Status, needs and perspectives in measuring of pressure drops

Nuclear Engineering and Design 354 (2019) 110218

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Status, needs and perspectives in measuring of pressure drops ⁎

Klaus Umminger , Lars Dennhardt, Peter Burwitz, Simon Schollenberger

⁎

T

Framatome GmbH, Erlangen, Germany

A R T I C LE I N FO

A B S T R A C T

Keywords: Pressure drop measurements Thermal-hydraulics Nuclear power plants Integral test facilities Accident behavior Reactor safety

Together with the measurement of temperature, the measurement of pressure or pressure drop is one of the first and most used measurements in science and plays a central role in the detection and recording of flow processes in thermal-hydraulic systems. In nuclear engineering this not only applies to detection and recording of the processes in the nuclear power plants (NPP) themselves but also to the test facilities used for analysis and interpretation of important thermal-hydraulic processes. While pressure drop measurements – belonging to the so called ‘conventional instrumentation’ – have been primarily used for single phase flow situations, the benefit of its use and application in two-phase flow conditions can also be clearly demonstrated. A comprehensive review of two-phase instrumentation methods, including pressure drop, has been provided by Hewitt in the seventies and eighties of the last century (Hewitt, 1978, 1982). Even though the development of new ‘advanced instrumentation’ has made enormous progress in recent years, pressure drop measurements will remain very important in the future. Pressure drop measurements cover a real broad spectrum of application, so that not all aspects can be addressed in this paper. Under consideration of the information provided in the book ‘Thermal Hydraulics in Water-Cooled Nuclear Reactors’ (Umminger and D’Auria, 2017), this paper will focus on pressure drop measurements in integral test facilities for PWR with some hints to applications in NPPs. Status and use will be discussed on the basis of some typical examples. Important issues, further needs and perspectives in measuring of pressure drops will also be outlined.

1. Introduction Pressure is a scalar physical variable, which describes the force per unit area exerted by a fluid on a surface. Together with the measurement of temperature, the measurement of pressure or pressure difference is one of the first and most used measurements in science. In the middle of the 17th century, the Italian physicist Evangelista Torricelli invented the barometer for the evaluation of the atmospheric pressure. The technology for pressure measurements has evolved over time and since 1930 first pressure transducers with diaphragms reacting with a displacement on a pressure change and sensors converting these displacements into electrical signals have been in use. Since beginning of this century, the piezo-resistive sensor technology is the most common one. Different instrumentation methods including pressure drop

measurements in multi-phase systems have been comprehensively reviewed by Hewitt around 40 years ago (Hewitt, 1978, 1982). Pressure measurement, which is one of the so-called “conventional measurements,” continues to play a central role in the detection and recording of flow processes in thermal-hydraulic systems such as nuclear power plants and corresponding test facilities operated to investigate in detail relevant T-H phenomena expected to occur in assumed NPP accident scenarios In NPPs pressure drop measurements are mainly used for operational- and safety-related level detection and in many cases the measurement signals are employed for control purposes or even for initiation of countermeasures under accident situations. The qualification of measurement installations in general – and the Δp sensors in particular due to their presumed proximity to the relevant components – is a precisely defined procedure and, depending on the safety classification

Abbreviations: BWR, boiling water reactor; BIC, boundary and initial conditions; CCFL, counter current flow limitation; CFD, computational fluid dynamics; ECC, emergency core cooling; HPSI, high pressure safety injection; ITF, integral test facility; LB-LOCA, large break loss of coolant accident; LOOP, loss of offsite power; NC, natural circulation; NPP, nuclear power plant; PRZ, pressurizer; PWR, pressurized water reactor; RCP, reactor coolant pump; RCS, reactor coolant system; RHRS, residual heat removal system; RPV, reactor pressure vessel; SB-LOCA, small break loss of coolant accident; SETF, separate effect test facility; SG, steam generator; SGTR, steam generator tube rupture; SOAR, state of the art report; T-H, thermal hydraulics; UP, upper plenum ⁎ Corresponding authors. E-mail addresses: [email protected] (K. Umminger), [email protected] (S. Schollenberger). https://doi.org/10.1016/j.nucengdes.2019.110218 Received 26 June 2019; Received in revised form 18 July 2019; Accepted 18 July 2019 Available online 07 August 2019 0029-5493/ © 2019 Elsevier B.V. All rights reserved.

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Nomenclature g p v Δp ρ

d e f l ref st t

gravitational acceleration pressure velocity pressure drop density

dynamic elevation friction local reference static total

Subscripts a

acceleration

drop measurements also play an important role in connection with computational fluid dynamics (CFD) codes. Depending on the boundary conditions, pressure drops may be highly affected by turbulence, both in single- and two-phase flows, therefore pressure drop measurements are also useful for turbulence model assessment. Pressure drop across spacer grids in a fuel rod bundle characterized by local turbulences is only one typical example. As reported in the literature, this field of application has been addressed within numerous research activities in recent years and is not subject of this paper. The pressure drop measurement under complex two-phase flow conditions and especially the interpretation of the gained signals is surely still a challenge and needs further developments e.g. with respect to further experimental campaigns and improved transducer arrangements. So that despite the promising new developments of other twophase flow measurement techniques (as illustrated during the Specialist Workshop on Advanced Instrumentation and Measurement Techniques for Nuclear Reactor Thermal Hydraulics, SWINTH, organized in Livorno, Italy, June 2016) the pressure drop measurement will – partly in combination with other measurements – remain of great importance in the future in this field.

