

Two cases are developed and simulated for the RCS loss scenarios given in Table 1. In the 

Case 1 (C1) it is assumed that the emergency diesel generators (EDG) started and normally 

operated for one hour after the loss of offsite power. Case 2 (C2) assumes that the extended 

SBO occurrence is concurrent with LOOP. For Case 1 availability of all safety systems is 

assumed one hour, before the extended SBO is assumed. This time is assumed because in 

Fukushima event the EDGs were running for 45 minutes until tsunami arrived. LOOP of offsite 

power causes reactor trip and after one hour the decay heat level already significantly drops.

For each type of scenarios different time delays between the extended SBO start and start of 

the pump injections to SG were considered and analysed in Step 2. For case C1 in all six type 

scenarios the start of pump injection was delayed 0, 0.5 h, 1 h, 2 h, 3 h, 4 h and 5 h (42 

scenarios in total). For case C2 the time delays 0, 0.5 h, 1 h, 2 h and 3 h were considered (30 

scenarios in total). The scenario name is composed from scenario type name and the delay 

time (e.g. SLD_LOSS type scenario with 4 h delay of steam generator feeding is labelled as 

SLD_LOSS_4). It should be noted that in addition to assumed delays the no delay scenario was 

analysed for base case to verify if for selected scenario type the core heatup could be prevented 

through cooling the RCS by secondary side (e.g. RCS makeup would be needed in case of 

large leaks).

The scenarios were simulated until 72 h after SBO occurrence or core heatup, whichever is 

earlier. In the simple case having two time intervals with constant flow the first pump operational 

interval selected is 24 hours reduced for start delay and the second time interval selected is 

from 24 h to 72 h. The constant flows for these two intervals were then used for verification 

calculations.
9

4. 9BRESULTS

Two kinds of SBO calculations for two-loop PWR are performed, base and verification. Base 

calculations are needed to determine necessary flowrate for SGs feeding in such a way that it is 

not overfilled or emptied. The verification calculations are then performed to verify if the 

determined minimum flowrates are sufficient to prevent core heatup.

4.1 15BBase Calculations

All six types of scenarios shown in Table 1 were simulated. For SL_LOSS type scenarios, as 

shown on Figure 3, core heatup starts well before 24 hours. 

Figure 3 Influence of SG Injection Delay on Average Rod Cladding Temperature for

SL_LOSS Scenario Type: (a) Case C1, (b) Case C2

Obtained results are expected considering large letdown loss of RCS coolant, and 

consequential core uncovery. This means that for these type scenarios cooling of RCS by 

400

600

800

1000

1200

1400

1600

0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15

Temperature (K)

Time (h)

SL_LOSS_0 (C1)

SL_LOSS_0.5 (C1)

SL_LOSS_1 (C1)

SL_LOSS_2 (C1)

SL_LOSS_3 (C1)

SL_LOSS_4 (C1)

SL_LOSS_5 (C1)

(a)

SBO start

400

600

800

1000

1200

1400

1600

0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15

Temperature (K)

Time (h)

SL_LOSS_0 (C2)

SL_LOSS_0.5 (C2)

SL_LOSS_1 (C2)

SL_LOSS_2 (C2)

SL_LOSS_3 (C2)

(b)
10

secondary side is not sufficient and that RCS injection is also needed. Figure 4 shows the RCS 

pressure. In SL_LOSS type scenarios the pressure initially drops due to RCS loss. Later there is 

plateau due to coupling of primary and secondary pressure. When injection to SGs is started, 

the RCS pressures drop but later start to increase again due to natural circulation lost. 

Therefore the core boil-off started and when the core is uncovered it starts to heatup.

Figure 4 Influence of SG Injection Delay on RCS Pressure for SL_LOSS Scenario Type:

(a) Case C1, (b) Case C2

For other than SL_LOSS type scenarios there is in general no core heatup if injection to SGs is 

not delayed too much. For the case C1 with EDG running one hour after reactor trip on LOOP, 

the time 5 h time delay in injection causes core heatup, while in C2 case with reactor trip on 

SBO without running EDG the 3h time delay in injection causes core heatup (see Figure 6). If 

0

2

4

6

8

10

12

14

16

18

20

0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15

Pressure (MPa)

Time (h)

SL_LOSS_0 (C1)

SL_LOSS_0.5 (C1)

SL_LOSS_1 (C1)

