Category Archives: Natural circulation data and methods for advanced water cooled nuclear power plant designs

ERHRAC and PCCAC-2D,3D computer codes

The ERHRAC computer code is a special computer code developed by Nuclear Power Institute of China (NPIC).

ERHRAC can be used to analyze and calculate the following natural circulation flow characteristics of ERHR system for AC600/1000 PWR:

— The steady state and transient thermal hydraulic behaviour of primary coolant loop, secondary side cycle of SG and air cycle;

— Start-up mode research of triggering natural circulation flow, specially for secondary side cycle of SG;

— The conditions for establishing natural circulation flow and residual heat removal ability.

Emergency residual heat removal (ERHR) system of China advanced AC600/1000 PWR NPP China consists of a tandem three-loop system including primary loop, secondary side cycle of SG and air cycle. After accident such as station black out, the decay heat of reactor core can be removed to atmosphere by natural circulation flow through the mentioned three cycles.

PCCAC-2D, 3D are special two and three dimension computer codes developed by NPIC.

The containment of AC600/1000 is a two-shell structure. Between inner steel shell and outer reinforced cylindrical concrete shell, there is a baffle to form an annular wind channel. In the top of containment, there is a large water storage tank that can meet the requirement of 72 hours water spraying on the steel shell after LOCA.

PCCAC-2D, 3D can be used to predict the pressure and temperature of mixing gas in the containment after the accident of primary pipe rupture or main steam line rupture. The heat removal characteristics from inside containment to atmosphere through the water film on the out surface of steel shell and natural circulation flow of air can be also simulated and calculated by PCCAC-2D or PCCAC-3D.

ERHRAC and PCCAC computer codes will be verified and improved through use of experiments done in NPIC facilities.


Figure 5 shows the basic design of the SWR 1000 scram systems. The control rods can be inserted into the core in about 70 s by electric drives, or rapid control rod insertion can be implemented within 3 s by means of the hydraulic scram system. In addition there is a boron system which pumps a pentaborate solution into the RPV over a period of some 30 minutes.

The SWR 1000 has four scram tanks operating at a pressure of about 140 bar. In the existing Siemens BWR design this high pressure is generated by nitrogen. The first of the two valves downstream of the tanks is normally in the closed position while the second valve is open. To actuate a scram, the first valve is opened in a very short time and the water is routed from the tanks to the piston drives of the control rods. Several seconds after scram initiation the second valve must be closed. Otherwise the entire water inventory of the tank would be discharged into the RPV, and after that the nitrogen would follow.

It is not allowable for large quantities of nitrogen to ingress the RPV of the SWR 1000, as the emergency condensers would become ineffective given a certain content of non-condensable gases. Therefore, steam serves as the high-pressure medium inside the scram tanks, generated by electrical heating of the upper part of the water inventory such that the upper part of the scram tank is filled with saturated steam and saturated water. Due to the stable stratification no convection occurs inside the tank, and the main inventory of the tank remains cold.

In the event of a scram, the first valve is opened. The cold water flows from the tank to the piston drives of the control rods and the pressure inside the tank decreases. But with decreasing pressure, new steam is generated out of the saturated water inventory such that the pressure reduction is comparably small. Normally the second valve closes after several seconds, as otherwise hot water and steam could ingress the scram system which would generate thermal stresses in the piping which is not designed for strong temperature gradients.

Boron shut down system


FIG. 5. Conceptual arrangement of the SWR 1000 scram systems.

To avoid thermal stressing in the event that the second valve fails to close, a skirt is integrated into the scram tank. The thermally stratified water behind the skirt remains at its given elevational level after the water level inside the skirt drops due to the outflow of cold water. After a certain time the driving steam comes in contact with the cold parts of the skirt and condenses (see Fig. 6). Thus, the pressure inside the scram tank is reduced by condensation of steam. When the pressures in the RPV and in the scram tank are the same, flow within the scram system ceases despite the fact that both valves are still open.

