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Feedwater Flow Measurement with Venturi and Comparison to the other Parameters in NPP Krško

Vinko Planinc, Aljoša Šumlaj, Robert Rostohar, Dejvi Kadivnik

Nuclear Power Plant Krško Vrbina 12, SI-8270 Krško, Slovenia

vinko.planinc@nek.si, aljosa.sumlaj@nek.si, robert.rostohar@nek.si, dejvi.kadivnik@nek.si

ABSTRACT

Feedwater flow is one of the inputs to the secondary calorimetric calculation to determine the reactor thermal power in a nuclear power plant. Measurement of the mass flow is normally performed with one venturi nozzle per loop. The basic principle is to measure pressure difference between the nozzle inlet and the throat. The venturi nozzle is sensitive to the fouling which is changing throughout the fuel cycle and from cycle to cycle. These changes can be noticed by comparing different parameters which are correlated to the reactor thermal power. The main contributor is fouling which can occur inside of the flow elements and can give incorrect measurements. For a nuclear power plant it is important to precisely measure the feedwater mass flow, due to the fact that it is in direct correlation to the thermal output of the primary side. Every deviation of the mass flow measurement also means deviation of the output of the main electrical generator. The design uncertainty of the feedwater mass flow measurement is 2% and this error is also incorporated into the safety analyses. To confirm the appearance of the fouling on venturi, the flow can be independently measured. Among all the options the ultrasonic measurement and validation of parameters are most widely used in nuclear power plants. This paper also sheds light on some short-term and long-term actions which will confirm fouling and eliminate drift in feedwater flow measurement.

1 INTRODUCTION

After comparing some of the key parameters to the previous cycle values Krško NPP has noticed that the actual feedwater flow could be lower than indicated in the last fuel cycle. Following the modernization and uprate in 2000, Krško NPP made some investment in reliability and efficiency of the secondary cycle. The new Low-Pressure Turbines and High-Pressure Feedwater heaters were installed during the outage 2006 and the Plant performance test was performed; its results showed an increment in the generator gross output of 20 MWe.

The next improvement was the replacement of Moisture Separator Reheaters in 2007 which resulted in better efficiency of moisture removal, lower pressure drop and higher temperature of reheat steam. The generator output was expected to increase by at least 4 MWe.

When the plant reached full power and stabilization it was noticed that some of the key parameters, which are related to the thermal power of the reactor, were lower than during the

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previous cycle. A detailed analysis was performed to find the cause for discrepancy in thermal power. Different parameters were compared with the previous data.

2 BACKGROUND

Reactor thermal power (RTP) is directly proportional to the feedwater flow (feedwater

mass flow) for pressurized water reactors (PWR’s) as defined by the following equation: Reactor Thermal Power = qm × (hs – hfw) + (heat gain/loss) [W] (1) where: qm = feedwater mass flow [kg/s] hs = steam enthalpy, dependent on temperature, pressure, and moisture [J/kg] hfw = feed enthalpy, dependent on temperature and pressure [J/kg] heat gain/loss = from reactor-attached piping and pumps [W]

The key parameter in Eq. (1) is feedwater mass flow which is also the main contributor to the total uncertainty of computations of reactor thermal power in a PWR plant. Basic principle of calorimetric measurement is shown in Figure 1 [1].

Figure 1: Principle of calorimetric measurement

2.1 Measurement principle of feedwater flow

The calorimetric measurement on the secondarz side is used to determine the reactor

thermal power. Pressure and temperature can be precisely measured and the enthalpy can be calculated from those measurements. The measurement of fluid mass flow in a feedwater line is more complex.

The technology in common use for measuring feedwater flow in nuclear plants can be

divided into two categories: methods based on the measurement of differential pressure created by an inserted device (orifice plate, venturi, or nozzle) and methods based on the measurement of high frequency sound waves (ultrasonic technology) [2].

Proceedings of the International Conference Nuclear Energy for New Europe, Portorož, Slovenia, Sept. 8-11, 2008

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2.1.1 Differential pressure measurement

Differential pressure (DP) technology has been the industry standard for feedwater flow measurement throughout the development of the nuclear power industry. These instruments calculate mass flow from the measurement of differential pressure across the flow constriction and from the static pressure and temperature of a flowing fluid. Types of differential pressure flow elements include venturies, flow nozzles, and orifices. Each of these elements is based on the same operating principle; however, they differ in the geometry of the flow restriction, the extent of irreversible pressure loss, and code acceptance.

