Overview of BARC severe accident analyses for VVER · PDF fileOverview of BARC severe accident...

4th European Review meeting on Severe Accident Research (ERMSAR-2010) Bologna (Italy), May 11-12, 2010

Overview of BARC severe accident analyses for VVER-1000 and

Pressurised Heavy Water Reactors

MITHILESH KUMAR, P. MAJUMDAR, B. CHATTERJEE, D. MUKHOPADHYAY, H. G. LELE,

A. K. GHOSH AND H. S. KUSHWAHA

Reactor Safety Division, Bhabha Atomic Research Centre (BARC)

Mumbai 400085 - India

ABSTRACT

There is an Indo-French bilateral collaboration programme for joint R&D studies in Nuclear Safety. BARC and IRSN have identified a common area of interest for R&D in the field of Severe Core Damage Studies for Nuclear Reactors. Program involves development, validation and application of severe accident codes like ASTEC, ICARE, ICARE/CATHARE with emphasis on both PWR and PHWR reactors.

The ASTEC [1, 2] code, jointly developed by IRSN (France) and GRS (Germany) has detailed models of core degradation, fission product transport in circuit and containment, iodine chemistry, corium concrete interaction, hydrogen recombiners etc. which are currently undergoing intense development and validation.

Several analyses are carried out for VVER 1000 reactor with objective of evaluation of severe core damage progression, source term estimation and development of severe accident management

PHWR analysis is done for Limited Core Damage Accident (LCDA), results are compared with RELAP/SCDAP calculations and limitations are brought out.

This paper describes some of the analyses pertaining to VVER and PHWR using ASTEC and inter-code comparisons.

1 INTRODUCTION

India has number of reactors with diversity in their characteristics. Numbers of activities are underway in the field of experimental and analytical research for both designs basis and beyond design basis accident including severe accidents covering different nuclear power plants under design, construction, and operation and planning. This paper highlights salient predicted research findings obtained using French code ASTEC. It describes severe accident analyses performed for PHWR and VVER 1000. Objective of severe accident analyses for PHWR is inter-code comparison which may serve as input to program for adaptation of ASTEC for PHWR analysis. Some of the analyses for VVER 1000 are presented to so as to cover accidents in all categories and examine severe accident management strategies:

2 VVER ANALYSIS

VVER analysis involves high pressure scenario, low pressure scenario with spectrum of break sizes, interfacing LOCA like steam generator tube rupture considering influence of late high

Session 2 “ASTEC”, Paper 2.2 1/19


pressure injection and severe accident measures like secondary side injection at different entry temperatures.

2-1 Brief Description of VVER 1000

A VVER-1000 is a 3000 MWth Pressure vessel type reactor with four loop system. The reactor core consists of 163 hexahedral Fuel Assemblies (FAs) and each FA comprises 312 fuel elements and 18 control rods. There are 61 Control Rods Assembly (CRA). The primary circuit consists of reactor core, downcomer, lower plenum, upper plenum and four loops. Each loop contains a reactor coolant pump (RCP), hot leg, cold leg and horizontal steam generator (SG). The pressuriser is connected to hot leg of 3rd loop and the spray lines of pressuriser are connected to cold leg of 4th loop. Three Pulse Safety Devices (PSDs), one control and two safety PSDs, are mounted on the pressuriser relief line. The secondary circuit consists of four loops and each loop is having a horizontal SG and four steamlines. On each steamline two Pulse Safety Devices (PSDs), one Main Steam Isolation Valve (MSIV) and one atmospheric steam Discharge valve (BRUA) are mounted. These valves (BRUAs) dump steam in the atmosphere at a pressure threshold of 7.16 MPa and close at a pressure of 6.13 MPa.

2.2 Computer Code ASTEC and VVER-Plant Model

The computer code ASTEC V1, jointly developed by IRSN, France and GRS, Germany, is used for analyzing the transient. ASTEC V1 is used as reference code for French PSA2 studies on PWR900 and 1300 reactors. However, because of its general structure, it can be applied on other reactor types like German KONVOI or Russian VVER 440 and VVER1000 as shown in the SARNET European network of excellence [1, 2]. Because of its modularity, it is also able to simulate many experiments like PHEBUS, CORA and QUENCH for instance [3]. ASTEC V1 is composed of a set of modules which can be used either in stand-alone mode or coupled together. CESAR module does the thermal hydraulic calculations in RCS. DIVA module calculates the core degradation. ELSA module computes the Fission Product (FP) release from the core. SOPHAEROS deals with the FP transport (aerosol and vapour) within the RCS. The containment thermal-hydraulics and fission product transport are calculated with CPA module and other ex-vessel phenomena are simulated with MEDICIS and IODE.