SL_LOSS_2 (C1)

SL_LOSS_3 (C1)

SL_LOSS_4 (C1)

SL_LOSS_5 (C1)

safety valve setpoint

SBO start

(a)

0

2

4

6

8

10

12

14

16

18

20

0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15

Pressure (MPa)

Time (h)

SL_LOSS_0 (C2)

SL_LOSS_0.5 (C2)

SL_LOSS_1 (C2)

SL_LOSS_2 (C2)

SL_LOSS_3 (C2)

safety valve setpoint

(b)
11

delays are one hour less, only for C1 and C2 S_LOSS_4 scenarios the core heatup started after 

61 h and 48 h, as shown in Figure 5.

Figure 5 Comparison of Scenario Types for Average Rod Cladding Temperature:

(a) Case C1 With 4 H Delay, (b) Case C2 With 2 H Delay

Figure 6 shows that core heatup is not prevented for C1 scenarios with 5 hours delay and for C2 

scenarios with 3 hours delay. The reason is that the steam generators are too long dry without 

makeup water and so core cooling through secondary side is lost. Due to insufficient cooling, 

the pressurizer safety valves start to discharge the RCS mass. The large loss of RCS mass 

results in the core uncovery and heatup after few hours.

400

500

600

700

800

0 5 10 15 20 25 30 35 40 45 50 55 60 65 70

Temperature (K)

Time (h)

NO_LOSS_4 (C1)

N_LOSS_4 (C1)

S_LOSS_4 (C1)

SLD_LOSS_4 (C1)

NSLD_LOSS_4 (C1)

(a)

SBO start

400

500

600

700

800

0 5 10 15 20 25 30 35 40 45 50 55 60 65 70

Temperature (K)

Time (h)

NO_LOSS_2 (C2)

N_LOSS_2 (C2)

S_LOSS_2 (C2)

SLD_LOSS_2 (C2)

NSLD_LOSS_2 (C2)

(b)
12

Figure 6 Comparison of Scenario Types for Average Rod Cladding Temperature:

(a) Case C1 with 5 h Delay, (b) Case C2 with 3 h Delay

4.2 16BAssessment of Necessary Flowrates

The assessed necessary flowrates in the first 24 h for the scenarios not leading to core heatup 

(for both Case 1 and 2) are given in Table 2. The first row in Table 2 specifies the RCS coolant 

loss scenario defined in Section 3.3. The first column contains the delay time, in hours, of the 

pump start following the extended SBO. The pump constant flows in first 24 hours are given in 

the remaining columns for Case 1 and Case 2 scenarios. The largest minimum required flows are 

400

500

600

700

800

900

0 1 2 3 4 5 6

Temperature (K)

Time (h)

NO_LOSS_5 (C1)

N_LOSS_5 (C1)

S_LOSS_5 (C1)

SL_LOSS_5 (C1)

SLD_LOSS_5 (C1)

NSLD_LOSS_5 (C1)

(a)

SBO start

400

500

600

700

800

900

0 1 2 3 4 5 6

Temperature (K)

Time (h)

NO_LOSS_3 (C2)

N_LOSS_3 (C2)

S_LOSS_3 (C2)

SL_LOSS_3 (C2)

SLD_LOSS_3 (C2)

NSLD_LOSS_3 (C2)

(b)
13

obtained for SLD_LOSS type scenarios and the smallest for S_LOSS type scenarios. For all 

SL_LOSS scenarios, as shown on Figure 3, core heatup starts before 24 hours. Obtained results 

are expected considering large letdown loss of RCS coolant, which consequently cause core 

uncovery and heatup.

Table 2 Minimum Necessary Constant Flows within the 24 h After SBO Start

Scenario type NO_LOSS N_LOSS S_LOSS SLD_LOSS NSLD_LOSS

Case 1, EDG running 1 h

Delay (h) Flow (kg/s) Flow (kg/s) Flow (kg/s) Flow (kg/s) Flow (kg/s)

0 4.79 4.77 4.68 5.51 5.50

0.5 4.90 4.87 4.78 5.63 5.62

1 4.99 4.98 4.88 5.75 5.74

2 5.22 5.20 5.11 6.00 5.99

3 5.47 5.45 5.34 6.25 6.25

4 5.63 5.62 5.51 6.48 6.47

Case 2, EDG not running

Delay (h) Flow (kg/s) Flow (kg/s) Flow (kg/s) Flow (kg/s) Flow (kg/s)

0 5.50 5.46 5.38 6.21 6.19

0.5 5.60 5.58 5.49 6.34 6.33

1 5.72 5.70 5.61 6.46 6.45

2 5.86 5.83 5.72 6.54 6.57

For the time interval from 24 h to 72 h the minimum necessary flowrates assessed were 3.0 kg/s 

for NO_LOSS, SLD_LOSS and NSLD_LOSS cases and 2.8 kg/s for S_LOSS and N_LOSS.

