STUDY ON PHASE CHANGE IN THE CORE OF NUCLEAR REACTOR

Thus, based on the material of the national projects code DTDL.2011-G/82 and KC.05.26/11-15 it is proposed a motivation of the study with the objective to predict void fraction predictio

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1 Introduction

Void fraction plays an important role in modeling of two phase flow

at component scale In CTF code, during solving the conservation equations, the void fraction is calculated Then, the flow regime is defined based on value of void fraction For example, flow regimes are determined based on the range of void fraction in normal wall models as illustrated in Figure 2.8 of this thesis The normal wall flow regime map includes the following flow regimes:

 Small-bubble defined by void fraction below 0.2

 Small-to-large bubble (Slug) defined by void fraction in range (0.2, 0.5)

 Churn/turbulent defined by void fraction in range (0.5 αcrit)

 Annular/mist defined by void fraction greater αcrit

Then each of the individual flow regimes of the normal wall map, the interfacial area, interfacial drag and interfacial heat transfer are defined differently

Figure 2.8 CTF normal-wall flow regime maps (source [38])

1.1 Thesis objectives

Void fraction study on flow through a channel with heating wall is important in thermal hydraulics analysis with a lot of experiments and codes developments Thus, based on the material of the national projects (code DTDL.2011-G/82 and KC.05.26/11-15) it is proposed

a motivation of the study with the objective to predict void fraction prediction in core of the VVER-1000/V392 nuclear reactor using

 To consider a combination of CTF and CFX (Ansys CFX) codes to improve void fraction predicted by CTF in specific timing within the transient period

In the issues mentioned above, the consideration of utilization of CTF and CFX codes to improve void fraction prediction in core is a new one As usually, CTF is used to predict void fraction during transient time It is expected that CFX with meso scale for void fraction prediction will give an improvement to predict for steady state in specific timing

1.1.1 Studied object

The void fraction in hot channel of VVER-1000/V392 reactor is predicted with different scales during 40 seconds of transient condition at the beginning of LOCAs with different break sizes

1.1.2 Scope study

It is also limited the scope of the study due to complexity of the two phase flow The investigated two phase flow through core sub channels is vertical flow with the specific regime such as bubbly, slug, churn and annular

1.2 Thesis outline

Thus, the thesis includes four chapters and the conclusion at the last The chapter 1 mentions about introduction that leads to motivation of this study with following arguments:

 Status of nuclear power in the World and Vietnam

 Brief overview of nuclear safety

 Core thermal hydraulics safety analysis in transient condition

 VVER technologies understanding in Vietnam related to this study

 Thesis objectives

 Thesis outline

Chapter 2 presents the methodology that is related to multi scale analysis along with the code theories at different scale for RELAP5,

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CTF and CFX with emphasize to phase change models in several items below:

 Multi scale approach to LWR thermal hydraulic simulation

 System code RELAP5

 Sub channel code CTF

 Meso scale code CFX

 Conclusions

The verification and assessment of modeling used in these codes in this study versus experiment data are presented in chapter 3 The system simulation results are compared with those in SAR documents (Belene project) The assessment of CTF code is implemented by simulation BM ENTEK experiment tests which is

an International Standard Benchmark to investigate boiling flow through Russian fuel bundle of RBMK reactor The meso scale code CFX is verified with PSBT single sub channel which is also an International Standard Benchmark Therefore the contents of chapter

3 are presented as following:

 Brief information of VVER-1000/V392 nuclear reactor

 Verification of RELAP5 simulation models for 1000/V392 reactor with SAR

VVER- Verification and assessment of CTF models with BM ENTEK experiment tests

 Verification CFX models with PSBT sub channel experiment tests

 Conclusions

The tasks related to thesis objectives mentioned in chapter 1 are solved in chapter 4 with following steps:

 Calculation Diagram

 Power distribution calculation by MCNP5 code

 LOCAs simulation by RELAP5 code

 Transient simulation in LOCAs by CTF code

 Steady state simulation specific timing of LOCAs by CFX code

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2.1 Multi codes and multi scales approach to PWR thermal hydraulic simulation

2.1.1 Neutron codes and thermal hydraulics codes

2.1.2 The different scale of thermal hydraulic codes

2.1.3 The different thermal hydraulic modeling approaches

2.2 Phase change models in system code RELAP5

2.3 Phase change models in sub channel code CTF

2.3.1 Evaporation and condensation induced by thermal phase change

2.3.2 Evaporation and condensation induced by turbulent mixing and void drift

2.4 Phase change models in meso scale code CFX

2.4.1 Evaporation at the wall

2.4.2 Condensation model in bulk of liquid

2.5 Conclusions

In summary, this chapter shows the detail of multi code and multi scale for PWR thermal hydraulics simulation, in especially for void fraction prediction The phase change of the codes used to predict void fraction in this thesis including RELAP5, CTF and CFX are briefly presented