The ASTEC specific VVER-1000 plant model has been developed with modules namely CESAR, DIVA, SOPHAEROS and CPA. No other modules are involved as the study is specific to in-vessel degradation. The modelling is divided into two parts, namely geometric and physical modelling, that are described below.

2.2.1 Geometrical Modelling

The input model includes the description of the following nuclear plant “sub-components” for CESAR module. Nodalisations of RCS and secondary side are given in Fig.1.

Primary loops: Four loops have been modelled independently. Each loop has been modelled by 7 volumes and 8 junctions representing hot leg (HL), SG hot collector (SG_HCOL), SG tubes (SG_TUB) and SG Cold collector (SG_CCOL), cold leg (CL-P1, CL-P2 CL-P3) and a Main Circulating Pump (ULEGP).

(i) Pressuriser: Pressuriser (PRZ), surge lines (SRGLIN1) and its connections to loop1 hot leg have been modelled with the appropriate number of volumes and junctions. Three relief valves connected to pressuriser have been modelled.



(ii) SG secondary side: All four Steam Generators (SGs), Steam lines, Steam Header, Feed water pumps and BRU-A valves have been modelled. These valves dump steam in the atmosphere at a pressure threshold of 7.16 MPa and close at a pressure of 6.13 MPa.

(iii) Upper Plenum: Upper plenum has been modelled by 2 volumes (UPPLE1&2).

(iv) Downcomer: Downcomer has been modelled by one separate volume (DOWNCO1).

(v) Hydro Accumulators: Two accumulators have been represented by accu1&2 and connected to upper plenum. Two other accumulators represented by accu3&4 have been connected to downcomer.

Core_A

to Core_J

UPPle1

Dow

nco1

CL1P3

CL1P2

CL1P1

SG1_C

CO

L

SG1_TUB

HL1

SR

G_LIN

E

PRZ

UPPle2

core_vin

Dow-Cor

Loop1j8 (8.523)

Loop1j6

Loop1j4

Loop1j3

Loop1j3

Loop1j2

Loop1j1 INRJ7

Upp_cor (6.733)

SRGLIN1

SRGLIN2

ULEGP1(Pump)

REG_PR

volinf

SG

1_HC

OL

Loop1j7

Loop1j5

PRZ Relief Valves

To Loop 2, 3, 4 From Loop 2,

3, 4

ACCU3&4 ACCU1&2

BRU-A

To Turbine

SG1_SEC(SG)

SG2_SEC (SG)

SG3_SEC (SG)

SG4_SEC (SG)

BRU-A

Main Steam

Feed Flow

Main Steam Line

Safety Valves (SG-PSDs)

BRU-K



BRU-A BRU-A

Fig.1: Nodalisation of Primary and Secondary System

The reactor vessel structures are modelled with the DIVA module which includes reactor core, baffle, the cylindrical part of the barrel, vessel and fuel assembly supports. Following are the description of the models of each component.

(i) Reactor Core: The core consists of 163 hexagonal fuel assemblies. These fuel assemblies have been distributed in 5 radial rings with a radial power profile. Each ring contains different numbers of fuel rods, guide tubes with and without control rods as furnished in Table 1. The reactor core has been discretised axially into ten identical meshes with an axial power profile for the fuel rods.

Table 1: Core Zone Model

Fuel Rods

Guide Tubes with Control Rods

Guide Tubes without Control Rods

Radial Power Profile

ZONE-1 10171 220 400 1.0

ZONE-2 10171 220 400 0.928

ZONE-3 10171 220 400 1.203

ZONE-4 10171 220 400 1.045

ZONE-5 10172 218 399 1.045

(ii) Baffle & Barrel: The annular cylindrical baffle with small axial penetrations facilitates the coolant bypass (3% of total flow) to take place. The bypass coolant path is modelled. with



the thermal-hydraulic components of the DIVA module The structural material of the baffle and barrel are modelled with the shroud macro component

(iii) Vessel: This consists of three parts namely upper head, cylindrical portion and the lower head. The cylindrical part which extends from lower grid plate to the fuel top elevation is modelled with shroud macro component where as the lower head is modelled with ‘lowple’ macro-component.