Figure 2: Differential pressure measurement –by VENTURI nozzle DP flow meters reduce the flow cross-sectional area, resulting in an increase in fluid

velocity through the throat of the flow element. The increased velocity causes a measurable difference in pressure between locations upstream and downstream of the flow element. The difference between the pressure upstream of the contraction and the pressure downstream of the contraction is measured with taps bored into the pipe or the flow element. These taps can be located at several locations, determined by the particular style and method of determining flow. In a nozzle, the low-pressure (downstream) can be measured at the throat or downstream of the throat, again determined by the style of device (for example, throat tap or wall tap). The flow meter measures the differential pressure produced, and then flow is calculated using Eq. (2), which is derived from the Bernoulli equation.

ρεπ⋅Δ⋅⋅⋅⋅⋅

⋅= pECdqm 2

4

2

(2)

where: qm = mass flow [kg/s] C = discharge coefficient [/]

E = )1(

14β−

= velocity of approach factor

d = diameter of flow restrictor [m] D = pipe inside diameter [m] ε = expansion factor [/] Δp = differential pressure [Pa] ρ = fluid density [kg/m3] β = d/D = beta ratio[/]

Proceedings of the International Conference Nuclear Energy for New Europe, Portorož, Slovenia, Sept. 8-11, 2008

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As documented in the EPRI “Nuclear Feedwater Flow Measurement Application Guide” [2], the main errors related to differential pressure flow measurement can be grouped into the following categories: density errors, errors in differential pressure measurement, errors due to the plant computer, Errors due to impulse line condition, Errors due to thermal expansion, errors in the discharge coefficient and errors due to changes of the internal conditions.

Changes of the internal conditions of a differential pressure flow meter affect the

accuracy of the feedwater flow calculation. Erosion, corrosion, corrosion product buildup, and cleaning can alter the internal condition of the meter. Deposition of corrosion products in front of the throat, in the throat section, and in the recovery cone of the throat tap flow meter can increase the pressure drop across the meter. This maz cause an erroneously high mass flow indication.

2.1.2 Ultrasonic cross flow measurement

An ultrasonic transit time flow meter measures the time necessary for an ultrasonic pulse to travel across the pipe from one transducer to another along the path that is diagonal to the fluid flow. The difference in times of flight for pulses traveling with and against the fluid flow is proportional to the fluid velocity.

Figure 3: Ultrasonic measurement - Cross flow principle M = axVAPF ⋅⋅⋅ρ (3) Where: M = mass flow [kg/s] ρ = density [kg/m3] PF = velocity profile factor [/] A = area of cross section [m2] Vax = average axial velocity [m/s] Technical errors potentially affecting ultrasonic flow measurement include the

following: dimensional errors, density errors, timing errors, profile factor errors (experimental uncertainty, piping configuration, pipe roughness etc.), miscellaneous errors including sensitivity to the transducer equipment [2].

A typical uncertainty of cross flow measurement is less than 0.7%. The uncertainty

calculated using actual pipe dimensions and observed variations is as low as 0.5% per pipe [2].

Proceedings of the International Conference Nuclear Energy for New Europe, Portorož, Slovenia, Sept. 8-11, 2008

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2.2 Parameters related to the thermal power

Some of the parameters in nuclear power plant are directly or indirectly related to the reactor thermal power. These important parameters are listed in Table 1. Comparison of parameters between the current and previous fuel cycle can show if there is any change in feedwater flow measurement behavior. As an example, cold leg temperatures of reactor coolant are presented in Figure 4.

Table 1: Important parameters related to the thermal power

Parameter Relation - Description Feedwater flow Measured with venturi meter, used for calorimetric

calculation

1st stage impulse pressure Turbine power indication, used for determining referenced average temperature

Core temperature difference Measured in the primary legs. Linearly proportional to the thermal reactor power.

Reactor coolant temperatures Hot legs, Cold legs and Average reactor coolant temperature are used for reactor protection.

Steam flow Measured on the SG’s outlet, independed of feedwater flow measurement

Opening of turbine control valves Position of control valves, defines the steam flow, with respect to the SG pressure

Final feedwater temperature Measured with thermo element, used to calculate density and enthalpy. Increased temperature means higher power.

Steam generator pressure Used to calculate steam enthalpy. Increased pressure of main steam means lower power.

