Evaluation of the potential for containment bypass due to steam generator tube rupture in VVER-1000/V320 reactor during extended SBO sequence using SCDAP/RELAP5 code

Nuclear Science and Technology, Vol.9, No. 4 (2019), pp. 09-15

Evaluation of the Potential for Containment Bypass due to

Steam Generator Tube Rupture in VVER-1000/V320 Reactor

during Extended SBO sequence using SCDAP/RELAP5 code

Nguyen Van Thai^1,*, Doan Manh Long², Tran Chi Thanh²

¹Department of Nuclear Engineering and Environmental Physics, School of Engineering Physics,

Hanoi University of Science and Technology

²Vietnam Atomic Energy Institute

E-mail: [email protected]

(Received 05 November 2019, accepted 30 December 2019)

Abstract: A severe accident-induced of a Steam Generator (SG) tube releases radioactivity from

the Reactor Coolant System (RCS) into the SG secondary coolant system from where it may escape

to the environment through the pressure relief valves and an environmental release in this manner is

called “Containment Bypass”. This study aims to evaluate the potential for “Containment Bypass”

in VVER/V320 reactor during extended Station Blackout (SBO) scenarios that challenge the tubes

by primarily involving a natural circulation of superheated steam inside the piping loop and then

induce creep rupture tube failure. Assessments are made of SCDAP/RELAP5 code capabilities for

predicting the plant behavior during an SBO event and estimates are made of the uncertainties

associated with the SCDAP/RELAP5 predictions for key fluid and components condition and for

the SG tube failure margins.

Keywords: Containment Bypass, SBO, SCDAP/RELAP5.

failure of steam generator tubes prior to the

failure of one of other components (hot leg

piping, pressurizer surge-line piping, and the

reactor vessel) leads to discharge of some

fission products into the steam generator

secondary system from where they may be

discharged to the environment through the

pressure-relief valves. This sequence is

potentially more risk-significant since it

involves a containment bypass scenario. The

relative timing of these structural failures

therefore affects the event sequence and

whether the containment is bypassed.

I. INTRODUCTION

In 1990s, U.S. Nuclear Regulatory

Commission (NRC) and the Russian Federal

Nuclear and Radiation Safety Authority

Gosatomnadzor (GAN) agreed to work

together in BETA Project to perform a

probabilistic risk assessment (PRA) of Unit I

of the Kalinin Nuclear Power Station (KNPS),

which is a VVER1000/V-338. Analysis within

the BETA Project involves different levels of a

wide-scope PRA and the PRA model

represents the set of accident sequences

following the initial events (IEs) up to the end

state of each sequence. The reports indicated

that the most significant release category for

offsite consequences is containment failure as a

result of an IE with a leak from the primary to

the secondary circuit [1]. In this aspect, steam

generator tubes comprise a majority of the

reactor coolant system pressure boundary and

This paper present thermal-hydraulic

evaluations of VVER-1000/V-320 during

extended

SBO

event

using

the

SCDAP/RELAP5 systems analysis code. In

general, the design features of VVER-1000/V-

338 are similar to the standard VVER-1000/V-

320, excepted of the main circulation loops

EVALUATION OF THE POTENTIAL FOR CONTAINMENT BYPASS DUE TO STEAM…

configuration which is equipped with main

gate valves on the cold and hot legs (CLs and

HLs) [2]. The potential for “Containment

Bypass” is associated with natural circulation

of superheated steam inside the piping loop

that might induce creep rupture tube failure.

SCDAP/RELAP5 predictions provide key fluid

and components condition for estimation of the

SG tube failure margins.