4.3 17BVerification Calculations

The third step in the methodology presented in Section 2 is to verify the calculated minimum 

required pump flows. The verification calculations are performed for all scenarios shown in Table 

2 in which core heat up is prevented in the first 24 h, except for zero delay. Namely, it is assumed 

that at least half an hour will be needed to start the portable pump.

The main results are shown in Figures 7 through 11. The obtained results show that if pump, TDAFW or mobile, injects the assessed minimum necessary flows then effective cooling of reactor 

core is provided and SGs are not overfilled.

Results of the verification runs for SLD_LOSS type scenarios for C1 and C2 cases are given in 

Figure 7. Figure 7(b2) shows that in SLD_LOSS_2, Case 2 scenario with minimum required pump 

flow of 6.54 kg/s will result in core heat up. The TD-AFW is operating at maximum flow in the 

initial period of base calculation presented in previous section resulting in fast recovery of the SG 

inventory. The calculated constant flow of 6.54 kg/s is substantially smaller comparing to mass 

flow of 45 kg/s in base calculation in this initial period, therefore verification calculation resulted in 

insufficient cooling and core heat up. When the constant flow of 10 kg/s is assumed (verification 

calculation SLD_LOSS_2f1 (C2)) in the first 4 h and 5.77 kg/s in the remaining 18 h, core heatup 

is prevented as shown in Figure 7(b2). Figures 7(a1) and (a2) show that the RCS pressure drops 

very quickly after the start of the secondary side depressurization but then increase again, if 

secondary side injection is not started in due time. The RCS pressure in SLD_LOSS_4 scenario, 

Case 1 and SLD_LOSS_2, Case 2 reached the pressurizer safety valves setpoint 17.2 MPa 

resulting in RCS mass loss through the pressurizer safety valve. This RCS mass loss causes core 

uncovery resulting in heatup in SLD_LOSS_2, Case 2 as shown on Figure 7(b2). The steam 
14

generator wide range levels on Figures 7(c1), (c2), (d1) and (d2) show the steam generator levels 

start to increase gradually after injection with no overfill in the first 24 h.

Results of the verification runs for NSLD_LOSS type scenarios for C1 and C2 cases are shown 

in Figure 8. There are very similar to SLD_LOSS type scenarios because due to 

depressurization the contribution of normal leakage loss is rather small (up to 5 tonnes RCS 

mass loss in 24 h). Slightly larger loss resulted in slightly earlier core heatup when comparing 

SLD_LOSS_2 (C2) and NSLD_LOSS_2 (C2) calculations.

Figure 9 shows the RCS pressure, core collapsed liquid level and both steam generator wide 

range levels for S_LOSS type scenarios. Figures 9(b1) and (b2) show the benefit of letdown 

isolation and limitation of RCS loss to the RCP seals. Figures 9(c1) to (d2) show that in certain 

periods only one out of two steam generators is filling and vice versa. The small oscillations 

present in the SG level trends are due to relief valves discharge at the discrete periods.

Figure 10 shows results for N_LOSS type scenarios for C1 and C2 cases. Results of N_LOSS

type scenarios for first 72 h period are similar as for S_LOCA type, except that slightly larger 

(around 1.5%) feed flow to SGs is needed due to smaller RCS loss in N_LOSS type scenarios.

Figure 11 shows results for NO_LOSS type scenarios for C1 and C2 cases. In the 72 h period, 

as shown on Figures 11(b1) and (b2), core is covered as a result of the effective cooling. In 

NO_LOSS scenarios the loss of secondary side heat sink results in discharge of RCS coolant 

through pressurizer safety valves when safety valves setpoint is reached (see Figures 11(a1)

and (a2)). The duration of the pressurizer safety valves opening and by this the mass of 

discharged RCS coolant depends on the delay of pump start and establishment of SG feed flow. 

