Plant Transients under Small Abnormalit ies of FBR … Transients under Small Abnormalit ies of FBR...

57 ISSN-1883-9894/10 © 2010 – JSM and the authors. All rights reserved.

E-Journal of Advanced Maintenance Vol. 6 (2014) 57-70 Japan Society of Maintenology

Plant Transients under Small Abnormalities of FBR “Monju” Calculated by a Plant System Code Hiroyasu MOCHIZUKI1,* 1 Research Institute of Nuclear Engineering, University of Fukui, 1-2-4 Kanawa-cho, Tsuruga, Fukui 914-0055, Japan ABSTRACT The objectives of the present study are to analyze plant transients caused by small abnormalities and to find plant parameters by which operators can recognize these small abnormalities. In order to evaluate the plant transient during an abnormal situation in the water system using the plant system code NETFLOW++, the turbine and feedwater (FW) systems should be analyzed with good precision. The code is validated using the measured data at “Monju”. Several abnormalities in the water system are candidates of the present study, e.g., FW control valve degradation, FW pump degradation, heat transfer degradation due to fouling on heat transfer tubes of the evaporator, loss-of-feedwater-heating, etc. All major components in the tertiary system are included in the calculation model such as the steam generators, the high-pressure turbine, the deaerator, the FW pump, the FW heaters, the FW control valves, the steam control valve, extraction lines and drainpipes. In case of a malfunction of a FW control valve resulting in low flow rate, a large temperature increase at the outlet of the evaporator is observed. On the other hand, a temperature decrease at the outlet of the evaporator occurs if heat transfer tubes in the evaporator have fouling. As a result of the calculations, it was determined that temperature at the outlet of the evaporator is a good indicator to detect abnormal situations.

* Corresponding author, E-mail: [email protected]

KEYWORDS

Monju, FBR, Plant system analysis, Water system, Slight abnormality, Abnormality indicators

ARTICLE INFORMATION

Article history: Received 15 January 2014 Accepted 20 October 2014

1. Introduction

When a large transient like a main pump trip occurs in a Fast Breeder Reactor (FBR), the reactor should be scrammed within a couple of seconds. However, if the pump undergoes a small degradation, plant parameters will change gradually and may reach the scram level in some cases. The objectives of the present study are to analyze plant transients caused by small abnormalities and to deliver the results to the group members who are studying the detection of the plant abnormality. The other objective of the present study is to find plant parameters which operators can use to recognize these abnormalities. In case of the FBR “Monju”, the fluctuation of the process data in the primary system at 45% thermal power (40% electric power) were studied by Udagawa et al. [1]. However, since the operation analyzed in reference [1] was steady state, meaningful results concerning off-normal situations were not obtained. Santhosh et al. [2] studied process data in order to develop a system that would assist the operator to identify abnormal transients at the earliest stages of their development. They focused on a large break LOCA. The present study is focusing on abnormal predictors which operators can use to mitigate the conditions.

There are few studies related to the water loop for FBRs. Therefore, the present paper focuses mainly on the abnormalities in the water loop consisting of turbine and feedwater (FW) systems. In order to evaluate the plant transient under an abnormal transient in the water system using a plant system code, the turbine and FW systems should be analyzed with good precision. Mochizuki [3] developed the system code NETFLOW++ which can be applied to the above situation and Mochizuki and Tsukamoto [4] analyzed the turbine and FW systems of the “Fugen” reactor which was a prototype heavy-water-moderated, light-water-cooled pressure-tube-type reactor in Japan. The following abnormalities in the water system are candidates of the present study: 1) FW control valve

H. Mochizuki / Plant Transients under Small Abnormalities of FBR “Monju” Calculated by a Plant System Code

58

degradation, 2) FW pump degradation, 3) steam control valve degradation, 4) heat transfer degradation due to fouling on heat transfer tubes of the evaporator, 5) degradation of FW heating. Relating the sodium system, a main pump abnormality is simulated. 2. Calculation Code and Model 2.1 Brief code description

The system analysis 1D code NETFLOW++ (Mochizuki [3]) can calculate single-phase and two-phase flows of water. For liquid sodium, one-dimensional single-phase incompressible flow is assumed. The followings are momentum and continuity equations for flow segment i. The following is the brief introduction of the basic equations to calculate the flow in the plant system. Please refer to the above paper in order to understand the code precisely. 2.1.1 Momentum equation

Because of non-compressible flow, the momentum equation is simplified as follows.

Piiiii

ii

i

iiiidiu

ii gHgH

AWW

Delpp

dtdWL ρρ

ρlζ +−

+−−= 2,, 2

)( (1),

If there is no pump in the interested segment, the last term should be omitted. The pump characteristics such as rotation speed, head and flow rate under steady state and transient conditions are solved by a pump kinetic equation with inertia, the torque of the motor and frictional torque using a characteristic curve between flow rate and head and an efficiency curve. 2.1.2 Continuity equation

Mass flow rate should be preserved at each main joint (junction) j.