(iv) Fuel Assembly Supports: 163 fuel assembly supports are modelled with ‘cylinder’ macro-components in the vessel lower plenum.

2.2.2 Physical Modelling

Conductive Heat transfer in the solid components is 2D and activated through ‘COND’ block data. In this block data, faces of the same or different macro components are defined which are involved in the heat exchange. Similarly the fluid channels and solid component faces involved in convective heat exchange have been defined in the ‘CONV’ block data. Radiative heat exchange is modelled through DIVA specific ‘RADC’ block data for large core radiation. In this model, it is assumed that the core is discretized into many concentric rings and the fuel rods are the most numerous and control rods are sparsely distributed. The radiative and conductive heat exchange in the gap between fuel and clad and in the gap between the control rod clad and guide tube are taken into account through the ‘GAP’ block data. Physical models used for simulating different phenomena and reactions have been given in Table 2.

Table 2: Physical models

Components

Involved Correlation

Zircaloy Oxidation by Steam

Clad, control rod guide tube, grid spacers

URBANIC

UO2 and Zirconia dissolution by molten Zircaloy

Fuel Element

KIM-CONV

Chemical kinetics

Steel oxidation by steam Control rod cladding, Lower grid plate

MATPRO (J.F. WHITE Correlation)

Corium relocation Candling velocity Fuel, clad, control

rods 0.6 m/s

Fuel rod cladding melting Zircaloy clad 2400 K

Fuel relocation occurs Fuel 2550 K *

Relocation of control rod cladding Control rods 1730 K

Vessel bottom head failure Lower head 1473 K

Integrity Rules

Vessel bottom head failure Lower head 150 MPa



* This temperature, about 300 to 400 K lower than the pure material melting one, is issued from the interpretation of the PHEBUS FP integral experiments.

Material Movements: materials in the molten state are first transferred radially to the external surface of structures before they flow down along these structures. Radial movement is defined through DECA block data and axial candling through CAND block data. Following radial movements are considered in the modelling (Table 3).

Table 3: Radial material movement

Radial Material movement

From To

Fuel Clad External surface

Control Rod, Control rod clad and Guide tube Guide tube external surface

Lower head Corium layers in lower plenum

Fission product release has been modelled by ELSA module of ASTEC. Fission product transport in the Primary loops has been modelled by SOPHAEROS module. Steam and hydrogen have been defined as carrier fluid for fission products.

2.3 Results and Discussions

Following case have been analysed covering high pressure scenario, low pressure scenario and containment bypass scenario:

- LOCA with Spectrum of Break sizes from 50 mm to 850 mm in cold leg

- LOCA with Spectrum of Break sizes from 50 mm to 850 mm with HA injection

- Steam Generator Tube rupture

- Station Blackout (SBO)

- SBO with steam generator flooding at core exit temperature of 650 0C

- SBO with steam generator flooding at core exit temperature of 980 0C

- SBO with pressuriser valve stuck open

- SBO with pressuriser valve stuck open and Accumulator injection

LOCA scenarios are simulated along with station blackout. Scenario involves initiation of break, tripping of primary and secondary pumps, failure of Secondary atmospheric discharge (BRU A), actuation of safety valve and initiation and stopping of passive accumulators. Spectrum of breaks in cold leg is evaluated. Scenarios are evaluated with and without accumulator injection. Behaviour of thermal hydraulic parameters like system pressure, temperature and flows, and core degradation parameters like hydrogen generation, corium mass and its composition, aerosol generation and transport, timings of vessel failure are obtained and compared. Extent of core degradation for different breaks at entry condition to SAMG is also evaluated and compared for different break sizes. Details are given in Ref. [4,5]. Steam Generator Tube Rupture considers single tube rupture, reactor trip and subsequent primary and secondary pump trips at about 600 seconds and stuck open atmospheric discharge valves. Its thermal hydraulics and core degradation behaviour is slower than 50 mm break in old leg. It involves fission product release bypassing containment and retention in SG tubes.



Behaviour of corium mass and hydrogen generation for spectrum of breaks in cold leg with and without accumulators is indicated in Fig. 2 to Fig. 5. Core degradation at entry condition to SAMG, time margins available and recommendations are summarized in Table 4. Different core degradation parameters for SGTR are summarized in General Table 5, while typical fission product release through BRU A is given in Fig. 6.

