Water level measurement uncertainty during BWR instability

ELSEVIER

Nuclear Engineering and Design 151 (1994) ! 73-184

Nuclear Engineed.ng andDeslgn

Water level neasurement uncertainty during BWR instability R.C. Torok a, T.C. Derbidge b, J.M. Healzer c Electric Power Research Institute, PO Box 10412. Palo Alto, CA 94303, USA b Tile Research Partnership, 814 Henrietta Avenue, Sunnyvale, CA 94086. USA c S. Levy Incorporated, 3425 South Bascom Avenue, Campbell, CA 95008-7006. USA

Received 25 October 1993; revised 3 March 1994

Abstract This paper addresses the performance of the water-level measurement system in a boiling water reactor (BWR) during severe instability oscillations which, under some circumstances, can occur during an anticipated transient without SCRAM (ATWS). Test data from a prototypical mock-up of the water-level measurement system was used to refine and calibrate a water-level measurement system model. The model was then used to predict level measurement system response, using as boundary conditions vessel pressures calculated by RETRAN for an ATWS/ins~ability event. The results of the study indicate that rapid pressure changes in the reactor pressure vessel which cause oscillations in downcomer water level, coupled with differences in instrument line lengths, can produce errors in the sensed water level. Using nominal parameters for the measurement system components, a severe instability transient which produced a 0.2 m peak-to-minimum water-level oscillation in the vessel downcomer was predicted to produce pressure difference equivalent to a 0.7 m level oscillation at the input to the differential pressure transmitter, 0.5 m oscillation at the output of the transmitter, and an oscillation of 0.3 m on the water-level indicator in the control room. The level measurement system error, caused by downcomer water-level oscillations and instrument line length differential, is mitigated by damping both in the differential pressure transmitter used to infer level and in the control room display instrument.

1. !ntroduetion Recent t h e r m a l - h y d r a u l i c instability events at domestic and foreign B W R s have led to renewed interest in instability phenomena. Several new studies have been conducted to predict system behavior during such an event (Cheng, 1993; GE, 1992a; Wulff, 1991). T o apply these analyses to plant operation, the effects o f the unstable behavior on the plant instrumentation and on the control r o o m water-level display device must be understood.

Downcomer water level is an important performance indicator, particularly during an ATWS, because the emergency operating procedures (EOPs) use water level both as an input and as a controlled parameter. In some ATWS scenarios, the EOPs require the operator to control water level manually at a predetermined target elevation. It is important that the control r o o m waterlevel indication be accurate er that pessib!e sources o f error be understood, particularly during any operator action based on water level.

0029-5493/94/$07.00 ~(?:1994 Elsevier Science S.A. AI! rights reserved SSD! 0029-5493( 94)00856-T

R.C. Torok et al./ Nuclear Engineering and Design 151 (1994) 173-184

174

Water-level signals are also used to activate a number of safety systems automatically; for example, emergency core cooling. Measurement errors during an instability event, if sufficiently large, could result in inappropriate actuation of such systems. The BWR water-level measurement system, shown schematically in Fig. l, infers the downcomer level from differential pressure (d.p.) measurements between vessel pressure taps located above and below the water level. The vessel taps are connected to a d.p. transmitter through waterfilled instrument lines that may be up to 60 m long and are nearly a!ways of different lengths, the line from the condensing chamber being longer. The d.p transmitter output is used to drive control room displays and provide input to emergency systems that activate automatically on high- or low-water-level indication. The focus o" the present study is the water-level measurement system performance during a BWR instability event, while the primary system remains pressurized. There are other water-level measurement concern= when the primary system is depressurized, including flashing and release of non-condensible gases in the instrument lines. Since instrument lines are not insulated, the possibility of flashing is minimized. Absorption of noncondensible gases is a problem in the instrument line connected to the condensing chamber, but