The steam generator levels recover in 24 h and the level is not overfilled in the 72 h period as 

shown in Figures 11(c1) to (d2).
15

Figure 7 Main Parameters for SLD_LOSS Verification Calculations: (a) RCS Pressure,

(b) Average Rod Cladding Temperature, (c) SG No. 1 WR Level,

(d) SG No. 2 WR Level.

0

2

4

6

8

10

12

14

16

18

20

0 10 20 30 40 50 60 70

Pressure (MPa)

Time (h)

SLD_LOSS_0.5 (C1)

SLD_LOSS_1 (C1)

SLD_LOSS_2 (C1)

SLD_LOSS_3 (C1)

SLD_LOSS_4 (C1)

safety valve setpoint

(a1)

SBO start 0

2

4

6

8

10

12

14

16

18

20

0 10 20 30 40 50 60 70

Pressure (MPa)

Time (h)

SLD_LOSS_0.5 (C2)

SLD_LOSS_1 (C2)

SLD_LOSS_2 (C2)

SLD_LOSS_2f1 (C2)

safety valve setpoint

(a2)

0

20

40

60

80

100

120

0 10 20 30 40 50 60 70

Level (%)

Time (h)

SLD_LOSS_0.5 (C1)

SLD_LOSS_1 (C1)

SLD_LOSS_2 (C1)

SLD_LOSS_3 (C1)

SLD_LOSS_4 (C1)

(b1)

SBO start

0

20

40

60

80

100

120

0 10 20 30 40 50 60 70

Level (%)

Time (h)

SLD_LOSS_0.5 (C2)

SLD_LOSS_1 (C2)

SLD_LOSS_2 (C2)

SLD_LOSS_2f1 (C2)

(b2)

0

10

20

30

40

50

60

70

80

90

100

0 10 20 30 40 50 60 70

Level (%)

Time (h)

SLD_LOSS_0.5 (C1)

SLD_LOSS_1 (C1)

SLD_LOSS_2 (C1)

SLD_LOSS_3 (C1)

SLD_LOSS_4 (C1)

(c1)

SBO start

0

10

20

30

40

50

60

70

80

90

100

0 10 20 30 40 50 60 70

Level (%)

Time (h)

SLD_LOSS_0.5 (C2)

SLD_LOSS_1 (C2)

SLD_LOSS_2 (C2)

SLD_LOSS_2f1 (C2)

(c2)

0

10

20

30

40

50

60

70

80

90

100

0 10 20 30 40 50 60 70

Level (%)

Time (h)

SLD_LOSS_0.5 (C1)

SLD_LOSS_1 (C1)

SLD_LOSS_2 (C1)

SLD_LOSS_3 (C1)

SLD_LOSS_4 (C1)

(d1)

SBO start

0

10

20

30

40

50

60

70

80

90

100

0 10 20 30 40 50 60 70

Level (%)

Time (h)

SLD_LOSS_0.5 (C2)

SLD_LOSS_1 (C2)

SLD_LOSS_2 (C2)

SLD_LOSS_2f1 (C2)

(d2)
16

Figure 8 Main Parameters for NSLD_LOSS Verification Calculations: (a) RCS Pressure,

(b) Average Rod Cladding Temperature, (c) SG No. 1 WR Level,

(d) SG No. 2 WR Level.

0

2

4

6

8

10

12

14

16

18

20

0 10 20 30 40 50 60 70

Pressure (MPa)

Time (h)

NSLD_LOSS_0.5 (C1)

NSLD_LOSS_1 (C1)

NSLD_LOSS_2 (C1)

NSLD_LOSS_3 (C1)

NSLD_LOSS_4 (C1)

safety valve setpoint

(a1)

SBO start 0

2

4

6

8

10

12

14

16

18

20

0 10 20 30 40 50 60 70

Pressure (MPa)

Time (h)

NSLD_LOSS_0.5 (C2)

NSLD_LOSS_1 (C2)

NSLD_LOSS_2 (C2)

NSLD_LOSS_2f1 (C2)

safety valve setpoint

(a2)

400

600

800

1000

1200

1400

0 10 20 30 40 50 60 70

Temperature (K)

Time (h)

NSLD_LOSS_0.5 (C1)

NSLD_LOSS_1 (C1)

NSLD_LOSS_2 (C1)

NSLD_LOSS_3 (C1)

NSLD_LOSS_4 (C1)

(b1)

