NUREG/IA-0249

 International 

Agreement Report

Loss of External Load Analysis with

RELAP5/MOD3.3 Patch 03 Code

Prepared by:

A. Prošek, B. Mavko

Jožef Stefan Institute

Jamova cesta 39

SI-1000 Ljubljana, Slovenia 

A. Calvo, NRC Project Manager 

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

December 2010 

Prepared as part of

The Agreement on Research Participation and Technical Exchange 

Under the International Code Assessment and Maintenance Program (CAMP)

 

Published by

U.S. Nuclear Regulatory Commission
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NUREG/IA-0249

Loss of external load analysis with

RELAP5/MOD3.3 Patch 03 code

Manuscript Completed: September 2010

Date Published: December 2010

Prepared by: Andrej Prošek, Borut Mavko

Jožef Stefan Institute

Jamova cesta 39

SI-1000 Ljubljana, Slovenia

A. Calvo, NRC Project Manager

Prepared as part of

The Agreement on Research Participation and Technical Exchange 

Under the International Code Assessment and Maintenance Program (CAMP)

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

iii

ABSTRACT

By removing resistance temperature detectors (RTD) bypass manifold system with fastresponse thermowell-mounted RTDs the reactor coolant system temperature measurement time 

may change, what influences the time delays for overtemperature ΔT (OTΔT) and overpower 

ΔT (OPΔT) trips. The total time delays should not exceed the value assumed in the accident 

analyses.

In this paper we are presenting Krško nuclear power plant specific analysis of loss of external 

load varying the total delay time for OTΔT trip. For simulations the latest RELAP5/MOD3.3 

Patch 03 thermal hydraulic computer code was used. The verified standard RELAP5/MOD3.3 

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

calculation was compared to safety analysis report calculation to qualify the RELAP5 calculated 

results before the sensitivity analyses were performed.

v

CONTENTS

Page

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

Executive Summary...................................................................................................................ix

Acknowledgements ....................................................................................................................x

Abbreviations .............................................................................................................................xi

1. Introduction ...........................................................................................................................1

2. Plant Description...................................................................................................................3

3. Input Model Description .......................................................................................................5

4. Scenario Description ............................................................................................................7

5. Results ...................................................................................................................................9

5.1 Base Calculation ...............................................................................................................9

5.2 Sensitivity Calculations....................................................................................................15

5.2.1 Variation of Initial and Boundary Conditions........................................................15

5.2.2 Variation of Measurement Delay Time for Case 1...............................................23

5.2.3 Variation of Measurement Delay Time for Case 2...............................................28

5.2.4 Variation of Measurement Delay Time for Case 3...............................................33

5.2.5 Summary of Results ............................................................................................38

5.3 Results Discussion ..........................................................................................................39

6. Run Statistics ......................................................................................................................41

7. Conclusions.........................................................................................................................43

8. References...........................................................................................................................45
vi

Figures

Page

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

2. Reactor power - base case comparison between RELAP5 and USAR .................................10

3. Pressurizer pressure - base case comparison between RELAP5 and USAR........................10

4. Pressurizer volume - base case comparison between RELAP5 and USAR ..........................11

5. RCS average temperature and measured RCS average temperature - base case comparison 

between RELAP5 and USAR................................................................................................11

6. Hot leg no. 1 temperature - base case comparison between RELAP5 and USAR ................12

7. Cold leg no. 1 temperature - base case comparison between RELAP5 and USAR ..............12

8. Pressurizer relief valves flow - base case comparison between RELAP5 and USAR ...........13

9. Steam generator no. 1 pressure - base case comparison between RELAP5 and USAR ......13

10. Steam generator no. 1 relief valves mass flow - base case comparison between RELAP5 

and USAR .............................................................................................................................14

11. Departure from nucleate boiling ratio (critical heat flux ratio) - base case comparison 

between RELAP5 and USAR................................................................................................14

12. Reactor power – influence of I&B conditions ........................................................................16

13. Pressurizer pressure - influence of I&B conditions ...............................................................16

14. Pressurizer volume - influence of I&B conditions..................................................................17

15. RCS average temperature and measured RCS average temperature - influence of I&B 

conditions..............................................................................................................................17

16. Hot leg no. 1 temperature - influence of I&B conditions........................................................18

17. Cold leg no. 1 temperature - influence of I&B conditions ......................................................18

18. Pressurizer relief valves flow - influence of I&B conditions ...................................................19

19. Steam generator no. 1 pressure - influence of I&B conditions..............................................19

20. Steam generator no. 1 relief valves mass flow - influence of I&B conditions........................20

21. Mass flow through hottest axial location in the core - influence of I&B conditions ................20

22. Equilibrium quality at the hottest axial location in the core - influence of I&B conditions ......21

23. Critical heat flux ratio - influence of I&B conditions ...............................................................21

24. Overtemperature ΔT setpoint calculation and measured delta T - influence of I&B conditions

..............................................................................................................................................22

25. Pressurizer pressure - influence of measurement delay for case 1 ......................................23

26. RCS average temperature and measured RCS average temperature - influence of 

measurement delay for case 1..............................................................................................24

27. Mass flow through hottest axial location in the core - influence of measurement delay for 

case 1....................................................................................................................................24

28. Equilibrium quality at the hottest axial location in the core - influence of measurement delay 

for case 1 ..............................................................................................................................25

29. Critical heat flux ratio - influence of measurement delay for case 1......................................25

30. K2 temperature term calculation - influence of measurement delay for case 1 ....................26

31. K3 pressure term calculation - influence of measurement delay for case 1..........................26

32. Overtemperature ΔT setpoint calculation - influence of measurement delay for case 1.......27

33. Measured delta T - influence of measurement delay for case 1 ...........................................27

34. Pressurizer pressure - influence of measurement delay for case 2 ......................................28

35. RCS average temperature and measured RCS average temperature - influence of 

measurement delay for case 2..............................................................................................29

36. Mass flow through hottest axial location in the core - influence of measurement delay for 

case 2....................................................................................................................................29
vii

37. Equilibrium quality at the hottest axial location in the core - influence of measurement delay 

for case 2 ..............................................................................................................................30