0=∑∈ jCi

iW (2),

jCi ∈ : a set of flow passage i linked to junction j 2.1.3 Energy equation of the secondary fluid in heat exchanger

( ) 2

2

zTkTT

AK

zTW

AC

tTC s

lsts

sss

s

ssss ∂

∂+−

′=

∂∂

′+

∂∂ρ (3),

In case of flow in a pipe, this equation is neglected. 2.1.4 Energy equation of heat transfer tube

( ) ( )tst

stp

t

pttt TT

AKTT

AK

tTC −

′+−

′=

∂∂ρ (4),

pfpsp

p

tpp

tubep

hdddd

khd

NK

,

12ln

211

+

++

=p (5),


59

sfss

sp

tss

tubes

hdddd

khd

NK

,

12

ln211

+

++

=p (6),

ttt

tss

AAAA

θθ

secsec

=′=′ (7),

Temperatures are calculated at the half thickness of the shell or pipe. In case of flow in a pipe, the above equation should be transformed to the following equation taking into account the heat loss from the pipe.

( )t

tpt

pttt A

qTTAK

tTC

′−−=

∂∂ρ (8),

( )atLoss TTUq −=′ (9).

The second term of the right hand side represents heat loss from the pipe to the environment.

The equation for the shell of the heat exchanger is the same. An overall heat transfer coefficient ULoss should be calculated from the pipe to the environment through insulator in the similar way as Eq. (5). 2.1.5 Energy equation of the primary fluid

( ) 2

2

zT

kTTAK

zT

WAC

tT

C plpt

p

ppp

p

pppp ∂

∂+−=

∂∂

+∂

∂ρ (10),

The second term expresses heat conduction in flow direction (similar to Eq. (3)). These

equations are discretized using the implicit method. Models in the code are verified using test results of mock-ups of a boiling water reactor. Thermal-hydraulics and neutronics of the coupled system of the core, heat transport systems and the turbine system can be calculated by this code. The models in the code have been developed further in order to simulate characteristics specific to liquid metal cooled fast reactors. These models include inter-subassembly heat transfer that becomes non-negligible under natural circulation conditions, the heat transfer and heat transfer models for air coolers, the heat transfer model for the IHX, the model of the upper plenum, etc.

2.2 Calculation Model of “Monju”

The individual models incorporated in the NETFLOW++ code have been verified through the

experimental results and validated using data from nuclear plants like “Joyo”, “Monju”, “Phenix” and “EBR-II”. Therefore, it is confirmed that this code can be applied to both the loop-type and pool-type FBRs. The latest validation has been conducted through the IAEA benchmark (Mochizuki et al. [5]). In the present study, the heat transport systems of the “Monju” prototype FBR are modelled and analyzed. A schematic diagram of the reactor heat transport systems is illustrated in Fig. 1. It is a loop type reactor which consists of the primary, secondary and tertiary systems. It has three loops in the primary and secondary heat transport systems (HTSs), respectively. Heat generated in the reactor core is transferred to the secondary HTS via three intermediate heat exchangers (IHXs). Each secondary HTS has a steam generator consisting of an evaporator (EV) and a super-heater (SH). Hot sodium from the core enters


60

into the SH, then flows into the EV. The tertiary system consists of a high-pressure turbine, low-pressure turbines, a condenser, a series of feedwater heaters (FWHs), a deaerator, FW pumps, flow control valves, etc. FW is heated by a bank of FW heaters, controlled by the flow rate control valve, and is supplied to the EV, then flows into the SH. The dryout of water occurs near the outlet of the EV. A steam separator is installed at the outlet of the EV to prevent liquid droplets from flowing into the SH. Dried steam at around 13MPa is supplied to the high pressure turbine through a steam control valve. Exhausted low-pressure steam is supplied to the low pressure turbines and condensed by the condenser situated beneath the low pressure turbines.

Primary heat transport system Turbine and feedwater system

IHX

Primary pump

Core

Air cooler

Evaporator(EV)

TurbineGenerator

Feedwaterpump

Seawaterpump

Condenser

Secondary HTS

Secondarypump

Super-heater(SH)

Fig. 1. Schematic diagram of heat transport systems of “Monju” FBR reactor

In order to analyze the turbine and FW systems together with the primary and secondary HTSs,

the calculation models illustrated in Fig. 2 and Fig. 3 are used. Figure 2 shows the whole plant model and Fig. 3 shows the detailed model of the turbine and FW systems. Although condensed water is heated by three low-pressure FWHs, these are omitted in the main analysis because the effect on the FW temperature by these heaters is small. Thus, the flow rate and temperature at the inlet of the deaerator are provided as constant boundary conditions. The outlet pressure at the high-pressure turbine is taken as a boundary condition. These simplifications are justified by taking into account the low-pressure turbine and a bank of the low-pressure FWHs. The calculation model connecting the sodium systems and whole water systems is unusual and challengeable one.