This scram tank design was experimentally tested at a VVT laboratory in Finland. Results demonstrated that this principle of pressure reduction is effective. Prior to scram initiation, strongly stratified conditions prevailed both inside and behind the skirt. Subsequent to scram, natural circulation developed both behind and inside the skirt. Again, the small pressure differentials of several Pa in the natural circulation circuits resulted in pressure reductions in the scram tanks of some 60 bar and more.



FIG. 6. Pressure reduction in the scram tank by steam condensation with falling water level.

Full pressure core make-up tank test

Passive core make-up water tank is used so as to eliminate high pressure safety injection pumps. Transient characteristic research of core make-up water tank in the case of small LOCA were performed. The break sizes are: 2, 6, 12, 18, 30 mm respectively. Thermal hydraulic behaviors of pressurizer and draining flow rate measurement from core make-up water tank to primary coolant system following small LOCA were researched and carried out in the test. A total of 80 sensors were used to measure temperatures, liquid level and flow rate.

The above figures show the pressure curves of CMT for all 5 break sizes. It is evident that there exists a pressure oscillation phase during pressure drop down, especially for f30mm break size and lasts from 50s to end of draining. In general, the blow down phase is longer, steam condensation in CMT is more. The draining behaviors of CMT are complex and change with break size. The draining mass flow rate increases rapidly at the beginning and has a low oscillation phase before stable draining phase. All break sizes have obviously a high flow rate peak. After about 500s, all flow fluxes are nearly the same.




Break size(mm)






Initial water storage







System pressure







Upper temp. of RPV







Bottom temp of RPV







Pressure at water injection







APROS analyses of the GDE-24, GDE-34 and GDE-43 experiments

The PACTEL GDE-24 experiment was calculated with APROS 4.02 code [18], [19], [20],

[21] . The models of the CMT, PBL and IL were included to the APROS model of PACTEL

created for SBLOCA calculations. The CMT was modelled with a standard pipe module of

APROS divided in 30 equal length nodes.

• APROS calculated successfully the recirculation phase.

• Due to numeric diffusion the temperature profile was not as steep as in the experiment.

• Problems occurred during the draining mode. Injection flow started to oscillate continuously never reaching the anticipated full magnitude. The vapour entering to the top node condensed directly to the subcooled water and caused a flow stagnation due to rapid pressure drop. The injection was possible only after the water had reached saturation temperature in the boundary node. Hence, the explanation for these flow oscillations was lack of continuous existence of saturated liquid layer, which would prevent direct contact of vapour and cold water.

• Oscillating characteristics caused delay to the timing of main events.

• Important parameters were maximum time step, amount of nodes and hydraulic diameter in the CMT. To reduce the condensation it was possible to manipulate the condensation heat transfer coefficient by giving higher values to the hydraulic diameter in the CMT nodes. To get more accurate calculation results, it was necessary to use dense CMT nodalization (30 nodes) and small maximum time step (0.025 s) during the first 1000 s transient period.

For the calculation of the GDE-34 experiment a smaller CMT model with 30 nodes was created and the CMT water was initiated with warm water. Some minor modifications for the PBL pipework were also made.

• The recirculation mode existed in the calculation though it was not observed in the experiment. Hence, the density difference between PBL and CMT was enough to initiate the flow in the calculation.

• Injection flow oscillated, but with smaller amplitude than in GDE-24, because of less condensation. The average injection flow rate was quite near to the measured one. Also, flow injection stagnated once due to temporary water level increase in the vertical section of cold leg 2, which is connected to the PBL.

• The increase of hydraulic diameter reduced condensation also in this case.

• The calculation was not very sensitive to any other parameters.

• Timing of the main events agreed well with the experiment.

For the calculation of the GDE-43 experiment a smaller CMT model with 30 nodes was also used.


The SWR 1000 (about 3000 MW^) from SIEMENS is equipped with emergency condensers. With the assumptions

same flow resistances as in the SWR 1000

200 tubes with the same material and geometries as used in NOKO

temperature in the Reactor Pressure Vessel about 100 C temperature of the outside pool 30°C

a total power of about 12 MW could be transferred.

This corresponds to a decay heat at about two days after scram.

A change in flow resistances, number of tubes and pool temperature would change the amount of energy transferred but not the mode of heat transfer.

For long periods it has to be considered that the water pool has to be cooled.