of the system to which they are assigned, a demanding one with respect to the sometimes harsh testing boundary conditions (e.g. LOCA conditions). All major suppliers or manufacturers of components for the nuclear industry set up laboratories dedicated to the conduction of qualification tests for sensor equipment. For the operation of nuclear power plants and conduction of experiments at test facilities a regular zero balance of the Δp sensors is a prerequisite to assure the significance of the measurement signal. For the PWR and the test facilities, this is done on a regular basis during outage or test preparations, respectively. The test cycles for the individual sensors are obtained from the component ageing monitoring and management concept implemented in many nuclear installations which is a prerequisite to maintain or extend the operation license in many countries. An increasing number of utilities implement computation-supported tools for the analysis and requirements on documentation in the frame of the component ageing management concepts. Test facilities are – compared with NPPs – normally even more intensively equipped with differential pressure measurements in order to capture in addition also the irreversible pressure drop in all relevant sections of the system under a wide range of boundary conditions. This is especially the case for integral test facilities used to investigate the thermal-hydraulic system behavior of PWR under accident situations and to analyze the relevant phenomena in different scenarios. In this context the importance of experimental results including pressure drop measurements for the validation of thermal-hydraulic computer codes which are again used for the extrapolation of experimental results to real plants has to be stressed. Even though the focus in this paper will be put on the use of pressure drop measurements in context with T-H system codes and corresponding integral test facilities, it should be mentioned that pressure

2. Measurement techniques Pressure is usually directly measured by the applied force, whereas in practice, pressure measurements give a pressure difference between two systems or between two points in a system. For pressure drop measurements in ITF and NPP the typical measurement configuration consists of two pressure tappings (points to be investigated) from where the pressures are forwarded by two hydraulic lines to the two channels of a pressure transducer (Fig. 1, left). The commonly used diaphragm sensor inside the transducer reacts on the pressure difference exerted on

Fig. 1. Typical arrangements for differential pressure measurements (schematically) (left: pressure drop in horizontal flow channel, right: level measurement). 2

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in cross-sectional area or change in density. In transient processes, another term representing the local acceleration over time has to be added to this convective acceleration. The pressure drop due to gravity and the acceleration pressure drop are reversible components, while the friction pressure loss represents the permanent or irreversible pressure drop. The local pressure loss consists of a reversible and an irreversible part whereas the irreversible part is of main technological relevance. The measurement of the pressure drop generally includes all of the above mentioned components. Under certain conditions or if certain influencing factors can be ignored or can be derived from other measurements, the pressure drop measurement can be used not only to determine the permanent pressure loss (friction and local), but other important physical variables can also be derived, such as:

it with a displacement or deformation which is again converted by an electrical sensor (inductive, capacitive, piezo-resistive, etc.) into electrical signals proportional to the differential pressure of interest. The pressure transmission lines (sometimes called pressure sensingor impulse-lines) have a diameter of 5–10 mm and are filled by incompressible fluid, typically liquid water. It must be ensured that liquid does not vaporize at the contact with the high temperature fluid and that (reasonably) non-condensable gases are not present to form bubbles or even are not dissolved in the fluid. Presence or occurrence of gas or steam bubbles in the pressure sensing lines may heavily affect the measured values. In some cases (especially when differential pressure measurement is used for level detection), condensate vessels are integrated in the pressure sensing lines in order to keep the level of water in the sensing line constant (see Fig. 1, right). Due to this measurement principal, pressure measurements generally give a pressure difference between two points or the difference in pressure between a measurement point and a reference pressure. Depending on the reference pressure, the measured pressure difference is defined as follows:

– Level (swell level/collapsed level). – Mean fluid density. – Mean void fraction. In steady-state flows of incompressible fluids in flow paths with constant cross-section, the acceleration term is omitted completely. For horizontal flows the head (gravity) component is also omitted and the measured differential pressure in fully developed flow is the remaining pressure loss in the flow path (Δpf + Δpl). For vertical flow sections the impact of the static head is not negligible and knowledge about the gravity term is required which is surely given under single phase conditions. Under 2-phase conditions additional information (measurements) about void fraction or density would be necessary. Besides the provision of information on pressure losses along flow paths which are usually difficult to assess by means of computational models alone, pressure drop measurements in the ITF are an easy way to provide information on the distribution of coolant and water within the primary or secondary-side circuit, respectively. Under stagnant flow conditions or at very low fluid velocities, the acceleration and the irreversible pressure loss components are negligible and the measured pressure drop represents the static head of the fluid columns. This situation allows – with knowledge of the density of the water and in case of the steam – the direct determination of the water level according to Eq. (3.2). Some remarks regarding the use of pressure drop measurements are given below:

– Absolute pressure: If the reference pressure is zero (i.e., complete vacuum). – Gauge pressure: if the reference pressure equals the atmospheric pressure. – Differential pressure: if both measuring points are located within the system (as shown in Fig. 1), i.e., the pressure drop along the flow path which is mainly of interest later. In flowing media the total pressure is made up of a static and a dynamic component. pt = pst + pd (2.1) pst – the hydrostatic pressure (static pressure) is the pressure which the observer would measure when moving in the fluid too. pd – the hydrodynamic pressure (velocity or dynamic pressure) results from the kinetic energy of a flowing liquid and increases quadratically with velocity (pd = ½ ρ v2) pt – the total pressure describes the pressure that would be measured if the flow is brought to rest isentropically. The individual components can be recorded by appropriate attachment/orientation of the pressure sensing lines. The static pressure is measured via a pressure sensing line which connects directly to the wall of the flow path, at which the opening is parallel to the flow. The pressure drop in a flow path is usually recorded by measuring the static pressure along the flow path or by measuring the difference in static pressure between 2 points according to the configuration shown in Fig. 1, left. If the opening of the pressure sensing line is perpendicular to the flow, the total pressure is given (with vacuum pressure as the reference). The dynamic pressure can be determined from the measurement of the difference between total pressure and the static pressure (Prandtl probe, Pitot tube, Fig. 2). With known or measured fluid density (ρ), the flow velocity can be determined from the dynamic pressure (ρ v2).