A number in circle in Fig. 2 indicates “joint” where a flow path may branch. The joint has no volume. Piping, indicated as a line called as a “link” in the model, connects these joints. The link has a cross-section and length which can be divided into several nodes. Small circles indicate sub-joints, where the diameter of the piping may change. The reactor core is modelled with ten different hydraulic channels and a bypass channel.

The turbine and FW systems are modelled in detail as shown in Fig. 3. All major components in the systems are included in the model such as the steam generators consisting of the EVs and SHs, the high-pressure turbine, the deaerator, the FW pump, the FWHs, the FW control valves, the steam control valve, extraction lines and drainpipes. The shell-side of the FWH is divided into three regions in order to have stable calculation. Nevertheless, the calculation in the water system is sometimes unstable dependent on the input conditions because of condensation on the shell-side and flashing in drainpipes connecting the higher and lower pressure FWHs. When the low-pressure system is connected, the calculation is much difficult compared to the case of high-pressure components because the pressure in the drainpipe of the FWH #1 is only 0.004 MPa. The pressure range in the water system is enormous. Therefore, the low-pressure side is neglected in the most calculations and constant FW flow rate from the low-pressure FWH is assumed. The pressure boundary condition is provided at the exhaust line to the low pressure turbines.


61

7

Evaporator Super-heater

[13]

[1]～[7]

[47]15

10-2

8 9

12

1

5

6

-1

[8] IHX

Air cooler

[28]

[29]

[30]

High-pressureplenum

Pump

Upper plenum

[31]

[32]

[33][15]

[16]

[34][35]

[17]

[14]to turbine

11[10]

[11]

[12]

[9]

Pump

3

Link 1-Link 6: 1st to 6th layer (Inner driver)Link 7: 7th & 8th layer (Outer driver)Link 8: 9th to 11th layer (Blanket)Link 9: Center CRLink 10: CRsLink 11: Bypass

[24]

[18]

[19]

[46]

[23]

19

[49]

1922

-4

2021

24

[42]

[43]

[44] [45]

23

[25]

[26]

[27]

13

[48]

1816

-3

14

18

[37]

[38]

[39] [40]

17

[20]

[21]

[22]

[36]

[41]to B-loop

to C-loop

A-loopB-loop

C-loop From B&C loops

14 16

４

15

2

2 3020

#2#3#4#5

25

#1

1

13

1

-1

Low-pressure turbine

Deaerator

Feedwaterpump

Condenser

High-pressure feedwater heaters Low-pressure feedwater heaters

Low-pressure side

1

1

16

16

14

14

Low-pressureplenum

Fig. 2. Whole plant calculation model of the “Monju” reactor system

2 3020

17 9 10 11 1235 24 29 34 35

#2#3#4#5

2 3 3

6 7 8

2 3 3

3 4 5

2 3 3

21 22 23

2 3 3

26 27 28

2 3 3

31 32 3325

#1

36

37

38

31

33

32

26

28 29

1

3 4

2

6

19

16

17

12

21

42 43 44 45

15

27

830 35

Condenser

Feedwater pump

Flow control valve

Low-pressure turbine

High-pressure turbine

5

46

14

14

13

1

151819

16

3934

40

13

18

22

23

24

25

20 9 10 11

Steam control valve

Loss

Super-heater

Evaporator

Deaerator

41

Turbine for feedwaterpump

47

48 49

36 37-1

15

High-pressure feedwater heaters Low-pressure feedwater heaters

Sodium flow

Low-pressure side

Fig. 3. Detailed model of turbine and FW systems

3. Validation of the Initial Conditions

In order to investigate the small disturbances in the FBR plant due to a small abnormality such as the FW control valve degradation, the initial conditions of the plant should be calculated with good precision. The heat balance confirmation test was conducted at 45% thermal power (40% electrical power) of “Monju” and the result was reported by Miyakawa et al. [6]. Although 40 % power is described in their paper, that value is electrical power and electrical power does not make sense in the plant analysis. Therefore the above expression in thermal power is used in the present paper. The operation mode of the FW pump is switched from an electric pump to a turbine driven pump at 45%