only when the condensing chamber is not vented, as shown in the level measurement system arrangement in Fig. 1. For those- BWRs where degassing is a problem, back-flush systems either are or will be installed to eliminate the build-up of non-condensible gases in the condensing chamber instrument line and resolve this problem. The present study examines level measurement errors which result from power oscillations in the reactor core, with attendant oscillations in flows and pressures throughout the reactor pressure vessel. These oscillations have a period of 2 to 5 s, and for severe instability produce rapid changes in the core inlet flow, and vessel pressurization rates in excess of 350 kPa s- ~. Under such conditions, the dynamic effects of accelerating the fluid in the downcomer can become large enough to significantly distort the level measurement inferred from the pressure taps on the vessel. Transmission of the rapidly changing pressures through instrument iines of differing length introduces different delay times, which further distort the level signal. The d.p. transmitter and control room display instruments can also alter the pressure signal, but they filter the pressure difference signal if it is changing too rapidly for the instruments to follow accurately. In an earlier study (Jensen, 1993) of level measurement system errors during system pressure oscillations, water-level oscillations were mug-

TAP "-"------~..,.= CONDENSING

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Fig. I. BWR level measurement system schematic.

R.C. Torok et al. ] Nuclear Engineering and Design 151 (1994) 173-184

nified and shifted in phase with respect to the actual water level. Calculations from this earlier study showed that 0.3 m oscillations in the downcomer water level corresponded to 0.7 m oscillations on the control room recorders. The difference between the actual (downcomer) oscillation amplitude and the indicated (control room) amplitude was attributed to two factors. First, the acceleration of the water in the downcomer that results from large pressure excursions in the reactor core produces an acceleration pressure component that affects the lower (submerged) vessel tap pressure, but not the upper tap pressure. Second, the rapid pressure changes in the vessel pressure are sensed at both vessel taps, but communicated to the d.p. trano,nitter through instrument lines of unequal lengths. This introduces in the pressure signals a timing offset that the transmitter sees as an additional presgure difference. The present ,;tudy was undertaken to provide confirmation ot the water-level measurement system model presented in Jensen and Healzer (1993). A representative water-level measurement system was constructed in a test facility, using hardware typical of that found in operating plants. The test facility was also modeled, including acoustic-wave effer'tS in the instrument lines and the response characteristics of the instruments. Data from the tests provided direct confirmation of the level-measurement errors caused by differences in the instrument line lengths. The tests also provided benchmarking data for the model, which was then used to predict the plant level measurement system response using the reactor pressure vessel thermal-hydraulic behavior under ATWS]instability conditions, from 2eao~en and Healzer (1993).

2. Water-level measurement system survey

To help ensure that the test and analysis results would be representative of operating plant installations, the water-level instrument line piping layouts of several operating BWRs were reviewed. The typical BWR has three different ranges of water-level measurement: a narrow range used during normal operations, a wide range and a fuel

175

zone range used during refueling. The first two ranges have two independent d.p. measurements that determine level; the fuel zone range usually determines level based on a single d.p. measurement. During an ATWS event, particularly if the water level was manually controlled, the widerange level range would be used. The survey therefore focused on the piping layout for the wide-range 'wel measurement system. Data were bathered for a total of ten operating plants. Complete piping layouts were available for three plants and less detailed information for seven others. The instrument line lengths varied from plant to plant, but the average piping run of thr: upper line, from the condensing chamber to the d.p. transmitter, was 46 m. The average length of' the lower line, from the lower vessel tap and the d.p. transmitter, was 21 m. In every case, the upper line was longer. Hardware components in the lines and the type of piping also varied. Most systems used l-in, Schedule 80 piping within the containment, and tubing outside, near the instrument rack. All systems included multiple elbows or bends, isolation valves and other fittings. As many as a dozen elbows or bends were not uncommon in the upper line. In addition to the d.p. transmitter, tees to other pressure sensors were present in most lines. For all systems, both upper and lower lines included excess-flow check valves (EFCVs) located just outside the containment penetration. A few systems also included flow-limiting orifices at these locations. It was not known if the large amplitude-pressure oscillations predicted for BWR instability conditions would be damped by the instrument line and its hardware. One objective of the experimental part of the study was to investigate this, using typical instrument line hardware. Of all the instrument line hardware, the EFCVs were judged to have the greatest potential for pressure drop and for damping of the pressure signal through the instrument lines. The EFCVs have small flow passages and relatively tortuous flow paths. To investigate the effect of these and other components in the instrament lines, the test apparatus included EFCVs, tees to parallel pressure sensors and several short radius elbows. While this assortment of instrument line hardware was not identi-