3 Verification and assessment of phase change models by numerical simulation

The chapter 3 presents the thesis’ study on verification and assessment of phase change models by numerical simulation codes with different scales such as RELAP5 for the whole system of VVER-1000/V392 reactor, CTF for ENTEK BM experiment and CFX for PSBT single channel Then the assessment of CFX and CTF for void prediction based on PSBT is presented In briefly conclusions, the findings and achievements of this chapter are presented below

3.1 Verification of RELAP5 simulation models for 1000/V392 reactor with SAR

VVER-The purpose of this section is verification of RELAP5 modeling developed by this study for VVER-1000/V392 reactor Therefore, simulation results of this study are compared with those in SAR for LOCAs scenarios The comparisons focus on steady state results,

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timing of transient events and behavior of peaking temperature of cladding

3.2.2 Verification of modeling through steady-state study

The results show that the steady-state calculations are matched and

acceptable with design values in compare with SAR calculation in

[35] since the deviation between these two values is within 4%

Table 3.2 Comparison of steady-state of VVER-1000/V392

value

SAR

results[35]

The present

study

Deviation

(Percent) Reactor thermal power MW 3000 3120 3120 4.00%

Reactor outlet pressure MPa 15.7 ± 3 16.0 15.8 0.64%

Reactor inlet temperature o K 564.15 +2

-5 566.15 561.8 0.42%

Reactor outlet temperature o K 594.15±5 599.15 591.8 0.40%

Pressurizer (PRZ) level M 8.17 8.17 8.18 0.12% Steam Generator (SG) inlet

pressure - primary side

MPa 15.64 ± 0.3 15.94 15.75 0.70% Total volumetric flow rate at

reactor inlet

m3/h 86000+2600 82200 86029 0.03% SGs wide range level

measurement

M 2.7 ± 0.05 2.65 2.63 2.59%

Feed water temperature o K 493.15± 5 498.15 493.15 0.00%

Maximum fuel temperature o K 2078.15 2189.15 1843.14 0.84% SGs pressure at steam collector

outlet

MPa 6.27 ± 0.1 6.37 6.27 0.00%

3.2.3 Verification through accident case study

 The comparison results of chronological sequence of the main events for Event 3 are given in Table 3.4

Table 3.4 chronological sequence of the Event 3 from SAR [35] and

The present study Dev

0.00 MCPL guillotine break at the

reactor inlet

And Loss of off-site power

supply:

- startup of DG and safety

systems according to stepwise

startup program (failure of one

DG to start up and one DG is in

- startup of DG and safety systems according to stepwise startup program (failure of one DG to start

up and one DG is in repair)

- Trip of the systems for normal operation;

0

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0.03 Scram signal generation due to

Pressure above the core is below

14.70 MPa

0.08 Scram signal generation due to Pressure above the core is below 14.70 MPa

1.90 SCRAM signal generation 1.90 SCRAM signal generation 0

7.00 Beginning of injection into the

reactor from HA-1

6.00 Beginning of injection into the reactor from HA-1

1

17.0 Opening of valves in the pipelines

connecting HA-2 to the core due

to primary pressure decreases up

to 1.50 MPa

15.00 Opening of valves in the pipelines connecting HA-2 to the core due to primary pressure decreases up to 1.50 MPa

15

63.00 Termination of boron solution

injection into the reactor from

HA-1

63.00 Termination of boron solution injection into the reactor from HA-1

0

117.00 Beginning of supply from HA-2 115.00 Beginning of supply from HA-2 2

500.00 End of the calculation 500.00 End of the calculation 0

 The comparison of peaking temperature (PCT)

Figure 3.4 (a) Cladding temperature from calculations, (b) Cladding

temperature from SAR

Figure 3.4 also show the study’s calculations are similar to SAR

results in term of peaking cladding temperature and timing to cool

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down cladding temperature Therefore, the simulation model of this

study is appropriate with reference model from the SAR

3.3 Verification and assessment of CTF models with BM ENTEK

tests

 Along channel void fraction distribution discussion

Table 3.6 and Table 3.7 show the void fraction distribution

calculation results versus experiment distribution along the channel

Table 3.8 shows the deviation between void fraction distribution

calculation results versus experiment distribution It is observed that

CTF’s void fraction distribution predictions for base cases are good

agreement with experiment distribution with mainly deviation

around 0.03 of void The maximum deviations with value around 0.1

are occurred just one or two locations of the tests T04 and T08

Especially, for the five tests at 7 MPa (T17, T18, T22, T24 and T25),

the very good void fraction distribution calculations are agreed with

experiment distribution with deviation not more than 0.03 void along

the channel

Table 3.6 Base case void fraction distribution calculations versus

experiment for cases at 3MPa

Table 3.7 Base case void fraction distribution calculations versus

experiment for cases at 7MPa

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5.448 0.033 0 0.485 0.503 0.654 0.694 0.3973 0.406 0.4021 0.364 6.135 0.079 0.056 0.553 0.585 0.733 0.75 0.4834 0.512 0.5178 0.591 6.76 0.194 0.17 0.612 0.628 0.79 0.781 0.5585 0.564 0.6398 0.67