62

thermal power. The test conditions are calculated by the NETFLOW++ code over a period of 20,000 seconds to get a good steady state condition. The calculated results are compared with the test results in Table 1. Since all water is evaporated in the heat transfer tubes near the outlet of the EV, i.e., dryout, the outlet temperature is higher than the saturation temperature. The outlet temperature cannot be predicted well if the dryout position is not predicted well. Other than this temperature, the super-heater outlet temperature, are used as indicators of the scram inter-locks. Therefore, the absolute values of these temperatures are important parameters. Major plant parameters are calculated within 2.25% error. The error is calculated as the fractional difference between calculated and measured data. In the case of temperature, the error is dependent on unit such as Celsius and Kelvin. Therefore, the value of the error is shown as a subtracted value from the melting point in order to be independent from the unit. Since “Monju” has never been operated at full power condition, there is no measurement data for plant parameters at full power. The calculated conditions are compared with the design values as shown in Table 2. Good agreement is also obtained for the 100% full power condition. Even in the case of a large transient such as the turbine trip event from 45% thermal power, the major parameters of the plant could be evaluated with good precision [3]. Therefore, it is concluded that the small disturbances caused by the various abnormalities can be calculated with good accuracy by the NETFLOW++ code.

Table 1 Comparison of plant parameters between measured data at “Monju” and calculation at 45% thermal power

Item Measured data Calculated result Error* (%) Reactor vessel inlet temperature (ºC) 364 372.2 3.11 Reactor vessel outlet temperature (ºC) 486 491.3 1.37 IHX primary outlet temperature (ºC) 367 372.2 1.95 IHX secondary inlet temperature (ºC) 282 278.8 -1.76 IHX secondary outlet temperature (ºC) 486 491.0 1.29 B-loop primary flow rate (kg/s) 702.4 706.4 0.569 B-loop secondary flow rate (kg/s) 398.6 391.5 -1.53 Feedwater (FW) flow rate (kg/s) 43.33 42.20 -2.58 FW temperature (ºC) 196.9 194.2 -1.32 Evaporator outlet steam temperature (ºC) 370 365.9 -1.11 Super-heater outlet steam temperature (ºC) 483.1 486.9 0.787

*Error= (Calculation result - Measured data)/(Measured data) for flow rate, and Error= (Calculation result - measured data)/(Measured data - melting point) for temperature

Table 2 Comparison of plant parameters between design data of “Monju” and calculation at 100% thermal power

Item Design data Calculated result Error* (%) Reactor vessel inlet temperature (ºC) 397 391.4 -1.89 Reactor vessel outlet temperature (ºC) 529 523.0 -1.40 IHX primary outlet temperature (ºC) 397 391.4 -1.89 IHX secondary inlet temperature (ºC) 325 328.4 1.51 IHX secondary outlet temperature (ºC) 505 508.6 0.88 Primary loop flow rate (kg/s/loop) 1422.2 1425.8 0.25 Secondary loop flow rate (kg/s/loop) 1030.6 1035.1 0.44 FW flow rate (kg/s) 105.6 106.7 1.04 FW temperature (ºC) 240 238.2 -0.75 Evaporator outlet steam temperature (ºC) 369 368.8 -0.05 Super-heater outlet steam temperature (ºC) 487 485.0 -0.41

*Error= (Calculated result - Design data)/(Design data) for flow rate, and Error= (Calculation result - Design data)/(Design data - melting point) for temperature

4. Simulation

4.1 Decrease of FW flow rate

The plant is scrammed if the FW flow rate decreases to 20% of rated value. However, the


63

temperatures around the steam generators may rise up to the trip level even if the flow rate is only slightly degraded. Temperatures and FW flow rate for plant trip are listed in Table 3. Two direct causes for a decrease of the FW flow rate are discussed in the present chapter, although there are possible causes, e.g., malfunction of a FW control valve, degradation of a FW pump, a blockage of flow passage by a loose part, etc.

Regarding the FW system, a FW control system is provided in order to keep the EV outlet temperature at the set point temperature. This system controls the FW control valve directly under the normal conditions. The other controlling is to keep the pressure difference of the FW control valve at the set point pressure by changing the speed of revolution of the FW pump. However, above mentioned control system does not work properly for the abnormalities of FW system.

Table 3 Relevant plant trip conditions

Items Conditions Trip activation

Feed water (FW) flow rate Less than 20% of the rated condition under the evaporator steam temperature more than 325 ºC

Secondary pump trip

FW temperature Less than 153 ºC FW pump trip Evaporator outlet steam temperature Higher than 417 ºC Secondary pump trip Evaporator outlet sodium temperature Higher than 347 ºC Scram Super-heater outlet sodium temperature Higher than 497 ºC Secondary pump trip