Table 4: Different Inputs to SAMG from analysis

Cases Beginning of Core Oxidation (s)

Time to attain 650°C (s)

Extent of Core Damage at time 650°C is reached

(% liquefied UO2)

RCS Pressure

(MPa)

Time margin to start SAMG (s)

Comments

on SAMG

With Hydro-accumulators

850 mm Break 502 1582 3% 0.24 -1080

150 mm Break 6184 1611 0% 1.06 4573

50 mm Break 9341 5865 0% 4.02 3476

Without Hydro-accumulators

850 mm Break 50 550 1.6% 0.193 -500

150 mm Break 611 822 0% 0.7 211

50 mm Break 3956 4987 0% 3.7 1031

(a) In core Pumped Injection Requires Min Head of 4.06 MPa

(b) Time Margin for In-core injection is very less for Large LOCA

0 4000 8000 12000 16000 20000

0

20

40

60

80

100

Mas

s (t.

)

Time (s)

50 mm w/o HAs 150 mm w/o HAs

850 mm w/o HAs

Fig.2: Cumulative Corium Mass in Lower Header (w/o HAs)



0 5000 10000 15000 200000

100

200

300

400

500

Mas

s (k

g)

Time (s)

50 mm w/o HAs 150 mm w/o HAs 850 mm w/o HAs

0 6000 12000 18000 24000 30000 36000

0

100

200

300

400

500

600

Mas

s (k

g)

Time (s)

50 mm with HAs 150 mm with HAs 850 mm with HAs

Fig. 3: Cumulative H2 Gen. w/o HA Fig. 4: Cumulative H2 Gen. with HA

0 6000 12000 18000 24000 30000 36000

0

20

40

60

80

100

Mas

s (t.

)

Time (s)

50 mm with HAs 150 mm with HAs 850 mm with HAs

30000 40000 50000 600001E-6

1E-5

1E-4

1E-3

0.01

0.1

1

10

100

Mas

s (k

g)

Time (s)

Ba (Release from BRUA valve) Mo (Release from BRUA valve) Ba (Core release) Mo (Core release)

Fig. 5 Corium Mass in Lower Head (with HA) Fig. 6: FP Release for SGTR

The Station Blackout scenario comprises of initiation of transient with tripping of primary and secondary pumps, reactor trip, primary and secondary pressurization, opening and closing of pressuriser relief valve and core degradation and vessel failure. The analyses have been done with pressuriser valve stuck open also. Influence of steam rich environment [6] and implementation of severe accident measures at different entry criteria are also examined. Fig. 7 shows system pressure behaviour for station blackout. Core Temperature pattern and core exit temperature are depicted in Fig. 8 and 9. Influence of different entry criteria to SAMG on hydrogen generation can be seen from Fig. 10 and Fig. 11.




0 10000 20000 30000 40000 500004

6

8

10

12

14

16

18

20

Pres

sure

(MPa

)

Time (s)

RCS Pressure SG Pressure

SG Boil Off

RCS Pressurisation

Chattering of PRZ Relief Valve

Lower Head

Core Temperature FieldDegraded Fuel RPV Wall

Core Barrel

Corium

Lower Head


Core Barrel

Corium

Lower Head


Core Barrel

Corium

Fig.7: System Pressures Fig.8: System Temperatures

0 10000 20000 30000 40000 50000 60000 70000400

500

600

700

800

900

1000

Tem

pera

ture

(K)

Time (s)

Core Exit Temperature

0 20000 40000 60000 80000

0

10

20

30

40

Mas

s (k

g)

Time (s)

Cumulative Hydrogen Generation

Fig. 9: Core Exit Temperatures Fig. 10: Hydrogen Gen. (6500C SAM entry)

0 10000 20000 30000 40000 50000 60000

0

20

40

60

80

100

Mas

s (k

g)

TIME

Cumulative Hydrogen Generation

Fig.11: Hydrogen Generation (9800C SAM Entry)


Table 5: Comparison of Different Severe Accident Parameters during Different Scenarios