3.1. Filling level Using the static pressure of water columns to take differential pressure measurements of the water filling level or water reservoirs is a common method used both in nuclear plants and also associated testing systems. If tanks are filled with water and steam (idle system) and the

3. Use of pressure drop measurements in nuclear T-H According to the basic theorem of conservation of momentum, the measured static pressure drop which a fluid is exposed to flowing through a flow path is made of several components including gravity, acceleration and pressure losses (which are again composed of frictional and local pressure losses), (Umminger and D’Auria, 2017) Δ p= p1 − p2 = Δpe + Δpa + Δploss , with Δploss = Δpf + Δpl (3.1) where Δpe is the pressure drop due to gravity and the result of the weights of fluid columns Δpe = ρ2 ·g· h2 − −ρ1 ·g· h1 (3.2)Δpa is the acceleration pressure drop due to change in flow velocity along the flow path as a result of change

Fig. 2. Pitot tube for the determination of fluid flow velocities by comparison of total pressure (velocity pressure plus static pressure) with static pressure alone. 3

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due to dynamic effects (e.g., wall friction, interfacial two-phase friction, local pressure drop). This often applies to flow conditions in which a specific steam quantity flows from bottom to top through standing water in tanks or vertical pipes. As an example, Fig. 4 shows the determination of the mean densities and the trend of the swell level derived from differential pressure measurements by linking several Δp sectors along the core of the RPV in the PKL test facility.

two phases exist in a separated form, the differential pressure measurement is a simple and very precise method for determining the filling level; this method requires knowledge of the densities of the two phases (e.g., by means of additional absolute pressure and temperature measurements). The water filling level can be determined according to Eq. (3.2). Wherever there are high pressure levels, the static pressure of the steam can no longer be neglected and should be taken into account when determining the water filling level. Under flow conditions, the impact of the dynamic components can – depending on the mass flow rates or fluid velocities – no longer be neglected. In ITF, this dynamic impact may be however – with knowledge of the mass flow rates, the geometrical configuration and the usually known pressure loss coefficients – eliminated. The way to correct the originally measured pressure drop should be considered in the measurement error documentation. Under two-phase conditions, with the knowledge on density in the components, Δp may be converted to collapsed level indicators. As an example, Fig. 3 shows the determination of the collapsed level (considering also the reference columns in the pressure sensing lines) realized in a primary side component of the PKL test facility (Umminger et al., 2012).

3.3. Friction and local pressure drops in ITF The reactor-typical simulation in ITF of relevant (pressure drop dominated) phenomena under NPP accident situations is evidently facilitated by a 1:1 replication at least of the single phase pressure drops of the reference plant for the entire system and as far as possible for the individual components (OECD/NEA/CSNI, 2016; Umminger et al., 2012; Nakamura et al., 2009). Characterization tests on pressure drops in ITF, ideally performed under different flow conditions (mass flow rates) provide a useful set of input data for the computer codes. The knowledge of results from such characterization tests is also important for the analysis and interpretation of the essential phenomena observed in the experiments. Fig. 5 shows the evolution of the different pressure drop components along a single loop as derived from measurements in the PKL test facility for full flow conditions (RCPs in operation) and low flow (typical for natural circulation) conditions. More detailed explanations can be found below.

3.2. Swell level It is not possible to directly determine the swell level (separation between the two-phase mixture and the phase with a predominant steam proportion) by measuring the differential pressure alone; additional measuring information is required. However, conducting several differential pressure measurements in succession can pinpoint the position of the swell level to at least a segment of the corresponding differential pressure measurement. Also, if the mixture level changes (increase/decrease), it is possible to at least trace the trend of the temporal progression of the mixture level. For instance, usually information can be derived from the Δp measuring signal curve about the passing of the swell level at a pressure tapping point and therefore the transition to a predominant steam proportion within the associated Δp measurement (when the swell level decreases). If a measuring range section is completely filled with a two-phase mixture and if the water and steam density is known, it is possible to derive from the measured differential pressure a constant assumed mean density of the mixture or the mean void fraction. Strictly speaking, these considerations only apply to states or processes in which the pressure drop can be neglected

3.4. Pressure drops in RCS of a PWR Pressure drops in the primary circuit of a PWR are constituted from the primary components (most notably the RPV and the SG) where a sometimes complex conduction of flow is a result of the requirements (i.e. operational and safety-related functionalities) to the component. The sum of pressure drops along the primary loop and the core flow required for removal of the design thermal core power determines the required pumping power of the RCPs and – in case of RCP failure or other accidental operating condition – the natural circulation performance, i.e. the ability for the primary side to remove the decay power from the core under natural circulation (convective flow condition) to the steam generator secondary sides. Fig. 5 depicts the evolution of the Δp along a primary loop in qualitative comparison of contributions

Fig. 3. Example for measurement of Δp in the PKL ITF and the conversion to the collapsed water level (L) in the component (hot leg and SG inlet plenum in this case). 4

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Fig. 4. Decreasing swell level (mid diagram, blue filled curve) during SB-LOCA as derived from discrete points when density measurements (converted Δp readings, top diagram) turn close to zero. With core uncover detected by rod cladding temperature measurements (Tcladding) at different elevations (bottom diagram).