 The turbine-driven auxiliary feedwater

(TDAFW) system is assumed to independently

fail, so no MFW or auxiliary feedwater (AFW)

system is available;

 A station battery life of four hours is

assumed; after that time all automatic and

operator control of the pressurizer PORVs and

SG secondary system PORVs is lost;

During the initial portion of the accident

scenario, buoyancy-driven coolant-loop natural

circulation carries hot water from the core

through the SGs, transferring heat to the SG

secondary water inventory. The SG water

inventory is boiled and the steam is released

through the SG PORVs. The secondary water

inventory declines and is eventually fully

depleted since the MFW and AFW systems are

not operative. After that time, the core decay

power heats and swells the RCS water,

increasing its temperature and pressure. The

basic physical processes during this regard the

transport of hot steam from the core outward

into the other regions of the reactor vessel

(RV) and coolant loops. Two coolant loop

natural circulation flow patterns that may be

encountered subsequent to the uncovering and

heat-up of the reactor core, based upon whether

or not the loop seals (the cold leg piping

connecting the outlets of the SGs to the inlets

of the RCPs) remain liquid-plugged.

II. DESCRIPTIONS OF ACCIDENT

SCENARIO AND PLANT BEHAVIOR

The low-probability SBO base case

accident event scenario results in a severe

accident because none of the systems that

normally provide core cooling are assumed to

be operable nor is any alternate equipment

(e.g., security-related mitigation methods)

assumed to be available [4]. The accident event

is initiated by a loss of off-site alternating

current (AC) power, which immediately results

in reactor and turbine trips and the coast-down

of the four reactor coolant pumps (RCPs).

Also, assumptions are made as follows:

 The diesel-electric generators fail to

start and all AC power sources are lost;

 Letdown flow is isolated and the

pressurizer level control and RCP seal injection

functions of the charging system are lost;

 The high-pressure and low-pressure

safety injection systems are unavailable as a

result of the AC power loss;

If liquid is cleared from a loop seal

(along with the region of the RV lower plenum

that extends above the bottom of the core

barrel, the "downcomer skirt"), hot steam is

transported from the core through the HL, SG

tubes and cold legs in the normal direction of

flow (i.e., that seen during plant operation).

This flow pattern transports the hot steam

directly (without benefit of mixing) through all

of the SG tubes, leading to SG tube failure

 The accumulator systems (four HAs)

are available for injecting coolant into the cold

legs should the RCS pressure fall below the

initial accumulator pressure, 5.9 MPa;

 The main feedwater (MFW) flow stops

and the motor-driven auxiliary feedwater

(MDAFW) system is unavailable as a result of

the AC power loss;

10

NGUYEN VAN THAI et al.

prior to HL or pressurizer surge line failure.

However, if a loop seal remains liquid-

plugged, the more complex flow pattern

develops instead. Hot steam is transported

through the upper portion of the HL cross

section to the SG inlet plenum, where it is

mixed with cooler steam emanating from

circulations set up within the SGs, with some

of the tubes flowing in the normal direction

and the remaining tubes flowing in the reverse

direction. The mixing process within the SG

inlet plenum determines the temperatures of

the steam entering the SG tubes and the steam

that is returned to the RV through the lower

portion of the HL cross section. Fluid mixing

in the SG inlet plenum buffers the entry of hot

steam into the SG tubes, thus delaying SG tube

failure and making it more likely that some

other component (HL, pressurizer surge line or

RV) will be the first to fail.

The issues of primary interest for

containment bypass are: (1) do the loop seals in

all coolant loops remain liquid plugged and if

so (2) does the fluid mixing in the SG inlet

plenum sufficiently slow the SG tube heat-up

process so that the HL, pressurizer surge line

or RV will fail prior to a SG tube? [4]. The

analyses in this paper address these issues.

Fig. 1. RCS Layout of VVER-100/V-320 [3]

