NUREG/IA-0182

International 

Agreement Report 

Application of 

RELAP5/MOD3.2 to the 

Loss-of-Residual-Heat-Removal 

Event Under Shutdown Condition 

Prepared by 

K.W Seul, Y.S. Bang, H.J. Kim, KINS 

Korea Institute of Nuclear Safety 

PO. Box 114 

Yusung, Taejon 

305-600, Korea 

Office of Nuclear Regulatory Research 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-0001 

April 2000 

Prepared as part of 

The Agreement on Research Participation and Technical Exchange 

under the International Code Application and Maintenance Program (CAMP) 

Published by 

U.S. Nuclear Regulatory Commission
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NUREG/IA-0182

International 

Agreement Report 

Application of 

RELAP5/MOD3.2 to the 

Loss-of-Residual-Heat-Removal 

Event Under Shutdown Condition 

Prepared by 

K.W Seul, Y.S. Bang, H.J. Kim, KINS 

Korea Institute of Nuclear Safety 

PO. Box 114 

Yusung, Taejon 

305-600, Korea 

Office of Nuclear Regulatory Research 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-0001 

April 2000 

Prepared as part of 

The Agreement on Research Participation and Technical Exchange 

under the International Code Application and Maintenance Program (CAMP) 

Published by 

U.S. Nuclear Regulatory Commission
Application of RELAP5IMOD3.2 to the Loss-of-Residual-Heat

Removal Event under Shutdown Condition 

Abstract 

The long-term transient following a loss-of-residual-heat-removal (loss-of-RHR) event 

during reactor shutdown was analyzed to determine the containment closure time (CCT) to 

prevent the release of fission products to environment and the gravity-injection path and rate 

(GIPR) to effectively cool the core. The thermal-hydraulic analysis was carried out using the 

RELAP5/MOD3.2 code and relevant modeling scheme, which were assessed with the LSTF 

experiment in a previous study (NUREG/IA-0143). Based on the plant-specific geometry data 

including various operating conditions, the possible event sequences were identified for the 

Yonggwang Units 3&4 plant (YGN 3/4), which is CE-typed PWR of 2,815 MW thermal 

power in Korea. As a result, the real plant simulation gives the similar calculation 

characteristics to the previous LSTF simulation, and then it was found that the 

RELAP5/MOD3.2 code is capable of appropriately simulating the loss-of-RHR event of the 

real plant. 

From the five cases of the CCT analyses, it was estimated that the containment closure 

should be achieved within about 40 minutes to prevent the release of fission products in the 

large cold-leg opening case under the worst event sequence. However, it was also found that 

the first core uncovery could occur in the early phase of the event by the loop seal clearing 

phenomenon in the crossover leg. From the six cases of the GIPR analyses, it was revealed 

that the system was well depressurized and the core boiling was successfully prevented by the 

gravity-injection in cases with the injection point and opening on the different leg side. 

However, it was also found that the gravity-injection process could be ineffective in the case 

of relatively high pressurizer-manway opening because of the water holdup phenomena in the 

pressurizer. Also, it was estimated that about 54 kg/s of minimum injection rate was required 

to maintain core cooling and the core cooling could be provided for about 10.6 hours with the 

nominal water capacity of refueling water storage tank (RWST). 

These results will provide useful information to operators to cope with the event. And, to 

apply them to the emergency and recovery procedures against the event, additional case 

studies will be needed for wide range of operating conditions such as reactor coolant system 

inventory, RWST water temperature, and core decay heat rate.

iii
Contents

Abstract ......... 

Contents ........ 

List of Tables ..... 

List of Figures .... 

Executive Summary 

Acknowledgments. 

Nomenclature ....

I. Introduction .................... 

II. Description of Plant Configurations 

II.A. Possible Event Sequences .... 

II.B. Plant Configurations ........ 

II.C. Analysis Cases ............. 

II.C.1 The CCT Analysis .... 

II.C.2. The GIPR Analysis .. 

III. Description of Analysis Model .....

iii 

vii 

vii 

ix 

xi 

xnl 

.1 

.5 

.5 

.8 

11 

11 

11 

15

IV. Analysis Results and Discussion ............................................ 19 

IV.A. The CCT Analysis Results ........................................... 19 

IV.A. 1 Analysis Results for the Hot-leg Side Opening ...................... 19 

IV.A.2. Analysis Results for the Cold-leg Side Opening ..................... 25 

IV.A.3. Discussion on the Containment Closure Time (CCT) ................. 29 

IV .B. The GIPR Analysis Results ........................................... 31 

IV.B. 1 Analysis Results for the Hot-leg Injection Cases ..................... 31 

IV.B.2. Analysis Results for the Cold-leg Injection Cases ................... 37 

IV.B.3. Discussion on the Gravity-Injection Path and Rate (GIPR) ............ 40 

IV .C . Run Statistics ...................................................... 43 

V . Conclusions ........................................................... 47 

R eferences ............................................................. 49 

Appendix A Input Decks for the Steady State Calculations ....................... 51 

Appendix B Input Decks for the Transient Calculations .......................... 53
List of Tables

Table 1. Potential PCS openings during shutdown operation of YGN 3/4 .......................... 8 

Table 2. Major elevation and reference point of YGN 3/4 ...................................... 9 

Table 3. Initial conditions for transient analysis ............................................. 18 

Table 4. List of the major parameters represented in the figures ................................. 18 

Table 5. Results of transient analyses for the YGN 3/4 ....................................... 29 

List of Figures 

Fig. 1. Possible event sequences following a loss-of-RHR event ................................. 7 

Fig. 2. RCS configurations and gravity-injection paths in the YGN 3/4 .......................... 13 

Fig. 3. RELAP5 nodalization for simulation of a Loss-of-RHR event in YGN 3/4 plants ............ 17 

Fig. 4. Pressure behavior in the upper plenum (cases 1,2 and 3) ................................ 21 

Fig. 5. Total heat trnsfer through the intact-loop SG (case I and 2) ............................. 21 

Fig. 6. Discharging flow rate through the opening (cases 1, 2 and 3) ............................ 22 

Fig. 7. Collapsed water level in the pressurizer (cases I and 2) ................................. 22 

Fig. 8. Water temperature in the hot and cold legs (cases 2) ................................... 23 

Fig. 9. Void fraction in the hot and cold legs (case 2) ......................................... 23 

Fig. 10. Collapsed water level in the RPV (cases 1,2 and 3) ................................... 24 

Fig. 11. Fuel cladding temperature (cases 1, 2 and 3) ......................................... 24 

Fig. 12. Pressure behavior in the upper plenum (cases 4 and 5) ................................. 26 

Fig. 13. Collapsed water level in the RPV (cases 4 and 5) ..................................... 26 

Fig. 14. Void fraction in the hot and cold legs (case 4) ........................................ 27 

Fig. 15. Pressure difference in the crossover leg (case 4) ...................................... 27 

Fig. 16. Discharging flow rate through the opening (cases 4 and 5) .............................. 28 

Fig. 17. Fuel cladding temperature (cases 4 and 5) ........................................... 28 

Fig. 18. Pressure behavior in the upper plenum (cases A, B and C) .............................. 33 

Fig. 19. Mass flow rate from the RWST (cases A, B and C) ................................... 33 

Fig. 20. Collapsed water level in the pressurizer (cases A) ..................................... 34 

Fig. 21. Discharging flow rate through the opening (cases AB and C) ...................... 34 

Fig. 22. Void fraction in the hot and cold legs (case A) ........................................ 35 

Fig. 23. Water temperature above the core region (cases A, B and C) ........................... 35 

Fig. 24. Collapsed water level in the RPV (cases A, B and C) .................................. 36

vii
Fig. 25. Pressure behavior in the upper plenum (cases D, E and F) .............................. 38 

Fig. 26. Water temperature above the core region (cases D, E and F) ............................ 38 

Fig. 27. Mass flow rate from the RWST (cases D, E and F) ................................... 39 

Fig. 28. Collapsed water level in the RPV (cases D, E and F) ................................... 39 

Fig. 29. Mass flow rate from the RWST (sensitivity) ......................................... 41 

Fig. 30. Discharging flow rate through the opening (sensitivity) ................................ 41 

Fig. 31. Water temperature above the core region (sensitivity) .................................. 42 

Fig. 32. Comparison of CPU time ........................................................ 45 

Fig. 33. Estim ated mass error ............................................................ 45 

Fig. 34. Advanced time step (case A) ...................................................... 46

viii
Executive Summary

To investigate the mitigation measures following a loss-of-RHR event during reactor 

shutdown, the plant operating conditions of the Yonggwang Units 3 and 4 (YGN 3/4), PWR 

type of 2,815 MW thermal power in Korea, were reviewed. The possible event sequences 

were identified and the long-term transient analyses were performed using RELAP5iMOD3.2 

code. This analysis is finally to determine the containment closure time (CCT) to prevent the 

release of fission products to environment and the gravity-injection path and rate (GIPR) to 

effectively cool the core after event. To do this, based on the combination of the RCS opening 

and the SG secondary water level condition, the five cases of typical RCS configurations were 

identified for the CCT analysis as follows: 

Case 1. a pressurizer-manway-opening (PMO) case with water-filled SG 

Case 2. a pressurizer-manway-opening (PMO) case with emptied SG 

Case 3. a SG-inlet-plenum-manway-opening (SMO) case with emptied SG 

Case 4. a small cold-leg-opening (CLO) case with emptied SG 

Case 5. a large cold-leg-opening (CLO) case with emptied SG 

Also, based on two available gravity-injection lines from the cold water of refuel water 

storage tank (RWST) and three of large RCS openings as a RCS drain path, the six cases of 

the injection paths were identified for the GIPR analysis as follows: 