38. Critical heat flux ratio - influence of measurement delay for case 2......................................30

39. K2 temperature term calculation - influence of measurement delay for case 2 ....................31

40. K3 pressure term calculation - influence of measurement delay for case 2..........................31

41. Overtemperature ΔT setpoint calculation - influence of measurement delay for case 2.......32

42. Measured delta T - influence of measurement delay for case 2 ...........................................32

43. Pressurizer pressure - influence of measurement delay for case 3 ......................................33

44. RCS average temperature and measured RCS average temperature - influence of 

measurement delay for case 3..............................................................................................34

45. Mass flow through hottest axial location in the core - influence of measurement delay for 

case 3....................................................................................................................................34

46. Equilibrium quality at the hottest axial location in the core - influence of measurement delay 

for case 3 ..............................................................................................................................35

47. Critical heat flux ratio - influence of measurement delay for case 3......................................35

48. K2 temperature term calculation - influence of measurement delay for case 3 ....................36

49. K3 pressure term calculation - influence of measurement delay for case 3..........................36

50. Overtemperature ΔT setpoint calculation - influence of measurement delay for case 3.......37

51. Measured delta T - influence of measurement delay for case 3 ...........................................37

Tables

Page

1. Comparison between USAR and RELAP5 initial conditions ...................................................8

2. Sensitivity of OTΔT trip time and minimum CHFR on RCS temperature measurement 

response time........................................................................................................................38

3. Run statistics.........................................................................................................................41

ix

EXECUTIVE SUMMARY

Several nuclear power plants remove resistance temperature detectors (RTD) bypass manifold 

system with fast-response thermowell-mounted RTDs. The reactor coolant system temperature 

measurement time may therefore change, what influences the total response time for 

overtemperature ΔT (OTΔT) and overpower ΔT (OPΔT) trip signals. The total response time 

should not exceed the time delay for trip assumed in the accident analyses. The objective of this 

study was to investigate the influence of measurement delay on OTΔT trip protection during loss 

of external load analysis. For simulations the latest RELAP5/MOD3.3 Patch 03 thermal 

hydraulic computer code was used. The verified standard RELAP5/MOD3.3 input model from 

2008 (cycle 23) was delivered by Krško nuclear power plant. First comparison with the original 

updated safety analysis report (USAR) calculation was done to show that main plant parameters 

agree with USAR calculation and that the quantitative differences are understood. The 

differences were mainly due to different computer codes, as the code used for USAR calculation 

was not best-estimate code. Then sensitivity analyses were performed. The results showed that 

when the temperature measurement delay was rather small (e.g. ± 2 seconds of the value 

assumed in USAR), also the trip time on OTΔT was similarly delayed. The study also 

demonstrated that the trends for departure from nuclear boiling ratio can be evaluated by 

RELAP5/MOD3.3 Patch 03 using the critical heat flux ratio for the average core condition, while 

for the licensing calculation of minimum DNBR the coupling of RELAP5 code with the code 

having capability to calculate transient DNBR for hot rod is required, taken into account local 

mass flux and local quality.
x

ACKNOWLEDGEMENTS

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

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

budget by the Slovenian Research Agency (program no. P2-0026). The RELAP5/MOD3.3 base 

input model of Krško nuclear power plant is the courtesy of Krško Nuclear Power Plant. Finally, 

special thanks to M.Sc. Božidar Krajnc, director of engineering services division in Krško 

nuclear power plant, for expressing the need for such an analysis and for permission to publish 

the results from updated safety analysis report.
xi

ABBREVIATIONS

ACC accumulator

AFW auxiliary feedwater

AP1000 advanced passive 1000

CHFR critical heat flux ratio

CPU central processing unit

CVCS chemical and volume control system

DNB departure from nucleate boiling

DNBR departure from nucleate boiling ratio

ECCS emergency core cooling system

HPSI high-pressure safety injection

I&B initial and boundary

LPSI low-pressure safety injection

MD motor driven

MFW main feedwater

MPa megapascal

MWt megawatt thermal

NPP nuclear power plant

OPΔT overpower delta T

OTΔT overtemperature delta T

PORV power operated relief valve

PRZ pressurizer

PWR pressurized water reactor

RCP reactor coolant pump

RCS reactor coolant system

RPV reactor pressure vessel

RTD resistance temperature detector

SG steam generator

SL surge line

TD turbine driven

USAR updated safety analysis report

1

1. INTRODUCTION

Several nuclear power plants remove resistance temperature detectors (RTD) bypass manifold 

system with fast-response thermowell-mounted RTDs. The total response time for 

overtemperature ΔT (OTΔT) and overpower ΔT (OPΔT) may affect the accident analyses. Fast 

acting RTD/thermowell response time is slower than response of direct immersion RTD, but 

there is no loop transport or thermal lag. Depending on the delay of fast acting RTD, new 

analysis may be needed or not. The total response time for OTΔT is specified in the plant 

technical specifications. When new total response time is bounded by technical specification 

response time, safety analyses are not affected. The influence of measurement delay on 

departure from nucleate boiling (DNB), which is protected by OTΔT trip, was investigated for the 

selected loss of external load transient. For simulations the latest RELAP5/MOD3.3 Patch 03 

thermal hydraulic computer code was used. The verified standard RELAP5/MOD3.3 input model 

from 2008 (cycle 23) was delivered by Krško nuclear power plant. The analysis focused on the 

DNB, while overpressure analysis is not influenced by OTΔT response time.

The core average DNBR is not a safety related item as it is not directly related to minimum 

DNBR in the core, which occurs at some elevation in the limiting flow channel. Similarly, the 

DNBR at the hot spot is not directly related. The minimum DNBR in the limiting flow channel will 

be downstream of the peak heat flux location (hot spot) due to the increased downstream 

enthalpy. The LOFTRAN computer code, which was used in the updated safety analysis report 

(USAR) analysis, has the capability of calculating transient value of DNBR. On the other hand, 

the RELAP5 is limited to calculate the departure from nucleate boiling ratio (DNBR). Also, the 

study assessing the CHF predictive methods accessible in the open literature, including 

RELAP5 methods, showed that no satisfactory model to predict critical heat flux for non-uniform 

heat flux exists (Ref. 1). In RELAP5 the heat transfer critical heat flux ratio (CHFR) can be 

obtained by division of critical heat flux by heat transfer coefficient during subcooled or 

saturated boiling. This ratio has been added to the list of plot variables in the RELAP5/MOD3.3.

