CHAPTER 1. INTRODUCTION
1.6. Structure of this report
This chapter provides a brief discussion of the background for this CRP, the CRP objectives and lists the participating institutes. Chapter 2 provides a summary of important and relevant thermohydraulic phenomena for advanced water cooled reactors on the basis of previous work by the international community. Chapter 3 provides details of the database for critical heat flux, and recommends a prediction method which has been established through international co-operation and assessed within this CRP. Chapter 4 provides details of the database for film boiling heat transfer, and presents three methods for predicting film boiling heat transfer coefficients developed by institutes participating in this CRP. Chapter 5 compiles a range of pressure drop correlations, and reviews assessments of these relations and the resulting recommendations. Chapter 6 provides general remarks and conclusions, and comments on future research needs in thermohydraulics of advanced water cooled reactors.
Nomenclature is provided at the beginning of each chapter for which it is necessary, and references are provided at the end of each chapter. Chapter appendices present relevant information in more detail. The report contains two annexes. Annex A identifies the contents of a base of thermohydraulics data which has been contributed by institutes participating in the CRP and made openly available on the internet site which is maintained by International Nuclear Safety Center at Argonne National Laboratory. Annex B discusses a methodology to select the range of interest for parameters affecting CHF, film boiling and pressure drop in advanced water cooled reactors.
REFERENCES TO CHAPTER 1
IAEA, 1998, Energy, Electricity and Nuclear Power Estimates for the Period up to 2020, Reference Data Series No. 1, Vienna.
IAEA, 1997a, Terms for Describing New, Advanced Nuclear Power Plants, IAEA-TECDOC-936, Vienna.
IAEA, 1997b, Thermophysical Properties of Materials for Water Cooled Reactors, IAEA-TECDOC-949, Vienna.
IAEA, 1997c, Status of Advanced Light Water Cooled Reactor Designs: 1996, IAEA-TECDOC-968, Vienna.
IAEA, 1997d, Advances in Heavy Water Reactor Technology, IAEA-TECDOC-984, Vienna.
OECD-NEA-CSNI, 1987, CSNI Code Validation Matrix of Thermal-Hydraulic Codes for LWR LOCA and Transients, Report No. 132, Paris.
OECD-NEA-CSNI, 1989, Thermohydraulics of Emergency Core Cooling in Light Water Reactors, a State-of-the-Art Report, (SOAR) by a Group of Experts of the NEA Committee on the Safety of Nuclear Installations, Report No. 161, Paris.
OECD-NEA-CSNI, 1992, Wolfert, K., Glaeser, H., Aksan, N., CSNI validation matrix for PWR and BWR Codes, RL(92)12 (Proc. CSNI-Specialists Mtg on Transient Two-Phase flow, Aix-en-Provence), M. Reocreux, M.C. Rubinstein, eds.
OECD-NEA-CSNI, 1994, Aksan N., et al., Separate Effects Test Matrix for Thermal-Hydraulic Code Validation, R(93)14, Part 1 and Part 2, Volume I: Phenomena Characterization and Selection of Facilities and Tests, Volume 2: Facility and Experiments Characteristics.
OECDNEA-CSNI, 1995, Städtke H., User on the transient system code calculations, R(94)35.
OECD-NEA-CSNI, 1996a, Annunziato A., et al., CSNI Integral Test Facility Validation Matrix for the Assessment of Thermal-Hydraulic Codes for LWR LOCA and Transients, R(96)17.
OECD-NEA-CSNI, 1996b, Lessons Learned from OECD/CSNI ISP on Small Break LOCA, R(96)20, OECD/GD(97)10.
.
Chapter 2
THERMOHYDRAULIC PHENOMENA OF INTEREST TO ADVANCED WATER COOLED REACTORS
This chapter will provide a short summary on the important and relevant thermal-hydraulic phenomena for advanced water cooled reactor designs in addition to the relevant thermal-hydraulic phenomena identified for the current generation of light water reactors (LWRs). The purpose of these relevant phenomena lists is that they can provide some guidance in development of research plans for considering further code development and assessment needs, and for the planning of experimental programs.
All ALWRs incorporate significant design simplifications, increased design margins, and various technical and operational procedure improvements, including better fuel performance and higher burnup, a better man-machine interface using computers and improved information displays, greater plant standardization, improved constructability and maintainability, and better operator qualification and simulator training.