176

R.C. Torok et al./ Nuclear Engineering and Design 151 (1994) 173-184

The d.p. transmitter was a Rosemount Model 1153, Series D, Code 5 variable-capacitance differential-pressure transmitter, typical of those used in level-measurement systems. A Tracor Westronics E3 Series strip chart recorder was used to represent the control room output device. Pressures during the tests were recorded using a digital data acquisition system with a 0.002 s sampling period. The instrument lines were stainless steel tubing of 9.5 mm diameter (3/8 in) with a 0.9 mm wall thickness. For the baseline tests, line lengths of 23 m and 46 m were used to represent the upper and lower instrument lines. The effect of tube wall flexibility on the propagation of the pressure signals was small both in this thin-wall tubing and in a typical level-system instrument line that uses 1-in, Schedule 80 pipe. As already noted, baseline tests included several short radius elbows and EFCVs located near line mid-points, as identified in the measurement system survey dis,:ussed earlier. To fill, the system was first evacuated with a vacuum pump and then filled with deaerated water. The system was then held under pressure for several days prior to testing to ensure that no residual gas remained. While it is unlikely that

cal to any particular level-measurement system from the survey, it was representative of the type of hardware found in the instrument lines. A more detailed description of the experimental apparatus is provided in the next section.

3. Experimental apparatus A schematic of the experimental apparatus is shown in Fig. 2. A pressure generator, designed to produce a ramp forcing function with a pressurization rate representative of those predicted for an ATWS/instability event, was used to represent the reactor pressure vessel for a series of transient response tests. The pressure generator consisted of a cylinder which was partially filled with water and connected at the bottom to the instrument lines. To generate the desired pressure ramp, the gas volume was rapidly charged with nitrogen from a high-pressure reservoir through a fast-acting shut-off valve. The instrument lines, which were of different length, terminated at the two input ports of the d.p. transmitter (DP). In addition to the d.p. measurement, gage pressures (PI, P2 and P3) at the pressure generator and the d.p. transmitter input ports were recorded.

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R.C. Torok et al. / Nuclear Engineering and Design 151 (1994) 173-184

similar care would be taken to evacuate and fill the level-measurement system in a plant, operation at high pressure for extended periods would ensure tha'~ any residual gas left after filling would be dissolved. A manually operated pistor~ was used to adjust system pressure during fill and leak cheek operations and in pressure-set tests, which were conducted to measure the system stiffness. A screw-driven piston was used to pressurized a small cylinder connected to the instrument lines. This cylinder was valved out of the system during the dynamic tests. A second pressure generator, which consisted of a motor-driven reciprocating piston in a cylinder, was used to produce a variable frequency sinusoidal driving function for oscillatory tests. In these tests, very short lines were used to connect the pressure generator to the d.p. transmitter to eliminate low-frequency instrument line acoustic response. The measured frequency response characteristics of the d.p. transmitter and the strip chart recorder from these oscillatory tests provided benchmark data for modeling these components.

4. Test results

Three types of test were conducted in this study. 4.1. Pressure-set tests

These tests were conducted to measure the system stiffness. Pressure-set tests were conducted after every system fill to measure system stiffness and confirm the absence of non-condensible gases. A manually adjusted piston was used to apply a known volume displacement to the system and the pressure change resulting from this volume displacement was measured. The repeatability of these tests was an important indicator of the condition of the system prior to the transientresponse tests. Typical results from pressure-set tests are shown in Fig. 3. The steepest curve (water only) corresponds to the compressibility of the water

177

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volume in the test apparatus, based on handbook values of the bulk modulus for water and assuming the piping system is perfectly rigid. The other curves show the actual system compressibility, both with and without the d.p. transmitter in the system. Note that, when the transmitter is ineluded in the system, there is a considerable change in system stiffness at low pressures. At higher pressures, the pressure vs. displacement curves become parallel, indicating similar system stiffness. The effect of the differential-pressure tran,,~,~il~.er on the system response at low pressures was also apparent in the transient-response tests. To obtain realistic results in the transientresponse tests, the system must behave as it would at normal plant operating pressures, near 6.9 MPa. Results in Fig. 3 indicate that this is achieved if the transient-response tests are performed at pressures on the straight part of the stiffness curve, above 1400 kPa. 4.2. Transient-response tests