Another possibility would be to bring the Reactor Pressure Vessel to a pressure below 1 MPa and allow boiling of the water pool. Then energy would be removed from the water pool via the building condensers to a pool outside the containment.


A test with the NOKO facility has confirmed that the Emergency Condensers as proposed for the SWR 1000 from SIEMENS can transfer decay heat produced in the core region to an outside water pool. The capacity, however, is not high enough to cope with the total amount of decay heat immediately after scram. However, after some time — depending on the actual design — this heat transfer mode is effective.

[1] IAEA-TECDOC-936, Terms for Describing New, Advanced Nuclear Power Plants.

[2] IAEA-TECDOC-626 : Safety Related Terms For Advanced Nuclear Plants.


Consider a simple nonuniform diameter natural circulation loop as shown in Fig. 1 with a horizontal heat source at the bottom and a horizontal heat sink at the top. The heat sink is maintained by providing cooling water to the secondary side of the cooler at a specified inlet temperature of Ts. In this analysis, the secondary side temperature is assumed to remain constant. The heat flux at the heat source is maintained constant. Assuming the loop to be filled with an incompressible fluid of constant properties except density (Boussinesq approximation where density is assumed to vary as p=pr[1-P(T-Tr)]) with negligible heat losses, axial conduction and viscous heating effects, the governing differential equations can be written as



The prediction of nuclear reactor thermal-hydraulic behaviour under operational and accidental conditions can potentially be made by

(a) The analysis of the thermal-hydraulic behaviour in smaller scale facilities and extrapolation to real plants, or

(b) The application of computer codes simulating component or system behaviour to plant situations.

The development of nuclear reactor designs has shown a change from small scale experiments in the early days of nuclear reactor development to the application of qualified versions of large computer codes nowadays supported by operational data from real plants. It is generally understood that the assessment under high accidental loads, e. g. in case of loss-of-coolant accidents or even for core melt sequences can only be based on code calculations. Thus the confidence in the code predictions is highly important. It has to be recognized that the requirements on the quality of new experimental data and of code calculations have increased during recent years; this implies the use of actual geometries, materials and thermal-hydraulic boundary conditions — if possible — as well as validation, documentation and uncertainty analysis.


(a) The consideration of advanced designs is especially important to the future of the nuclear power industry. Advanced reactors in general have the following goals:

— Satisfying stringent safety requirements (e. g. by use of a combination of both active and passive safety systems),

— Improving the economics of power generation

(b) TCM participants concluded that additional analytical and experimental work is highly recommended both for the development of new, and the validation of current, computer codes. Basic research in natural circulation phenomena could help to improve codes and design capabilities. In addition, large-scale tests of natural circulation systems could provide direct substantiation of their functioning under the real plant conditions;

(c) The papers presented to this TCM show that a large scope of analytical and experimental investigations has been performed and are being planned in Member States in relation to natural circulation systems and equipment.

. Single-Phase NC (SPNC)

SPNC regime implies no void occurrence in the upper plenum of the system. Therefore, coolant at the core outlet shall be subcooled up to nearly saturated. Core flowrate derives from the balance between driving and resistant forces. Driving forces are the result of fluid density differences occurring between [descending side of U-Tubes & vessel downcomer] and [core & ascending side of U-Tubes]. Resistant forces are due to irreversible friction pressure drops along the entire loop. Resulting fluid velocities are sufficient for removing core power in (subcooled) nucleate boiling or forced convection heat transfer regimes: no film boiling condition is experienced in the core. It may be noted that the secondary side of SG is also a natural circulation system working in two-phase conditions. SPNC may occur at any primary system pressure, consistently with SG pressure. However, typical primary system pressures range between 8 and 16 MPa with secondary pressure close to the nominal operating condition. It is expected from the NPP design that SPNC, provided the availability of SG cooling, is capable to remove the nuclear heat decay from the core. Experimental database, including NPP tests, confirms this capability.