SBO start

400

600

800

1000

1200

1400

0 10 20 30 40 50 60 70

Temperature (K)

Time (h)

NSLD_LOSS_0.5 (C2)

NSLD_LOSS_1 (C2)

NSLD_LOSS_2 (C2)

NSLD_LOSS_2f1 (C2)

(b2)

0

10

20

30

40

50

60

70

80

90

100

0 10 20 30 40 50 60 70

Level (%)

Time (h)

NSLD_LOSS_0.5 (C1)

NSLD_LOSS_1 (C1)

NSLD_LOSS_2 (C1)

NSLD_LOSS_3 (C1)

NSLD_LOSS_4 (C1)

(c1)

SBO start

0

10

20

30

40

50

60

70

80

90

100

0 10 20 30 40 50 60 70

Level (%)

Time (h)

NSLD_LOSS_0.5 (C2)

NSLD_LOSS_1 (C2)

NSLD_LOSS_2 (C2)

NSLD_LOSS_2f1 (C2)

(c2)

0

10

20

30

40

50

60

70

80

90

100

0 10 20 30 40 50 60 70

Level (%)

Time (h)

NSLD_LOSS_0.5 (C1)

NSLD_LOSS_1 (C1)

NSLD_LOSS_2 (C1)

NSLD_LOSS_3 (C1)

NSLD_LOSS_4 (C1)

(d1)

SBO start

0

10

20

30

40

50

60

70

80

90

100

0 10 20 30 40 50 60 70

Level (%)

Time (h)

NSLD_LOSS_0.5 (C2)

NSLD_LOSS_1 (C2)

NSLD_LOSS_2 (C2)

NSLD_LOSS_2f1 (C2)

(d2)
17

Figure 9 Main Parameters for S_LOSS Verification Calculations: (a) RCS Pressure, 

(b) Core Collapsed Liquid Level, (c) SG No.1 WR Level,

(d) SG No.2 WR Level.

0

2

4

6

8

10

12

14

16

18

20

0 10 20 30 40 50 60 70

Pressure (MPa)

Time (h)

S_LOSS_0.5 (C1)

S_LOSS_1 (C1)

S_LOSS_2 (C1)

S_LOSS_3 (C1)

S_LOSS_4 (C1)

safety valve setpoint

(a1)

SBO start 0

2

4

6

8

10

12

14

16

18

20

0 10 20 30 40 50 60 70

Pressure (MPa)

Time (h)

S_LOSS_0.5 (C2)

S_LOSS_1 (C2)

S_LOSS_2 (C2)

safety valve setpoint

(a2)

0

20

40

60

80

100

120

0 10 20 30 40 50 60 70

Level (%)

Time (h)

S_LOSS_0.5 (C1)

S_LOSS_1 (C1)

S_LOSS_2 (C1)

S_LOSS_3 (C1)

S_LOSS_4 (C1)

(b1)

SBO start

0

20

40

60

80

100

120

0 10 20 30 40 50 60 70

Level (%)

Time (h)

S_LOSS_0.5 (C2)

S_LOSS_1 (C2)

S_LOSS_2 (C2)

(b2)

0

10

20

30

40

50

60

70

80

90

0 10 20 30 40 50 60 70

Level (%)

Time (h)

S_LOSS_0.5 (C1)

S_LOSS_1 (C1)

S_LOSS_2 (C1)

S_LOSS_3 (C1)

S_LOSS_4 (C1)

(c1)

SBO start

0

10

20

30

40

50

60

70

80

90

0 10 20 30 40 50 60 70

Level (%)

Time (h)

S_LOSS_0.5 (C2)

S_LOSS_1 (C2)

S_LOSS_2 (C2)

(c2)

0

10

20

30

40

50

60

70

80

90

0 10 20 30 40 50 60 70

Level (%)

Time (h)

S_LOSS_0.5 (C1)

S_LOSS_1 (C1)

S_LOSS_2 (C1)

S_LOSS_3 (C1)

S_LOSS_4 (C1)

(d1)

SBO start

0

10

20

30

40

50

60

70

80

90

0 10 20 30 40 50 60 70

Level (%)

Time (h)

S_LOSS_0.5 (C2)

S_LOSS_1 (C2)

S_LOSS_2 (C2)

(d2)
18

Figure 10 Main Parameters for N_LOSS Verification Calculations: (a) RCS Pressure, 

(b) Core Collapsed Liquid Level, (c) SG No. 1 WR Level,

(d) SG No. 2 WR Level.