The hot spot was modeled by hot rod to evaluate the DNBR trend by RELAP5/MOD3.3. The 

purpose of the study was to perform sensitivity analysis by varying temperature measurement 

delay ± 2 seconds. To confirm the correct trend of DNBR evaluation, first comparison of 

RELAP5 results with the results from the original USAR analysis was done to show that main 

plant parameters including DNBR trend agree and that the quantitative differences are 

understood. 

3

2. PLANT DESCRIPTION

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

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

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

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

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

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

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

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

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

loops.

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

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

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

plant modernization in 2000. 

For more detailed description of the plant the reader is referred to Ref. 2.

5

3. INPUT MODEL DESCRIPTION

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

―Master input deck‖, which have been used for several analyses, including reference 

calculations for Krško full scope simulator verification (Refs. 3, 4, 5). The analysis was 

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

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

The model consists of 469 control volumes, 497 junctions and 378 heat structures with 2107 

radial mesh points. The nodalization scheme of Krško NPP for RELAP5 it is shown in Figure 1. 

The scheme was generated by SNAP through importing ASCII input deck and then manually 

rearranging the position and size of the volumes and junctions.

Modeling of the primary side without the reactor vessel and both loops includes the pressurizer 

(PRZ) vessel, pressurizer surge line (SL), pressurizer spray lines and valves, two pressurizer 

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

control system (CVCS) charging and letdown flow, and RCP seal flow. The reactor vessel 

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

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

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

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

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

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

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

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

low-pressure safety injection (LPSI) pumps. 

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

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

safety valves, main feedwater (MFW) piping, and auxiliary feedwater (AFW) piping from the 

header to the SG. The AFW injects above the SG riser. The main steam no. 1 has same 

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

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

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

dependent volumes, representing the condensate storage tank. For AFW pumps (two motor 

driven and one turbine driven), recirculation flow is modeled too.

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

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

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

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

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

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

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

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

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

control systems schemes the reader can refer to Reference 2. 
6

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

4. SCENARIO DESCRIPTION

Loss of load can result from loss of external electrical load due to some electrical system 

disturbance. Offsite power remains available to operate plant components such as the reactor 

coolant pumps. Following the loss of generator load, the turbine stop valves immediately close. 

This causes sudden reduction in steam flow, resulting in increase in the temperature and 

pressure in the steam generator. Heat transfer rate is reduced, causing the reactor coolant 

temperature to increase, which in turn causes reactor coolant expansion, pressurizer insurge 

and primary pressure rise. For loss of load without subsequent turbine trip, no direct reactor trip 

would be generated, since the plant has full load rejection capability. The plant is expected to 

trip from the reactor protection system signals (pressurizer pressure high, pressurizer water 

level high, overtemperature ΔT, overpower ΔT and steam generator water level low-low). It is 

assumed that plant auxiliary loads require 5% steam flow. Since the reactor trip on turbine trip is

not credited and the steam dump fails by assumption, the primary and secondary side pressure

increase is protected by pressurizer and steam generator safety valves.

The main conservative assumptions in the USAR analysis, which leads to the minimum 

departure from boiling ratio (DNBR), were the following:

reactor trip on high pressurizer pressure setpoint increased (to delay reactor trip),

reactor trip on turbine trip is not credited (reactor trip on other signal is later),

steam dump system not credited (failure of steam dump system could lead to a 

significant increase in a secondary side pressure and a heatup of the primary side),

rod control system not credited (to increase heatup and maximize power),

steam generator power operated relief valves not credited (to increase a secondary side 

pressure and a heatup of the primary side),

feedwater flow follows 5% steam flow until reactor trip (this adds considerably less 

feedwater than in reality when SG level control would attempt to compensate for water 

level decrease so the cooldown is minimized),

pressurizer pressure control assumed to function (to delay reactor trip and to minimize 

pressure for DNB calculation - pressurizer spray and PORVs reduce or limit the 

pressure, while pressurizer heaters are not simulated),

minimum reactivity feedback (to slowdown the power decrease),

high RCS average temperature case, 5% plugging level (less subcooling, smaller mass 

flow).

In the first phase (base case analysis), the loss of external load analysis with the same initial 

and boundary conditions (see Table 1), and assumptions as specified above was analyzed. In

general the initial conditions agree well. Due to different code and geometry, there was some 

difference in the secondary pressure. The aim of this analysis was to compare RELAP5 

calculations using NPP Krško standard input deck against USAR calculation to find possible 

differences in response due to different geometry (known for pressurizer water volume and 

steam generator mass) and the computer code (note that LOFTRAN code is not a best-estimate 

code). The modeling simplifications in LOFTRAN are such that they are conservative for safety 

analysis. The base case scenario was performed with compensation of heat transfer on the 

primary side in order to as closely as possible to reproduce the USAR primary pressure and 

temperature increase, at which DNBR was calculated. In this way the conditions for DNBR

evaluation were similar to the USAR conditions with limitation that local conditions in hot 

channel were not considered in the RELAP5 (local mass flow and local quality).
8

Table 1 Comparison between USAR and RELAP5 initial conditions

Variable USAR RELAP5 remark

Nuclear steam supply system power (MW) 2000 2000

RCS average temperature (K) 580.39 580.39 high Tavg

Minimum measured flow (per loop) m

3

/h 6.151 6.151 5% plugging

Pressurizer pressure (MPa) 15.513 15.513

Pressurizer water volume (m3) 18.2 18.4 correspond to 67% level

(with 5% uncertainty)

Steam generator pressure (MPa) 6.46 6.64

Steam flow (kg/s) 1090 1090

Feedwater temperature (K) 565.55 565.55 constant

Steam generator mass (per SG) (kg) 42210 44350 90% of nominal

In the second phase, the sensitivity analyses were performed to study the influence of 

temperature measurement time delay on OTΔT signal causing reactor trip and DNBR evaluation 

by CHFR. In order to study temperature measurement delay on OTΔT signal, the reactor trip on 

high pressurizer pressure and high pressurizer level were disabled. It should be pointed out that 

such scenarios are not much probable and are not analyzed in the safety analysis reports.