Fig. 5. Schematic comparison of evolutions of pressure components along a single developed primary loop as measured in PKL (considered representative of PWR). 5

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tests under full flow condition constitutes an ideal base for the modelling of the ITF (1:1 replication of single phase pressure drops in the entire primary system and in the individual components is an important scaling target in most ITF (OECD/NEA/CSNI, 2016; Umminger et al., 2012; Nakamura et al., 2009)). If the thermal-ydraulic behavior of the ITF is then due to the applied scaling concept correctly replicated for the BIC available and considered representative of PWR within the boundary conditions of interest, the pressure-loss data measured in the ITF may be used as input data for the validation of T-H system code models trying to describe the natural circulation behavior under various operating states. Apart from pressure drops found under steady-state flow conditions, depending on the operational or accidental transients, additional pressure drops that evolve at different dynamics may also develop in case of void generation in parts of the RCS or as results of increasing present heat sinks (e.g. SG cool-down gradients) or inducing additional heat sinks (e.g. breaks or condensation at ECC injection points). These pressure drops sometimes induce their own characteristic phenomena (e.g. loop-seal clearing) and may be accompanied by strong temporary pressure drops. The capabilities of the computational models in the prediction of pressure drops in particular or PWR T-H-behavior in general may be assessed in the frame of so-called benchmark activities, where test data is closely compared to pre- or blind calculations on a pre-defined scenario on basis of BIC input data.

from pressure drops induced by losses due to friction and local discontinuities, gravity and acceleration to the local total pressure for forced circulation (A) and natural circulation (B). Under forced circulation (Fig. 5, right hand side) the head delivered by the pump is consumed by the pressure drop long the components with significant contribution from the RPV and SG. Under forced circulation gravitational heads may provide a certain contribution to the overall Δp, especially when important level changes are involved. However their impact is much lower and not as dominant as under NC conditions. In natural circulation condition, flow and pressure drops are significantly smaller and the gravitational terms are driving and dominating the flow rates. In NC-condition significant contributions to pressure drops are usually found at the RPV, the SG and the RCPs at standstill. Under natural circulation, the irreversible pressure drops composed of friction and local pressure drops determine the effectiveness of natural circulation in heat removal (i.e. smaller pressure losses allow higher flow rates which in turn yield a smaller temperature spread (heat-up span across core, cool-down span in the SGs) in the RCS. The pressure drop distribution for a PWR is usually determined in the first plant of its type during the commissioning tests under full flow condition. Special instrumentation is installed temporarily for this specific test runs to capture the pressure drops for each component for a cross check with existing simulations and future development of the PWR-type related models in thermal–hydraulic system codes. For the integral test facilities (ITF) aiming at the replication of the behavior of the real (reference) PWR under operational and/or accident transients, the data on pressure losses from the PWR commissioning

Fig. 6. PKL III – CCFL in SG U-tubes during reflux condenser mode of operation, (Umminger and D’Auria, 2017). 6

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4. Typical examples

the mixture level in the RPV was between the core exit and the lower edge of the RCL and there was no water accumulation, neither in the hot legs nor in the SG inlet chamber or in the U-tubes. Under these conditions, the measured Δp on the inlet side (counter current flow) and outlet side (co-current flow) is close to zero. The pressure drop due to water/steam flow under co-current (outlet side) and counter-current (inlet side) under ‘pure’ reflux condenser conditions is obviously marginal (compared to the pressure drops observed in the later stage of the transient). The onset of CCFL in the U-tubes is clearly indicated by a significant increase in Δp on the SG inlet side and by an increase in Δp between the SG inlet and outlet chamber at a SG load of 6%. In addition the test results show that the amount of water hold-up in the tubes does not increase continuously with specific SG load. In the described PKL test, maximum water hold-up occurred at a specific SG-load of 12%. At higher power levels, some of the tubes are blown out and water is displaced towards the RPV (e.g. tube 19 at 16%). Fig. 6 also reveals a rather heterogeneous flow behavior among the individual U-tubes. The longer tubes experience water hold up leading to a redistribution of steam flow towards the smaller tubes. The increase in steam velocity in the smaller tubes results in condensate carryover to the outlet side in these U-tubes (tube 1) creating a pressure drop (due to high steam/water velocities) from the SG-inlet to the top of the U-tubes and from the top to the SG-outlet chamber (co-current water/steam flow). This phenomenon is also indicated by the negative value for the measured pressure difference in the outlet side of the short tube. Obviously, the steam is not condensed completely in the shorter tubes, and the remaining steam flows via the SG outlet chamber into the outlet sides of other tubes (which are experiencing CCFL in their inlet sides). As a result, CCFL may also occur in the outlet sides of longer tubes if this steam flow rate is large enough. The entire pressure drops on the inlet side (carry-over of condensate) and the outlet side (co-current flow of steam and condensate) corresponds to the differences in pressure drops between the inlet and outlet side of the longer tubes which are mainly caused by the hydrostatic head of the accumulated water. The mass of water accumulated in the U-tubes cannot directly in qualitative terms be derived from the measured pressure difference, because not all of the U-tubes are equipped with Δp sensors. Furthermore, the influence of the friction pressure drop due to wall and interfacial shear stress that is included in the measured pressure difference cannot easily be determined. The onset of CCFL and the amount of water in the U-tubes can however be derived and confirmed by a mass balance for all the primary side components (with the exception of the SGs) on the basis of pressure drop measurements. In this context, the PKL tests on CCFL provide also a source of data for checking the applicability of calculation models for the determination of pressure drops due to wall friction and interfacial friction in vertical tubes or for comparison of the experimental results on CCFL with empirical correlations (e.g. Wallis), (Trewin et al., 1994).