0

2

4

6

8

10

12

14

16

18

20

0 10 20 30 40 50 60 70

Pressure (MPa)

Time (h)

N_LOSS_0.5 (C1)

N_LOSS_1 (C1)

N_LOSS_2 (C1)

N_LOSS_3 (C1)

N_LOSS_4 (C1)

safety valve setpoint

(a1)

SBO start 0

2

4

6

8

10

12

14

16

18

20

0 10 20 30 40 50 60 70

Pressure (MPa)

Time (h)

N_LOSS_0.5 (C2)

N_LOSS_1 (C2)

N_LOSS_2 (C2)

safety valve setpoint

(a2)

0

20

40

60

80

100

120

0 10 20 30 40 50 60 70

Level (%)

Time (h)

N_LOSS_0.5 (C1)

N_LOSS_1 (C1)

N_LOSS_2 (C1)

N_LOSS_3 (C1)

N_LOSS_4 (C1)

(b1)

SBO start

0

20

40

60

80

100

120

0 10 20 30 40 50 60 70

Level (%)

Time (h)

N_LOSS_0.5 (C2)

N_LOSS_1 (C2)

N_LOSS_2 (C2)

(b2)

0

10

20

30

40

50

60

70

80

90

0 10 20 30 40 50 60 70

Level (%)

Time (h)

N_LOSS_0.5 (C1)

N_LOSS_1 (C1)

N_LOSS_2 (C1)

N_LOSS_3 (C1)

N_LOSS_4 (C1)

(c1)

SBO start

0

10

20

30

40

50

60

70

80

90

0 10 20 30 40 50 60 70

Level (%)

Time (h)

N_LOSS_0.5 (C2)

N_LOSS_1 (C2)

N_LOSS_2 (C2)

(c2)

0

10

20

30

40

50

60

70

80

90

0 10 20 30 40 50 60 70

Level (%)

Time (h)

N_LOSS_0.5 (C1)

N_LOSS_1 (C1)

N_LOSS_2 (C1)

N_LOSS_3 (C1)

N_LOSS_4 (C1)

(d1)

SBO start

0

10

20

30

40

50

60

70

80

90

0 10 20 30 40 50 60 70

Level (%)

Time (h)

N_LOSS_0.5 (C2)

N_LOSS_1 (C2)

N_LOSS_2 (C2)

(d2)
19

Figure 11 Main Parameters for NO_LOSS Verification Calculations: (a) RCS Pressure,

(b) Core Collapsed Liquid Level, (c) SG No. 1 WR Level,

(d) SG No. 2 WR Level.

0

2

4

6

8

10

12

14

16

18

20

0 10 20 30 40 50 60 70

Pressure (MPa)

Time (h)

NO_LOSS_0.5 (C1)

NO_LOSS_1 (C1)

NO_LOSS_2 (C1)

NO_LOSS_3 (C1)

NO_LOSS_4 (C1)

safety valve setpoint

(a1)

SBO start

0

2

4

6

8

10

12

14

16

18

20

0 10 20 30 40 50 60 70

Pressure (MPa)

Time (h)

NO_LOSS_0.5 (C2)

NO_LOSS_1 (C2)

NO_LOSS_2 (C2)

safety valve setpoint

(a2)

0

20

40

60

80

100

120

0 10 20 30 40 50 60 70

Level (%)

Time (h)

NO_LOSS_0.5 (C1)

NO_LOSS_1 (C1)

NO_LOSS_2 (C1)

NO_LOSS_3 (C1)

NO_LOSS_4 (C1)

(b1)

SBO start

0

20

40

60

80

100

120

0 10 20 30 40 50 60 70

Level (%)

Time (h)

NO_LOSS_0.5 (C2)

NO_LOSS_1 (C2)

NO_LOSS_2 (C2)

(b2)

0

10

20

30

40

50

60

70

80

90

0 10 20 30 40 50 60 70

Level (%)

Time (h)

NO_LOSS_0.5 (C1)

NO_LOSS_1 (C1)

NO_LOSS_2 (C1)

NO_LOSS_3 (C1)

NO_LOSS_4 (C1)

(c1)

SBO start

0

10

20

30

40

50

60

70

80

90

0 10 20 30 40 50 60 70

Level (%)

Time (h)