Three different cases were considered in which the temperature measurement delay was varied 

± 2 seconds from the nominal delay: a) case 1 using same initial and boundary conditions as in 

the base case with primary side heat transfer compensation, b) case 2 being same as the case 

1 except that best estimate sprays flow and PORVs flow were considered (i.e. no primary side 

heat transfer compensation), including one pressurizer PORV to be rate sensitive like in the 

plant, and c) case 3 being same as case 2 except that normal measured reactor coolant system 

flow and pressurizer level were considered (6.22 m3/h and 62%). It should be noted that base 

case scenario was compensated for the reason of different codes. Therefore case 2 was 

analyzed having no such compensation. Finally, the case 3 was performed by normal measured 

flow to see its influence on the results.
9

5. RESULTS

5.1 Base Calculation

The results of base case RELAP5 calculation compared to USAR calculation are shown in 

Figures 2 through 11. For DNBR calculation the initial 20 s are important. Figures show that the 

main variables of base case calculation qualitatively and quantitatively agree well with the 

USAR calculation. As the LOFTRAN is conservative code, in the RELAP5 calculation the 

conservatism in the primary side heatup was compensated by reduced cooling of the primary 

side to get quantitative agreement in the initial period of 20 s, when minimum DNBR occurs. 

Similar conservatism existed also for the secondary side heatup due to reduced steam flow, but 

no compensation was made because by steam generator relief valves initially such 

compensation is not feasible. Later, after minimum DNBR occurrence (slightly after reactor trip,

see Figure 2) heat transfer on the secondary side has no more influence on minimum DNBR 

calculation. The heat transfer on the primary side was compensated on the primary side by 

reduced spray and pressurizer PORVs flow to be more conservative in RCS heatup (in the 

opposite no reactor trip on high pressurizer pressure would be generated and the transient 

progression would be different - the reactor would trip on high pressurizer level). In the period 

with no discharge through pressurizer relief valves and spray flow (after 17 s), the compensation 

is no more present and the RELAP5 calculated pressurizer pressure drops more slowly than in 

USAR analysis, as can be seen from Figure 3. Also in the RCS temperature differences occurs 

in the period with no heat transfer compensation. Faster cooling in the RELAP5 case causes 

contraction of coolant in the pressurizer (see Figure 4). The reactor coolant system average 

temperature, the hot leg temperature and the cold leg temperature are shown in Figures 5, 6

and 7, respectively. The rate of temperature increase is similar due to compensation. 

Nevertheless, in the USAR calculation the temperature started to increase earlier and this is 

attributed to the computer code. The reduced pressurizer PORVs flow can be seen from Figure 

8 showing pressurizer relief valves flow. First discharge comes from PORVs with flow rate 

around 38 kg/s, while spike indicates safety valves opening. On the secondary side no heat 

transfer compensation was made, therefore conservative code used in USAR analysis causes 

faster secondary pressure increase, which is shown in Figure 9. The differences in the 

secondary side relief valves flow and closure time (see Figure 10) influence the later period of 

analysis. The focus of the analysis was to reproduce the initial phase until minimum DNBR is 

reached. Finally, Figure 11 comparing DNBR and CHFR shows that LOFTRAN and RELAP5 

trends are quite similar. It should be noted that the values could not be compared directly as 

CHFR was based on RELAP5 lookup table and minimum CHFR limit is not known like minimum 

DNBR limit in USAR calculation. However, the trends may give us some information in the 

sensitivity analysis of temperature measurement delay. Finally, during transient in total around 

20% of DNBR margin was lost in USAR calculation comparing to DNBR margin in normal 

operation.
10

0

500

1000

1500

2000

0 20 40 60 80 100

Time (s)

Power (MW)

USAR

RELAP5

Figure 2 Reactor power - base case comparison between RELAP5 and USAR

12

13

14

15

16

17

0 20 40 60 80 100

Time (s)

p (MPa)

USAR

RELAP5

Figure 3 Pressurizer pressure - base case comparison between RELAP5 and USAR
11

14

17

20

23

26

0 20 40 60 80 100

Time (s)

Volume (m

3

)

USAR

RELAP5

Figure 4 Pressurizer volume - base case comparison between RELAP5 and USAR

570

576

582

588

594

600

0 20 40 60 80 100

Time (s)

Temperature (K)

USAR

RELAP5

RELAP5 (meas.)

Figure 5 RCS average temperature and measured RCS average temperature - base 

case comparison between RELAP5 and USAR
12

570

575

580

585

590

595

600

605

610

615

0 20 40 60 80 100

Time (s)

Temperature (K)

USAR

RELAP5

Figure 6 Hot leg no. 1 temperature - base case comparison between RELAP5 and 

USAR

555

560

565

570

575

580

585

0 20 40 60 80 100

Time (s)

Temperature (K)

USAR

RELAP5

Figure 7 Cold leg no. 1 temperature - base case comparison between RELAP5 and 

USAR
13

0

30

60

90

120

150

0 20 40 60 80 100

Time (s)

Mass flow (kg/s)

USAR

RELAP5

Figure 8 Pressurizer relief valves flow - base case comparison between RELAP5 

and USAR

6.0

7.0

8.0

9.0

0 20 40 60 80 100

Time (s)

p (MPa)

USAR

RELAP5

Figure 9 Steam generator no. 1 pressure - base case comparison between RELAP5 

and USAR
14

0

120

240

360

480

600

0 20 40 60 80 100

Time (s)

Mass flow (kg/s)

USAR

RELAP5

Figure 10 Steam generator no. 1 relief valves mass flow - base case comparison 

between RELAP5 and USAR

1.0

1.5

2.0

2.5

3.0

3.5

4.0

0 5 10 15 20 25

Time (s)

DNBR (-)

USAR

RELAP5

not calculated 

DNBR limit

Figure 11 Departure from nucleate boiling ratio (critical heat flux ratio) - base case 

comparison between RELAP5 and USAR
15

5.2 Sensitivity Calculations

In the Figures 12 through 24 three cases were compared, in which initial and boundary 

conditions were varied while the measurement delay was for the existent RTD (total delay time 

of OTΔT trip signal is less than 6 s assumed in the safety analysis). Then for each case the 

measurement delay time was varied ± 2 seconds (representing the analyses with total delay 

time of OTΔT trip signal of 4 and 8 seconds to be assumed in the safety analysis). The influence 

of temperature measurement delay on parameters used for CHFR calculation, OTΔT trip signal 

and CHFR calculation is shown in Figures 25 through 33, Figures 34 through 42, and Figures 

43 through 51 for case 1, case 2 and case 3, respectively. Finally, Table 2 shows sensitivity of 

OTΔT trip time and minimum CHFR on RCS temperature measurement response time.