Accident Chronology

850 mm 150 mm 50 mm Phenomena

w/o HAs with HAs

w/o HAs

with HAs

w/o

HAs

with HAs

SGTR SBO SBO PRZ valve open

SBO with PRZ valve open and HA

SBO with SAM entry at 6500C

SBO with SAM Entry at 9800C

Beginning of oxidation (s)

50 502 611 6184 3956 9341 30589.0 14114 9664 9664

Total core uncovery (s)

18 502 1046 10252 5187 11584 35256.9 25116 10342 29499

1st fuel cladding rupture (s)

174 635 903 6760 4801 9341 33248.0 17660 10477 10915

Start of FPs release from fuel pellets (s)

174 636 905 6761 4803 9342 33248.9 17660 10479 10917

1st slump of molten materials in vessel lower head (s)

420 790 1704 8869 5808 11489 35773.0 25595 11286 29533

1st slump of corium with FPs in lower plenum (s)

711 1208 2168 9798 6759 16270 35773.0




Table 5: Comparison of Different Severe Accident Parameters during Different Scenarios (continued)

Core Degradation

850 mm

150 mm

50 mm SGTR SBO SBO PRZ valve open

SBO with PRZ valve open and HA



Parameters

w/o HAs

with HAs

w/o HAs

with HAs

w/o

HAs

with HAs

Hydrogen Generation (kg)

52.4 249.7 259.8 385.4 416.4 578.3 590.8 742.0 274.5 468.5 40.0 96.0

Corium mass in lower head (t.)

87.12 77.7 86.81 51.3 85.7 80.3 84.4 33

(76 %)

69.68 64.86 0.0 0.043

(0.1 %)

Aerosols mass produced (kg)

937.4 950.5 959.9 866.1 959.9 902.5 1201.9 885 212.6 320.0

Lower head vessel failure (s)

7520 10255 10897 23580 16605 35824 55127 48678 24509 46801 No vessel failure

No vessel failure


2-2 Concluding remark for VVER Analysis

Core degradation process is slower for high pressure scenario and it takes longer time for vessel failure. Corium formation is also smaller compared to low pressure scenario (LOCA) however hydrogen generation is very large. By implementation of SAMG like injection of water in the secondary side, at specified core exit temperature, it is possible to avoid or restrict corium formation and reduce hydrogen generation substantially.

In case of LOCA scenario, hydrogen generation is smaller and corium mass is larger. Corium mass formed is increases with break size but is of similar order for break sizes from 500 mm to 850 mm. However vessel failure time and hydrogen generation varies inversely with break size. Fission product release is also smaller for larger breaks due to smaller flows during degradation phase.

Accumulator injection provides steam rich environment resulting in increase in hydrogen production. It also delays vessel failure.

3 PHWR ANALYSIS

ASTEC code was being developed for Light Water Reactors, hence its present version has a limited scope. DIVA/ICARE modules of ASTEC do not have appropriate modules to simulate the thermo-mechanical deformation of Reactor Channels components and their horizontal geometry. Hence its is has limitations in simulating 2-D temperature distribution for Reactor Channels, Clad Ballooning, High temperature metal–steam Reactions & Fission Product Release.

However CESAR, SOPHAEROS and CPA modules are expected to simulate Thermal-Hydraulics of Parallel Channel Loop of the Circuit, FP transport in parallel channel loops of the circuit, containment Thermal-Hydraulics and aerosol transport. In this section predictions for LOCA with loss of ECCS scenario for CESAR simulation are compared with a RELAP5 analysis. Results for standalone simulation of DIVA for same scenario are also presented.

3-1 Brief Description of PHWR

The Primary Heat Transport (PHT) System of the Indian PHWRs (Fig. 12) of 220 MWe design is in the shape of a “Figure of Eight“ with each leg having an inlet header providing coolant to half of the channels of the reactor core and an outlet header receiving the hot coolant from the reactor channels. There are 306 horizontal coolant channels in the Indian PHWRs of 220 MWe design. Each Channel consists of 12 fuel bundles contained in a pressure tube, which is surrounded by a calandria tube. The annulus between the pressure tube and the calandria tube contains gas, which provides the required thermal insulation under normal operating conditions. The calandria tube is submerged in the relatively cold moderator. The primary coolant flows inside the pressure tube containing the 19-rod fuel bundle.