In the transient-response tests, the instrument lane and d.p. transmitter responses to a pressure ramp from the pressure generator were measured. Since both instrument lines were connected to the pressure generator, there was no difference between the line pressures at the pressure generator. If the ramp was applied slowly, the pressures at the end of the instrument lines connected to the

R.C. Torok e! al. / Nuclear Eng#1eering and Design 151 (1994) ! 73-184

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d.p. transmitter would also rise together and there would, for all practical purposes, be zero differential pressure. For ramp rates typical of those predicted for severe instability, a pressure difference at the d.p. transmitter was observed. The difference in transient time of the pressure signal, due to the difference in the line lengths, resulted in differential pressure which was osci|latory in behavior. If one of the lines were infinitely long or blocked, the d.p. transmitter would measure the full pressure rise from the unblocked line. For the pressure oscillations predicted for the ATWS/ instability event used in this study, the pressure rise would be as large as 300 kPa, which would be equivalent to a water-level change of nearly 30 m. Fig. 4 shows the ramp driving pressure for a typical transient response test and the pressures at the d.p. transmitter ends of the two lines. The pressure ramp is delayed in arriving at the line ends and exhibits a step-like behavior. This is characteristic of the transient response of a closed-end line being rapidly pressurized. The steps are produced by the superposition of the reflected pressure waves in the lines. At a closed end the wave is reflected and is doubled in magnitude. At the reservoir end it is inverted. The step duration corresponds to the round-trip acoustic travel time between the driver and the ends of the lines. The two lines therefore have different step duration, proportional to their lengths. The difference between the two line end pressures is shown in Fig. 5. The damped sinusoidal

behavior is a function of the pressure ramp rate, the line length and length difference between the lines. In the baseline tests, the lines were 23 m and 46 m in length, typical of the operating-plant instrument lines surveyed in this study. The peak differential pressure induced by the ramp is approximately + 10 kPa. This pressure difference is equivalent to a change of + 1 m in the indicated level. Baseline transient-response tests were conducted with nominal instrument line lengths, with the EFCVs and elbows in the lines, at a pressure of 1400 kPa and without the d.p. transmitter. Additional parametric tests were performed with the d.p. tr~,nsmitter in the system, without EFCVs and elbows, at different system pressures and with different lit0:: lengths. A complete reporting of all the test results can be found in Derbridge and Healzer (1993). Fig. 6 shows the line end pressure difference with the d.p. transmitter included in the system. Comparison of the two figures shows that the transmitter provided additional damping. In Fig. 6, the peaks in the end pressure difference signal are slightly lower, and the oscillation damps out much more quickly than in the test without the d.p. transmitter (Fig. 5). Also shown in Fig. 6 is the pressure difference from the output of the d.p. transmitter. The transmitter output shows even greater damping, with significant attenuation of the peaks and a slight time delay. Other tests conducted with clean instrument lines (no check valves, elbows, etc.) demonstrated

R.C. Torok et al. [ Nuclear Engineering and Design 151 (1994) 173-184 16-

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that the instrument line hardware had only a minor effect on system response. Tests with line lengths increased to 46 m and 91 m (Fig. 7) showed that line end pressure difference was increased and the frequency was approximately halved. In summary, the transient-response tests provided data at typical instrument line lengths which demonstrated that pressure ramps typical of those predicted during severe instability events will induce pressure oscillations which will be interpreted as changes in water level by the level measurement system. The tests also demonstrated the damping effect of the d.p. transmitter on the

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Fig. 8. Results from pressure oscillator tests.

level-measurement system response, the negligible damping effects of instrument line hardware, and the increased amplitude resulting from longer instrument lines and greater line length differences. 4.3. Oscillator), tests

These tests were used to measure the frequency response characteristics of the d.p. transmitter and the strip chart recorder. The oscillatory pressure generator was connected to one side of the d.p. transmitter, with the other side pressurized by a reservoir. Fig. 8 shows typical results from these tests. The pressures driving the transmitter are shown, along with the transmitter output. The amplitude of the input oscillation is approximately constant, while the amplitude of the transmitter output decreases with increasing frequency. Results from these tests provided input for modeling the d.p. transmitter.