NO_LOSS_0.5 (C2)

NO_LOSS_1 (C2)

NO_LOSS_2 (C2)

(c2)

0

10

20

30

40

50

60

70

80

90

0 10 20 30 40 50 60 70

Level (%)

Time (h)

NO_LOSS_0.5 (C1)

NO_LOSS_1 (C1)

NO_LOSS_2 (C1)

NO_LOSS_3 (C1)

NO_LOSS_4 (C1)

(d1)

SBO start

0

10

20

30

40

50

60

70

80

90

0 10 20 30 40 50 60 70

Level (%)

Time (h)

NO_LOSS_0.5 (C2)

NO_LOSS_1 (C2)

NO_LOSS_2 (C2)

(d2)
20
21

5. 10BCONCLUSIONS

Application of the RELAP5/MOD3.3 Patch 4 for study of the utilization of pump for mitigation of 

the extended blackout condition is investigated. The methodology for the assessment of the

necessary flowrates to be injected into steam generators for mitigation of extended blackout event 

is proposed. The assessed flows minimize the required number of flow changes and operators 

errors during required manipulations. The verification calculations are performed with pump 

flowrate equal to assessed flowrate to assure effective core cooling without overfilling the steam 

generators.

Six different scenario types of reactor coolant loss are developed and analyzed. Two cases are 

defined considering the operation of the emergency diesel generators. Different delays of the 

pump injection start following the station blackout are assumed and analyzed.

Obtained results demonstrated the need for utilization of deterministic system computer code for 

assessment of the pump flows. Obtained results show that typical pressurized water reactor with 

the leak tight reactor coolant pump seals with small seal loss or depressurization of the reactor 

coolant system can cope first 72 h of extended SBO if pump start to inject on secondary side 

within 2 h of the SBO start. Operation of the emergency diesel generator (EDG) for one hour 

extends the available time for the start of pump on 4 h. Failure to isolate letdown results in core 

damage before 24 h in all analyzed scenarios. Obtained results show that limitation of RCS loss 

prolongs the time before heat up.

One of the main conclusions in the study is that availability of equipment is prerequisite but not 

guarantee of successful mitigation, if it is not done in due time. The supporting analyses are 

necessary in order to verify the effectiveness of the proposed strategy.

23

6. 1BREFERENCES

1. Volkanovski, A., Prošek, A., 2013. Extension of station blackout coping capability and 

implications on nuclear safety. Nuclear Engineering and Design 255, 16-27.

2. ENSREG, 2011. Stress tests performed on European nuclear power plants - EU "Stress 

tests" specifications. European Nuclear Safety Regulators Group.

3. ENSREG, 2012. Stress tests performed on European nuclear power plants - Peer review 

report.

4. Volkanovski, A., Prošek, A., 2013. Extension of station blackout coping capability and 

implications on nuclear safety. Nuclear Engineering and Design 255, 16-27.

5. NEI, 2012. Diverse and flexible coping strategies (FLEX) implementation guide. Nuclear 

Energy Institute, Washington.

6. A. Prošek, B. Mavko, “Reactor trip analysis at Krško Nuclear Power Plant”, International 

agreement report NUREG/IA, 0221, Office of Nuclear Regulatory Research, U.S. Nuclear 

Regulatory Commission, Washington, 2010, pp. 1-57.

7. A. Prošek, I. Parzer, and B. Krajnc, “Simulation of hypothetical small-break loss-of-coolant 

accident in modernized nuclear power plant”, Electrotechnical Review, 71(4), 2004, pp. 

191-196.

8. 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, 49(9), 

2003, pp. 430-444.

9. Krajnc, B., Glaser, B., Jalovec, R., Špalj, S., 2011. MAAP Station Blackout Analyses as a 

Support for the NPP Krško STORE (Safety Terms of Reference) Actions, New Energy for 

New Europe Slovenia.

10. Westinghouse, 2012. PRA Model for the Westinghouse Shutdown Seal, WCAP-17100-

NP Supplement 1, Revision 0.

11. Imbro, E.V., Lanning, Wayne D., 1980. AEOD Observations and Recommendations 

Concerning the Problem of Steam Generator Overfill and Combined Primary and 

Secondary Side Blowdown. U.S. Nuclear Regulatory Commission, Washington.







NUREG/IA-0467 RELAP5 Analysis of Mitigation Strategy for Extended Blackout Power 

Condition in PWR