The analysis documented in the previous section has been extended. Loop seal piping and pumps available in the PWR-1 nodalisation have been modified and deleted, respectively. The loop seal, ‘U-shaped’ cold leg piping connecting SG outlet with the MCP, has been substituted by a ‘L-shaped’ piping connecting the SG outlet and the horizontal part of the cold leg. New calculations have been performed with results documented in Table VII and in Figs 9 and 10.

Table VII has the same structure as Table VI. However, the columns reporting primary system pressure, upper plenum fluid temperature and upper plenum void fraction in Table VI have been substituted by columns giving the steam line flowrate, the maximum rod surface temperature and the maximum void fraction in the core, respectively. Thermalhydraulic conditions in the core (void fraction and maximum cladding temperature) and at the outlet of the SG (dome temperature and steam line flowrate) can be seen in Figs 9 and 10, respectively.

Starting from the situation depicted by the data in Table VI, a number of strategies can be pursued to design a natural circulation PWR, taking the goal of full core power and allowing for two-phase flow in the core. For instance, system geometry (number of SG U-Tubes, mutual position between RPV and SG, core hydraulic diameter) can be substantially modified, steam generator pressure can be lowered, feedwater temperature can be lowered, primary system pressure can be increased. Strategies requiring (almost) no system geometry changes and minimizing the necessary variation in the values of operational parameters have been followed:

a) The occurrence of dryout, (stable condition with limited temperature excursion) is allowed.

b) The feedwater temperature and the operating pressure of the SG are decreased, thus allowing a loss of thermal efficiency of the plant.

TABLE VII. REMOVABLE POWER BY NATURAL CIRCULATION IN PWR-1 WITHOUT PUMPS AND LOOP SEALS.

No.	ID.	P	G	SG PRE/Tsat	RM	GSL	PCT/Tsat	MCV	G/P	RM/V
		MW/%	(Kg/s)/%	MPa/K	KgE5/%	(Kg/s)/%	K	—	Kg/sMW	Kg/m3
1	KK01	1876/100	9037/100	6.1/550	1.08/100	1030/100	615/618	0.	4.82	647
2	KN08	1126/60	1428/15.8	6.0/549*	0.89/82.9	620/60	627/620	0.69	1.27	536
3Л	NC00	1126/60	1880/20.8	6.0/549*	0.92/89.3	620/60	627/620	0.44	1.67	586
4Л	ND03	1876/100	1955/21.6	6.0/549*	0.79/76.7	1030/100	791/620	0.83	1.04	500
5Л	NI03	1876/100	2083/23	3.5/516**	0.93/90.3	1000/97	630/620	0.60	1.11	591

Nomenclature: See Table VI. In addition, GSL Steam line flowrate § Dryout occurrence MCV Maximum Void Fraction in the core * Feedwater temperature same as in nominal condition PCT Peak Cladding Temperature ** Feedwater temperature set at 393 K Л Modified PWR (pump and loop seals removed).

FIG. 9. Proposed NC system: Maximum values of core void fraction and of rod surface temperature calculated at stable operating conditions in code runs ND03 and NI03.

FIG. 10. Proposed NC system: Fluid temperature in SG dome and steam line flowrate calculated at stable operating conditions in code runs ND03 and NI03.

The database supplied by the two calculations identified as ND03 and NI03 in Table VII, constitutes the result of the activity. System configurations considered as input for calculations ND03 and NI03 comply with the strategies at items a) and b) above. In the Table VII, calculations KK01 and KN08 are taken from Table VI in order to set reference values for the considered parameters.

The comparison between results of calculations KN08 and NC00 shows the advantage derived from adopting the ‘L-shaped’ cold leg at the outlet of the SG instead of the ‘U — shaped’ cold leg. Core flowrate in NC conditions at 60% core nominal power increases from about 1400 to about 1800 Kg/s showing that pumps at rest conditions introduce a noticeable pressure drop in the loop. Keeping the ‘L-shaped’ configuration and all operational parameters of run NC00 (or KN08), it has been shown that 85% core power can be removed by NC without the occurrence of dryout. This value can be compared with the 70% limit applicable to the ‘U-shaped’ cold leg configuration, (run KN10 in Table VI).