TURBINE

196

194

MSIV

187

MSIV

287

MSIV

387

MSIV

487

186

286

386

486

490

190

489

BRU-A1

189

BRU-A1

182

184

484

482

REG

180

BRU-In1

191

BRU-In1

491

192

Valve Valve Valve Valve

185 285 385 485

492

480

Steam

dome

Steam

dome

169

168

167

469

468

467

Separator

Volume above

perforated sheets

Volume above

perforated sheets

508

177

477

ACC

152

452

PORV

Volume between

tube bundles and

perforated sheets

Volume between

tube bundles and

perforated sheets

506

511

466

176

175

174

136

166

125

124

123

122

121

116

416

415

414

413

412

411

410

436

435

476

JNG80

JNG60

087

JNG70

088

JNG50

086

165

115

465

464

463

462

461

425

424

423

179

178

173

172

171

135

134

133

479 475

089

164

114

434

433

432

431

430

478

473

474

504

450

150

163

113

162

112

422

421

132

131

130

472

471

161

111

502

110

56

54

100

400

3

102

61-62

1-2

49-50

45-48

41-44

51-52

63-64

3-4

402

8

7

4

3

2

1

2

4

142

144

5-6

7-8

444

442

141

441

11

1

11

36

37-40

9

9-11

30 32 33 31

26 28 29 27

10-12

34

4

1

2

3

1

6

9

8

7

6

5

2

3

140

440

35

4

3

2

1

4

2

1

4

5

6

23-25

22-24

18-20

5

6

19-21

15-17

14-16

Fig. 2. Nodalization scheme of VVER-1000/V-320 Plant

11

EVALUATION OF THE POTENTIAL FOR CONTAINMENT BYPASS DUE TO STEAM…

zero, the structure has not experienced any

creep damage. If the value is one, the structure

has failed due to creep damage. Two different

theories (Larson Miller and Manson-Haferd)

are applied in SCDAP/RELAP5 which are

dependent upon the structural materials and the

range of stress and summarized in Table I.

III. PLANT NODALIZATION AND

PARAMETERS SETUP WITH

SCDAP/RELAP5

The SCDAP/RELAP5 computer code

calculates the overall RCS thermal-hydraulic

response for severe accident situations that

include core damage progression and RV heat-

up and damage. The computer code is the result

of a merging of the RELAP5 and SCDAP

computer codes. Prediction of structural failure

are made based on the structure configuration,

its material properties and the fluid conditions

that are locally present. The code also includes

models for calculating the creepture failure of

structural components and these are used to

predict failure time for the hot legs, pressurizer

surge line, and SG tubes [5].

Table I. Equations for the time to creep rupture [5]

Since SG tubes are made from heat-

resistant steel (08X18H10T) which is similar to

heat-resistant Stainless Steel, equations for

creep rupture time of 316 Stainless Steel in

Table i are selected in this study. Estimated

stress in SG tubes (cylindrical shaped

structure) is calculated as follows:

The

SCDAP/RELAP5

VVER-

1000/V320 plant model represents the fluid

volumes and structures in the core, RV and

primary and secondary coolant system regions

in the plant as shown in Fig 1. The nodalization

diagrams for the final SCDAP/RELAP5

VVER-1000/V-320 four-loop plant model are

provided in Fig. 2.

(P r  P r_o)

i

rⁱ_o r_i^o

 

(2)

In this work, creep rupture failure

calculations are performed for the hot average SG

tubes in which pressure and temperature

conditions of fluid and heat structures in four SGs

are used as inputs for the analyses. A parameter

that measure creep damage is calculated each time

step for each structure being monitored for creep

rupture by following equation:

Fig. 3. Model ofSG ubes

IV. RESULTS AND DISCUSSION

A. Plant behaviors

t

D_c



t  t



 D_c



t





(1)

t_r

t

 

The SCDAP/RELAP5 plant model was

run to a steady solution. The plant model using

the nodalization scheme to establish full-power

steady-state conditions from which the SBO

transient accident sequence is initiated

(Figures.4&5).

Where D_c

problem time t, t is time step at current

problem time and t_ris time required for the



t



is creep damage at

 

t

structure to fail by creep rupture at current state

of temperature and stress. If the value of D_cis

12

NGUYEN VAN THAI et al.