Case A. a hot-leg injection and a PMO discharge 

Case B. a hot-leg injection and a SMO discharge 

Case C. a hot-leg injection and a small CLO discharge 

Case D. a cold-leg injection and a PMO discharge 

Case E. a cold-leg injection and a SMO discharge 

Case F. a cold-leg injection and a large CLO discharge 

The applicability of the code to the loss-of-RHR event under shutdown conditions was 

assessed in a previous study (NUREG/IA-0143), which was based on the comparison of the 

calculation with the experiment simulating the event during mid-loop operation in the Large 

Scale Test Facility (LSTF). The same code and consistent modeling scheme were used in this 

calculation. The major findings from the long-term transient analysis of the loss-of-RHR 

event for the real nuclear power plant were as follows: 

(1) The real plant simulation of the YGN 3/4 plant gives the similar calculation 

characteristics to the LSTF simulation in the required CPU time, computational time step, and 

system mass error. Thus, it was found that the RELAP5/MOD3.2 code was capable of 

appropriately simulating the loss-of-RHR event of the real plant.

ix
(2) From the CCT analysis for the five cases of typical RCS configurations with no RCS 

makeup and unavailable secondary cooling, the time to boil after event was estimated to be 

about 10 to 13 minutes regardless of the opening size and location. Also, the time to core 

uncovery was estimated to be about 40 to 183 minutes depending on the elevation and size of 

the opening and the SG secondary water level condition. Particularly, in case with the water

filled SG, it was delayed about 100 minutes by the reflux condensation on SG U-tubes, as 

compared to the emptied SG case. However, it indicated that the first core uncovery could 

occur in the early phase of the event by the loop seal clearing phenomenon in the crossover 

leg for the cold-leg opening case. As a result, it was found that the earliest CCT was 40 

minutes after event for the SG-inlet-plenum-manway opening or the large cold-leg opening 

cases. Beside, the containment closure is needed to initiate before the boiling time because the 

discharge via the opening after boiling could threaten the workers in the containment. 

(3) From the GIPR analysis for the six possible gravity-injection paths from the RWST, the 

following conclusions were obtained. In cases with the PMO, where was located at higher 

elevation than the RWST water level, the system pressure continued increasing due to the 

water holdup phenomenon in the pressurizer and then the core was uncovered at about 96.6 

minutes despite of the gravity-injection process. In cases with the injection point and opening 

on the same leg side, the core cooling was dependent on the core flow. Particularly, in the 

cases with the injection point and opening on the different leg side, the RCS was well 

depressurized and the core boiling was successfully prevented for a long-term transient. As a 

result, those injection paths were estimated to be the most suitable paths in avoiding core 

boiling for a long-term period. In addition, about 54 kg/s of minimum injection rate was 

required to maintain core cooling. Such an injection rate was capable of providing the core 

cooling for about 10.6 hours if 70% of the RWST water was available. 

(4) These results will provide useful information to operators to understand the plant 

behavior and to cope with the event in timely manners. And, to apply them to the emergency 

and recovery procedures against the loss-of-RHR event, additional case studies will be needed 

in the future for wide range of operating conditions such as RCS inventory, RWST water 

temperature, and core decay heat rate.

x
Acknowledgements

This study was carried out under the sponsorship of Ministry of Science and Technology 

(MOST) in Korea. The President, S. J. Kim in the Korea Institute of Nuclear Safety (KINS), 

contributed significantly to the administration of this project. Authors express an appreciation 

to him and related staff. And also, authors appreciate to Dr. F. Eltawila and related staff of the 

U.S. Nuclear Regulatory Commission who manage the CAMP project.

xi
Nomenclature

CCFL Counter Current Flow Limitation 

CCT Containment Closure Time 

CLO Cold-leg Opening 

ECCS Emergency Core Cooling System 

GIPR Gravity-Injection Path and Rate 

LSC Loop Seal Clearing 

LSTF Large Scale Test Facility 

NOO Non-Opening 

NPP Nuclear Power Plant 

PMO Pressurizer Manway Opening 

PWR Pressurized Water Reactor 

Pzr Pressurizer 

RCP Reactor Coolant Pump 

RCS Reactor Coolant System 

RHRS Residual Heat Removal System 

RPV Reactor Pressure Vessel 

RWST Refueling Water Storage Tank 

SG Steam Generator 

SMO SG-inlet-plenum Manway Opening 

YGN 3/4 Yonggwang Units 3 and 4 Plants

xiii
I. INTRODUCTION

The residual heat removal (RHR) system is used to remove a core decay heat under 

shutdown condition or refueling stage in pressurized water reactor (PWR). It is operated, in 

some occasions, with the reactor inventory reduced to a mid-water level of the primary loop, 

which is called mid-loop operation, for a maintenance or inspection of components such as 

steam generator U-tubes and reactor coolant pumps during a plant outage. Recently, the loss

of-RHR event during the mid-loop operation was of great concern, since there have occurred 

many events associated with it and the potential for the significant risk has been identified. 

During shutdown operation or refueling, three initiators are considered as major causes of 

the loss-of-RHR event: a loss of reactor coolant system (RCS) inventory, a loss of RHR flow, 

and a loss of support systems. The loss of RCS inventory could be caused by overdrain while 

going to reduce the RCS inventory or by failure to maintain a water level during reduced 

inventory such as mid-loop operation. It would eventually lead to a loss of RHR flow due to 

the reduction in RHR pump suction head. A loss of offsite power or a failure of RHR heat 

exchanger or pump would also cause a failure of the RHR system. Recently, the loss-of-RHR 

events have been experienced several times in PWR plant [1, 2]. The causes of the events 

were found mainly to be a failure of RHR pump or a loss of vital ac power [3]. Although the 

plant was recovered within a proper time after event, the continued recurrence of the events 

raised the issues on the reliability of the RHR system and the importance of the plant recovery 

measures. 

In order to cope with the event in a timely manner, abnormal operating procedure should 

be prepared based on the plant responses to the event. Generally, it includes two types of 

actions to mitigate the event. One is to protect the personnel working in the containment and 

to prevent an uncontrolled release of fission products to atmosphere, which includes an 

evacuation of nonessential personnel from the containment and a closure of the containment 

openings. After the loss of the RHR flinction, the coolant heat-up would lead to core boil-off 

and core damage. In particular, for the case that the RCS is open, the RCS coolant could be 

discharged into containment and threaten the personnel working in the containment. In 

addition, if there is. containment opening, such as personnel or equipment hatches, an 

uncontrolled release of fission products to environment would be possible in the early phase 

of the event. Thus, the timing to take actions for the containment closure is important in 

mitigating the event and it should be determined from the detailed transient analysis under 

various plant conditions and operating states.

1
The other is to restore the removal capability of the core decay heat, which includes a 

recovery of the RHR system and/or an alignment of the water-feed line for an alternate core 

cooling. Naff, et al. [4] investigated the important thermal-hydraulic processes following the 

event during reduced inventory operation and discussed the recovery measures. Particularly, 

they analyzed two types of alternate cooling methods for the decay heat removal in the 

absence of the RHR system. One is a reflux condensation cooling in a closed RCS using 

steam generators (SG) as a heat sink and the other is a gravity-injection cooling using water of 

the refueling water storage tank (RWST). They concluded that the condensation cooling was a 

viable strategy to maintain core cooling after event, however the integrity of temporary RCS 

closures such as nozzle dams could be threatened due to the high system pressure. Meanwhile, 

they reported that the gravity-injection cooling could be an effective measure to maintain core 

cooling under the open RCS conditions. In practice, the RCS has various openings for 

maintenance during plant outages. Also, the gravity-injection processes are very different from 

plant to plant. Thus, an appropriate injection path and rate to mitigate the event should be 

determined from the detailed transient analysis based on the plant-specific conditions and 

configurations. 

Parrish and Till [5] analyzed the loss-of-RHR event for the Palo Verde Nuclear Generation 

Station to investigate the plant recovery measures. They used the RETRAN code and a very 

simple nodalization to simulate the event, and also used a hand calculation to reduce the 

computational time. The results of the simulation indicated that the power station could 

undergo the event without experiencing a rapid core uncovery due to water ejection through 

opening. However, due to the simple models and conservative assumptions, they could not 

simulate some dominant thermal-hydraulic phenomena, such as a loop seal clearing in the 

crossover leg, a liquid holdup in the pressurizer, and non-condensable gas effect. Also, the 

long-term behavior up to the core heat-up could not be simulated due to the calculation 

limitation. Hassan and Raja [6] also performed the transient analysis of typical four-loop 

PWR using the RELAP5/MOD3 code for two cases, non-opening and pressurizer-vent 

opening cases. They reported that the fuel cladding temperature was below the accepted safety 

limits throughout one hour of transient. However, depending on the plant configurations such 

as the size and location of the RCS opening and secondary water condition, the core could be 

damaged even in the early phase of transient. In addition, the core cooling capability will be 

significantly dependent on the RCS water makeup process. 