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In an earlier water-level instrument measurement system study (Jensen, 1993), predictions of the instrument line response were made using a method of characteristics (MOC) model based on material from Wylie and Streeter (1978). With this model, continuity and momentum equations

R.C, Torok et al. Nuclear Engineering and Design 151 (1994) 173-184

180

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of the following form are used:

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=-8Itlul In an early series of shakedown tests, reflections of a pressure impulse were found to decay much more rapidly than was predicted, which indicated that damping was not properly represented in the model. Several authors have studied damping at near-zero flow and have suggested various modeling methods. For the present study, a method by Zielke (1968) was adopted. This method relates the transient wall shear to the instantaneous mean velocity and to the weighted past velocity changes.

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where H = (P/pg + :) is the piezometrtc head, - is the elevation change, p is the fluid density, u is the average velocity in the pipe, c is the acoustic velocity, D is the pipe diameter, g is the acceleration of gravity and rw is the wall shear. In this formulation, the only damping is due to wall shear. For the very low velocities in the instrument lines, the flow will be laminar, and under these conditions the wall shear term is usually expressed in terms of the average pipe velocity and I~, the viscosity:

TES~S W/91,m & 46m LINES NO SYSTEM HARDWARE 140(2kPo SYSTEM PRESSURE DP TRANSMITTER NOT INCLUDED

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tween the model and the data is good. Predictions and data from the transient response tests with shorter line lengths and without the d.p. transmitter are shown in Fig. 10. The differential-pressure response is not as great in these tests. Note that the initial pressure peaks are well predicted, but the model predictions show less damping than the data. Predictions and data for ~e~ts with the d.p. transmitter in the system are shown in Fig. 11. Prediction of this data was more difficult because of the damping added by the transmitter. To model the added damping, a pressurized surge tank was added to the model, connected t o th~ end of each instrument line through an orifice. This modeling change provided added flexibility

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W is an infinite series of exponential functions that was fitted by Zielke with a polynomial. The integral in the added term is replaced with a series when it is implemented in the instrument line model. With this change in the wall shear, the instrument line model was able to predict the decay of pressure pulses in the test facility shakedown tests and was also benchmarked against other independent test data (Derbridge, 1993). Fig. 9 compares data from the transient response tests with instrument line lengths of 46 m and 91 m to predictions by the model. These tests provide the greatest differential pressure response and were the easiest to predict; agreement be-

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5.0

R.C. Torok et al. / Nuclear Engineering and Design 151 (1994) 173-184 80

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and damping to the system. Like the predictions without the transmitter, the initial pressure peaks are well predicted, but the damping is under-predicted. Fig. 11 also shows data and predictions of the d.p. transmitter output. For these predictions, the transmitter input pressure difference was filtered with a first-order Butterworth filter with a cut-off frequency determined from the oscillatory tests. Based on these comparisons, it was concluded that the water-level measurement system model was capable of predicting the initial over- and under-pressure response of the measurement system ar.d a similar damped response following the initia! pressure peaks. Model predictions in which the line hardware components (EFCVs and elbows) were represented as local orifice-type losses showed results similar to the test data. The added hardware increased the high-frequency response and the damping of the system, but did not affect the initial over- and under-pressure peaks in the data.

6. Predictions of BWR ATWS instability response To extend the predictions to BWR ATWS instability, the vessel tap pressures calculated by RETRAN-03 (Paulson, 1991) from the earlier study (Jensen, 1993) were used. This RETRAN

0

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Fig. 12. RETRAN predictions of BWR ATWS transient.

analysis used a point-kinetics core neutronics model and a single-average-channel core hydraulics model. Predictions of power, system pressure and downcomer level are shown in Fig. 12. More detailed analyses, such as the General Electric study (GE, 1992a), have used more detailed modeling, with a three-dimensional core neutronics model and a multi-channel core hydraulics model. The GE study shows power oscillations which va.~, in amplitude and period from cycle to cycle, not as regular as the RETRAN results, but still similar in overall behavior. To model the level measurement system, nominal instrument line lengths were used, along with d.p. transmitter and the strip chart recorder models based on the test results described earlier. To provide boundary conditions for the level measurement system model, one cycle of the system pressure at each of the vessel pressure tap locations was used to form a repeating pressure forcing function. For the cycle chosen, the relative power peak was eight times rated power, the vessel pressure change was about 275 kPa and the downcomer level variation was about 0.2 m. The rate of pressure increase during the pressure rise was about 350 kPa s - ' , similar to pressure rise rates tested. Fig. 13 shows predicted pressure differences at various locations in the water-level measurement system using these boundary, conditions. The lowermost curve shows the static head pressure ch,,.-ge that would result from the downcomer

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R.C. Torok et al. / Nuclear Engineering and Design 151 (1994)j .,3- y-~.,

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Fig. 13. Predictions of level measurement system pressure differences - - input pressure difference to t t z D P transmitter.