Fig. 4. Steady-state conditions (Pressure) at primary

Fig. 6. Transient condition of reactor power and

and secondary loops

pressure in PRZ

Fig. 7. Transient condition of mass flow rate

through the primary loops and SG levels

Fig. 5. Steady-state conditions (Mass Flow Rate

and Temperature) at primary and secondary loops

Buoyancy-driven coolant-loop natural

circulation carries hot water from the core

through the SGs, transferring heat to the SG

secondary water inventory until the SG heat

sink is lost around 11000s (Fig. 7). It can be

seen that after the SG heat sink is lost, the

pressure the cooling afforded by system heat

loss to containment and RCP shaft seal leak

flow is insufficient to remove the RCS heat

load, causing the RCS and PRZ pressure to

increase (Fig. 6). After SG dry-out, the RCS

pressure increase is limited by multiple cycling

of the PORVs and by two cycles of the SRVs

during the period with the most-challenging

RCS pressurization conditions. This challenge

The SBO base case event sequence was

simulated with SCDAP/RELAP5, starting from

time 250s when the loss of off-site power

occurs. Figure 5 showed the transient condition

of reactor power and pressure level in PRZ.

After a short period when the PRZ

pressure initially falls and rises slightly due to

the effects of the reactor and turbine trips, the

PRZ pressure declines in response to the

cooling provided by heat removed to the SGs.

The RCS fluid mass lost through the

pressurizer PORVs and SRVs and through the

RCP shaft seal leakage paths depletes the

RCS inventory.

13

EVALUATION OF THE POTENTIAL FOR CONTAINMENT BYPASS DUE TO STEAM…

is presented when the increasing temperatures

cause the RCS fluid to swell, completely filling

the pressurizer with water. The fuel

temperature increases after the heat sink is lost

and the onset of fuel rod oxidation as well as

fuel rod cladding rupture occurred as an

unvoidable consequence (Fig. 8).

occur after the time when SG tube structural

failures is experienced.

Fig. 9. Structure temperature and creep

damage index

V. CONCLUSIONS

The

paper

presents

preliminary

Fig. 8. Temperature of fuel pins and hot legs

assessments the VVER-1000/V-320 plant

behavior during an SBO event with the

SCDAP/RELAP5 for key fluid and

components condition and for the SG tube

failure margins. Further analyses will be

performed to evaluate the potential for

“Containment Bypass” in VVER-1000/V-320

reactor during extended Station Blackout

(SBO) scenarios.

B. Creep Rupture Analysis of SG Tubes

Creep rupture model allows one to

specify a "stress multiplier" in following

meaning: a multiplier of 1.0 provides a creep

rupture failure prediction based on no

degradation of the structural strength of the

material and multipliers greater than 1.0

represent degraded structural strengths. In this

study, a stress multiplier from 1.0 to 5.0 (in

increments of 0.1) are used for SG tube

rupture prediction to investigate the spectrum

of material strengths, from undergraded to

highly-degraded.

ACKNOWLEGEMENTS

This work was supported by the

Ministry of Science and Technology (MOST)

R&D Project with contract number

05/HĐ/ĐTCB signed on 05/01/2018.

Typical results on Larson-Miller creep

rupture damage index for the hottest SG tubes

is presented in Fig. 9. It was found that the

average-tube predicted failure margin is 2.3

and the hottest-tube predicted failure margin is

1.7. Also, SG tube failure margin are

insensitive to variations in the fuel damage

progression behavior primarily because these

.REFERENCES

[1]. Federal Nuclear and Radiation Safety

Authority of the Russian Federation, Kalinin

VVER-1000 Nuclear Power Station Unit 1

PRA (Beta Project), NUREG/IA-0212,

published by US NRC, 1992.

14

NGUYEN VAN THAI et al.

[2]. Design, Safety Technology and Operability

Features of Advanced VVERs, OKB

“GIDROPRESS’’, 2011.

[4]. C.D., Fletcher et al. SCDAP/RELAP5 Thermal-

Hydraulic Evaluations of the Potential for

Containment Bypass During Extended Station

Blackout Severe Accident Sequences in a

Westinghouse Four-Loop PWR. NUREG/CR-

6995, published by US NRC, 2010.

[3]. S.,

Pylev.

Assessment

Study

of

RELAP5/MOD3.2 Based on the Kalinin NPP

Unit-1 Stop of Feedwater Supply to the Steam

Generator No. 4, NUREG/IA-0167, published

by US NRC, 1999.

[5]. C. Allison et al., “SCDAP/RELAP5/MOD2

Code Manual”, Volume: I, II and III,

NUREG/CR-5237, EGG-2555, June 1989.

15