4.1.1. Malfunction of the FW control valve

What will happen when the FW control valve in the A-loop undergoes a small malfunction? One can easily imagine that the location where dryout occurs in the heat transfer tubes of the EV moves upstream. The assumption is that the FW flow rate decreases to 90% of the rated value at t=1,000 seconds and the plant is operating at 45% thermal power. The effect on important plant parameters on the water side and the sodium side are shown in Fig. 4. The system pressure of the steam generator at the initial conditions is approximately 13MPa, and the saturation temperature is 330 ºC. Since two-phase flow causes dryout and is super-heated near the outlet of the EV, the steam temperature at the outlet of the EV increases from 366 ºC of the initial temperature to the trip temperature of the secondary main pump at 417ºC within 1,100 seconds. In the present study, the calculation is continued without reactor trip. This time domain is shown as a shadowed area. During this transient, the steam pressure at the outlet of the SH is almost constant. Similar trends are obtained for the operation at 100 % thermal power. Although the sodium temperature at the outlet of the SH increases due to a decrease of heat transfer from sodium to steam in the SH, sodium temperatures at the SH outlet and the EV outlet do not reach the trip levels. It is concluded that the EV outlet steam temperature is a good indicator to detect long-term flow degradation due to malfunction of the FW control valve.

Fig. 4. Evolution of plant parameters in the tertiary (left) and the secondary (right) heat transport system

during an A-loop FW control valve malfunction at 45% thermal power

0

100

200

300

400

500

600

0

50

100

150

0 500 1000 1500 2000 2500 3000

SH outletEV-A outletEV-B outletEV-C outletFeedwater

A-loop flowrateB-loop flowrateSH outlet pressure

Wat

er te

mpe

ratu

re (o C

)

Time (sec)

Feed

wat

er fl

ow ra

te (k

g/s/

loop

), P

ress

ure

(bar

)

417oC trip temp. at 2100 sec.

Degradation of feedwatercontrol valve A

200

250

300

350

400

450

500

550

600

0 500 1000 1500 2000 2500 3000

SH-A outletSH-B outletSH-C outletEV-A outletEV-B outletEV-C outletS

odiu

m te

mpe

ratu

re (o C

)

Time (sec)

Trip level of evaporatoroutlet sodium temperature

Trip level of super-heateroutlet sodium temperature


64

4.1.2. Degradation of the FW pump

It is assumed that the plant operates at 45% thermal power. At t=1,000 s the pump head of the FW pump is degraded by 10% in a stepwise manner. This causes a reduction of the FW pump flow rate. The flow rates in the three loops decrease by approximately 8% as shown in Fig. 5. All steam temperatures at the outlet of the EV increase and reach the trip temperature of 417ºC at t=1373 s. The steam pressure at the outlet of the SH decreases at the same time. In case of the periodic FW flow rate fluctuation, the temperature at the EV outlet does not change. This is confirmed by the test and calculation. Although the sodium temperature at the SH outlet increases, the temperature does not reach the trip level before the reactor trip by the steam temperature at the EV outlet. It is concluded that the EV steam and SH sodium outlet temperatures and the system pressure are good indicators to detect long-term flow degradation due to degradation of the FW pump.

Similar trends are found in the case of the 100% rated power condition. The result is illustrated in Fig. 6. The temperature at the outlet of the EV reaches the trip temperature at around t=1900 s. As a result of these analyses, it is obvious that a small degradation of the FW pump causes plant trip in a relatively short period. It will be difficult for operators to avoid this trip.

Fig. 5. Evolution of plant parameters in the tertiary (left) and sodium (right) heat transport system

during a FW pump degradation at 45% thermal power

Fig. 6. Evolution of plant parameters in the tertiary heat transport system during a FW pump

degradation at 100% thermal power 4.2 Pressure increase due to steam control valve malfunction

The plant behavior is investigated when the steam control valve causes malfunction resulting in the system pressure increase. The FW control system to keep the EV outlet temperature is assumed to be out of order. Figure 7 shows the temperature behavior in the steam generator and the system

0

100

200

300

400

500

600

0

50

100

150

0 500 1000 1500 2000 2500 3000

SH outletEV outletFeedwater

Feedwater flowrateSH outlet pressure

Wat

er te

mpe

ratu

re (o C

)

Time (sec)

Feed

wat

er fl

ow ra

te (k

g/s/

loop

), P

ress

ure

(bar

)

Degradation of feedwater pump


200

250

300

350

400

450

500

550

600

0 500 1000 1500 2000 2500 3000

Upper plenumIHX secondary outletSH outletIHX primary outletEV outlet

Sod

ium

tem

pera

ture

(o C)

Time (sec)

Trip level of evaporatoroutlet sodium temperature

Trip level of super-heateroutlet sodium temperature

0

100

200

300

400

500

600

0

50

100

150

0 500 1000 1500 2000 2500 3000



Wat

er te

mpe

ratu

re (o C

)

Time (sec)

Feed

wat

er fl

ow ra

te (k

g/s/

loop

), P

ress

ure

(bar

)

Degradation offeedwater pump



65

pressure, when the plant operates at 45% thermal power. Pressure loss at the steam flow control valve is approximately 9MPa under the normal operating conditions (0 to 1,000 seconds). The left figure shows an example when the system pressure increases by 0.5 MPa, for instance due to the valve stick at a small throttling position without FW flow rate control. This event results in a decrease of the FW flow rate. The temperature at the EV outlet increases rapidly and reaches the trip temperature at 417 ºC within approximately 10 minutes. Similar trends are found for the 100% rated operation condition. However, the plant has a control system to keep the EV outlet temperature at a constant value. As a result, the evolution of the plant parameters on the right figure is obtained. It is seen that the malfunction, i.e., a slight closure of the steam control valve, can only be detected by the increase of the system pressure.