The present report describes two long-term transient analyses after the loss-of-RHR event 

for the various real plant configurations, the containment closure time (CCT) analysis and the

2
gravity-injection path and rate (GIPR) analysis. The CCT analysis is to determine the 

containment closure time to prevent an uncontrolled release of fission products to atmosphere 

under the situation that the recovery of the decay heat removal is delayed for a long-term 

period. The GIPR analysis is to determine the gravity-injection path for an effective core 

cooling and the injection rate needed to prevent core boiling under the situation that the RCS 

makeup is available in a proper time. In general, the plant responses to the event are strongly 

dependent on the operating states and the plant configurations. Thus, the real plant conditions 

of the Yonggwang Units 3 and 4 (YGN 3/4) are reviewed to identify the possible event 

sequences. The plant is CE-typed PWR plant of 2,815 MW thermal power begun commercial 

operation in 1995 in Korea. Particularly, the location of RCS openings and the water-injection 

paths are investigated. The transient analyses for the identified event sequences are carried out 

using the best-estimate system transient analysis code, the standard version of 

RELAP5/MOD3.2, with detailed simulation of the real plant. Based on the calculation results, 

the applicability of the code to the event of the real plant is also discussed.

3
II. DESCRIPTIONS OF PLANT CONFIGURATIONS

II.A. Possible Event Sequences 

When the loss-of-RHR occurs as an initiating event during shutdown operation, various 

scenarios following the event are possible according to the operating states and plant geometry 

conditions. Particularly, during the plant outages, the RCS could be in different operating 

states from the full power operation, and the major safety systems could be in an inoperable 

state. Thus, the possible and prominent event sequences could be identified depending on the 

mitigation measures against the event [7]. Based on these event scenarios, nine possible event 

sequences could be derived in the YGN 3/4 plant as shown in Fig. 1. 

If the RHR function is recovered quickly after the event, the core decay heat would be 

successfully removed, and the plant would reach a safe condition (sequence 1). But, if the 

recovery of the RHR system is delayed for a long time, the plant behavior is generally divided 

into two main paths, one with an open RCS and the other with a closed RCS. If the RCS is 

open, the secondary cooling cannot be provided, but a bleed path can be established. Thus, if 

the RCS inventory makeup is available for a long term, the core decay heat would be 

successfully removed (sequence 4) and, if not available, the core would be damaged (sequence 

2). Also, if the long-term recirculation in the RCS is not available even with inventory 

makeup, the core could be challenged (sequence 3). In the case of the closed RCS, if the 

secondary cooling with available SGs (sequence 5) or the RCS inventory makeup and long

term recirculation (sequence 9) is available, then the core decay heat would be also removed. 

However, with the secondary cooling not available due to an installation of nozzle dams or a 

maintenance of auxiliary feedwater system, if either the water feeding into the RCS or the 

steam bleeding outside the RCS is inoperable, the core would be uncovered and damaged 

(sequences 6, 7, and 8). As a result, Fig. 1 shows that the core could be damaged for the five 

event sequences out of nine. 

In general, the closed RCS provides an additional barrier of fission products and a longer 

time for the personnel to work in the containment after event. If the RCS is open, the fission 

products would be released into the containment through the opening and then jeopardize the 

personnel working in the containment. In addition, if the containment is open, the radiological 

materials would be released to atmosphere. Therefore, the event with the RCS openings may 

result in serious consequences. In the present study, the sequence 2 is selected as the worst 

event sequence for the CCT analysis because the core damage and the containment challenge

5
are expected to take place in the earlier phase of the transient. There is no RCS inventory 

makeup and SG secondary cooling by natural circulation throughout the transient. Also, the 

sequences 3 and 4 are selected for the GIPR analysis to evaluate the core cooling capability 

after event. In these cases, the RCS inventory makeup can be provided with two types of 

water-feeding means, i.e., pump and gravity feeding from the water sources such as the 

RWST. In general, the forced-injection process using the active pump requires the off-site 

power supply, and if it is possible, it is expected to sufficiently make up the coolant inventory 

and maintain the core cooling after the event. Thus, the gravity-injection process, which 

possible without an off-site power, is evaluated in this study as an alternate core cooling 

measures.

6
a! SYMBOLS 

i~f-• :Successful Heat 

Removal 

* : Core Damage 

----- ---- ----- -------- --------

Fig. 1. Possible event sequences following a loss-of-RHR event

7
ll.B. Plant Configurations

During plant outages, the RCS has various openings. For instance, a pressurizer manway is 

open for the RCS coolant drain, SG manways are open for the SG U-tube inspections after 

installing nozzle dams, and a reactor coolant pump (RCP) seal or impeller assembly can be 

removed for a repair. Also, the drain or vent paths out of the RCS, such as pressurizer safety 

valves or reactor vessel head vents, could be open during the water-feeding processes. 

In general, the location and size of the RCS openings are known to significantly affect the 

plant behavior and the thermal-hydraulic processes after event [4]. Table 1 represents the 

potential RCS openings during shutdown operation and refueling of the YGN 3/4 plant. Table 

2 shows the elevation of the openings and reference height of the major location. Among 

those RCS openings, the largest opening size is the manways in the SG plenum and the top of 

the pressurizer, that is 16 inches inner diameter, except the opening of the reactor vessel head

off. The cold-leg opening size is assumed 5 to 30% of the cold-leg cross sectional area while 

the RCP seal or impeller assembly is removed. The highest elevation of the opening is about 

17.6 m of the pressurizer manway above the centerline of the hot leg, and the lowest opening 

is on the cold leg. 

Table 1. Potential RCS openings during shutdown operation of YGN 3/4 

RCS Openings Diameter (in.) Remarks 

"* Pressurizer manway 16 - for RCS coolant draining 

"* Primary SG inlet plenum manways 16 - for SG U-tube inspection 

"* Primary SG outlet plenum manways 16 - for SG U-tube inspection 

"* Pressurizer safety relief valves 6 - for maintenance 

"• Vessel upper head vent 3/4 - for venting 

"• Pressurizer vent line 1 - for venting 

"* Reactor vessel head off - - for refueling 

"* RCP seal or impeller 5%-30% of - while repaired 

cold leg area

8
Table 2. Major elevation and reference point of YGN 3/4

Elevation Reference Point Remark 

* 161 ft 3 in - Top of the pressurizer - Pressurizer manway 

* 137 ft - Top of the RWST 

* 129 ft - 10 % of the pressurizer water level 

* 116 ft 3 in - Flange of the reactor pressure vessel 

* 109 ft 3.5 in - Inlet of SG U-tube - SG inlet/outlet plenum 

* 105 ft I in - Top of the hot-leg manways 

* 103 ft 4 in - Centerline of the hot and cold legs - Cold-leg opening 

Besides, two sources of cold water may be available for the RCS makeup during the plant 

outages; the accumulators and the RWST. The gravity-injection from the accumulators is not 

a practical method because it is difficult to manually control the injection flow, even they 

could be pressurized by the gas. Meanwhile, the gravity-injection from the RWST could be an 

effective measure if there is a net positive differential pressure between the RWST and the 

RCS. During mid-loop operation, the RWST water level is generally higher than the RCS 

water level, resulting in providing the net positive elevation head. However, because the 

injection paths are long and complex, and fitted with various components such as flow

orifices, valves, pumps, or heat exchangers, it is important to identify the possible and 

effective paths for the gravity injection. The YGN 3/4 plants have multiple injection paths 

from the RWST to the RCS as follows: 

"* The cold-leg injection path through a high pressure safety injection system 

"* The cold-leg injection path through a charging and letdown system 

"* The cold-leg injection path through a RHR system 

"* The hot-leg injection path through RHR suction lines, etc. 

Among those flow paths, the hot-leg injection path via the RHR suction lines has relatively 

low hydraulic resistance because there are no pumps and a few numbers of valves on the flow 

path.

9
II.C. Analysis Cases

11. C. 1. The Containment Closure Time (CCT) Analysis 

For the CCT analysis, three locations of the RCS openings are selected based on the above 

typical plant configurations: pressurizer-manway opening (PMO), SG-inlet-plenum-manway 

opening (SMO), and cold-leg opening (CLO). When there are these openings during the 

event, it is expected that the reactor coolant would be discharged into the containment via the 

openings from the early phase of the event. In addition, the secondary water level condition 

will affect the thermal-hydraulic process in the RCS. Thus, based on the combination of the 

RCS openings and the SG secondary water level condition, the five cases of typical RCS 

configurations are identified to analyze in detail the plant behavior. 

1. a pressurizer-manway-opening (PMO) case with water-filled SGs (Case 1) 

2. a pressurizer-manway-opening (PMO) case with emptied SGs (Case 2) 

3. a SG-inlet-plenum-manway-opening (SMO) case with emptied SGs (Case 3) 

4. a small cold-leg-opening (CLO) case with emptied SGs (Case 4) 

5. a large cold-leg-opening (CLO) case with emptied SGs (Case 5) 

The CCT could be determined based on the time to boil, the time to core uncovery, and the 

time to core heatup after event obtained from the detailed transient analyses. The time to boil 

is defined as the time for the water in the reactor vessel upper head to reach a saturation 

temperature under atmospheric pressure. The time to core uncovery is defined as the time for 

the collapsed water level to be below the top of the core. The time to core heatup is defined as 

the time for the fuel surface temperature to begin to rapidly increase. Generally, to prevent the 

release of fission products to environment, the containment closure needs to be initiated 

before the time to boil and completed before the time to core uncovery. 