Reactor

IH IHO H O H

text

SteamG enerators

M oderatorPum p

M oderator HeatExchanger

IH – Inlet HeaderO H – O utlet Header

Calandria

Steam pipes

Fuel

Heat TransportPum p

Header

Fig. 12: Schematic of Pressurised Heavy Water Reactor

3-2 Modelling Using RELAP5 AND ASTEC codes

The reactor Primary Heat Transport (PHT) system, as shown in Fig. 13, has been modelled in to the one average path and one hot channel path in each pass. Out of the 153 channels in a pass, 152 channels are clubbed to form an average channel and a single hot channel producing 3.2 MW power is separate channel. The other pipe components between RIH (Reactor Inlet Head) and ROH (Reactor Outlet Head) such as inlet and outlet feeders are clubbed in accordance to the above numbers. Proper care has been taken to input hydrodynamic parameters such as hydraulic diameter, flow area, volume, bends, elevation and length. Pressure drops arising due to spacers in the fuel bundle and complex flow geometry at either ends of the reactor are modelled with proper loss coefficient values used in the respective junctions. The two parallel paths between ROH and RIH have been lumped into an equivalent single path. The two pumps and the two steam generators present in the system are clubbed into an equivalent single pump and a single steam generator.

The Steam Generator (SG), as shown in Fig. 14, riser portion is modelled with 5 vertical volumes. The economiser is modelled with a single volume. The separator is modelled with RELAP5 specific separator model. The downcomer portion is divided in 4-volume using pipe component. Steam drum is modelled with pipe component having 5 volumes.

Nodalisation for standalone simulation of DIVA and SCDAP is given in Fig. 17.



100(1)

103(10)

200(1)

107(1) 109(1)

300(1)

117(1) 119(1)

114(10)

400(1)

102(1) 104(1)

106(2)

108(

4)

110(2)

111(1)

112(1)

113(1)115(1)

116(2)

118(

4)

120(2)

121(1)

122(1)

301

302 303

304

305

306

307308

309

312

313

314315

316

317

318

319320

321

322

905

907

955

956

957

958

123

124

904

906

908

959

214(10)

203(10)215(1) 213(1)

202(1)

204(1)327

328329

330

324 325

326323

600Containment

Volume

Break Junction201

Fig. 13: Primary Heat Transport System Nodalisation for RELAP5

701

704 712

703-1

703-2

703-4

108

705

702710

706-1 711

107 109

SDVs

756751

754

755

752753

701

704 712

703-1

703-2

703-4

108

705

702710

706-1 711

107 109

SDVs

756751

754

755

752753

Fig. 14: Nodalisation scheme for Steam generator of PHWR in RELAP5



Results

Scenarios

Code Developmental Activitiers

SG Tubes

Inlet and OutletPlenum

Pump

Inlet Header

Outlet Header

FeedBleed

chan

nel

Inle

t Fe

eder

outle

t Fe

eder

Fig. 15: Primary Heat Transport System Nodalisation for ASTEC

Feed

W

ater

To

Turb

ine

Steam DrumVolumes

Steam Lines

Fig. 16 Nodalisation scheme for Steam generator of PHWR in ASTEC



Calandria Tube

Pressure Tube

Fuel Bundle

Power Distribution in Rings

Inner : 1

Middle : 1.1

Outer : 1.4

Fig. 17: DIVA and SCDAP Simulation of LCDA

3-3 Results and Discussions

Various assumptions made during analysis are:

- Uniform channel power profile,

- In absence of reactor kinetics in ASTEC, the power transient is obtained from RELAP5 analysis of 200 % inlet header break and used in ASTEC,

- Actual PHT pump characteristics were not simulated. Instead it is assumed that pumps stop developing head in 140 sec after trip,

- Emergency Core Cooling System is not modelled,

- Clad is not simulated,

- No structural heat loss.

The postulated scenario involves the break initialization and tripping of primary circulating and Steam generator feed pumps after steady state for 60 seconds, reactor trip on high power and stopping of primary feed pump after emptying of storage tank.

Reactor power increases to about 2.6 times from its steady state value and decreases after reactor trip as seen in Fig.17. The power is issued as a boundary condition for both analyses. Fig. 18 and Fig. 19 compare predictions for break flow and system pressure. Trends are found to be similar. Differences may be due to different critical discharge models.