5.2.1 Variation of Initial and Boundary Conditions

The reactor power shown in Figure 12 shows, that OTΔT reactor trip signal in the case 1 with 

reduced spray and PORV flow occurs later than in the cases 2 and 3 assuming best-estimate 

spray and PORVs flow. The reason is the higher pressure (see Figure 13) which is less 

conservative for CHFR calculation. The differences in the pressure are due to best estimate 

modeling. In the initial seconds due to higher spray flow the pressure increase is less steep. 

Nevertheless, since in the cases 2 and 3 one pressurizer PORV is modeled as rate sensitive, it 

opens earlier than case 1 PORVs which both open on setpoint pressure 16.2 MPa. For this 

reason the pressure increase in cases 2 and 3 is smaller, besides the best-estimate spray and 

PORVs flow. Pressurizer volume shown in Figure 14 is significantly different for case 3, 

because the uncertainty in the initial pressurizer level was not taken into account. Figure 15

shows the RCS average temperature and its measured signal. It may be seen that temperature 

increases until reactor trip. Later is the reactor trip, higher it is. The RCS average temperature is 

calculated from the hot and cold leg temperature, shown in Figures 16 and 17, respectively.

Before the reactor trip, the differences are negligible. After the reactor trip the differences are 

due to different trip times and persist until hot and cold leg temperatures approach to each 

other. In Figure 18 is shown the pressurizer relief valves flow. It was already mentioned that rate 

sensitive valve open first in cases 2 and 3. The spike in cases 2 and 3 indicates opening of the 

second PORV on setpoint 16.2 MPa. In the case 1 both PORVs open at the same time, 

however their total flow is reduced. In case 1 the spike indicates the opening of the pressurizer 

safety valves. Steam generator no. 1 pressure is shown in Figure 19. It may be seen that the 

shape of cold leg temperature follows the steam generator pressure shape, which indicates that 

heat transfer rate to steam generators is reduced. Due to increasing pressure the steam 

generator safety valves open as can be seen from Figure 20. There are five safety valves per 

steam generator. One may see that after 25 s the valves start to close. It is clearly seen when 

the last two valves close. At maximum pressure four out of five steam generator safety valves 

were open. In Figure 21 is shown the reactor coolant system flow. It may be seen that in cases 

1 and 2 the initial flow is lower than in case 3, because minimum measured flow was assumed. 

As the density started to decrease with increasing RCS temperature, and volumetric flow is 

constant, the mass flow started to decrease too. After the reactor trip the RCS mass flow started 

to increase. Figure 22 shows the equilibrium quality at the hottest axial location of the hot rod in 

the core. Increasing quality decreases the CHFR value. The CHFR trend is shown in Figure 23. 

It may be seen that in the period with high pressure, the CHF is set to zero and therefore CHFR 

is not calculated. Finally, the measured delta T and calculated OTΔT shown in Figure 24 are

used for tripping the reactor. The reactor trip signal is generated when the measured delta T 

value reaches the calculated setpoint for OTΔT. The measured delta T does not change much 

with changing the initial and boundary condition, while calculated OTΔT setpoint is influenced.
16

0

500

1000

1500

2000

2500

0 20 40 60 80 100

Time (s)

Power (MW)

case 1

case 2

case 3

Figure 12 Reactor power – influence of I&B conditions

12

13

14

15

16

17

0 20 40 60 80 100

Time (s)

p (MPa)

case 1

case 2

case 3

Figure 13 Pressurizer pressure - influence of I&B conditions
17

12

15

18

21

24

27

30

0 20 40 60 80 100

Time (s)

Volume (m

3

)

case 1

case 2

case 3

Figure 14 Pressurizer volume - influence of I&B conditions

570

575

580

585

590

595

600

0 20 40 60 80 100

Time (s)

Temperature (K)

case 1

case 2

case 3

case 1 (meas.)

case 2 (meas.)

case 3 (meas.)

Figure 15 RCS average temperature and measured RCS average temperature -

influence of I&B conditions
18

570

580

590

600

610

620

0 20 40 60 80 100

Time (s)

Temperature (K)

case 1

case 2

case 3

Figure 16 Hot leg no. 1 temperature - influence of I&B conditions

560

565

570

575

580

585

0 20 40 60 80 100

Time (s)

Temperature (K)

case 1

case 2

case 3

Figure 17 Cold leg no. 1 temperature - influence of I&B conditions
19

0

20

40

60

80

100

120

0 20 40 60 80 100

Time (s)

Mass flow (kg/s)

case 1

case 2

case 3

Figure 18 Pressurizer relief valves flow - influence of I&B conditions

6.5

7

7.5

8

8.5

9

9.5

0 20 40 60 80 100

Time (s)

p (MPa)

case 1

case 2

case 3

Figure 19 Steam generator no. 1 pressure - influence of I&B conditions
20

0

100

200

300

400

500

600

0 20 40 60 80 100

Time (s)

Mass flow (kg/s)

case 1

case 2

case 3

Figure 20 Steam generator no. 1 relief valves mass flow - influence of I&B 

conditions

8300

8400

8500

8600

8700

8800

8900

9000

0 20 40 60 80 100

Time (s)

Mass flow (kg/s)

case 1

case 2

case 3

Figure 21 Mass flow through hottest axial location in the core - influence of I&B 

conditions
21

-0.25

-0.20

-0.15

-0.10

0 20 40 60 80 100

Time (s)

Quality (-)

case 1

case 2

case 3

Figure 22 Equilibrium quality at the hottest axial location in the core - influence of 

I&B conditions

1

2

3

4

5

0 10 20 30 40

Time (s)

CHFR (-)

case 1

case 2

case 3

not calculated 

later not 

calculated 

Figure 23 Critical heat flux ratio - influence of I&B conditions
22

0

0.5

1

1.5

0 10 20 30 40

Time (s)

OTdT setpoint/dT measured (-)

case 1 (OTdT sp.)

case 2 (OTdT sp.)

case 3 (OTdT sp.)

case 1 (dT meas.)

case 2 (dT meas.)

case 3 (dT meas.)