Design features proposed for the ALWRs include in some cases the use of passive, gravity-fed water supplies for emergency core cooling and natural circulation decay heat removal. This is the case, for example, for the AP600 and ESBWR. Further, natural circulation cooling is used for the ESBWR core for all conditions. Both plants also employ automatic depressurization systems (ADSs), the operation of which are essential during a range of accidents to allow adequate emergency core coolant injection from the lower pressure passive safety systems. The low flow regimes associated with these designs will involve natural circulation flow paths not typical of current LWRs. These ALWR designs emphasize enhanced safety by means of improved safety system reliability and performance. These objectives are achieved by means of safety system simplification and reliance on immutable natural forces for system operation. Simulating the performance of these safety systems is central to analytical safety evaluation of advanced LWRs with passive safety systems.
Specifically, the passive safety principles of the next generation ALWR designs include:
(1) low volumetric heat generation rates,
(2) reliance solely on natural forces, such as gravity and gas pressurization, for safety system operation,
(3) dependence on natural phenomena, such as natural convection and condensation, for safety system performance.
The engineered safety features which incorporate these passive safety principles achieve increased reliability by means of system redundancy, minimization of system components, non-reliance on external power sources, and integral long term decay heat removal and containment cooling systems. In the design of the current generation of operating reactors, redundancy and independence have been designed into the protection systems so that no single failure results in loss of the protection function. Since the new ALWR designs incorporate significant changes from the familiar current LWR designs and place higher reliance on individual systems, a thorough understanding of these designs is needed with respect to system interactions. These interactions may occur between the passive safety systems e.g. the core makeup tanks and accumulators in the AP600, and the ADS system and
isolation condensers in the ESBWR. In addition, there is a close coupling in both plant designs between the reactor coolant system and the containment during an accident.
It can also be noted that in order to fully profit from the safety benefits due to the introduction of the passive safety systems, the behaviour of plants in which engineering safety features involving active components have been replaced with completely passive devices must be carefully studied to ensure the adequacy of the new design concepts for a wide spectrum of accident conditions. In fact, choice of passivity is an advantage in reducing the probability of the wrong operator interventions, especially in the short-term period after an accident, although passive systems require more sophisticated modelling techniques to ascertain that the natural driving forces that come into play can adequately accomplish the intended safety functions. Hence, there is also the need for an in-depth study of the basic phenomena concerning the design of ALWRs which make use of passive safety features.
Thermalhydraulic phenomena relevant to the evolutionary type ALWRs can be considered the same as those valid for the current generation LWRs. A suitable review of applicable phenomena can be found in [OECD-NE-CSNI (1987, 1989, 1994 and 1996)] and [NUREG (1987)]. For completeness, the list is reported in Table 2.1. A limited specific research activity in this area appears necessary, if one excludes new domains like Accident Management and special topics like instability in boiling channels where the interest is common to the present generation reactors.
In the case of advanced cooled reactors the foreseeable relevant thermalhydraulic phenomena can be grouped into two categories [see OECD-NEA-CSNI (1996)]:
a) phenomena that are relevant also to the present generation reactors (Table 2.1) b) new kinds of phenomena and/or scenarios.
For the category a) the same considerations apply as for the evolutionary ALWRs and the phenomena of concern are therefore well documented in [OECD-NEA-CSNI (1987 and 1989)] and [NUREG (1987)]. However, it has to be noted that significance of various phenomena may be different for the passive and advanced reactors. Nevertheless, it is believed that the data base, understanding and modelling capabilities acquired for the current reactors are adequate for phenomena in category a.
Phenomena of the category b can be subdivided into three classes:
b1) phenomena related to the containment processes and interactions with the reactor coolant system
b2) low pressure phenomena
b3) phenomena related specifically to new components, systems or reactor configurations In current generation LWRs the thermalhydraulic behaviour of the containment system and of the primary system are studied separately. This is not any more possible in most of the new design concepts; suitable tools must be developed to predict the performance of the integrated system.
A speciality common to almost all the advanced design reactors is the presence of devices that depressurize the primary loop essentially to allow the exploitation of large amount of liquid at atmospheric pressure and to minimize the risk of high pressure core melt.