. C.C2. The Gravity-Injection Path and Rate (GIPR) Analysis 

For the GIPR analysis, the six cases of the gravity-injection paths are identified to evaluate 

the core cooling capability after event. It is based on two available gravity-injection lines, the 

cold-leg and the hot-leg injections, and three of large RCS openings as a RCS drain path, 

which are the same openings as the CCT analysis. The six cases of the identified injection 

paths are as follows;

11
1. a hot-leg injection and a PMO discharge (Case A) 

2. a hot-leg injection and a SMO discharge (Case B) 

3. a hot-leg injection and a small CLO discharge (Case C) 

4. a cold-leg injection and a PMO discharge (Case D) 

5. a cold-leg injection and a SMO discharge (Case E) 

6. a cold-leg injection and a large CLO discharge (Case F) 

Especially, in the YGN 3/4 plant design, the total water volume of the RWST as a water 

source for core cooling is 2978 m3 [8]. When 70% of the RWST water is available during 

plant outages, the water level of the RWST is about 7.0 m above the hot-leg centerline. The 

pressure and the water temperature in the RWST are assumed to be atmospheric and 307 K, 

respectively. 

The diameter of the pipe from the RWST to the RCS injection point is assumed 25.4 cm, 

based on the pipe diameter of safety injection system. In practice, the pipe size varies 

depending on the flow path, and also the injection flow is constrained by a hydraulic 

resistance on the flow path. Thus, the sensitivity on the gravity-injection flow rate is studied 

and discussed in Section IV.B.3. Figure 2 represents the possible gravity-injection paths and 

the locations of the RCS openings in the YGN 3/4 plant.

12
Pressurizer 

manway 161.3 ft

RPV

Fig. 2. RCS configurations and gravity-injection paths in the YGN 3/4

13
III. DESCRIPTIONS OF ANALYSIS MODEL

To analyze the thermal-hydraulic behavior following the event, the system transient 

analysis code, the standard version of RELAP5/MOD3.2 released by the U.S. Nuclear 

Regulatory Commission (NRC) [9], is used. The code is run on a DEC 5000/240 workstation. 

The applicability of the code to the loss-of-RHR event under shutdown conditions was 

assessed in a previous study [10, 11]. The assessment was based on the comparison of the 

calculation with a Rig of Safety Assessment-IV/Large Scale Test Facility (ROSA-IV/LSTF) 

experiment simulating the event during mid-loop operation [12]. It revealed that the code was 

capable of simulating appropriately the major thermal-hydraulic processes following the 

event, including the coolant boiling, the system pressurization, the steam condensation on SG 

U-tube wall, the loop seal clearing in the crossover leg, the water holdup in the pressurizer, 

and the non-condensable gas behavior. Also, Banerjee et al. [13] reported that the code gave a 

good qualitative agreement with the same experiment data. Recently, Femg and Lee [14] also 

indicated that the code and models behaved well in capturing the overall system responses to 

the event through the assessment of the RELAP5 code with Taiwan's INER Integral System 

Test (IIST) data for the loss-of-RHR event. 

However, there have been many difficulties in getting convergence of the transient 

calculation following the loss-of-RHR event during the mid-loop operation. Particularly, it 

was difficult to calculate the transport process of the mixture including non-condensable gas 

under the low flow and low pressure conditions. When the non-condensable gas enters a 

hydrodynamic volume filled with steam, an extremely small size of time step was required 

and a long CPU time was consumed. Recently, Seul. et al. [10] reported that the 

RELAP5/MOD3.2 version, which incoporated new models and improvements related to the 

analysis of the loss-of-RHR event, was capable of simulating the event with an appropriate 

time step and CPU time. They also indicated that the long-term behavior of the transient 

including the non-condensable gas behavior could be reasonably predicted. 

In this calculation, the same code and consistent models as the previous study are used 

except the difference of the plant geometrical conditions. The nodalization for simulation of 

the plant is consisted of~- 236 hydrodynamic volumes connected by - 263 junctions and - 228 

heat structures, as shown in Fig. 3. The RPV elements (volumes 100 to 130) include the 

volumes corresponding to the downcomer, the lower plenum and upper plenum, the core, the 

upper head, and the guide thimble channel. The core is modeled as two-channel core with 12 

volumes and heat structures per each channel connected by crossflow junctions. This

15
arrangement is adopted to compensate the multi-dimensional effect such as an natural 

circulation flow in the core region. Also, to simulate the steam bypass flow from the upper 

plenum to the upper head, the guide thimble tube is modeled. 

The two loops are consisted of an intact loop (volumes 200 to 299) and a broken loop 

(volumes 400 to 499) in a nearly symmetrical way. Each loop is composed of a hot leg, SG 

inlet and outlet plena, a SG U-tube, two crossover legs and RCPs, and two cold legs. 

Especially, the crossover legs are modeled as 9 nodes to accurately simulate the loop seal 

clearing phenomena. The pressurizer is connected to the hot leg in intact loop through the 

surge line elements. The secondary sides of two SGs (volumes 300 to 399 and 500 to 599) are 

simulated with a downcomer, boiling section, a steam separator and a steam dome. The RHR 

system is modeled by time dependent volumes and junctions connected to the hot leg and the 

cold leg in both loops. The RHR lines are needed to calculate the steady-state conditions 

during the mid-loop operation of the plant. Basically, in the current YGN 3/4 transient 

analysis, the same options related to the volume and junction and the same special models 

such as the CCFL model, the cross flow model and the control logic including the time step 

control are used as the previous LSTF assessment [ 10]. 

The initial conditions used in the calculation are represented in Table 3. They are nearly 

same for the CCT analysis and GIPR analysis. The decay heat rate depending on the time after 

reactor shutdown is conservatively assumed to remain 0.5% of full power throughout the 

transient. The RCS water level is assumed to be in mid-level of the hot leg. The pressure in 

the primary and secondary systems remains atmosphere, and the gas space is filled with non

condensable gas of air. The steady state conditions are obtained from new transient run up to 

1000 s and the loss-of-RHR events are initiated by isolating the RHR flow and opening the 

PMO, the SMO, or the CLO. In the GIPR transient analysis, the gravity-injection from the 

RWST is assumed to begin at 20 minutes after event, based on the typical operator action 

time. 

The input decks for the steady state and transient calculations of the event in the YGN 3/4 

plants are attached to this report as an Appendix A and B. In the following section, the 

analysis results are discussed separately for the CCT and GIPR calculations. Table 4 indicates 

the list of major parameters represented in the figures of this report.

16
J590

J915 J645

Pressurizer Manway

RWST

311

RHR

RHR

Broken Loop

Cold leg Opening

Fig. 3. RELAP5 nodalization for simulation of a Loss-of-RHR event in YGN 3/4 plants

Downcomer 

Feedwater

I.-. J417

Downcomer 

Feedwater

Intact Loop

J390
Table 3. Initial conditions for transient analysis

Major Parameters

"* Core power (MW thermal) 

"* Primary and secondary pressures 

"* Hot leg and cold leg water temperature (K) 

"* SG secondary water temperature (K) 

"* Water level in primary side 

"* Water level in SG secondary side (m) 

"* Initial mass inventory (kg) 

"* Non-condensable gas 

"* RWST water level 

"* RWST water temperature (K) 

"* PMO and SMO area (M2) 

" 5 % and 30 % of CLO area (m2

)

4- Simulation Conditions

* 14.125 (0.5 % of full power) 

* Atmospheric 

* 327.6 and 313.1 

o313.1 

"* Mid-level of loop 

" empty or 11.0 

* 104618 

* Air 

* 70 % of full height 

* 307.0 

*0.13 

* 0.0228 and 0.1368

Table 4. List of the major parameters represented in the figures

* DP: Differential pressure, BL: Broken loop, IL: Intact loop

18

Major Parameters Figures Calculated Parameters 

"* Pressure at upper plenum Fig.4, 12, 18, 25 p-120010000 

"* DP at crossover leg downside-BL Fig. 15 cntrlvar-755 

"* DP at crossover leg upside-BL Fig. 15 cntrlvar-753 

"* Water temperature at hot leg-IL Fig.8 tempf-200010000 

"* Water temperature at cold leg-IL Fig.8 tempf-253030000 

"* Water temperature at upper plenum Fig.23, 26, 31 tempf-120010000 

"* Fuel cladding temperature at top Fig. 11, 17 httemp-001001201 

"* Collapsed water level in RPV Fig. 10, 13, 24, 28 cntrlvar 125 

"* Collapsed water level in pressurizer Fig.7, 20 cntrlvar 126 

"* Flow rate through opening Fig.6, 16, 21, 30 mflowj-915000000 

"* Flow rate from RWST Fig. 19, 27, 29 mflowj-887000000 

"* Void fraction at hot leg-IL Fig.9, 14, 22 voidg-200010000 

"* Void fraction at cold leg-IL Fig.9, 14, 22 voidg-253030000 

"* Total heat transfer-IL Fig.5 cntrlvar 63 

"* CPU time Fig.32 cputime-0 

"* Calculated time step Fig.34 dt-0 

"* Courant time step Fig.34 dtcmt-0 

"* Estimated mass error of primary Fig.33 sysmer-1
IV. ANALYSIS RESULTS AND DISCUSSIONS

IV.A. The CCT Analysis Results 

IV.A. 1. Analysis Results for the Hot-Leg Side Opening (Cases 1, 2, and 3) 

The plant response following the loss-of-RHR event is generally dependent on the location 

of the RCS opening, especially whether it is on the hot-leg side or on the cold-leg side. For the 

hot-leg side opening cases, as shown in Fig. 4, the pressure in the upper plenum rapidly 

increases in the early transient phase due to the coolant heat-up and boiling after the event. 