Figure 24 Overtemperature ΔT setpoint calculation and measured delta T -

influence of I&B conditions
23

5.2.2 Variation of Measurement Delay Time for Case 1

In Figures 25 through 33 are shown the most important variables used for OTΔT setpoint and 

CHFR calculation for different measurement delays. The measurement time delay initially

influences the reactor trip time on OTΔT signal, which in turn influences the other variables, 

used for CHFR calculation. After the reactor trip the pressurizer pressure (see Figure 25) drops 

significantly. Similar is the situation for RCS average temperature shown in Figure 26. Important 

is the fact that the larger the measurement delay is, the larger the temperature increase is. With 

increasing RCS average temperature the mass flow is decreasing due to lower density of the 

coolant as can be seen from Figure 27. Also it can be seen that the mass flow decreases with 

increasing measurement delay. The equilibrium quality (see Figure 28) initially decreases when 

the pressure is increasing and later increases due to dropping pressure. The CHFR shown in 

Figure 29 initially increases due to pressure increase. After pressure drop it started to decrease

almost linearly. This means the later is the trip, the smaller is the minimum CHFR. Finally, the 

temperature and pressure terms for OTΔT setpoint calculation are shown in Figures 30 and 31, 

respectively. The equation for OTΔT setpoints has a constant 1.25 from which the temperature 

term K2 is subtracted and the pressure term K3 is added. This means that in the initial period 

the temperature decreases the setpoint while the pressure increases it. After 10 seconds the 

temperature term K2 becomes dominating. From Figure 32 it can be seen how the OTΔT 

setpoint decrease is delayed proportionally to the measurement delay time. Finally, also 

measured delta T is influenced by measurement delay. Nevertheless, at the time of reactor trip 

the measurement delay is no more influential. The results of sensitivity analysis showed that 

when the temperature measurement delay is rather small (e.g. ± 2 seconds with respect to base 

case), also the reactor trip on OTΔT signal is similarly delayed. The delays impact the curves of 

important plant variables in such a way that the trend direction is continued for delay time 

difference and after the reactor trip the curves are time shifted, when compared to each other. 

As CHFR calculation depends on pressure, mass flux and equilibrium quality, the same is true 

for CHFR as for these parameters.

12

13

14

15

16

17

0 20 40 60 80 100

Time (s)

p (MPa)

4 s

6 s

8 s

Figure 25 Pressurizer pressure - influence of measurement delay for case 1
24

570

575

580

585

590

595

600

0 20 40 60 80 100

Time (s)

Temperature (K)

4 s

6 s

8 s

4 s (meas.)

6 s (meas.)

8 s (meas.)

Figure 26 RCS average temperature and measured RCS average temperature -

influence of measurement delay for case 1

8200

8400

8600

8800

9000

0 20 40 60 80 100

Time (s)

Mass flow (kg/s)

4 s

6 s

8 s

Figure 27 Mass flow through hottest axial location in the core - influence of 

measurement delay for case 1
25

-0.3

-0.2

-0.1

0

0 20 40 60 80 100

Time (s)

Quality (-)

4 s

6 s

8 s

Figure 28 Equilibrium quality at the hottest axial location in the core - influence of 

measurement delay for case 1

1

2

3

4

5

0 10 20 30 40

Time (s)

CHFR (-)

4 s

6 s

8 s

later not calculated 

Figure 29 Critical heat flux ratio - influence of measurement delay for case 1
26

-0.5

0

0.5

1

1.5

0 10 20 30 40 50

Time (s)

K2 term (-)

4 s

6 s

8 s

Figure 30 K2 temperature term calculation - influence of measurement delay for 

case 1

-0.5

0

0.5

1

1.5

0 10 20 30 40 50

Time (s)

K3 term (-)

4 s

6 s

8 s

Figure 31 K3 pressure term calculation - influence of measurement delay for case 1
27

-0.5

0

0.5

1

1.5

0 10 20 30 40 50

Time (s)

OTdT setpoint (-)

4 s

6 s

8 s

Figure 32 Overtemperature ΔT setpoint calculation - influence of measurement 

delay for case 1

-0.5

0

0.5

1

1.5

0 10 20 30 40 50

Time (s)

delta T (-)

4 s

6 s

8 s

Figure 33 Measured delta T - influence of measurement delay for case 1
28

5.2.3 Variation of Measurement Delay Time for Case 2

Figures 34 through 42 show the case 2 calculation, in which primary side heat transfer is not 

compensated like in base calculation. The results show that for case 2 same can be said as for 

case 1. When the temperature measurement delay is rather small (e.g. ± 2 seconds with 

respect to base case), also the trip on OTΔT is similarly delayed. 

To additionally confirm this, the delay +8 seconds with respect to base case is presented. From 

Figure 35 it can be seen that the measured RCS average temperature is time shifted and the 

peak temperature is higher when measurement delay is larger. On the other hand the mass flow 

decreases (see Figure 36) more, while the equilibrium quality (see Figure 37) and RCS average 

temperature (see Figure 36) increase with larger time delay. The results for delay +8 seconds 

(see Figures 38 through 42) also support the statement that the larger the measurement delay

is, the later is the reactor trip on OTΔT and smaller is the minimal CHFR.

12

13

14

15

16

17

0 20 40 60 80 100

Time (s)

p (MPa)

4 s

6 s

8 s

14 s

Figure 34 Pressurizer pressure - influence of measurement delay for case 2
29

570

575

580

585

590

595

600

0 20 40 60 80 100

Time (s)

Temperature (K)

4 s

6 s

8 s

14 s

4 s (meas.)

6 s (meas.)

8 s (meas.)

14 s (meas.)