There is a large number of examples of important thermal hydraulic phenomena expected to occur under accident conditions in PWR that can be demonstrated and reproduced in ITF experiments with the help of differential pressure measurements. Such results significantly contribute to the understanding of complex thermal hydraulic processes. The observed phenomena can very often also be extrapolated to PWR conditions at least in qualitative terms. The quantitative extrapolation generally requires the support of computer codes whereas the experiments provide the experimental data base for the validation of the used models in the codes. A selection of some typical thermal hydraulic phenomena governed by the prevailing pressure drops during transients simulating accident sequences in NPP and analyzed in ITF experiments on the basis of pressure drop measurements, usually in combination with other measurements, are: – – – –

Loop seal clearing. Reverse flow in SG U-tubes under NC conditions. Counter current flow limitation (CCFL). Main phenomena in LB-LOCA: quench front propagation, stagnation point, steam binding, level suppression, core/downcomer oscillations.

The CCFL behavior and the use of differential pressure measurement inside and around the SG for the highly complex phenomena will be explained in more detail (Umminger and D’Auria, 2017), followed by some remarks on bypass flows in the RPV. 4.1. Pressure drops and flooding at CCFL location Counter-current flow (CCF) with upward oriented steam flow and gravity-driven water down flow in vertical flow sections are expected to occur in many accident sequences. CCF may also appear in inclined or even horizontal flow sections if there is a connection with inclined or vertical sections experiencing CCF. At high steam flow rates or steam velocities the water flow in the opposite direction may be impeded or even terminated/diverted in the same direction as the steam flow (cocurrent flow) due to momentum exchange between the phases. This phenomenon, called counter-current-flow limitation (CCFL), may lead to accumulation of water, i.e. flooding or water hold-up at the CCFL location. CCFL can be considered as one of the key thermal hydraulic phenomena in reactor safety analyses because it can prevent or at least deteriorate the backflow of water to the reactor core. The occurrence of CCFL is generally connected with an increase in pressure drop at the CCFL location. It therefore does not only lead to a displacement of water from the core to the CCFL location but it may also cause a certain suppression of the core level. CCFL under reflux condenser conditions has been extensively investigated for instance in the PKL III test facility (Trewin et al., 1994). Within parametric studies, the influences of the specific SG load (heat to be removed by the SG) and the primary pressure on the occurrence of CCFL and on the amount of water accumulated in the U-tubes due to CCFL was systematically analyzed under quasi-steady state and transient conditions (secondary side cool down). As an example Fig. 6 shows typical results from an experiment on CCFL under reflux condenser conditions performed with a stepwise increase of the SG load from 4% to 20% of the nominal SG power at a constant primary pressure of 40 bar (realized by controlling the secondary side pressure). The measured Δp, between the in- or outlet chamber and the top of the Utubes for three tubes with different heights (assumed to be representative for the existing 28 tubes in the PKL test facility) as well as the measured Δp between the SG inlet and outlet chamber are show in Fig. 6. The test started at t = 0 in “pure” reflux condenser condition, i.e.

4.2. Bypass flows in the RPV If correctly scaled at the points of interest, an ITF or SETF may produce data on the impact of certain pressure losses on the overall T-H behavior of a PWR where T-H codes face some difficulties, either as a result of insufficient modelling of phenomenon-related geometries in the nodalisations of primary components or inadequate computational models not able to correctly capture relevant phenomena. For instance, the impact of possible flow paths and associated pressure losses along the path from the upper plenum via the controlrod guide-assemblies (CRGA) to the RPV dome and further into the upper downcomer (upper head bypass flow) has a significant impact on the development of the temperature field in the RPV dome which determines the point in time with the formation of the upper head void in the course of events that include cool-down in natural circulation (e.g. cool-down after SGTR concurrent with LOOP). From several tests on the SB-LOCA scenario it became obvious that 7

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5.1. The first category

the magnitude of the upper head bypass pressure losses is a key parameter to influence the point in time for the occurrence of loop-seal clearing, which may be – in case of application of design-extension conditions (failure of HPSI) – the first mechanism to temporarily rewet the core during the cool down transient.

The first category includes errors caused by the sensor itself and the connected data processing system (e.g. electronic equipment with amplifier, analog–digital converter, multiplexer, etc.). Important specification features for the sensors are hereby: range of measurement, repeatability, hysteresis, non-linearity, frequency response, sensitivity and independence from environmental conditions (temperature, humidity, vibration, etc.). The predominant part of the currently used pressure transducers are characterized by high reliability and rather high accuracy with maximum errors in the order of 1% (or less) of the full scale within the adjusted measurement range. This of course only refers to the pressure transducer itself (including data processing system). A general issue of pressure drop measurement is the fact that the range of variation expected for pressure drop measurement may be several orders of magnitude lower than the absolute pressure acting in the system at the location of pressure tappings (i.e., the port connecting the pressure sensing lines and the concerned “high” pressure system). In addition, high sensitivity of the transducer and high measurement range can often not be fulfilled at the same time. Compromises have to be accepted or on occasion several instruments (transducers) may be connected in parallel to cover the entire range to be investigated within the required accuracy. For example, in transient tests in ITFs with RCPs in operation at the beginning, and subsequent coastdown of the RCPs with transition to natural circulation, the use of several instruments in parallel with sufficient different ranges at least for some relevant test sections is required to cover the extremely large pressure drop range (partly factor more than 1:1000) and to measure the pressure drop in these sections over all phases of the transient with the required accuracy.