Particularly, the Case 2 with the PMO shows the rapid pressure increase, eventually reaching a 

maximum of 240 kPa. The Case I also shows the same rate of pressure increase in the early 

phase, and thereafter a moderately increasing rate throughout the transient. This difference is 

because the Case 1 with the water-filled SG transfers more decay heat into the secondary side 

by a reflux condensation on the SG U-tubes than the Case 2 with emptied SG. Figure 5 shows 

the total heat transfer through the SG for the Cases 1 and 2. More than 8 MW among the total 

core power of 14.125 MW is removed through the both SGs. Nearly same amount of heat 

removal by the SGs was also reported by Hassan and Raja [6]. Meanwhile, the Case 3 with 

the SMO, located at relatively low elevation, shows much less pressurization. It is because the 

coolant is discharged much earlier than that for the PMO cases. 

Figure 6 shows the discharging flow and its initiating time through the opening into the 

containment. For the Cases 1 and 2, the discharging flow has a similar pattern with a time 

delay of - 7000 s due to the different pressurization rate. The Case 3 shows that the two-phase 

mixture is vigorously discharged in the earlier phase of the transient. It implies that the 

coolant discharge behavior is strongly dependent on the location of opening as well as the SG 

secondary water level condition. 

In addition, in the PMO case, the reactor coolant is held up due to flooding in the bottom of 

the pressurizer. Figure 7 shows that the collapsed water level in the pressurizer increases and 

stabilizes depending on the pressure behavior for the Cases 1 and 2. Particularly, for the Case 

2, Figures 8 and 9 show that the water in the hot leg reaches a saturated temperature and 

voided in a short time of transient. After the hot leg is emptied, the increase of the water level 

in the pressurizer stops, and thereafter the significant discharging flow via the PMO is 

established. It indicates that the water holdup phenomena in the pressurizer and the 

discharging flow significantly affects the behavior of the coolant inventory in the RPV. As

19
shown in Fig. 10, for the Cases 1 and 2, the RPV water level slowly decreases before the 

significant discharging flow. However, it rapidly decreases after the coolant discharge. 

Especially, the Case 3 indicates the rapidly decreasing water level in the earlier phase of the 

transient because of the early discharge. 

As the water level in the RPV is reduced to the top of the core, the core uncovery occurs, 

and eventually the core heat-up is initiated. From this water level behavior, the time to core 

uncovery is estimated to be - 12000 s for the Case 1, - 6000 s for the Case 2, and - 3600 s for 

the Case 3. And, Fig. 11 shows that the core heat-up is initiated at - 2000 s after the core 

uncovery. It implies that the core damage time is strongly affected by the location of opening 

and SG secondary water level.

20
0 2000 4000 6000 8000 10000 12000 14000 16000 

Time [s] 

Fig. 4. Pressure behavior in the upper plenum (cases 1, 2, and 3)

0 2000 4000 6000 8000 

Time [s]

10000 12000 14000 16000

Fig. 5. Total heat transfer through the intact-loop SG (cases 1 and 2)

21

350

300 

250 

q

S200 

"1) 

150 

100 

50

12000 

10000 

8000 

6000 

4000 

2000 

0 

-2000

"cc M

M)
0 2000 4000 6000 8000 

Time [s]

10000 12000 14000 16000

Fig. 6. Discharging flow through the opening (cases 1, 2, and 3)

0 2000 4000 6000 8000 

Time [s]

10000 12000 14000 16000

Fig. 7. Collapsed water level in the pressurizer (cases 1 and 2)

22

LI) 

0) 

LL 

0 cc) 

C,

50 

40 

30 

20 

10 

0 

-10

T 

7-

10 

8 

6 

4 

2 

0 

-2
2000 4000 6000 8000 

Time [s] 

Fig. 8. Water temperature in the hot and cold legs (case 2)

2000 4000 6000 8000 

Time [s] 

Fig. 9. Void fraction in the hot and cold legs (case 2)

23

480

440 

400 

E 360 

320 

280

0

0

a 

0 

LL

1.4 

1.2 

1 

0.8 

0.6 

0.4 

0.2 

0

10000

10000
0 2000 4000 6000 8000 10000 12000 14000 16000 

Time [s] 

Fig. 10- Collapsed water level in the RPV (cases 1, 2, and 3)

0 2000 4000 6000 8000 

Time [s]

10000 12000 14000 16000

Fig. 11. Fuel cladding temperature (cases 1, 2, and 3)

24

10 

8

CU

6 

4

2 

0 

900 

800 

700

S600 

0. 

E 

I

500 

400 

300
IVA.2. Analysis Results for the Cold-Leg Side Opening (Cases 4 and 5) 

The cases with the cold-leg side opening show somewhat the different plant behavior from 

the hot-leg side opening cases. In particular, the reactor coolant in the cold leg begins to 

discharge to the containment in the earlier phase of the transient because the opening is 

located at relatively lower elevation. As shown in Fig. 12, the system pressure rapidly 

increases after boiling at - 1610 s, and then the pressure difference between the hot leg and 

the cold leg becomes increasing. The high differential pressure expels the water in the 

crossover leg and RPV toward the cold leg with the opening, and eventually the coolant is 

discharged via the opening to the containment. Then, the water level in the RPV decreases 

below the top of the core in the early phase of the boiling, resulting in the first core uncovery, 

as shown in Fig. 13. When the pressure reaches a maximum of 153 kPa for the Case 4 and 135 

kPa for the Case 5, the water in the crossover leg is immediately cleared, which is called a 

loop seal clearing (LSC). Simultaneously, the pressure in the hot leg and upper plenum drops 

to the cold leg pressure and such a pressure drop quickly increases the water level in the RPV 

because the compression force of the steam space in the upper plenum is disappeared. 

Eventually, the core is covered again by the coolant. 

Figures 14 and 15 show the voiding in the hot leg and the complete LSC in the crossover 

leg for the Case 4. Figure 16 also indicates that the coolant is significantly discharged via the 

opening before the LSC. After the LSC, because of the continuous steaming in the core 

region, the system pressure again increases. The Case 4 with the small opening shows a higher 

pressurization rate than the Case 5. As the RPV is pressurized, the steam is again discharged 

through the opening and then the water level in the RPV decreases moderately, as shown in 

Fig. 13. Eventually, the core is uncovered again at - 5500 s for the small CLO case and 

3400 s for the large CLO case. Thereafter, the water inventory further decreases and the core 

heat-up is initiated, as shown in Fig. 17. As a result, the calculation indicates that the first core 

uncovery is initiated from - 1600 s regardless of the opening size, whereas the second core 

uncovery time is dependent on the opening size.

25
2000 4000 6000 8000 

Time [s]

Fig. 12. Pressure behavior in the upper plenum (cases 4 and 5)

2000 4000 6000 8000 

Time [s] 

Fig. 13. Collapsed water level in the RPV (cases 4 and 5)

26

300

250 

_200 

"U 150 

100 

50

0 10000

8 

6 

E 

4 

2 

0

0 10000
2000 4000 6000 8000 

Time [s] 

Fig. 14. Void fraction in the hot and cold legs (case 4)

2000 4000 6000 8000 

Time [s] 

Fig. 15. Pressure difference in the crossover leg (case 4)

27

0 

U--

1.6 

1.4 

1.2 

1 

0.8 

0.6 

0.4 

0.2 

0

0

0

40 

30 

CL 

•. 20 

a) 

10 

0 

-10

10000

10000
2000 4000 6000 8000 

Time [s]

Fig. 16. Discharging flow through the opening (cases 4 and 5)

2000 4000 6000 8000 10000 

Time [s] 

Fig. 17. Fuel cladding temperature (cases 4 and 5)

28

cc 

o 

IT

120 

100 

80 

60 

40 

20 

0 

-20

0 10000

800 

700 

7O 

600 

cc 

0. 

E 500 

400 

300

0 12000
IV.A.3. Discussion on the Containment Closure Time (CCT)

The results of CCT analyses for the five cases of typical RCS configurations under the 

worst event sequence were summarized in Table 5 and compared with some available data of 

the CE-typed PWR [15]. It indicates that the time to boil-off and the time to core uncovery of 

the YGN 3/4 are estimated to be slightly earlier than that of the CE-typed PWR. It is due to 

the difference of the calculation models and plant geometry. In the YGN 3/4 simulation, the 

time to boil-off is estimated to be about 10 to 13 minutes regardless of the opening location 

and the SG secondary water level condition. It is because the time to boil-off is strongly 

dependent on the initial conditions such as the decay heat load and the amount of reactor 

coolant in and above the core. This result agrees well with 12 minutes of Parrish and Till's 

calculation [5]. 

On the contrary, the time to core uncovery and the time to core heat-up are strongly 

dependent on the size and location of the opening and the secondary water condition. The 

Case 5 with the large CLO and emptied SG indicates the earliest time to core uncovery of 40 

minutes after event. The Case 3 with the SMO also indicates 42.5 minutes of the core 

uncovery time and 85.7 minutes of the core heat-up time, which is the earliest core heat-up in 

the five cases. Also, the Case 1 with the water-filled SG shows that the time to core uncovery 

is delayed up to 183.3 minutes due to the secondary side cooling. 