When the feed water system is normal, the FW flow rate is increased in order to keep the constant EV outlet temperature. As a result, all the plant parameters except the system pressure show constant.

Fig. 7. Evolution of plant parameters in the tertiary heat transport system during a steam control valve

malfunction without (left) and with (right) FW flow control at 45% thermal power 4.3 Heat transfer degradation due to fouling on heat transfer tubes of evaporators

Water contains many impurities which cause fouling on the inside surface of heat transfer tubes. The heat transfer tubes in the EVs are susceptible to catch fouling due to the boiling of water. Since super-heated steam is supplied to the SH from the EV, it is assumed that the fouling will not deposit on the inner surface of the heat transfer tubes of SH. The fouling causes a decrease of the overall heat transfer coefficient. Every steam generator will encounter the above situation in the long run. We assume a stepwise increase of the thermal resistance due to fouling (0.01 to 0.05 m2K/kW: inverse value of the heat transfer coefficient) at t=1,000 s. The evolution of the plant parameters is shown in Fig. 8.

The abnormality of the FW system is also assumed. In this event, the SH outlet sodium temperature decreases. The EV outlet sodium temperature is almost constant. Regarding temperatures in the tertiary system, only the EV outlet temperature decreases to the saturation temperature corresponding to the system pressure which is not affected strongly by fouling. The transient behavior is similar for the case under 100% power conditions. In this event, the two-phase flow does not cause dryout near the outlet of the EV. As a result, lower energy steam and a large amount of droplets will flow into the SH. Because of this, the heat transfer rate in the SH increases and the sodium temperature at the outlet of the SH decreases. Therefore, the heat transfer degradation event in the EV can be detected by the decrease of the steam temperature signals at the outlet of the EV, as well as the decrease of the sodium temperatures at the outlet of the SH.

In practice, the FW control system acts to keep the EV outlet temperature at the set point. Therefore, FW flow rate is decreased automatically by the system in a long period. Therefore, operators may not aware of the heat transfer degradation because of the slow event.

0

100

200

300

400

500

600

0

50

100

150

0 500 1000 1500 2000 2500 3000


FeedwaterSH outlet pressure

Wat

er te

mpe

ratu

re (o C

)

Time (sec)

Feed

wat

er fl

ow ra

te (k

g/s/

loop

), P

ress

ure

(bar

)

Degradation of steam control valve


0

100

200

300

400

500

600

0

50

100

150

0 1000 2000 3000 4000 5000 6000


FeedwaterSH outlet pressure

Wat

r tem

pera

ture

(o C)

Time (sec)

Feed

wat

er fl

ow ra

te (k

g/s/

loop

), S

team

pre

ssur

e (b

ar)

Degradation of fsteam control valve


66


when the thermal resistance increases to 0.05m2K/kW due to fouling at 45% thermal power

4.4 Degradation of FW heating

There are several reasons which cause degradation of FW heating. One is fouling on the inside and outside of the heat transfer tubes. Another reason is the loss of extracted steam flow from the turbine system. These events may cause a decrease of the temperature of the FW supplied to the EV. Figure 9 shows the transient where a valve on the extraction line from the high-pressure turbine to the last stage high-pressure FWH fails. As a result of the failure, the FW temperature decreases and the EV outlet temperature decreases. The FW flow rate is decreased by the control system in order to maintain the EV outlet temperature at the present level. Therefore, non-stationary signals are the FW temperature, FW flow rate and the system pressure. The temperatures in the secondary system decrease slightly as shown in the right side of figure of Fig. 9. These trends are almost same as the results using a full scope simulator of the “Monju” reactor.