4.3. Practical applications for differential pressure measurement in NPP Almost all of the PWR or BWR saw the application of Δp measurement for operational- and safety-relevant level detection, most notably for the steam generator (SG), pressurizer (PRZ) or reactor pressure vessel (RPV). This measurement application – if used as an input parameter for level control circuits, for instance, must be combined with an input parameter which reflects the current power load situation. As the Δp measurement cannot selectively identify the relevant geodetic head imposed by the water level from the overall sum of heads (hydraulic head, friction head and additional friction head stemming from two-phase flow), a (computational) processing of the recorded Δp is required to get a reliable water level signal, in particular during sharp load-changing transients (e.g. SG water level control under sudden RCP trip, or PRZ level measurement during primary-side bleed accident management procedure). This is even more important when certain component level measurements (steam generator, pressurizer) are also used as criterion for (automatic) instigation of emergency actions (reactor scram, start of safety systems such as emergency core coolant injection systems). Another example of suitable application of Δp measurement is the reactor coolant pump (RCP)-shaft rupture scenario where a Δp measurement is useful as a second (and essential) criterion, as the speed control of the RCP may not be able to capture the rupture of the shaft. In some PWR an additional level measurement on basis of Δp is introduced to the hot legs for special operating states (reduced primaryside water inventory) in preparation of unscrewing and lifting of the RPV vessel head e.g. in preparation for refueling during the regular outage. As for all Δp sensors the knowledge of the significance is a key to the interpretation of the signal readings. For instance while the hotleg level measurement allows for affine adjustment of the so-called 3/4loop operation under stable residual-heat-removal system (RHRS) operation, a failure of the RHRS may cause voiding in the RPV with the resultant formation of a two-phase flow in the hot legs. Under these circumstances, the significance of this level measurement is limited.

5.2. The second category The second category concerns the configuration and the state in the components between the pressure transducer and the flow path, i.e. the pressure sensing lines and the pressure tappings. The pressure difference which the transducer is exposed to and measured by the transducer is composed of the pressure difference between the points of interest at the tapping points and the pressure difference created by the hydrostatic heads in the pressure sensing lines. The state (or its knowledge) in the sensing lines (typically 5–10 mm internal diameter, sometimes more than 10 m long) is therefore of great importance, especially when the tapping points are on different vertical levels as it is the case for level measurements but partly also for friction or local pressure drop measurements. Usually, the pressure drop of interest at the tapping points is gained by the data processing by compensating the influence of the sensing lines under the assumption of completely water filled lines at known constant and homogeneous temperature (e.g. ambient temperature). Deviations from these assumptions (e.g. inhomogeneous temperature distribution, evaporation at the contact with high temperature fluid or flashing inside the lines due to depressurization during the transient, presence of gas bubbles from the beginning or due to degassing during the transient or small leakages) may lead to unqualified pressure drop results with very few possibilities to recover the right signal. The goal of the measurement is mostly the difference in static pressures, problems may also arise due to possible influence of dynamic pressure effect, e.g., caused by imperfectly aligned pressure tappings or by the creation of vortexes at the opening of the sensing lines. Long pressure sensing lines for transmitting pressure drop signal may generate errors in measurement during very fast transients.

5. Issues/challenges connected with pressure drop measurements ITF are in general intensively equipped with differential pressure measurements in all relevant components and subsystems in order to analyze inventory distribution, mass flow rates and partly void or density distribution as well as local and distributed pressure drops in individual sections of the facilities under single and two phase flow conditions. The use of the measurement results for the interpretation of thermal hydraulic processes observed in the experiments and especially the use for the assessment of thermal hydraulic computer codes requires a proper uncertainty analysis of the measurements. Even though the measurement of the pressure drop is based on a proven and reliable technology, a number of issues/challenges connected with pressure drop measurements such as measurement errors, reproducibility of test conditions and measured data and lack or insufficient knowledge of boundary conditions have to be considered. Three main sources for errors and uncertainties in the gained experimental results can be defined: 1. Apparatus error (error of transducer and data processing). 2. Divergent (from assumption) state and conditions in the components between transducer and flow path. 3. Lacking knowledge about boundary conditions in the flow path of interest.

5.3. The third category The third category addresses the most challenging part of issues 8

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lower plenum or UP in case of PWR RPV. In those situations it is impractical to install transducers along the actual flow path of the fluid. This gives origin to a hidden error when attempting to simulate fluid flow behavior in those conditions.

connected with pressure drop measurements and deals with the interpretation of the measured value and the knowledge of the boundary conditions in the flow path of interest. As described in chapter 3, the measured differential pressure includes in principle several components such as gravity, acceleration and pressure losses due to friction and local pressure losses. The individual components can be separated and derived from the pressure drop measurements under certain conditions, when the effect of the other components can be neglected or when the contribution of these components can be determined with the help of other information/measurements. For this purpose the knowledge of the boundary conditions in the flow path of interest is essential. While the error sources discussed for the first and second category are relevant for all pressure drop components, category 3 has to be individually discussed for the different pressure drop components:

Besides the identification of important thermal hydraulic phenomena on the basis of pressure drop measurements in transient tests of ITF, the experimental determination of frictional and local pressure drops is also used for the assessment of the calculation models used in the thermal hydraulic codes. A large number of methods for the prediction of pressure drops in single and two-phase flow can be found in the literature and is not subject to this paper. A comprehensive, still valid and applicable document, mostly focusing on local pressure drops, has been provided by Idel’chick (1979). More recent efforts summarizing current technological understanding in the area of water cooled nuclear reactors have been performed within the framework of the International Atomic Energy Agency (IAEA) activities (Aksan et al., 2001). It is common understanding that experimental data are needed to treat the limitations and deficiencies still connected with the current practice for calculating the pressure drop coefficients in the codes (Umminger and D’Auria, 2017). On the other hand side, the above discussed issues connected with experimental investigation of pressure drops require an interpretation of the gained results and the proper documentation of the boundary conditions by the experimentalists.