Table 5. Results of transient analyses for the YGN 3/4 

(unit: minutes) 

RCS Openings Time to Boil Time to Core Time to Core Uncovery Heat-up 

* Case 1; PMO with water-filled SGs 12.3 183.3 218.0 

* Case 2; PMO with emptied SGs 12.3 83.3 139.0 

* Case 3; SMO with emptied SGs 13.0 42.5 85.7 

* Case 4; Small CLO with emptied SGs 11.6 (13.6) 75.0 (91.5) 123.3 

* Case 5; Large CLO with emptied SGs 10.6 (13.6) 40.0 (59.9) 111.7 

* ( ) is CE-typed PWR data 

In general, time is required to completely close the containment openings such as personnel 

or equipment hatches. Thus, an initiating time and completion time of the containment closure 

must be estimated to ensure the habitability of the personnel in the containment. From these 

transient analysis results, it is found that the containment openings are needed to start closing

29
before coolant boiling because the high enthalpy of steam or two-phase mixture is discharged 

to the containment shortly after coolant boiling. It could result in jeopardizing the personnel 

working inside the containment. Particularly, the simulations for the CLO cases indicate that 

the core could be instantaneously uncovered and partially damaged in the early phase of the 

boiling. Thus, if the containment closure is initiated after coolant boiling, it is needed to 

evaluate the in-containment environment to ensure the survivability of the workers in the 

containment. Also, to prevent the uncontrolled release of fission products to atmosphere, the 

containment openings must be completely closed before time to core uncovery. If the 

containment closure is achieved after the core uncovery, the environmental impact is needed 

to analyze. Based on these evaluations, the CCT could be determined from the time to core 

uncovery. For example, if there is a large opening such as the SMO (Case 3) or the CLO (Case 

5), the containment closure must be achieved within - 40 minutes after event. Also, if the 

pressurizer manway is open (Case 2), the containment closure must be completed within - 83 

minutes. However, if the SGs are filled with water for the same opening (Case 1), it could be 

extended to - 183 minutes. 

As a result, it was found that the earliest CCT was 40 minutes after event for the SG-inlet

plenum-manway opening or the large cold-leg opening cases with the emptied SGs. Also, the 

containment closure is needed to initiate before boiling time because the discharge via the 

opening after boiling could threaten the workers in the containment.

30
IV.B. The GIPR Analysis Results

IVB. 1. Analysis Resflts for the Hot-Leg Injection Cases (Cases A, B, and C) 

In the GIPR analysis to investigate the core cooling capability by the RWST water, the 

plant behavior before the gravity-injection is similar to that of the CCT analysis. Following 

the loss of the residual heat removal function, the core coolant is heated up and boiled off at a 

saturation temperature, and the RPV is pressurized. Figure 18 shows the pressure behavior in 

the upper plenum for the Cases A, B, and C with the same hot-leg injection and the different 

RCS opening. After the gravity-injection, the Case A shows a continuous pressure increase, 

while the pressure in the Cases B and C remain nearly constant. Such a pressure difference 

results from the different injection rate depending on the location of the RCS opening. 

As shown in Fig. 19, the gravity-injection flow for the Case A completely stops at - 550 s 

after gravity-injection, while the injection flow for the Cases B and C continues to remain 

high. This is because the pressure in the Case A continuously increases due to the water 

holdup in the bottom of the pressurizer and it makes the gravity flow stopped when it reaches 

about 172 kPa corresponding to the hydrostatic elevation head between the RWST and the 

RCS water levels. The water holdup phenomenon was already discussed in the CCT analysis. 

Figure 20 shows the collapsed water level in the pressurizer for the Case A. It indicates that a 

large amount of water is held in the pressurizer. 

After the gravity-injection flow is lost, the pressure further increases because of the water 

boil-off in the core region. When the pressure reaches about 300 kPa at - 4700 s, the two

phase mixture begins to be discharged through the PMO, and thereafter, the pressure 

moderately decreases. As shown in Fig. 21, the discharging flow via opening immediately 

increases at the maximum pressure for the Case A. Also, Fig. 22 shows that the hot leg is 

voided after the significant discharging flow. Thereafter, the steam is steadily discharged and 

the water level in the pressurizer is stabilized. 

Meanwhile, in the Cases B and C with the openings at lower elevation than the RWST 

water level, the RCS outflow through the opening is well established as well as the gravity

injection flow. Especially, Fig. 21 indicates that the discharging flow for the Case C is very 

stable and steady. The Case B also indicates the sufficient discharging flow even with some 

flow oscillations. Eventually, the system pressure remains atmospheric for a long-term 

transient after gravity-injection. 

Figure 23 indicates that the water temperature above the core region immediately drops

31
due to mixing between the RCS coolant and the injected water when the gravity injection 

starts. Depending on the mixing effect, the core coolant remains either a subcooled or a 

saturated condition. The Case A shows that the temperature again increases up to the 

saturation within a short time after the gravity injection. As discussed above, it is because the 

core decay heat could not be removed due to the stop of the gravity-injection flow. 

Meanwhile, the Cases B and C show that the temperature remains subcooled for a long-term 

transient. It is because the core flow is sufficiently established by the gravity injection and the 

core decay heat is fully removed. As a result, it indicates that the core boiling after event is 

prevented in the Cases B and C by the gravity-injection using the RWST water, whereas the 

gravity-injection process in the Case A could be ineffective. 

Figure 24 shows that the collapsed water levels in the RPV rapidly increase for all the 

cases when the gravity injection starts. The Cases B and C, thereafter, show that the levels 

remain nearly constant due to the stable RCS inflow and outflow. However, the Case A shows 

the water level decreasing. The initially rapid decrease is due to the stopped RCS inflow and 

the movement of the RCS coolant toward the pressurizer, and the lately moderate decrease is 

due to the two-phase mixture discharging through the opening. The continuous discharging 

leads to reduce the water level below the top of the core, and then the core is uncovered at 

6800 s, that is 96.6 minutes after event.

32
2000 4000 6000 8000 

Time [s]

Fig. 18. Pressure behavior in the upper plenum (cases A, B and C)

0 2000 4000 6000 8000 

Time [s] 

Fig. 19. Mass flow rate from the RWST (cases A, B and C)

33

500

400 

S300. 

a) 

cn 

200 

100

0 10000

300 

250 

200 

150 

100 

50 

0 

-50

Ca 

FL 

0 to Mi

10000
2000 4000 6000 8000 10000

Time [s]

Fig. 20. Collapsed water level in the pressurizer (case A)

2000 4000 6000 8000 10000

Time [s]

Fig. 21. Discharging flow rate through the opening (cases A, B, and C)

34

20

16 

S12 

.-, 

S8 

4 

0

0

800 

600 

400 

200

0 

1 

Cu

0 

-200

0
2000 4000 6000 8000 

Time [s] 

Fig. 22. Void fraction in the hot and cold legs (case A)

10000

2000 4000 6000 8000 10000

Time [s]

Fig. 23. Water temperature above the core region (cases A, B and C)

35

0

1.6

1.4 

1.2 

,- 1 

0 

LL0.8 

0.6 

"0.4 

0.2 

0

500 

450 

2 400 

(D 0,. 

E 

(D 3

350

300

0
10

9 Case A: Hot-leg injection and PMO discharge 

Case B: Hot-leg injection and SMO discharge 

8 Case C: Hot-leg injection and small CLO discharge 

S7 Case B 

>1 

S6 -- 'Case C 

3 5 • CaseA 

4 4 I Ii.-. ,,Core Top 

Loss-of-RHR 

3 I Gravity Injection Start 

0 2000 4000 6000 8000 10000 

Time [s] 

Fig. 24. Collapsed water level in the RPV (cases A, B and C)

36
IVB.2. Analysis Results for the Cold-Leg Injection Cases (Cases D, E, and F) 

The Cases D, E and F with the same cold-leg injection and the different RCS opening have 

the thermal-hydraulic behavior similar to the cases of the hot-leg injection. As shown in Figs. 

25 and 26, the Case D with the PMO indicates that the system pressure continues increasing 

and the core is boiled off after gravity injection. As similar to the Case A, the gravity flow 

completely stops at - 500 s after gravity-injection, as shown in Fig. 27. Then, due to the 

continuous discharge via opening, the water level in the RPV decreases and the core uncovery 

occurs. As shown in Fig. 28, the core is uncovered at the nearly same time as the Case A. This 

result indicates that the core cooling for the PMO case could not be maintained by the gravity

injection process regardless of the injection paths. 

The Case E with the SMO shows that the system pressure remains low enough to maintain 

the injection flow rate as the Case C after gravity-injection. The core is also successfully 

cooled for a long-term transient by the well-established RCS inflow and outflow. As a result, 

it indicates that the Cases C and E with the injection and opening on the different leg side, 

such as the hot-leg side injection and cold-leg side discharge, are the most suitable gravity

injection paths to avoid the core boiling after event. 

The Case F with the CLO also indicates low pressure as the Case B, but the water in the 

core region is saturated and boiled off after gravity-injection as shown in Fig. 26. It is because 

most of the cold water injected through the cold leg is directly discharged through the CLO 

without passing the core region. Meanwhile, in the Case B with the hot-leg injection, part of 

the cold water injected passes the upper part of the core region and then the core boiling is 

prevented. As a result, it indicates that the Case B with the injection point and opening on the 

hot-leg side is a little more effective in core cooling after event than the Case F with the 

injection point and opening on the cold-leg side.