TABLE 2.1. RELEVANT THERMALHYDRAULIC PHENOMENA IDENTIFED FOR THE CURRENT GENERATION REACTORS*
0 BASIC PHENOMENA 1 Evaporation due to Depressurisation 2 Evaporation due to Heat Input 3 Condensation due to Pressurisation 4 Condensation due to Heat Removal 5 Interfacial Friction in Vertical Flow 6 Interfacial Friction in Horizontal Flow 7 Wall to Fluid Friction
8 Pressure Drops at Geometric Discontinuities 9 Pressure Wave Propagation
1 CRITICAL FLOW 1 Breaks
2 Valves
3 Pipes
2 PHASE SEPARATION/VERTICAL FLOW WITH AND WITHOUT
MIXTURE LEVEL 1 3 STRATIFICATION IN HORIZONTAL FLOW 1 Pipes 4 PHASE SEPARATION AT BRANCHES 1 Branches
5 ENTRAINMENT/DEENTRAINMENT 1 Core
2 Upper Plenum
3 Downcomer
4 Steam Generator Tube
5 Steam Generator Mixing Chamber (PWR) 6 Hot Leg with ECCI (PWR)
6 LIQUID-VAPOUR MIXING WITH
CONDENSATION 1
2 Core Downcomer
3 Upper Plenum
4 Lower Plenum
5 Steam Generator Mixing Chamber (PWR) 6 ECCI in Hot and Cold Leg (PWR) 7 CONDENSATION IN STRATIFIED
CONDITIONS
1 2
Pressuriser (PWR)
Steam Generator Primary Side (PWR) 3 Steam Generator Secondary Side (PWR)
4 Horizontal Pipes
8 SPRAY EFFECTS 1 Core (BWR)
2 Pressuriser (PWR)
3 Once-Through Steam Generator Secondary Side (PWR) 9 COUNTERCURRENT FLOW/
COUNTERCURRENT FLOW LIMITATION 1
2 Upper Tie Plate
Channel Inlet Orifices (BWR) 3 Hot and Cold Leg
4 Steam Generator Tube (PWR)
5 Downcomer 11 HEAT TRANSFER: NATURAL OR FORCED CONVECTION 1 Core, Steam Generator, Structures
SUBCOOLED/NUCLEATE BOILING 2 Core, Steam Generator, Structures DNB/DRYOUT 3 Core, Steam Generator, Strucutres POST CRITICAL HEAT FLUX 4 Core, Steam Generator, Strucutres
RADIATION 5 Core
CONDENSATION 6 Steam Generator, Structures 12 QUENCH FRONT PROPAGATION/REWET 1
2
Fuel Rods
Channel Walls and Water Rods (BWR) 13 LOWER PLENUM FLASHING
14 GUIDE TUBE FLASHING (BWR)
15 ONE AND TWO PHASE IMPELLER-PUMP BEHAVIOUR 16 ONE AND TWO PHASE JET-PUMP BEHAVIOUR (BWR)
17 SEPARATOR BEHAVIOUR
18 STEAM DRYER BEHAVIOUR
19 ACCUMULATOR BEHAVIOUR
20 LOOP SEAL FILLING AND CLEARANCE (PWR) 21 ECC BYPASS/DOWNCOMER PENETRATION 22 PARALLEL CHANNEL INSTABILITIES (BWR) 23 BORON MIXING AND TRANSPORT 24 NONCONDENSABLE GAS EFFECT (PWR) 25 LOWER PLENUM ENTRAINMENT
* This table is applicable to LWRs and is expected to be applicable to WWERs as well.
TABLE 2.2. RELEVANT THERMALHYDRAULIC PHENOMENA OF INTEREST IN THE ADVANCED WATER COOLED DESIGN REACTORS
b1. Phenomena occurring due to the interaction between primary system and containment 1. Behaviour of large pools of
liquid: - thermal stratification
- natural/forced convection and circulation - steam condensation (e.g. chugging, etc.)
- heat and mass transfer at the upper interface (e.g.
vaporization)
- liquid draining from small openings (steam and gas transport)
2. Tracking of
non-condensibles (essentially H2, N2, air):
- effect on mixture to wall heat transfer coeficient - mixing with liquid phase
- mixing with steam phase
- stratification in large volumes at very low velocities 3. Condensation on the
containment structures: - coupling with conduction in larger structures 4. Behaviour of containment
emergency systems (PCCS, external air cooling, etc.):
- interaction with primary cooling loops
5. Thermofluiddynamics and pressure drops in various geometrical configurations:
- 3-D large flow pths e.g. around open doors and stair wells, connection of big pipes with pools, etc.