Fig. 14. Predictions o f level measurement system pressure differences - - output from the D P transmitter and control room display.

level variation. This is the pressure difference signal that the d.p. transmitter would receive if there were no measurement error. The second curve shows the pressure difference at the vessel, taps. The difference between this and the bottom curve is due to the acceleration of the downcomer liquid by the core flow oscillations. During the level oscillation, when the water level changes direction, the acceleration of the downeomer liquid increases the lower tap pressure, resulting in a short-duration peak in the pressure difference. This peak occurs in every cycle, and just as the water level in the bottom curve changes direction. The top two curves represent the pressure difference at the d.p. transmitter and demonstrate the effect of instrument line length difference. One curve is the predicted pressure difference if the two instrument lines were nearly the same length. In this case, the pressure difference from the vessel taps has not been amplified, but the instrument line acoustic response has been added to the pressure signal. The other curve is for the instrument line le~.~gths typical of those identified in the survey. For this case, the pressure difference from the vessel taps has been amplified because of the sensing line length difference. Fig. 14 shows the damping effect of the level measurement system instruments. The bottom two curves are the same as in Fig. 13, but the top

two curves show the pressure difference after the d.p. transmitter and the pressnre difference corresponding to the level variation that would he displayed in the control room. The pressure difference after the d.p. transmitter is based on the transmitter input pressure difference, filtered through a low-pass filter with the characteristics determined in the oscillatory tests. At the transmitter output, the oscillation amplitude has been attenuated, but is still greater than the corresponding signal at the vessel. The control room display instrument was also modeled with a low9

1

LEVELS DISPLACED BY 1.Ore

~> t LEVEL SHOWN ON STRIP CHART RECORDER <

LEVEL BASED ON DP TRANSMITTER PRESS DIFF

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LEVEL FROM VESSEL PRESSURE TAPS

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DOWNCOMER LEVEL (RETRAN) 133''1

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3

I ....

4

f,,rrf

5

....

6

j ....

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7

R

TIME (5)

....

9

J ....

10

i,,,,~

11

....

12

i ....

13

14

Fig. 15. Prediction of level measurement system I,~vel indication.

R.C. Torok et al. / Nuclear Engineering and Design 151 (1994) 173-184

pass filter, with a response time of 1 s, typical for this type of instrument. This resulted in additional damping of the d.p. transmitter output. Fig. 15 shows the pressure differences converted to water level and compared to the level variation from the R E T R A N calculation, which was based on actual water inventory in the vessel downcomer. As noted earlier, if there was no waterlevel measurement error, the level variation shown in the bottommost curve would be displayed in the control room. Most of the water-level measurement distortion is evident at the vessel pressure taps and is due to acceleration of the downcomer liquid as discussed earlier. The length difference in the instrument lines further distorts the pressure difference signal, while the d.p. transmitter and strip chart recorder successively damp the signal. The pressure difference oscillations at the omput of the d.p. transmitter are significantly greater than would result from the actual level excursions in the downcomer, and the transmitter output is used by the reactor safety systems, so this pressure difference is important. The uppermost curve in Fig. 15 is the level indication displayed in the control room. By adjusting the display instrument damping, it would be possible to reduce the magnitude of the level oscillations back to that actually occurring in the vessel. As shown in the figure, for the expected display instrument response characteristics, the variation of level indicated in the control room would be not quite twice that of the actual level variation. Adding electronic or mechanical damping to the level measurement system might be useful in mitigating the effects of BWR instability on the water-level measurement. Filtering at the d.p. transmitter output signal could help preclude inadvertent actuation of safety systems that are activated based on water-level setpoints. At the same time, it might help avoid misleading the operator about the size of the water-level swings and the operability of the measurement system. Finally, it should be noted that these results are based on nominal instrument line lengths and instrument response characteristics. Changing any of these would change the water-level measurement system response. Also, a simplified represen-