The effect of thermal resistance in the FWHs was also investigated. The thermal resistance is an order of magnitude larger than in the case of fouling on the heat transfer tubes of the evaporators. However, the effect on the FW temperature is very small and there is sufficient margin to the FW pump trip due to low temperature. Therefore, detection of the fouling build-up on the inner and outer surfaces of the heat transfer tubes is rather difficult compared to the case of the evaporator.


when a valve on the extraction line fails at 45% thermal power

4.5 Degradation of Main pump

Even though there are not many active components in the primary system, the main pump is one of the active components. Three heat transport systems are provided in the “Monju” reactor, and each system is recirculated by one main pump. It is assumed that the pump in the A-loop causes

0

100

200

300

400

500

600

0

50

100

150

0 2000 4000 6000 8000 1 104



Wat

er te

mpe

ratu

re (o C

)

Time (sec)

Feed

wat

er fl

ow ra

te (k

g/s/

loop

), P

ress

ure

(bar

)

Degradation ofheat transfer

200

250

300

350

400

450

500

550

600

0 2000 4000 6000 8000 1 104

SH-A outletSH-B outletSH-C outletEV-A outletEV-B outletEV-C outlet

Sod

ium

tem

pera

ture

(o C)

Time (sec)

0

100

200

300

400

500

600

0

50

100

150

0 1000 2000 3000 4000 5000 6000

Super-heater outletEvaporator-A outletFeedwater

Feedwater flow rateMain steam pressure (bar)

Tem

pera

ture

(o C)

Time (sec)

Feed

wat

er fl

ow ra

te (k

g/s/

loop

), P

ress

ure

(bar

)

Abnormarity of extraction valve for #2 high-pressure feedwater heater

200

250

300

350

400

450

500

550

600

0 1000 2000 3000 4000 5000 6000

Super-heater inletSuper-heater outletEvaporator outlet

Sod

ium

tem

pera

ture

(o C)

Time (sec)


67

degradation at 1000 seconds. The evolution of the temperature and flow rate in the primary heat transport system is illustrated in Fig. 10. When the flow rate in the A-loop decreases, the flow rates in the other two loops increase as if to compensate the total flow rate. As a result, the outlet temperatures in the primary system change slightly, and inlet temperatures stay almost constant. The evolution of these plant parameters is quasi-static. This event can be detected by watching the flow rate difference between the loops.

100

200

300

400

500

600

700

0

500

1000

1500

0 1000 2000 3000 4000 5000 6000

IHX A inletIHX B inletIHX C inletIHX A outletIHX B outletIHX C outlet

A loopB loopC loop

Tem

pera

ture

(o C)

Time (sec)

Fow

rate

(kg/

s/lo

op)

Temperature Flowrate

Fig. 10. Evolution of plant parameters in the primary system during a degradation of a main pump in the

A-loop 5. Discussion

The above analyses are conducted with a model which neglects the low-pressure equipment such as low-pressure turbines, low-pressure FWHs etc. Therefore, the differences of analytical results using the simplified model and the detailed model should be discussed. In the simplified model, three FW heaters and the low-pressure turbine are omitted from the calculation model as mentioned in the previous section. The detailed model consists of whole plant system as shown in Fig. 2 and Fig. 3. Representative of the abnormalities, a degradation of the FW pump at 100% thermal power is discussed. A comparison of the analytical results between simplified and detailed models is shown in Fig. 11. There are minor differences of initial parameters between the two models. After the slight degradation of pump head, the plant parameters show similar trends between the two calculation models. However, the pressure at the outlet of the super-heater is higher by 1 MPa and the time to initiate the trip signal is earlier by 200 seconds for the detailed calculation model. The temperature difference at the EV outlet between the two models is 6.5 ºC. The difference of pressure between the simplified and detailed model may be caused by the inventory of vapor in the turbine system.

The above plant transient is almost independent of conditions of FW control system because the control system is inefficient due to the FW pump degradation. A malfunction of the steam control valve without FW control under the rated condition is also calculated using the simplified and detailed models as shown in Fig. 12. When the valve sticks at the small opening, the system pressure increases and FW flow rate decreases.

Therefore, the evaporator outlet temperature increases to the trip temperature level quickly. Results are very similar between the simplified and detailed models. The temperature difference at the EV outlet between the two models is 4.5 ºC. The two calculation cases show that the simplified model is adequate to calculate the effects due to minor abnormalities without large discrepancies compared to an analysis with a detailed analysis model. The similar result could be calculated for this event with the FW control system because the FW control system is inefficient due to the problem downstream of the water system.