37
2000 4000 6000 8000 10000

Time [s]

Fig. 25. Pressure behavior in the upper plenum (cases D, E and F)

2000 4000

Time [s]

6000 8000 10000

Fig. 26. Water temperaturesabove the core region (cases D, E and F)

38

500 

400

300

"0E wp 

5D

200 

100

0

500 

450 

S400 

CL 

E 

350 

300

I I Id-e I I I I I 

Case D: Cold-leg injection and PMO discharge 

Case E: Cold-leg injection and SMO discharge 

- Case F: Cold-leg injection and large CLO discharge 

Gravity Injection Start Case D 

Lo-f-H Case Ca-F 

0
2000 4000 6000 8000 

Time [s]

Fig. 27. Mass flow rate from the RWST (cases D, E and F)

2000 4000 6000 8000 

Time [s]

Fig. 28. Collapsed water level in the RPV (cases D, E and F)

39

300

250 

200 

S150 0.i 

0 i,' 100 

250

0 

-50

0 10000

0 

1 

0

10 

9 

8 

7 

6 

5 

4 

3 

2

0 10000
IV.B.3. Discussion on the Gravity-Injection Path and Rate (GIPR)

From the above GIPR analysis, it is found that the gravity-injection flow rate is dependent 

on the differential elevation head between the RWST and the RCS water levels and the RCS 

opening size and location. However, in practice, the gravity flow can be constrained by the 

hydraulic resistance of the flow path, such as valves, flow orifices, or pumps. In addition, the 

injection flow rate can be throttled by operator for a proper cooling or RCS inventory control. 

In the present study, to determine the minimum flow rate needed to prevent core boiling after 

event, additional calculations are performed with varying the injection line size for the Case C 

with the hot-leg injection and CLO discharge. The range of injection line size is 5 up to 10 

inches. The Case C was estimated to be the most suitable injection path to remove the decay 

heat after event in this transient analysis. 

Figure 29 shows that the gravity-injection flow rate depends on the injection line sizes. As 

the line size reduces, the RWST injection rate decreases. In particular, more than 6 inches of 

the line size indicates a uniform injection flow for a long-term transient after gravity-injection. 

As above discussed, the reason is that the RCS inflow from the RWST is balanced with the 

RCS outflow through the opening. Figure 30 shows the stable discharging flow for more than 

6 inches. However, for less than 5 inches diameter, the coolant in the core region continues 

boiling off because of the insufficient RCS inflow. Fig. 31 shows that the water temperature 

above the core region remains saturation condition despite of the gravity-injection. Eventually, 

it loses the gravity-injection flow at - 8000 s due to the system pressurization and the 

discharging flow is stopped after 9000 s. 

As a result, it indicates that the injection line with more than 6 inches diameter is effective 

to keep the reactor coolant subcooled by the gravity injection. The injection rate 

corresponding to the 6 inches diameter averages about 54 kg/s. That is a minimum gravity

injection rate needed to prevent the core boiling after event. Based on the minimum injection 

rate and the nominal capacity of the RWST of the YGN 3/4, the injection duration, which 

could delay the core boiling, is estimated to be about 10.6 hours if 70% of the RWST water is 

available. It indicates that the gravity-injection using the RWST water is capable of providing 

the core cooling for a sufficient long-term transient for the Case C after the event. The results 

are similar to the Case E with the cold-leg injection and the SMO discharge, even not 

presented in this report.

40
2000 4000 6000 8000 

Time [s] 

Fig. 29. Mass flow rate from the RWST (sensitivity)

2000 4000 6000 8000 

Time [s] 

Fig. 30. Discharging flow rate through the opening (sensitivity)

41

3

0) 

LL 

CO)

250 

200 

150 

100 

50 

0 

-50

0

0

200 

150 

S100 

0 

S50 

CO)

0 

-50

10000

10000
440

' 00 Start 

,•• 5 inches 

Eh 

i? 360 6 inches 

7 inches 

320 Loss-of-RHR 10 inches 

I , I ,I ,i, 

0 2000 4000 6000 8000 10000 

Time [s] 

Fig. 31. Water temperature above the core region (sensitivity)

42
IV.B. Run Statistics

In the previous study on the assessment of RELAP5/MOD3.2 code with the LSTF 

experiment for the loss-of-RHR event, it was reported that the MOD3.2 version was capable 

of simulating the transient following the event with appropriate time step and CPU time [10, 

11 ]. Even though there were some flow oscillations during boiling process in the core region, 

the major thermal-hydraulic phenomena such as a loop seal clearing, water holdup in the 

pressurizer, and a non-condensable gas migration were reasonably predicted with an 

appropriate computational time. In addition, the transient run was successfully performed 

without any failure. This plant application using the same code and modeling scheme as the 

LSTF simulation also gives a similar calculation characteristics. 

Figure 32 shows the required CPU times for simulating the transient in case of pressurizer 

manway open. The LSTF case represents the CPU time for the previous transient calculation 

of the LSTF simulation, and the YGN 3/4 cases represent the Case 1 of the CCT analysis and 

Case A of the GIPR analyses, respectively. It indicates that the similar CPU time is required 

for all the transient calculation. 

In addition, it was reported that the RELAP5/MOD3 code predicted too large system mass 

errors during the transient. In the previous calculation for the LSTF simulation, the mass error 

in the primary system was estimated about 90 - 100 kg for the transient of 2.8 hours, that was 

nearly 4% of initial coolant mass inventory. As similar to that, in the YGN 3/4 simulation, the 

mass error is estimated about 2,700 - 3,100 kg for the same duration, that is nearly 3% of 

initial coolant inventory. Figure 33 shows the estimated mass error behavior during the 

transient. It shows that the mass error is rapidly generated in the phase of coolant boiling and 

thereafter it gradually rises. Because the large mass error could significantly reduce the 

reliability of the calculation data, these mass errors should be reduced up to the negligible 

value. Thus, the efforts to reduce the system mass error are needed in the future to improve the 

reliability of the code. 

The main computer used in the calculation was a DEC workstation 5000/240 with UNIX 

operating system. During the transient, it indicates that the CCT and GIPR analyses require 

the similar calculation time steps. Figure 34 shows the Courant time limit and advanced time 

step size for the Case A of GIPR analysis. The maximum time step is 0.1 s. It shows that the 

calculation is reasonably conducted with appropriate time steps below the Courant Limit. In 

this case, the required CPU time to simulate the transient of 10,000 s is 152,094.1 s including 

10.4 s for input processing. For that time interval, the attempted advancement is 172,683 time

43
steps. Then, the grind time can be calculated as follows. It will be 4.10 CPU msec/vol/step. 

This value is similar to that of the LSTF simulation, 3.673 - 4.036 CPU msec/vol/step.

- CPU time 

- Number of time step 

- Number of Volume 

- Transient Real Time 

- Grind Time

CPU = 152,094.1 - 10.4 = 152,083.7 sec 

DT = 172,683 - 10,084 = 162,599 

C = 228 

RT = 10,000 sec 

GT = CPUxI000/(CxDT) = 4.10 CPU msec/vol/step

As a result, it is found that the real plant simulation of the RELAP5/MOD3.2 version gives 

the similar calculation characteristics to the LSTF simulation of the loss-of-RHR event in the 

CPU time, calculation time step, and estimated system mass error.

44
2000 4000 6000 8000 10000 

Time [s]

Fig. 32. Comparison of CPU time

0 2000 4000 6000 

Time [s

8000 10000 12000 14000

Fig. 33. Estimated mass error

45

200

160 

C, 

,: 120 x. 

0 

E 

S80 

CL

40 

0

0 12000

5000 

4000 

.'3000 

w 

S2000 

1000 

0
0 .8 I I I | 

0.7 YGN 3/4-GIPR CaseA 

0.6 U ' I "1i - ' 

I I l " 

I it 

rn 0.4 , I 1,,' 

('Di Courant Time Step 

II0.3II II I Ii' 

0.2 Advanced Timeý , I k +-I 

0. 

00 2000 4000 6000 8000 10000 

Time [s] 

Fig. 34. Advanced time step (case A)

46
V. CONCLUSIONS

To investigate tbe mitigation measures following the loss-of-RHR event during reactor 

shutdown, the plant operating conditions of the Yonggwang Units 3 and 4 (YGN 3/4), PWR 

type of 2,815 MW thermal power in Korea, were reviewed. The possible event sequences 

were identified and the long-term transient analyses were performed using the standard 

version of RELAP5/MOD3.2 code. This analysis is finally to determine the containment 

closure time (CCT) to prevent the release of fission products to environment and the gravity

injection path and rate (GIPR) to effectively cool the core after event. The findings from the 

transient analysis can be summarized as follows: 

(1) The loss-of-RHR event of the YGN 3/4 plant was analyzed using the 

RELAP5/MOD3.2 code and model which was assessed with the LSTF experiment in the 

previous study. As a result, the real plant simulation gives the similar calculation 

characteristics to the LSTF simulation in the required CPU time, computational time step, and 

system mass error. Thus, it was found that the RELAP5/MOD3.2 code was capable of 

appropriately simulating the loss-of-RHR event of the real plant. 

(2) From the CCT analysis for the five cases of typical RCS configurations with no RCS 

makeup and unavailable secondary cooling, the time to boil after event was estimated to be 

about 10 to 13 minutes regardless of the opening size and location. Meanwhile, the time to 

core uncovery was estimated to be about 40 to 183 minutes depending on the elevation and 

size of the opening and the SG secondary water level condition. Particularly, in case with the 

water-filled SG, it was delayed about 100 minutes by the reflux condensation on SG U-tubes, 

as compared to the emptied SG case. However, it also indicated that the first core uncovery 

could occur in the early phase of the event by the loop seal clearing phenomenon in the 

crossover leg for the cold-leg opening case. As a result, it was found that the earliest CCT was 

40 minutes after event for the SG-inlet-plenum-manway opening or the large cold-leg opening 

cases with the emptied SG. Besides, the containment closure is needed to initiate before the 

coolant boiling because the discharge via the opening after boiling could threaten the workers 

in the containment. 