The results from calculations ND03 and NI03 (Table VII and Figs 9 and 10) bring to the following remarks:

• The removal capability of 100% core power by NC has been demonstrated starting from a steady state condition for the system fixed at 3% core power with primary loop in subcooled conditions and pressurizer and SG at nominal pressure.

• Film boiling cooling establishes on the surface of fuel rods in a region of the core (calculation ND03). This conditions is characterized by a stable temperature jump (DTsat, difference between local clad temperature and fluid saturation temperature) of about 180 K. Maximum values of void fraction that are calculated for the core are typical for the nominal operation of Boiling Water Reactors.

• Excursion from the nucleate boiling is achieved at two of the ten axial levels of the core region (upper half) that are characterized by a power peak factor close to 1.5. A proper design of core power distribution, i. e. consistent with the moderation features of the new system, could reduce the spatial extension of the film boiling region.

• The decrease of steam temperature at the SG outlet of about 20 K, consequence of the pressure decrease, prevents dryout occurrence on the fuel rod clad surface, calculation NI03. In this case, no attempt has been made to optimize the combination between feedwater temperature and SG pressure keeping the goal of maintaining the core in nucleate boiling.

• The mass inventory for the primary loop stabilizes at about 80% and 93% of the initial value in the two operational configurations calculated by code runs ND03 and NI03, respectively. This includes the fluid in the pressuriser. Stable liquid level in the pressuriser, i. e. buoyant fluid in a dynamic condition, establishes as the result of the procedure considered for power rise. The level value is such to allow the operation of heaters, though they are not called into operation in the present framework.

• The G/P and the RM/V values that characterize the working points in the NCFM for code runs ND03 and NI03 lie in the area of the square symbols of Fig. 7, though the present values are not reported in that figure.

2. CONCLUSIONS

The database gathered in natural circulation tests performed in PWR simulators and the results of system code calculation related to the same experiments constituted the starting point for the present investigation. The experimental database has been utilized to set up a reference natural circulation flow map. This allowed the judgement of the performance, in natural circulation conditions, of current PWR systems not necessarily of the same type as those used for deriving the map.

It was found that the PWR equipped with Once-Through steam generators have a poor natural circulation performance when primary mass inventory is decreased. Otherwise, reasonable natural circulation performances of Russian designed reactors WWER were characterized. This is mainly true for the WWER-1000. Passive systems in the AP-600 innovative reactor are effective in keeping the primary system under single-phase natural circulation notwithstanding removal of coolant mass.

Power removal by natural circulation strongly depends upon primary and secondary system boundary condition. Two thermohydraulic thresholds have been considered and the neutronic — thermalhydraulic coupling has been neglected. The void formation in the core that prevents a PWR from being ‘pressurised’ and the technological limit constituted by the critical heat flux are the reference thresholds. When primary and secondary system boundary conditions are kept close to their nominal values, it was found that:

• single phase natural circulation is effective for removing up to around 20% of core nominal power,

• two phase natural circulation, with the primary system in a boiling condition, is effective for removing up to about 70% of core nominal power avoiding the occurrence of the dryout.

A deeper investigation showed that 100% core power can be removed in two-phase natural circulation, provided the steam generator pressure and the feedwater temperature are lowered to values of the order of 3 MPa and 100°C, respectively. An increase of primary pressure also brings to the increase of the power removal capability by natural circulation.

The substitution of the current ‘U-shaped’ loop seal with a ‘L-shaped’ pipe and the elimination of the pump component improved the natural circulation performance of the considered system. Two reference sets of natural circulation conditions suitable for additional investigation have been established at 100% reactor core power:

a) film boiling allowed within a limited extension core region;

b) reduced pressure in steam generator.

In the former case, roughly one-tenth of the fuel rod clad surface was calculated with a temperature 180 K larger than the saturation value. Steam conditions at the outlet of the steam generator are the same as in current design PWR. In the latter case, all fuel rod clads are predicted in the nucleate boiling heat transfer regime, though core average void fraction is close to 40%. Better core cooling conditions are obtained at the expenses of lower temperature of the steam and then of lower thermal efficiency for the plant.

Continuation of the study, if interest is shown from the technical community derived from the economic benefit of the proposed solutions, requires the investigation of the following aspects:

• role of the pressurizer;

• design of suitable neutron kinetic parameters for the core region and analysis of the thermalhydraulic-neutronic feedback;

• design of suitable start-up procedure;

• accident analysis.