40 45 50 55 60 65 70 75 80 85 90 95 100

0

200

400

600

800

1000

1200

1400

Rea

ctor

Pow

er (M

W)

Time (sec)

Fig. 17: Reactor Power

46 48 50 52 54 56 58 60 62 64 66 68 70

-2000

0

2000

4000

6000

8000

10000

12000 CESAR RELAP5

Bre

ak fl

ow ra

te (k

g/s)

Time (sec)

Fig. 18: Break Flow Rate Transient



46 48 50 52 54 56 58 60 62 64 66 68 700

2

4

6

8

10

12

14

Inle

t Hea

der P

ress

ure

(MPa

)

Time (sec)

CESAR RELAP

Fig. 19: Inlet Header Pressure Transient

Stand-alone simulation of a single PHWR hot channel for the above LCDA event is done using boundary conditions generated by global analysis. Predictions are made using both DIVA and SCDAP. Following assumptions are made in simulations:

- Calandria tube outer surface is maintained at 650C,

- System is at atmospheric pressure,

- Very low flow of steam (3g /s) condition,

- Decay power condition (2% of full power) and uniform power profile.

The various models used in the simulations are:

- Radiation between the fuel rings and inner surface of pressure tube,

- Radiation heat exchange between Pressure tube and calandria tube,

- Convective heat exchange with steam mixture,

- Metal water reaction and consequent hydrogen production,

- Solid and Liquid chemical Interaction between clad and fuel,

- Clad integrity.

The prediction of temperatures in different fuel rings and in pressure tube (PT) and calandria tube (CT), as computed by DIVA, are given in Fig. 21 and Fig. 22. Fig. 23 shows SCDAP predictions for maximum clad surface temperature and pressure tube and calandria temperatures. There are differences in the predictions between both codes that are being currently studied.



Fig. 21: Fuel ring temperatures using DIVA Fig. 22: PT and CT temperatures for DIVA

0 500 1000 1500 2000 2500300

400

500

600

700

800

900

1000

1100

1200

Tem

pera

ture

(K)

Time (s)

Fuel Pressure Tube Calendria Tube

Fig. 23: Maximum Fuel Ring Temperatures and PT and CT temperatures for SCDAP




4 CONCLUSIONS

A number of analyses have been performed for VVER-1000 reactor which can be helpful in understanding severe accident phenomena, source term estimation and evolving severe accident management strategies.

Limited core damage scenario for PHWR is simulated using both ASTEC and RELAP5 codes with certain set of assumptions. Results are comparable for the thermal hydraulic behaviour in the Primary Heat Transport system. Differences in the predictions for single channel behaviour are being studied.

REFERENCES

[1] J.P. Van Dorsselaere, C. Seropian, P. Chatelard, F. Jacq, J. Fleurot, P. Giordano, N. Reinke, B. Schwinges, H.J. Allelein, W. Luther, “The ASTEC integral code for Severe accident simulation”, Nuclear Technology, NT-3-0846, vol. 165, March 2009.

[2] J.P. Van Dorsselaere, J.C. Micaelli, H.J. Allelein, "ASTEC and SARNET - Integrating Severe Accident Research in Europe", ICAPP ’05, Seoul, Korea, May 15-19, 2005.

[3] G.Repetto, , B. Clement, et al., “Analysis of the FPT0, FPT1 and FPT2 Experiments of the PHEBUS FP program investigating in vessel phenomena during a LWR accident”, In: Proceedings of the NURETH-10, Seoul, Korea, 2003.

[4] B. Chatterjee, D. Mukhopadhyay, H.G. Lele, A.K. Ghosh, H.S. Kushwaha, P. Groudev, B. Atanasova, “Analyses for VVER-1000/320 reactor for spectrum of break sizes along with SBO”, Annals of Nuclear Energy 37 (2010) 359–370.

[5] B. Chatterjee, D. Mukhopadhyay, H.G. Lele, A.K. Ghosh, H.S. Kushwaha, P. Groudev, B. Atanasova, “Effect of steam environment on severe core damage behaviour for VVER-1000 with the ASTEC V1 code”, Nuclear Engineering and Design 239 (2009) 559-565.

[6] B. Chatterjee, D. Mukhopadhyay, H. G. Lele, A. K. Ghosh, P. Groudev, B. Atanasova, “Influence Of Steam Generator Inventory On Severe Core Damage Behaviour During SBO For VVER 1000 Reactor”, Nuclear Power for the People Conference, November 11-14, 2009, Veliko Tarnovo, Bulgaria

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