Figure 35 RCS average temperature and measured RCS average temperature -

influence of measurement delay for case 2

8200

8400

8600

8800

9000

0 20 40 60 80 100

Time (s)

Mass flow (kg/s)

4 s

6 s

8 s

14 s

Figure 36 Mass flow through hottest axial location in the core - influence of 

measurement delay for case 2
30

-0.3

-0.2

-0.1

0

0 20 40 60 80 100

Time (s)

Quality (-)

4 s

6 s

8 s

14 s

Figure 37 Equilibrium quality at the hottest axial location in the core - influence of 

measurement delay for case 2

1

2

3

4

5

0 10 20 30 40 50

Time (s)

CHFR (-)

4 s

6 s

8 s

14 s

later not calculated 

Figure 38 Critical heat flux ratio - influence of measurement delay for case 2
31

-0.5

0

0.5

1

1.5

0 10 20 30 40 50

Time (s)

K2 term (-)

4 s

6 s

8 s

14 s

Figure 39 K2 temperature term calculation - influence of measurement delay for 

case 2

-0.5

0

0.5

1

1.5

0 10 20 30 40 50

Time (s)

K3 term (-)

4 s

6 s

8 s

14 s

Figure 40 K3 pressure term calculation - influence of measurement delay for case 2
32

-0.5

0

0.5

1

1.5

0 10 20 30 40 50

Time (s)

OTdT setpoint (-)

4 s

6 s

8 s

14 s

Figure 41 Overtemperature ΔT setpoint calculation - influence of measurement 

delay for case 2

-0.5

0

0.5

1

1.5

0 10 20 30 40 50

Time (s)

delta T (-)

4 s

6 s

8 s

14 s

Figure 42 Measured delta T - influence of measurement delay for case 2
33

5.2.4 Variation of Measurement Delay Time for Case 3

Figures 43 through 51 show variation of time delay for case 3. Again it can be said that the 

delays impact the curves of important plant variables in such a way that the trend direction is 

continued for delay time difference and after the reactor trip the curves are time shifted, when 

compared to each other. As CHFR calculation depends on pressure, mass flux and equilibrium 

quality, the same is true for CHFR as for these parameters. The larger the measurement delay

is, the later the reactor trip on OTΔT is and the smaller the minimal CHFR is.

12

13

14

15

16

17

0 20 40 60 80 100

Time (s)

p (MPa)

4 s

6 s

8 s

Figure 43 Pressurizer pressure - influence of measurement delay for case 3
34

570

575

580

585

590

595

600

0 20 40 Time (s) 60 80 100

Temperature (K)

4 s

6 s

8 s

4 s (meas.)

6 s (meas.)

8 s (meas.)

Figure 44 RCS average temperature and measured RCS average temperature -

influence of measurement delay for case 3

8200

8400

8600

8800

9000

0 20 40 60 80 100

Time (s)

Mass flow (kg/s)

4 s

6 s

8 s

Figure 45 Mass flow through hottest axial location in the core - influence of 

measurement delay for case 3
35

-0.3

-0.2

-0.1

0

0 20 40 60 80 100

Time (s)

Quality (-)

4 s

6 s

8 s

Figure 46 Equilibrium quality at the hottest axial location in the core - influence of 

measurement delay for case 3

1

2

3

4

5

0 10 20 30 40 50

Time (s)

CHFR (-)

4 s

6 s

8 s

later not calculated 

Figure 47 Critical heat flux ratio - influence of measurement delay for case 3
36

-0.5

0

0.5

1

1.5

0 10 20 30 40 50

Time (s)

K2 term (-)

4 s

6 s

8 s

Figure 48 K2 temperature term calculation - influence of measurement delay for 

case 3

-0.5

0

0.5

1

1.5

0 10 20 30 40 50

Time (s)

K3 term (-)

4 s

6 s

8 s

Figure 49 K3 pressure term calculation - influence of measurement delay for case 3
37

-0.5

0

0.5

1

1.5

0 10 20 30 40 50

Time (s)

OTdT setpoint (-)

4 s

6 s

8 s

Figure 50 Overtemperature ΔT setpoint calculation - influence of measurement 

delay for case 3

-0.5

0

0.5

1

1.5

0 10 20 30 40 50

Time (s)

delta T (-)

4 s

6 s

8 s

Figure 51 Measured delta T - influence of measurement delay for case 3
38

5.2.5 Summary of Results

Table 2 shows sensitivity of OTΔT trip time and minimum CHFR on the temperature 

measurement delay for the three analyzed cases. In addition, the results for base calculation 

are added. It can be seen that due to disabling high pressurizer pressure and level trips (reactor 

trip on turbine trip also not assumed by assumption), in all cases the trip times are later due to 

slower OTΔT protection, causing smaller minimum CHFR than in the base calculation (see 

Section 5.1). Regarding temperature response time it can be seen that the delay in response 

almost proportionally delays the reactor trip time on OTΔT. The larger the delay, the smaller the 

minimum CHFR is. Also it can be seen that in the cases 2 and 3 the primary side cooldown is 

faster due to best estimate pressurizer sprays and PORVs flow. This minimizes primary 

pressure what is more conservative for CHFR calculation. On the other hand, also OTΔT trip 

signal protection responds faster. Finally, Table 2 shows that when comparing cases 2 and 3, 

the case with smaller RCS flow gives smaller minimum CHFR value.

Table 2 Sensitivity of OTΔT trip time and minimum CHFR on RCS temperature 

measurement response time

Base calculation (minimum measured RCS flow, reduced pressurizer sprays and PORVs 

flow)

Variable RELAP5 (6 s)

Time of high pressurizer pressure signal setpoint reached (s) 12.56

Minimum evaluated CHFR 2.28*

Case 1 (minimum measured RCS flow, reduced pressurizer sprays and PORVs flow, high 

pressurizer pressure and level trips disabled)

Variable RELAP5 (4 s) RELAP5 (6 s) RELAP5 (8 s)

Time of OTΔT signal setpoint reached (s) 16.62 18.51 20.28

Minimum evaluated CHFR 2.18 2.13 2.07

Case 2 (minimum measured RCS flow, best estimate pressurizer sprays and PORVs flow, 

high pressurizer pressure and level trips disabled)

Variable RELAP5 (4 s) RELAP5 (6 s) RELAP5 (8 s)

Time of OTΔT signal setpoint reached (s) 15.48 17.22 19.20

Minimum evaluated CHFR 2.13 2.09 2.02

Case 3 (normal RCS flow, best estimate pressurizer sprays and PORVs flow, pressurizer 

level w/o uncertainty, high pressurizer pressure and level trips disabled)