(3) From the GIPR analysis for the six possible gravity-injection paths from the RWST, the 

following conclusions were obtained. In cases with the PMO, located at higher elevation than

47
the RWST water level, the system pressure continued increasing due to the water holdup 

phenomenon in the pressurizer, and then the core was uncovered at about 96.6 minutes despite 

of the gravity-injection process. In cases with the injection point and opening on the same leg 

side, the core cooling was dependent on the core flow. For instance, as most of the water 

injected from the RWST bypassed the core region, the core cooling could not be effectively 

maintained. Meanwhile, in the cases with the injection point and opening on the different leg 

side, the RCS was well depressurized and the core boiling was successfully prevented for a 

long-term transient. As a result, these injection paths were estimated to be the most suitable 

paths in avoiding core boiling for a long-term period. In addition, about 54 kg/s of minimum 

injection rate was required to maintain core cooling. Such an injection rate was capable of 

providing the core cooling for about 10.6 hours if 70% of the RWST water was available. 

(4) These results will provide useful information to operators to understand the plant 

behavior and to cope with the event in timely manners. However, to apply them to the 

emergency and recovery procedures against the loss-of-RHR event during shutdown 

operation, additional case studies are needed in the future for wide range of operating 

conditions such as RCS inventory, RWST water temperature, and core decay heat rate.

48
REFERENCES

1. U.S. NRC, "Loss of Residual Heat Removal System-Diablo Canyon, Unit 2, April 10, 

1987," NUREG-1269, U.S. Nuclear Regulatory Commission (1987). 

2. U.S. NRC, "Loss of Vital AC Power and the Residual Heat Removal System During Mid

loop Operations at Vogtle Unit 1 on March 20, 1990," NUREG-1410, U.S. Nuclear 

Regulatory Commission (Jun. 1990). 

3. U.S. NRC, "Shutdown and Low Power Operation at Commercial Nuclear Power in the 

United States," NUREG-1449, U.S. Nuclear Regulatory Commission, (Sep. 1993). 

4. S.A. NAFF, et al., "Thermal-Hydraulic Processes During Reduced Inventory Operation 

with Loss of Residual Heat Removal," NUREG/CR-5855, EGG-2671, U.S. NRC, (Feb. 

1992). 

5. K.R. PARRISH AND H.A. TILL, "Palo Verde Nuclear Generating Station Midloop 

Condition RETRAN Model," Nucl. Technol., Vol. 93, pp. 53, (Jan. 1991). 

6. Y. A. HASSAN and L. L. RAJA, "Simulation of Loss of RHR During Midloop Operations 

and the Role of Steam Generators in Decay Heat Removal Using the RELAP5/MOD3 

Code," Nucl. Technol., Vol. 103, pp. 310 (Sep. 1993). 

7. M.B. SATTISON, et al., "Low-Power and Shutdown Models for the Accident Sequence 

Precursor (ASP) Program," Tran. 26th Water Reactor Safety Info. Mtg. (WRSIM), 

Bethesta, Maryland, (Oct. 1996). 

8. KEPCO, "Final Safety Analysis Report: Yonggwang Nuclear Power Plant Units 3 and 4," 

Korea Electric Power Company, (Mar. 1996). 

9. K. E. LARSON, et al., "RELAP5/MOD3 Code Manual: User's Guide and Input 

Requirements," NUREG/CR-5535, U.S. Nuclear Regulatory Commission, (INEL

95/0174), Vol. 1-5, (Aug. 1995). 

10. K. W. SEUL, Y. S. BANG, S. LEE, and H. J. KIM, "Assessment of RELAP5/MOD3.2 

with the LSTF Experiment Simulation a Loss of Residual Heat Removal Event During 

Mid-Loop Operation," NUREG/IA-0143, U.S. Nuclear Regulatory Commission, (Aug. 

1998). 

11. K. W. SEUL, Y. S. BANG, and H. J. KIM, "Plant Behavior Following a Loss-of

Residual-Heat-Removal Event under Shutdown Condition," Nucl. Technol., Vol. 126, pp. 

265, (Jun. 1999). 

12. H. NAKAMURA and Y. KUKITA, "PWR Thermal-hydraulic Phenomena following Loss 

of Residual Heat Removal (RHR) during Mid-Loop Operation," Proc. International 

Conference on New Trends in Nuclear System Thermohydraulics, Pisa, Italy, pp 77-86, 

(May-Jun. 1994). 

13. S. BANERJEE and Y. A. HASSAN, "RELAP5/MOD3 Simulation of the RHR System 

during Midloop Operation Experiment Conducted at the ROSA-IV Large Scale Test

49
Facility," Proc. International Conference on New Trends in Nuclear System 

Thermohydraulics, Pisa, Italy, pp. 407-414, (May-Jun. 1994). 

14. Y. FERNG and C, LEE, "Assessment of the Simulation Capability of RELAP5/ MOD3 

Compared with IIST Tests for Loss of the RHR System During Midloop Operation," Nucl. 

Technol., Vol. 116, pp. 19, (Oct. 1996). 

15. APS, "Safety Analysis Operational Data," Revision 4, SA-13-COO-95-005, Arizona Public 

Service Company, (May 1995).

50
Appendix A

Input Decks for the Steady State Calculation: 

Loss-of-RHR Event in Yonggwang Units 3 & 4 Plants 

1) Input Deck for the CCT Analysis 

2) Input Deck for the GIPR Analysis 

(Not Included by the Proprietary Information)

51
Appendix B

Input Decks for the Transient Calculation: 

Loss-of-RHR Event in Yonggwang Units 3 & 4 Plants 

1) Input Deck for the CCT Analysis 

2) Input Deck for the GIPR Analysis 

(Not Included by the Proprietary Information)

53
NRC FORM 336 U.S. NUCLEAR REGULATORY COMMISSION 1. REPORT NUMBER 

(24M (Assigned by NRC, Add Vol., ,ulp., Rev., NRCM o BIBUOGRAPHIC DATA SHEET and Addendum Numbe. f artany.) 

3201, 3202 BBIGAHCDT HE 

(See irnsruct~io on ate ievnwrs) 

2. TITLE AND SUBTITLE NUREG/IA-0182 

Application of RELAP5/MOD3.2-to the Loss-of-Residual-Heat-Removal Event 

Under Shutdown Condition 3. DATE REPORT PUBLISHED 

April 2000 

4. FIN OR GRANT NUMB•ER 

5. AUTHOR(S) 6. TYPE OF REPORT 

K.W. Seul, Y.S. Bang, H.J. Kim 

7. PERIOD COVERED (Inclusivee Das) 

8. PERFORMING ORGANIZATION - NAME AND ADDRESS (if NRC, provide Division, Office or Region, U.S. Nuclear Regulatry Comrission, and mailing address, f con&ac,, 

provide name and mailing address.) 

Korea Institute of Nuclear Safety 

P.O. Box 114 

Yusung, Taejon 

305-600, Korea

9. SPONSORING ORGANIZATION - NAME AND ADDRESS (if NRC, type 'Sameas above- if conactor, provide NRC Division, OficeorxRegion, U.S. Nuclear Regulatory Commission, 

and mailing addvass.)

Division of System Analysis and Regulatory Effectiveness 

Office of Nuclear Regulatory Research 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-0001

10. SUPPLEMENTARY NOTES

11. ABSTRACT (00 wa•ds or as) 

The long-term transient following a loss-of-residual-heat-removal (loss-of-RHR) event during reactor shutdown was analyzed to 

determine the containment closure time (CCT) to prevent the release of fission products to environment and the gravity-injection 

path and rate (GIPR) to effectively cool the core. The thermal-hydraulic analysis was carried out using the RELAP/5/MOD3.2 code 

and relevant modeling scheme, which was assessed with the LSTF experiment in a previous study (NUREG/IA-0143). Based on 

the plant-specific geometry data including various operating conditions, the possible event sequences were identified for the 

Yonggwang Units 3 and 4 plant (YGN3/4), which is CE-typed PWR of 2,815 MW thermal power in Korea. As a result, the real 

plant simulation gives the similar calculation characteristics to the previous LSTF simulation, and then it was found that the 

RELAP5/MOD3.2 code is capable of appropriately simulating the loss-of-RHR event of the real plant

12. KEY WORDSIDESCRIPTORS (List words oraplass hat will assist researchrs in ocating the repor.) 13. AVAILABILfTY STATEMENT 

RELAP5 unlimited 

Reflux Condensation 14. SECURITY CLASSIFICATION 

(This Page) 

unclassified 

(This Repo") 

unclassified 

15. NUMBER OF PAGES 

16. PRICE

NRC FORM 335 (2-89)
Federal Recycling Program
NUREG/IA- 0182

UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. 20555-0001

FIRST CLASS MAIL 

POSTAGE AND FEES PAID 

USNRC 

PERMIT NO. G-67

APPLICATION OF RELAP5/MOD3.2 TO THE 

LOSS-OF-RESIDUAL-HEAT-REMOVAL EVENT UNDER SHUTDOWN CONDITION

APRIL 2OUW