- gasliquid phase separation at low Re and in laminar flow - local pressure drops
b2. Phenomena occurring at atmospheric pressure
6. Natural circulation: - interaction among parallel circulation loops inside and outside the vessel
- influence of non-condensables 7. Steam liquid interaction: - direct condensation
- pressure waves due to condensation 8. Gravity driven reflood: - heat transfer coefficients
- pressure rise due to vaporization - consideration of a closed loop 9. Liquid temperature
stratification: - lower plenum of vessel - downcomer of vessel - horizontal/vertical piping
b3. Phenomena originated by the presence of new components and systems or special reactor configurations
10. Behaviour of density locks: - stability of the single interface (temperature and density distribution)
- interaction between two density locks 11. Behaviour of check valves: - opening/closure dynamics
- partial/total failure 12. Critical and supercritical
flow in discharge pipes and valves:
- shock waves
- supercritical flow in long pipes - behaviour of multiple critical section 13. Behaviour of Isolation
Condenser - low pressure phenomena
14. Stratification of boron: - interaction between chemical and thermohydraulic problems - time delay for the boron to become effective in the core
In this case, the phenomena may be similar (or the same) as those reported for current generation LWRs (Table 2.1) but the range of parameters and their safety relevance can be much different.
In addition to the concerns specific to light-water-cooled reactors, the following concerns are specific to HWRs:
I. Thermalhydraulics related to short fuel bundles located in long HWR-type horizontal channels, and on-line fuelling,
II. Thermalhydraulics related to radial and axial pressure-tube creep.
Finally, the presence of new systems or components and some geometric specialities of advanced design reactors require the evaluation of additional scenarios and phenomena.
A list of identified phenomena belonging to subclasses b1, b2 and b3 is given in Table 2.2.
REFERENCES TO CHAPTER 2
OECD-NEA-CSNI, 1987, CSNI Code validation Matrix of Thermal-Hydraulic Codes for LWR LOCA and Transients, Rep. No. 132, Paris, France.
OECD-NEA CSNI, 1989, Thermohydraulicsof Emergency Core Cooling in Light Water Reactors, a State-of-the-Art Report, (SOAR) by a Group of Experts of the NEA Committee on the Safety of Nuclear Installations, Rep. No. 161, Paris.
OECD-NEA-CSNI, 1994, Aksan, N., D'Auria, F., Glaeser, H., Pochard, R., Richards, C., Sjöberg, A., Separate Effects Test Matrix for Thermal-Hydraulic Code Validation, R(93)14, Vol. I: Phenomena Characterization and Selection of Facilities and Tests; Vol. 2: Facility and Experiments Characteristics.
OECD-NEA-CSNI, 1996, Aksan, N., D'Auria, F., Relevant Thermalhydraulic Aspects of Advanced Reactor Design, CSNI Status Report OCDE/GD(97)8, Paris.
NUREG, 1987, Compendium of ECCS Research for Realistic LOCA Analysis, USNRC Rep.
1230, Washington, DC.
.
Chapter 3
A GENERAL CHF PREDICTION METHOD FOR ADVANCED WATER COOLED REACTORS
NOMENCLATURE
C Constant -
CHF Critical heat flux kW/m2
Cp Specific heat kJ/(kg °C )
De, Dhy Hydraulic equivalent diameter m Dhe Heated equivalent diameter m
D Tube inside diameter m
d Fuel element diameter m
E Entrainment rate kg/(m2 s)
g Acceleration due to gravity m/s2
G Mass flux kg/(m2 s)
H Enthalpy kJ/kg
h Heat transfer coefficient k W/(m2 °C)
K, F Correction factor -
Lsp Distance to upstream spacer plane m
L Heated length m
P Pressure kPa
p Element pitch m
q Surface heat flux kW/m2
T Temperature °C
U Velocity m/s
X Quality -
Z Axial co-ordinate m
GREEK SYMBOLS
= Void fraction -
@ Inter element gap m
B Surface heat flux kW/m2
l Latent heat of evaporation kJ/kg
H Density kg/m3
I Surface tension N/m
y Dimensionless mass flux - ,0 Local subcooling, hs – h kJ/kg ,+ Bundle quality imbalance - ,6 Local subcooling, Ts – T °C
G Angle degrees
SUBSCRIPTS
a Actual value avg Average value
b Bubble, bulk, boiling BLA Boiling length average
c Critical, convection CHF Critical heat flux DO Dryout
f Saturated liquid value
fg Difference between saturated vapour and saturated liquid value h, H Heated
hom Homogeneous g Saturated vapour I, in Inside, inlet
l Liquid m Maximum max. Maximum min Minimum nu Non-uniform AFD P/B Pool boiling o Outside, outlet rad Radiation
s Saturation value sub Subcooling
U Uniform (AFD) v Vapour
ABBREVIATIONS
AFD Axial flux distribution BLA Boiling length average CHF Critical heat flux c/s Cross section
DNB Departure from nucleate boiling DO Dryout
FB Film boiling PDO Post-dryout
RFD Radial flux distribution
3.1. INTRODUCTION
The objective of this chapter is to recommend a validated CHF prediction method suitable for the assessment of critical power at both normal operating conditions and accident conditions in Advanced Water Cooled Reactors (AWCRs). This method can be implemented into systems codes such as RELAP, CATHARE, CATHENA as well as subchannel codes such as COBRA, ASSERT and ANTEO. The requirement of this prediction method has been discussed in more detail in previous CRP RCM meetings and expert meetings.