5.3.1. Level The general procedure for the determination of fill levels (including collapsed level, swell level, mean density, mean void fraction) on the basis of differential measurements, the limits and restrictions as well as the possibilities for correcting/eliminating the dynamic effects in the measured values have been discussed in chapter 3 and shall not be repeated here. The confirmation of the validity of the measured levels in ITF by mass balances is a useful tool to check the plausibility of the gained results.

6. Envisaged developments

5.3.2. Distributed pressure drop (friction and acceleration) In addition to the concerns discussed under category 1 and 2, the following main issues which may influence the quality of the determined distributed pressure drop components (acceleration pressure drop issues can be inferred as a consequence) and local pressure drops (at geometric discontinuities) have to be considered (Umminger and D’Auria, 2017):

As already mentioned in chapter 5, further developments in the area of pressure drop modelling especially under two-phase flow conditions are needed. The target for improvement of knowledge can directly be associated with the need for additional experimental campaigns in this field. Examples for desirable modelling improvements in T-H system codes and possible trends for CFD codes are discussed in (Umminger and D’Auria, 2017). Many of the discussed issues may be addressed by separate effect tests or basic tests with a large variation of the influencing parameters. Issues related to pressure drop predictions in T-H system codes and partly also in CFD codes may be also solved with support of system tests or ‘quasi separate effect tests’ or within parametric studies in ITF. Examples are:

– Elevation pressure drops. In vertical flow channels, it is necessary to separate the contribution of elevation pressure drops which is not straightforward when two-phase flows are involved (see below). – Compressible fluids including two phase. It is necessary to separate the contribution of (spatial) acceleration pressure drop which is not straightforward when two-phase flows are involved (see below). – Two-phase conditions. An interpretation model is needed to estimate the average void fraction between the two tappings. The model is necessarily imperfect and its quality depends upon the flow regime experienced by the process fluid. Another option is the use of an additional two-phase measurement technique for the detection of the void fraction

6.1. Flow reversal Pressure drops at discontinuities are well recognized as important parameters affecting the results of T-H system codes. For PWR systems, suitable information can be derived from ITF and partly from NPP measurements. The available information is related to forward (direct) flow connected with normal operating conditions. However flow reversal may occur under certain accident conditions. The predictability of flow reversal processes is linked with the knowledge of the corresponding loss coefficients at abrupt discontinuities which may be significantly different from those for forward flow. The provision of qualified experimental data from ITF including reverse pressure drops suitable for code assessment would be very useful.

5.3.3. Local pressure drops (at geometric discontinuities) The issues described for distributed friction pressure drops also apply to the measurement of local pressure drops at geometric discontinuities, e.g., signals used to characterize the K-factors. In addition, the following issues are specific for local pressure drop measurements (Umminger and D’Auria, 2017). – Any local pressure drop implies the occurrence of a reversible and an irreversible portion. Only the irreversible portion is of technological interest and contributes to the K-factor. Typically, the pressure drop composed of both parts occurs immediately downstream the geometric discontinuity and the reversible portion brings to a recovery in pressure at some distance downstream. The value of the “distance downstream” causes the issue: the distance is not known and depends upon the geometric configuration and the flow conditions. The measurement of the “distance downstream” is troublesome and is possible only in selected reference situations. Hence unavoidable error must be expected from measurements. – Special geometry. Local pressure drops also occur inside plena, e.g.,

6.2. Sump recirculation and partial core blockage In the long term cooling phase of a LOCA in a PWR, reactor coolant collected in the containment sump is recirculated by the residual heat removal pumps back to the reactor system. Certain amounts of the also accumulated debris in the containment may pass the sump strainers and reach the RCS leading to partial core blockage. Ballooning of fuel rods would result in a similar effect. Experiments simulating a well defined and potentially changeable partial core blockage could address this subject. Pressure and pressure 9

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7. Conclusions

drop measurement around the blockage in longitudinal and cross direction together with the measured impact on core cooling would provide a valuable data basis for computer codes for this specific kind of accidents.

Together with the measurement of temperature, the measurement of pressure or pressure drop is one of the first and most used measurements in science. Pressure drop measurement plays a central role in the detection and recording of flow processes in thermal hydraulic systems. In nuclear engineering this not only applies to the detection and recording of the processes in the nuclear plants themselves but also to the test facilities used for the examination of important phenomena expected to occur in NPP under accident situations. Integral test facilities, primarily designed to investigate the T-H system behavior of NPP under accident conditions, are in general intensively equipped with differential pressure measurements in all relevant components and subsystems in order to analyze inventory distribution, mass flow rates and partly void or density distribution as well as local and distributed pressure drops in individual sections of the facilities under single and two phase flow conditions. Experimental results derived from pressure drop measurements in integral test facilities (ITF) have significantly contributed to the better understanding of the sometimes complex thermal hydraulic phenomena. Insofar, these pressure drop measurements also represent a central part in the validation and improvement processes for the models applied in thermal hydraulic computer codes used for the prediction of the accidental behavior of NPP. Even though the measurement of the pressure drop is based on a proven and reliable technology, a number of issues/challenges connected with pressure drop measurements such as measurement errors, reproducibility of test conditions and measured data and lack or insufficient knowledge of boundary conditions leading to a potential misinterpretation of the measurement results have to be considered. To guaranty the provision of qualified test data, a careful examination of the result including plausibility checks and the documentation of the measurement errors is required. Pressure drop measurements are sometimes very complex and under certain boundary conditions particular measuring arrangements are needed to assure significant results. The description of the boundary conditions and detailed information about the realized configuration of the entire measuring system should be also part of the measurement documentation. The need for further developments in the area of pressure drop modelling in the computer codes is directly connected with the need for further experimental campaigns in this field. Possible areas of new investigations have been identified. Hereby, despite the promising developments of ‘advanced’ two-phase flow measurement techniques, the ‘conventional’ pressure drop measurement considering some improvements, such as higher spatial resolution or cross measurements in larger volumes, will remain of great importance for the investigation of new topics in the future. The combination of both is obviously the best solution.