183

tation of the pressure vessel behavior during the severe instability transient has been used in these predictions, and the actual level oscillations would not be as regular in amplitude or period. More detailed analysis of the instability event (Cheng, 1993; GE, 1992a) shows power and pressure oscillations that vary from cycle to cycle. Many of the power oscillations are smaller in amplitude than the eight times rated power used in this study, and a few are much greater. Furthermore, the severity of the instability oscillations may be reduced or avoided altogether by operator actions, as discussed in GE (1992b) and Jensen and Healzer (1992).

.

Conclusions

Rapid changes in reactor vessel pressure, typical of those predicted for a BWR ATWS/instability event, can produce errors in the level oscillations indicated by the level measurement system. • Level measurement errors are related to the rates of change of system pressure and to the instrument line lengths and difference in the lengths of the upper and lower instrument lines. • The d.p. transmitter and control room display instrument damp the measurement system response to the pressure transients, but, for the level system instruments used in this study, the indicated downcomer level oscillation amplitudes were still greater than the actual level oscillations. • For the instability transient used in this study, 0.2 m oscillations in the downcomer level were predicted to produce a 0.5 m level oscillation signal at the d.p. transmitter output and a 0.3 m level oscillation on the control room display instrument.

Acknowledgments The authors would like thank the editors of

Nuclear Engineering and Design for the opportunity to participate in this issue honoring Dr. Noyak Zuber. He has made enormous contributions

184

R.C Torok et aL / Nuclear Engineering and Design 151 (1994) 173-184

to the technology associated with the nuclear power industry over the past 40 years, ranging from his insightful critical heat flux and void-fraction models to his work to define a rational basis for scaling complex thermal-hydraulic experiments. As important as his technical achievements, his leadership in seeking practical and lastieg solutions to the difficult problems facing the nuclear industry in the United States is even more important. When honored with the ANS Technical Achievement Award a few years ago, Novak remarked that there were no present or future problems that the technical and scientific talent in our country could not solve, but we must have the determination to address the problem and maintain technical integrity in seeking the solutien. We look forward to many more important technical contributions from Novak and hope he will continue to remind us all of the need for determination and technical integrity in our work. References H.S. Cheng and U.S. Rohatgi, Instability due to a two recirculation pump trip in a BWR using RAMONA-4B computer code with 3D neutron kinetics, National Heat Transfer

Conf., August 8-11, 1993, Atlanta, GA, ANS Proc. 7 (1993) 265-271. T.C. Derbidge and J.M. Healzer, Water level measurement uncertainties during BWR instability events - - Test and an~!ysis, EPRI TR-103292, November 1993 (Electric Power Research Institute). GE, ATWS rule issues relative to BWR core thermal-hydraulics stability, NEDO-32047, February 1992a (General Electric Co.). GE, Mitigation of BWR core thermal-hydraulic instabilities, NEDO-32164, February 1992b (General Electric Co.). P.J. Jensen and J.M. Healzer. Water level reduction to suppress instability during a BWR ATWS, NSAC-163, October 1992 (Electric Power Research Institute). P.J. Jensen and J.M. Healzer, Water level instrument response during BWR instabilities, National Heat Transfer Conf., August 8-11, 1993, Atlanta, GA, ANS Proc. 7 (1993) 260-264" also see NSAC-162, October 1993 (Electric Power Research Institute). M.P. Paulson et al., RETRAN-03 - - A program for transient thermal-hydraulic analysis of complex fluid-flow systems, EPRI Report NP-7450-CCML, Vols. I-3, January 1991 (Electric Power Research Institute). W. Wulff, H.S.Cheng, A.N. Mallen and U.S. Rohatgi, BWR stability analysis with the BNL Engineering Plant Analyzer, BNL-NUREG-52312, November 1991 (Brookhaven National Laboratory, Upton, NY). E.B. Wylie and V.L. Streeter, Fluid Transients, McGraw-Hill, 1978, pp. 31-43. W. Zielke, Frequency-dependent friction in transient pipe flow, J. Basic Eng., ASME (March 1968) 109-115.