Nonuniform diameter loops (NDLs)

Most practical applications of natural circulation employ non-uniform diameter loops. Common examples are the nuclear reactor loop, solar water heater, etc. Most test facilities simulating nuclear reactor loops also use non-uniform diameter loops. Depending on the operating pressure, the non-uniform diameter loops can be categorised as High pressure loops and Low pressure loops. Most studies are conducted in the high-pressure test facilities simulating nuclear reactor loops. Typical examples are the SEMISCALE, LOBI, PKL, BETHSY, ROSA, RD-14, FISBE, etc. Some studies, however, are carried out in low-pressure facilities. Examples are the experiments carried out by Zvirin et al. (1981), Jeuck et al. (1981), Hallinan-Viskanta (1986), Vijayan (1988) and John et al. (1991). Most of the available experimental data in a usable form (i. e. full geometrical details are known) are from the low — pressure test facilities. High-pressure test data in a usable form was available only from FISBE. The data from NDLs are plotted in Fig. 10, which shows reasonable agreement with theoretical correlation in the laminar and the turbulent flow regions. Significant deviation is observed for the intermediate values of Grm/NG where the flow is neither fully laminar nor fully turbulent. Similar trend was observed in the case of UDLs. Interestingly, the parallel channel data of John et al. (1991) and parallel loop data of Jeuck et al. (1981) are also found to be in reasonable agreement with the theoretical correlation.

E. F. Hicken, H. Jaegers

Institute for Safety Research and Reactor Technology, Forschungszentrum Julich A. Schaffrath, F.-P. Weiss

Institute for Safety Research, Forschungszentrum Rossendorf Germany

Abstract. For the study of the effectiveness of passive safety systems a high pressure (up to 7 MPa) and high power (up to 4 MW) test facility — named NOKO — has been constructed and operated at the Forschungszentrum Julich. From 1996-1998 this facility was used for a project within the 4th FP of the EU "European BWR R&D Cluster for Innovative Passive Safety Systems". An overview and selected results are given for the tests with two bundles of the emergency condenser, with the building and plate condenser, with 4 different passive initiators, with a passive flooding system and with decay heat removal tests during shutdown. It has been decided to decrease substantially the safety research at the Forschungszentrum Julich; to maintain the experimental competence for two-phase flow the NOKO facility will be transferred to the Forschungszentrum Rossendorf by the end of the year 2000 up to the beginning of the year 2001. The facility will be named TOPFLOW; the main objectives of future tests will be oriented towards more generic research: investigation of steady state and transient two-phase flow phenomena especially transient two-phase flow patterns, the development of two-phase flow instrumentation, the generation of a data basis for Computati­onal Fluid Dynamic (CFD)-Code validation and testing of heat exchangers and safety systems. An overview will be given about the modifications and improvements related to the test facility and the planned tests.


It is a good demonstration of safety culture if vendors, utilities and licensing authorities equally make an effort to increase the safety level of Nuclear Power Plants — existing and future ones. Recognising that design and main licensing requirements were developed in the sixties and seventies it is appropriate now to develop new solutions as well as new licensing requirements, evaluate the feasibility of these solutions and possibly test their effectiveness.

The goals for new safety systems are evident: effective, simpler, more reliable, cheaper and licensable. Without major efforts, it can be stated that passive safety systems proposed up to now are simpler and are expected to be more reliable. They also seem to be licensable if the remaining uncertainties with respect to requirements for redundancy and diversity have been solved. An assessment of the costs is complex and cannot be discussed here.

During recent years and still ongoing are efforts to experimentally study the effectiveness of passive safety systems and compare the results with code calculations; due to the different operating conditions (e. g. small driving forces) as compared with active systems, some models in computer codes have to be improved. Therefore, it was decided to plan, construct and operate a facility at Forschungszentrum Julich to study experimentally and analytically the effectiveness of the emergency condensers planned to be installed in the SWR 1000. This facility was named NOKO.

Due to the decision by the board of directors of the Forschungszentrum Jiilich to reduce substantially the safety research for Nuclear Power Plants at Jiilich and the decision to maintain the experimental competence for two-phase flow relevant to reactor safety it has been decided to transfer the NOKO facility from the Forschungszentrum Jiilich to the Forschungszentrum Rossendorf by the end of the year 2000/the beginning of the year 2001. This facility will be named TOPFLOW.