Variable RELAP5 (4 s) RELAP5 (6 s) RELAP5 (8 s)

Time of OTΔT signal setpoint reached (s) 15.50 17.31 19.37

Minimum evaluated CHFR 2.16 2.09 2.04

*

- at nominal power the CHFR value is 2.48
39

5.3 Results Discussion

The results show that the selected loss of load transient with multiple failures of reactor 

protection system trip signals is not the most limiting for the DNB analysis. Based on the CHFR 

calculations the variations in the initial and boundary conditions and the temperature 

measurement delay do not much erode the DNBR margin for loss of external load transient. In 

spite of the fact that CHFR limit was not defined for the RELAP5 lookup tables, the trends 

between USAR and RELAP5 base calculation are so similar, that sensitivity study to show the 

influence of measurement delay time could be done based on the CHFR calculations. Having 

the DNBR trend one may linearly extrapolate the results in the case of larger measurement 

delays than assumed in safety analysis. The RELAP5 calculations for case 2 show that even 

when the delay is 8 seconds larger than assumed delay in USAR, the relations are quite linear.

Therefore it may be suggested that the presented methodology for CHFR could be used also for 

other concerned DNB transients relying on OTΔT protection, when informative safety analysis is 

needed. See Section 1 for explanation why RELAP5/MOD3.3 cannot be used for licensing 

analysis. For Krško NPP the challenging transient would be inadvertent opening of pressurizer 

relief or safety valve in which OTΔT protection is slightly faster than would be the low 

pressurizer pressure trip and the DNB margin is quite small. This means that for a few seconds 

larger measurement delays the low pressurizer pressure protection signal would be generated 

before OTΔT. The pressure trend for such transient is linear decreasing function in the case of 

Krško NPP and also in the case of AP1000 (Ref. 6). For Krško NPP the pressure trend is such 

that in approximately 2 seconds the pressurizer pressure would fall below the low pressurizer 

pressure trip setpoint. Nevertheless, in such cases the minimum DNBR limit would be also 

approached in few seconds. Based on the linear extrapolation of DNBR trend from USAR, up to 

4 seconds are remaining. It should be also noted that the OTΔT protection is provided if the 

transient is slow with respect to piping transient delays from the core to the temperature 

detectors and pressure is within the range between the high and low pressure reactor trips.

41

6. RUN STATISTICS

The calculations with the RELAP5/MOD3.3 Patch 03 computer code were performed on a 

Hewlett-Packard personal computer with Intel Core 2 Quad at 2.40 gigahertz under Microsoft 

Windows XP, Professional Version 2002, Service Pack 3. Table 3 shows the run statistics for 

base calculation. For other calculations the statistic is similar, the difference is less than 10%. 

The differences occur in the period between 10 and 20 s, in which case 1 calculations were the 

most demanding. For all calculations, the number of volumes was 469. The calculations run 

faster than real time. Steady-state calculations for all runs lasted 1,000 seconds and required 

314.0 seconds of central processing unit (CPU) and 31,427 steps.

Table 3 Run statistics

Transient Time

(s)

CPU Time

(s)

CPU/Transient 

Time

Number of Time 

Steps

25 18.52 0.74 1787

200 73.87 0.37 7185

43

7. CONCLUSIONS

The analyses of loss of external load were performed using RELAP5/MOD3.3 Patch 03 

computer code. The base calculation was performed with the same initial and boundary 

conditions and assumptions as in the USAR to the extent possible, to demonstrate that 

RELAP5/MOD3.3 Patch 03 can reproduce the USAR results. Due to non best estimate code 

used in USAR analysis, the compensation in primary side heatup was made in base calculation 

to obtain qualitative and quantitative agreement. This was important, because the RELAP5 is 

limited to calculate the departure from nucleate boiling ratio, which is the safety parameter of 

interest. It was important to compare the CHFR and DNBR trend in calculation, in which other 

important parameters quantitatively agree. The trends of CHFR calculated by RELAP5 and 

DNBR from USAR were similar, while the absolute values were different. 

Similar DNBR and CHFR trends makes meaningful to perform sensitivity analysis. The results 

of sensitivity analysis showed that when the temperature measurement delay is rather small 

(e.g. ± 2 seconds), the OTΔT trip additional delay is similar to additional measurement delay. 

The delays impact the curves of important plant variables in such a way that the trend direction 

is continued for delay time difference and after the reactor trip the curves are time shifted, when 

compared to each other. As CHFR calculation depends on pressure, mass flux and equilibrium 

quality, the same is true for CHFR as for these parameters. The larger the measurement delay

is, the later the reactor trip on OTΔT is and the smaller the minimal CHFR is. The results also 

showed that for small delays the relations are linear. It can be concluded that having the DNBR 

trend from original USAR calculation, one can thus by linear extrapolation of it qualitatively 

assess how small measurement delay time influence the DNBR during loss of external load.

Finally, it should be pointed out that scenarios where three different reactor trips are not 

functioning besides steam dump and we had to rely on the last reactor protection signal is not 

much probable and are not analyzed in the safety analysis reports.

45

8. REFERENCES

1. Ferrouka, M., S. Aissania, F. D’Auria, A. DelNevo, A. Bousbia Salah, ‖Assessment of 12 

CHF prediction methods, for an axially non-uniform heat flux distribution, with the RELAP5 

computer code‖, Nuclear Engineering and Design, 238, 2008, pp. 2718–2725.

2. Parzer, I., B. Krajnc, B. Mavko, ―Analyzing Operator Actions During Loss of AC Power 

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

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

accident in modernized nuclear power plant‖, Electrotechnical Review, Vol. 71, No. 4: 2004.

4. Parzer, I., B. Mavko, and B. Krajnc, ―Simulation of a hypothetical loss-of-feedwater accident 

in a modernized nuclear power plant‖, Journal of Mechanical Engineering, Vol. 49, No. 9: 

2003.

5. Parzer, I., ―Break model comparison in different RELAP5 versions‖, Proc. of International 

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

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

6. Westinghouse AP1000-DCD, Westinghouse AP1000 European Design Control Document, 

EPS-GW-GL-700, Rev. 0.