68

0

100

200

300

400

500

600

0

50

100

150

0 500 1000 1500 2000 2500

SH outlet (Simple)EV outlet (Simple)Feedwater (Simple)SH outlet (Detailed)EV outlet (Detailed)Feedwater (Detailed)

Feedwater flowrate (Simple)SH outlet pressure (Simple)Feedwater flowrate (Detailed)SH outlet prssure (Detailed)

Wat

er te

mpe

ratu

re (o C

)

Time (sec)

Feed

wat

er fl

ow ra

te (k

g/s/

loop

), P

ress

ure

(bar

)

Degradation offeedwater pump


Fig. 11. Comparison of evolutions of plant parameters in the tertiary heat transport system during a FW

pump degradation at 100% thermal power between simplified and detailed turbine and FW systems

0

100

200

300

400

500

600

0

50

100

150

0 500 1000 1500 2000 2500

SH outlet (Simple)EV outlet (Simple)Feedwater (Simple)SH outlet (Detailed)EV outlet (Detailed)Feedwater (Detailed)

Feedwatr flowrate (Simple)SH outlet pressure (Simple)Feedwater flowrate (Detailed)SH outlet pressure (Detailed)

Wat

er te

mpe

ratu

re (o C

)

Time (sec)

Feed

wat

er fl

ow ra

te (k

g/s/

loop

), P

ress

ure

(bar

)Degradation offeedwater pump


Fig. 12. Comparison of evolutions of plant parameters in the tertiary heat transport system during a

steam control valve malfunction at 100% thermal power between simplified and detailed turbine and FW systems

6. Conclusion

Minor abnormalities which do not cause a trip caused in the tertiary system of the prototype fast

reactor “Monju” are analyzed by the NETFLOW++ code using a whole plant model consisting of the primary, the secondary and the tertiary systems. A malfunction of a feedwater control valve and a degradation of a feedwater pump are selected as events of decreasing feedwater flow rate. In these events, the steam temperature at the outlet of the evaporator has a high sensitivity to the abnormalities. Fouling on the heat transfer tubes of evaporators is selected as an abnormality due to heat transfer in the EV. The EV outlet steam temperature has high sensitivity if the feedwater control system does not work. However, since the EV outlet temperature is controlled in order to maintain a constant temperature, detection of the heat transfer degradation of the EV is rather difficult. In the case of a pressure increase due to steam control valve malfunction, only the system pressure undergoes


69

sufficient change to detect the problem. Therefore, operators should pay attention to the pressure signal to detect the malfunction of the steam control valve.

Nomenclature

A: cross sectional area or heat transfer area (m2) C: specific heat (J/kg K) De: equivalent diameter (m) d: diameter (m) g: gravitational acceleration (m/s2) Hp: pump head if there is a pump in the interested link i. h: heat transfer coefficient (W/m2K) k: thermal conductivity (W/mK) L: inertia li／Ai (1/m), l: length (m), Ntube: number of tubes p: pressure (Pa) q’: heat loss (W/m2) T: temperature (K) t: time (s) U: overall heat transfer coefficient (W/m2K) W: flow rate (kg/s) z: coordinate (m) θ: angle of flow direction to the vertical direction (radian), ζ: local loss coefficient (-) l: pipe friction factor (-) ρ: density (kg/m3) Subscripts i; link number j: junction number f: fouling p: primary s :secondary t: tube. Acknowledgement

The present study includes the result of “R&D of an anomaly detection system by the advanced hybrid processing of plant data of “MONJU” entrusted to Okayama University by the Ministry of Education, Culture, Sports, Science and Technology of Japan (MEXT), and re-commissioned to University of Fukui. The authors would like to express their thanks to MEXT. The author would like to express his thanks to Mr. T. Shimizu and Mr. D. Nakamura who helped the author to adjust data of plant.

References [1] Udagawa K., et al., 2000. Evaluation of fluctuation signals on steady state in Monju, JNC Technical

Review, No. 9, pp. 1-10, (in Japanese). [2] Santhosh, T.V., et al., 2011. A diagnostic system for identifying accident conditions in a nuclear reactor,

Nuclear Engineering and Design, 241, pp. 177-184. [3] Mochizuki, H., 2010. Development of the plant dynamics analysis code NETFLOW++, Nuclear

Engineering and Design, 240, pp. 577-587. [4] Mochizuki, H. and Tsukamoto T., 2011. Network Analysis of Turbine and Feedwater Systems of the


70

‘Fugen’ Nuclear Power Plant, Journal of Nuclear Science and Technology, 48, 5, pp.786-796. [5] Mochizuki, H., et al., Computation of natural convection test at Phenix reactor using the NETFLOW++

code, Nuclear Engineering and Design, 262, (2013), pp.1-11. [6] Miyakawa, A. et al., 2005. The Prototype Fast Breeder Reactor Monju System Startup Tests; Report

Summary Report of the System Startup Tests; Criticality Test ~ Power Up Test(40% Power),Japan Nuclear Cycle Development Institute, JNC-TN2410-2005-002, p.267-269 (in Japanese: http://jolissrch-inter.tokai-sc.jaea.go.jp/pdfdata/JNC-TN2410-2005-002.pdf).

Plant Transients under Small Abnormalit ies of FBR … Transients under Small Abnormalit ies of FBR...

Documents

Transcript of Plant Transients under Small Abnormalit ies of FBR … Transients under Small Abnormalit ies of FBR...