Generator electrical output Electrical output corrected to the design vacuum. Moisture Separator Reheater outlet pressure

Increased outlet pressure means higher thermal power.

Figure 4: Measured cold leg temperature of reactor coolant

Proceedings of the International Conference Nuclear Energy for New Europe, Portorož, Slovenia, Sept. 8-11, 2008

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Some of the parameters are not affected by the fouling in the venturi flow element. One of such parameters is the turbine impulse pressure, which is proportional to the turbine power and can be correlated with the steam flow. New terms can be defined as Ratio of Steam Flow (RSF) and Relative Ratio of Steam Flow (RRSF).

RSF = Main steam flow/turbine impulse pressure (4) RRSF = RSFactual/RSFbaseline (5) Where: Main steam flow = difference between the feedwater and blowdown flow [kg/s] Turbine impulse pressure = 1st stage impulse pressure [Pa] RSFactual = actual ratio of steam flow RSFbaseline = baseline ratio of steam flow (obtained from Plant performance test after

Low-Pressure Turbine replacement) For monitoring purposes the easiest way is to monitor the trend of the relative ratio of

RSF, as presented in Figure 5.

Relative ratio of steam flow

0,98

0,99

1

1,01

1,02

14.01.04 01.08.04 17.02.05 05.09.05 24.03.06 10.10.06 28.04.07 14.11.07 01.06.08

RR

SF S

G1

& S

G2

(/)

RRPSG1RRPSG2

GC21 GC22 GC23GC20

Figure 5: Relative Ratio of Steam Flow over the time

2.3 Implementation of modification

In the past Krško NPP has implemented some changes which have positive effect on the secondary side efficiency. After the modernization and uprate in 2000, Krško NPP invested in reliability and efficiency of the secondary cycle. The new Low-Pressure Turbines and High-Pressure Feedwater heaters were installed during outage 2006 and Plant performance test was performed. The results have shown increment in Generator Gross Output of 20 MWe [3].

The next improvement took place in 2007 with Moisture Separator Reheaters’ (MSR) replacement, with better efficiency of moisture removal, lower pressure drop and higher temperature of reheat steam. The replacement of Low-Pressure Heaters in the condenser neck has minor influence on the efficiency.

Proceedings of the International Conference Nuclear Energy for New Europe, Portorož, Slovenia, Sept. 8-11, 2008

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After the replacement of Moisture Separator Reheaters, a performance test in accordance with ASME PTC 12.4 [4] (Power Test Code for Moisture Separator Reheater) was carried out. Pressure drop and outlet steam temperature have reached guaranteed values and the consumption of steam for reheat was bellow the design value, which means that more steam is available for turbine work. Outlet pressure of the superheated steam is lower than before the replacement and this can only be linked to the lower steam flow and consequently lower feedwater flow and therefore lower thermal power. Before the replacement, the generator electrical output increase of at least 4 MWe was anticipated.

In accordance with ASME PTC 12.1 [5] (Power Test Code for Feedwater Heaters), a performance test was carried out. The result showed that the heaters were operating as designed and there were no significant changes to other parameters.

All the modifications were reviewed and there is no negative effect on the generator electrical gross output. The net consumption was found to be increased by 0.15MW.

2.4 Determination of fouling thickness

Feedwater flow in Krško NPP is measured with venturi flow element on the principle of pressure drop as described in section 2.1. In case there are some deposits in the venturi the pressure drop will be higher and consequently the indicated feedwater flow will be higher than actual. A simple calculation was used to determine the necessary thickness for 0.5% error on measured feedwater mass flow.

Δqm = %1001

12 ⋅−

D

DD

qmqmqm (6)

Δqm represents the difference in flow due to fouling (qmD2) compared to flow with no

fouling (qmD1); expressed in % Using Eq. (2) with assumption that δ<<d we can obtain:

100

221

121 4

42

⋅

⎟⎟⎟⎟⎟⎟

⎠

⎞

⎜⎜⎜⎜⎜⎜

⎝

⎛

⎟⎟⎠

⎞⎜⎜⎝

⎛−−

−

−⋅⎟

⎠⎞

⎜⎝⎛ −

−=Δ

δδ

βδ

Ddd

dqm [%] (7)

Where: d......inside diameter of venturi = 207,2386 mm D.....pipe diameter = 363,5756 mm δ......scaling thickness β......beta ratio = d/D = 0,57 All other parameters are assumed to be constant for scaling thickness (δ)<<d

Proceedings of the International Conference Nuclear Energy for New Europe, Portorož, Slovenia, Sept. 8-11, 2008

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δ

Scaling thickness

Figure 6: Krško venturi flow element [6]

Results from Eq. (7) are presented in Figure 7.