The two main applications for CHF predictions are:
(i) to set the operating power with a comfortable margin to avoid CHF occurrence.
This margin to CHF can be expressed in terms of Minimum Critical Heat Flux Ratio (MCHFR, ratio of CHF to local heat flux for the same pressure, mass flux and quality), Minimum Critical Heat Flux Power Ratio (MCHFPR, the ratio of power at initial CHF occurrence to the operating power for the same pressure mass
flux and inlet temperature), or Minimum Critical Power Ratio (MCPR, the ratio of reactor or fuel channel power at initial dryout occurrence to normal operating power for the same system, pressure and inlet temperature); the definition of these ratios is illustrated in Fig. 3.1. A detailed discussion has been provided by Groeneveld (1996). Most CHF prediction methods address this concern; these prediction methods provide best-estimate values of the initial CHF occurrence in a reactor core or fuel bundle.
FIG. 3.1. Definition of margins to CHF as defined by MCHFR, MCHFPR and MCPR.
(ii) to evaluate the thermalhydraulic and neutronic response to CHF occurrence in a reactor core. This requires knowledge of how CHF spreads in the reactor core, which in turn requires a best-estimate prediction of the average CHF for a section of the core and/or prediction of the variation of fuel surface area in dryout as a function of power.
This chapter is subdivided as follows: Section 3.2 discusses various CHF mechanisms, followed by a description of the CHF database in Section 3.3. In Section 3.4, CHF prediction methodologies are reviewed for both tubes and bundle geometries, ranging from correlations, subchannel codes, analytical models and look up tables. In Section 3.5, the recommended prediction methods for CHF in AWCRs are described, together with correction factors to account for various CHF separate effects. The assessment of the accuracy of the recommended prediction method when applied to steady state conditions is described in Section 3.6. Finally in Section 3.7 the prediction of CHF during transients such as LOCAs, flow and power transients are discussed.
The topic of CHF has been extensively researched during the past 30 years. Excellent reviews may be found in text books by Collier (1981), Tong (1965), Tong and Weisman (1996), Hewitt (1970) and Hetsroni (1982), and review articles by Bergles (1977), Tong (1972), Groeneveld and Snoek (1986), Weisman (1992) and Katto (1994).
3.2. CHF MECHANISMS 3.2.1. General
In forced convective boiling, the boiling crisis2 occurs when the heat flux is raised to such a high level that the heated surface can no longer support continuous liquid contact. This heat flux is usually referred to as the critical heat flux (CHF). It is characterized either by a sudden rise in surface temperature caused by blanketing of the heated surface by a stable vapour layer, or by small surface temperature spikes corresponding to the appearance and disappearance of dry patches. The CHF normally limits the amount of power transferred, both in nuclear fuel bundles, and in conventional boilers. Failure of the heated surface may occur once the CHF is exceeded. This is especially true for highly subcooled CHF conditions. At high flows and positive dryout qualities, the post-dryout heat transfer is reasonably effective in keeping the heated surface temperatures at moderate levels, and operation in dryout may be sustained safely for some time.
In flow boiling the CHF mechanisms depend on the flow regimes and phase distributions, which in turn are controlled by pressure, mass flux and quality. For reactor
In flow boiling the CHF mechanisms depend on the flow regimes and phase distributions, which in turn are controlled by pressure, mass flux and quality. For reactor