Table 3.8 Deviation between void fraction distribution calculation results

versus experiment

Z D(T01)* Heat

mode D(T04)

Heat mode D(T08)

Heat mode D(T10)

Heat mode D(T14)

Heat mode

Z D(T017) Heat

mode D(T18)

Heat mode D(T22)

Heat mode D(T24)

Heat mode D(T25)

Heat mode

D (T01) = (T01C-T01X), c = calculation, x = experiment

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It is found that CTF tend to under prediction when experiment void

fraction below 0.2 where the CTF’s modeling for normal wall flow

regime map is small bubble At the nearly outlet of the channel

where the experiment data are more above 0.2 corresponding to heat

transfer in saturated mode, CTF tend to over prediction Thus, CTF

boiling model is still needed to be improved for both sub cooled and

nucleate boiling regimes in order to generate more void in sub cooled

region and reduce void at nucleate boiling region

3.4 Verification CFX models with PSBT sub channel tests

3.4.6 Assessment of CFX and CTF modeling results in comparison

with PSBT single channel

It is clear that CFX give better void fraction prediction in small

bubble flow regime corresponding with sub cooled heat transfer

mode in CTF It is noticed that the Nusselt number correlation used

in CFX models for this case is Ranz Marshall

Table 3.24 CFX and CTF results comparisons versus experiment

void fraction in small bubble and sub cooled region

Power (kW)

Inlet Temp

Table 3.25 shows the void fraction calculations by CTF and CFX

versus experiment in saturated region with measured void fraction in

range of small-to-large bubbles and pressure lower than 122 bar It is

observed that CTF give the over prediction in this region while CFX

give the under prediction It is noticed that Nusselt number

correlation used in CFX models in this case is Warierr

Table 3.25 Comparison of CFX and CTF results and experiment void

fraction in saturated region

Power (kW)

Inlet Temp

Table 3.28 shows the void fraction and temperature superheating

calculated before and after calibration presented in columns “Void”,

“Tsup” and “Void*”, “Tsup*”, respectively It is noticed that Nusselt

number correlation used in CFX models in this case is Kim and Park

It is summarized the overall conclusions for verification and

assessment of phase change as following

- The system code RELAP5 with capability of modeling of whole

system related to heat removal from the core to ultimate heat sink is

utilized to simulate the reactor VVER-1000/V392 and related

systems With the purpose of verification of our RELAP5 modeling,

a scenario of LBLOCA in the confident document, Safety Analysis

Report (SAR) for Belene (Bulgaria) project is used to compare

simulations results Based on the two following arguments: (a) the

deviations of timing in chronological sequence of main events in

Table 3.4 with maximum of 15 seconds; (b) the behavior of

maximum of peak cladding temperature (PCT) in the first duration of

300 seconds with similar maximum values less than 1200 oC and

timing of cool down around at 280 seconds as illustrated in Figure

3.4, it is shown that simulation results given by this study are good

agreed with the results presented in SAR

- With regard to void fraction prediction in the core, the system code

RELAP5 is not confident tool in case of high equivalent diameter of

channel This conclusion is exposed from two issues: (a) RELAP5 is

1D code, so that the average or hot channel in the core is simulated

as a pipe, that means the geometry of the core is not simulated as

reality and the flow regime map in modeling may be different from

- CFX gives the good prediction when void fraction below 0.2, corresponding with sub cooled heat transfer mode in CTF, with deviations around 0.03 of void So that, results in sub cooled region

it is recommended to use void fraction predicted by CFX instead of CTF

- For the saturated region corresponding with small-to-large bubble flow regime, CTF tends to give over void prediction and CFX tends

to give under void prediction, Then, it is considered CFX and CTF results as lower and upper bounds for void fraction prediction along the channel

- The improvement CFX simulation results in saturated region by scaling bubble departure diameter and maximum area fraction for quenching effect brings a new approach to continue development of RPI boiling models for saturated region

4 Void fraction prediction in hot channel of VVER-1000/V392 4.2 Power distribution calculation by MCNP5 code

The calculation results are based on whole core geometry simulation and the number of neutron histories equal 2.107 (with relative variation for keff around 10-5) The relative power for each assembly

in 1/6 of the core is presented in Figure 4.5 Thus, the hot channel is

an assembly with identification of 30A9P corresponding to maximum relative factor of 1.72 This value of power distribution is appropriate because it is within the range of (1.6, 1.8) mentioned in [34]