Feedwater flow decrease due to scaling buildup

0,00

0,20

0,40

0,60

0,80

1,00

1,20

0 0,05 0,1 0,15 0,2 0,25 0,3 0,35 0,4 0,45 0,5 0,55

Scaling thickness (mm)

Δqm

Cha

nge

in fe

edw

ater

flow

(%)

Figure 7: Feedwater venturi scaling thickness versus measured feedwater flow error

3 FUTURE ACTIVITIES

Evaluation of parameters after outage 2007 and their comparison with previous data

shows that there are certain changes in feedwater flow measurement. A change to the feedwater flow measurement technique and the uncertainty may be considered as design change. According to 10 CFR 50, Appendix K, the plant safety systems should be designed by assuming that the reactor is generating 102% of its licensed power. Its 2% is intended to cover the uncertainty of measurements used in calorimetric calculations.

Proceedings of the International Conference Nuclear Energy for New Europe, Portorož, Slovenia, Sept. 8-11, 2008

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Short-term activities will focus on confirmation of venturi fouling. Three different actions are proposed:

1. Development of a procedure to confirm that all important parameters are within the expected range, showing that when plant operates at full power, the thermal power is close to the design value of 1994 MW.

2. Krško NPP plans to perform one-time measurement of feedwater flow with radioactive tracer. The uncertainty of this method is below 0.5% or even less [7].

3. During the next outage venturi flow elements will be visually inspected. Long-term action:

1. Installation of permanent ultrasonic feedwater flow measurement. This will provide the basis for mini uprate up to 1.4%. The uprate is possible due to lower measurement uncertainty. This measurement can decrease uncertainty from 2% to the 0.5%.

2. Introducing Process Data Reconciliation according to VDI 2048 [8, 9] to decrease the feedwater flow measurement uncertainty. Uncertainty can be decreased to less than 0.5%. Some European nuclear power plants already implemented this method, e.g. Beznau [7].

4 CONCLUSIONS

Krško NPP implemented modifications and replaced some of the equipment on the secondary island to improve plant reliability and efficiency. The plant efficiency was increased after replacing the LP Turbines and HP Feedwater heaters in outage 2006. During outage 2007, two Moisture Separator Reheaters were replaced together with LP heaters in the condenser neck. It was expected that these changes would increase the generator output by at least 4 MW. However, after the plant stabilization at full power, the power output was not as high as expected. With the detailed analysis of several important plant aspects it can be concluded that fouling and scaling build-up inside venturies are the most likely reason for the missing MW’s. Supporting this idea, there are certain parameters independent from fouling, which exhibit lower current values compared to the previous cycle data. Krško NPP plans to perform actions to confirm the fouling and to prevent such situation in the future.

REFERENCES

[1] Plant Support Engineering: Methods for Monitoring and Adjustment of Reactor Power Measurement Drift. EPRI, Palo Alto, CA: 2007, 1014875

[2] Nuclear Feedwater Flow Measurement Application Guide: EPRI, Palo Alto, CA: 1999, TR-112118

[3] Thermal Kit New Low pressure Turbines T4-A6341, Rev. 9

[4] ASME PTC 12.4-1992 “Power Test Code for Moisture Separator Reheater”

[5] ASME PTC 12.1-2000 “Power Test Code for Feedwater Heater”

[6] Drawing 556-30275 “16" FEEDWATER FLOW ELEMENTS” rev. 2

Proceedings of the International Conference Nuclear Energy for New Europe, Portorož, Slovenia, Sept. 8-11, 2008

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[7] T. Hungerbühler, M. Langenstein, “Tracer test method and process data reconciliation based on VDI 2048 – Comparison of two methods for highly accurate determination of feed water mass flow at NPP Beznau”

[8] Verein deutscher Ingenieure (VDI); “Messung der Dampfnässe“; VDI 2043; 07/1999

[9] M. Langenstein, J. Jansky, “Process Data Reconciliation in Nuclear Power Plants”; BTB-Jansky GmbH

Proceedings of the International Conference Nuclear Energy for New Europe, Portorož, Slovenia, Sept. 8-11, 2008