6.3. Quench front propagation Quench front propagation during the core reflooding phase after a LB- or IB- LOCA is essential for the re-establishment of the core cooling. It is a complex process, coupled with other thermal hydraulic phenomena such as water entrainment and steam binding. Recent analyses on the quenching behavior in the core indicated an overestimation in the description of water entrainment with the steam flow. This is of relevance for the flooding and cooling of the core. New parametric studies with systematic variation of the influencing parameters (core power, primary pressure, initial core liquid level and ECC injection rate) in an integral test facility could provide the basis for the evaluation of calculation models employed for the simulation of the quench-front propagation in the core and the water entrainment to the SGs. The quench front propagation is generally detectable by pressure drop measurements. Additional instrumentation would be recommended to detect water entrainment 6.4. Interphase momentum exchange Another aspect, which should be covered by new experiments, is the provision of data for the validation of the calculation models on the interphase momentum exchange, especially for very low core powers and very low RCS pressures (around 1 bar). The models currently used in the T-H system codes are usually based on experiments performed at higher pressure levels. As part of the recommended investigations, phases of constant boundary conditions (core power, primary pressure) may be established to supply precise data on swell level heights in the RPV dependent (determined by pressure drop measurements) on the set of boundary conditions. 6.5. Stability of single-phase and two-phase flow Flow oscillations, i.e. stability issues, may occur in one-and twophase NC conditions in PWR. It is known from existing experiments in different integral test facilities that selected groups of U-tubes in SGs may experience flow oscillations including flow reversal which results in a reduction of the total core mass flow rate. Pressure drops and temperatures in the SG U-tubes are the key indicators of this phenomenon. The geometrical configuration (different length and positions) and the heterogeneous flow behavior in the SG inlet chamber have been identified as possible sources for the heterogeneous behavior of the Utubes, but this has not been finally confirmed up to know. Additional experiments with a broader range of parameter variations and improved measurement techniques (e.g. pressure drop measurements in cross direction in the SG inlet chamber to detect cross flow) would be certainly useful for the code assessment. For all these envisaged investigation, certain improvements regarding the pressure drop measurements may be considered. More local measurements with pressure tapping tubes leading into the interior of the components to realize a finer local resolution and to cover also radial pressure difference may be of interest. In some cases, the combination with other measurement systems, e.g. local advanced twophase flow measurement systems, like local void fraction measurements would improve the value of the experimental results. It has to be mentioned that the subjects on quench front propagation and interphase momentum exchange are planned to be addressed within the current OECD PKL4 project, the final boundary condition of these parametric studies will be agreed with the project partners. The other topics described in this chapter may be partly included in future PKL programs; discussions on this have been started.

Acknowledgements Some of the experimental results used in this paper have been derived from experiments in the PKL test facility. The support of the German Government, the German utilities and our national and international OECD-partners participating in the PKL Projects is explicitly acknowledged. References Aksan, S.N., D’Auria, F., Groeneveld, D., Kirillov, P., Saha, D., Badulescu, A., Cleveland, J. (Lead Authors), 2001. Thermohydraulic Relationships for Advanced Water Cooled Reactors. IAEA Tecdoc 1203, Vienna (A) (ISSN 1011-4289). Hewitt, G.F., 1978. Measurement of Two-Phase Flow Parameters. Academic Press, New York. Hewitt, G.F., 1982. Measurement Techniques, Chapt. 10 in Handbook of Multiphase Systems. Hemisphere, New York. Idel’chick, I.E., 1979. Handbook of Hydraulic Resistances. Hemisphere Publishing

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Company, New York. Nakamura, H., Watanabe, T., Takeda, T., Maruyama, Y., Suzuki, M., 2009. Overview of recent efforts through ROSA/LSTF experiments. Nucl. Eng. Technol. 41. OECD/NEA/CSNI, 2016. The Scaling State of the Art Report–The SOAR, NEA/CSNI/R (2016), Paris, France. Trewin, R., Weber, P., Mandl, R., Umminger, K., 1994. Countercurrent flow limitation and liquid entrainment in the steam generator U-tubes of a PWR. International Conference on New Trends in Nuclear Thermohydraulics, Pisa, Italy. Umminger, K., D’Auria, F., 2017. Pressure drops in nuclear thermal-hydraulics: principles, experiments, and modeling. In: D’Auria, F. (Ed.), Thermal-Hydraulics in WaterCooled Nuclear Reactors. Elsevier, Wood Head Publishing Duxford, England. Umminger, K., Dennhardt, L., Schollenberger, S., Schoen, B., 2012. Integral test facility PKL: experimental PWR accident investigation. Sci. Technol. Nucl. Install. 2012

Klaus Umminger After graduating from the Technical University of Munich Klaus Umminger joined Framatome (formerly AREVA and Siemens/KWU) in 1982, where he has been engaged in experimental projects in the field of reactor safety research and in the design of fluid instrumentation. Since 1995 he has been in charge of the large experimental program PKL. The PKL facility (operated at Framatome Germany) is used to investigate the T/H system behavior of PWRs during postulated accidents, since 2001 the PKL project has been continued as an international OECD project. His main areas of work include design, performance, interpretation of experiments for application in code validation, and experimental verification of AM measures for PWRs.

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