EFFICACY OF REACTOR VESSELS DESIGNED FOR NATURAL CIRCULATION

R. Z. Aminov, V. A. Khrustalev, A. ~. Borisenkov, and A. S. Dukhovenskii

UDC 621.039.524

Accident reports speak of the necessity for further increases in safety and reliability of nuclear power plants. In the first phase, this has to do with reactor installations as potential sources of radioactive contamination of the environment. Of some interest is a variant with integral core and once-through module steam generators [i, 2]. Such an arrange- ment is the basis of a W~R-1000 proposed at the I. V. Khurchatov Institute for Atomic Energy (IAE). The module lacks an external primary coolant loop and some of the elements of the steam generating plant, including the coolant circulating pump, the steam generator vessel, as well as other auxiliary systems. This allows one to eliminate the possibility of accidents such as a large leak, loss of coolant, and switching off of all the coolant pumps. This, plus the general increase in structural reliability and reduction in expenses, allows one to get an ecologically safer source of energy.

The individual capacity of such installations is limited by several factors; for in- stance, the size of the reactor vessel, fixed transport and technical conditions, and the possibility of reliable cooling of the core in all transition phases of the reactor. In these conditions, the thermal power of the reactor with natural circulation does not exceed 35-50% of the VVER-1000 thermal power.

The necessary coolant circulation pressure head with a moderate difference in height of core and steam generator is provided by a reduction of the parameters of the working fluid in the secondary system, as well as by small (up to 3%) steam content in the most energy inten- sive fuel element, which is permissible because of a reduction in energy stress in the core. Calculations conducted at the I. V. Kurchatov IAE revealed that for the output of an accept- able individual power unit, one can assume that the pressure in the secondary system equals 1.5 MPa, and the coolant temperature at the outlet of the steam generator is 220°C with a pressure of 8 MPa. The turbine stage for the indicated parameters is already designed (C. M. Kirov POAT KhTZproject for turbine K-I000-60/1500 with a double moisture separator).

The effectiveness of the investigated installation (in a future WER-B) as compared to a VVF/~-I000 is determined by the ratio of the derived expense to the unit installed power capacity. The principal factors are the structural reliability of the alternative configura- tion, the size and reliability of the power system, the conditions of energy consumption in it, specifying expenditures into a reserve for a guaranteed reliability of the power supply, the relationship of capital and fuel expenses, as well as the probability of failure of the core with great material and ecological damage.

From the point of view of structural reliability, a W~R-B will have a significant ad- vantage compared to traditional reactor installations, associated first of all with a reduc- tion in the number of operating plants and with the exclusion of possible structural failure of the plant. Low electrical power (350-500 MW) also lowers the expenses in a reserve elec- tric power system. The contributions to a reserve system were calculated for a given energy system reliability F = 0.999 and compared to i000 MW installed power. The calculation was carried out by the following method. We found the probable states and the appropriate power of the installation, measured at the input. In an isolated subsystem, every possible combina- tion of conditions was worked out. The electric power system input was defined by a set of states using equivalent characteristics [3]; qe = equivalent accident rate, and n e = equiva- lent number of plants. In the following step, via thorough investigation of subsystems and electric power systems, the complete group of conditions characterizing the new system was determined. Reserve is calculated by the equation

L

1--F=~ P e ( Y - N . , i , , ) l ( l - - N t - - ~ ~ / t | - - N • ~ ( l - N m | " ) / ( v - ~ " i ) r e s ~ • mill/

Translated from Atomnaya ~nergiya, Vol. 69, No. 4, pp. 207-211, October, 1990. Original article submitted May 16, 1989.

812 0038-531X/90/6904-0812512.50 © 1991 Plenum Publishing Corporation

\

0 / I o

e

U3-1 I I l / ' 2000 ¥OOO N, MW I

Fig. 1 Fig. 2

Fig. i. Savings in derived expenses for a given average ( ) and high ( .... ) reliability energy consumption with alternative input of the units with natural circulation and VVER-1000 as a function of input scale in the electrical power system (F = 0.999) for a single capacity W~R-B 350 (i, 2) and 500 MW (i', 2'). Characteristics and opeEating regime of the electric power system: i, i' - qe = 0.07, ~ = 0.9, Nmi n = 0.65; 2, 2' - 0.06, 0.85, 0.65 respectively, N e = 264 MW, n e = i00.

Fig. 2. Coolant diagram of the nuclear assembly for improved safety: 1 - reactor with natural coolant circulation; 2 - turbine.

where F = reliability index; L = number of electrical supply states; Pe and N£ = probability and relative power of the £-th state; 7 and Nmi n = density and relative minimum load of the electrical supply schedule; are s = reserve power fraction.

The derived cost of electrical energy produced by reserve plants was determined in ac- cordance with the annual number of hours of their use [4].

h =;Ere~es)87{;O, res

where Eres = expected annual output by the reserve plant, given by the formula

I E l'cNl

Eres~ S [ d N - I xdN-- l -{ -1" .

T h e r e l i a b i l i t y o f n u c l e a r p o w e r p l a n t e q u i p m e n t was c a l c u l a t e d on h i g h a n d a v e r a g e estimates in accordance with the findings of [5].

As indicated by an analysis of the results of the calculation, shown in Fig. l, the savings in derived expenses for the creation of reserve for a W~R-B compared to a W~R-1000 is determined by the following factors: increased structural reliability (absence of coolant pumps), less unit power, and characteristics of operation of the electrical sypply system (reliability and operating conditions). The savings in expenditures vary from 0.5 to 2.5 million rubles per year in installed power of 1000 MW. The greatest savings are attained with an average reliability of the equipment and compact load schedule on the supply system

( y = 0 . 9 ) .

The overexpenditure of nuclear fuel in a W~R-B is due, in particular, to low basic parameters of the working fluid in the steam cycle. An increase in fuel expenses (one mil- lion rubles/MW-year) in this case compared to a W~R-1000 is determined by the formula

bE F = I~" %~fd II0~ B B) -- f/(,IB)l/2~ , ( 1 )

where C n = unit cost of the nuclear charge in rubles/kg UO2; B and B B = fuel burn-up fraction in a WER-1000 and -B respectively, MW-days/kg U02; Teff = annual number of hours of power

utilization, h/yr; N8 and q = efficiency of a VV~R-B and W~R-1000.

A calculation was carried out with the following initial data: C n = 800 rubles/kg; B : B B = 40 MW × days/kg; Tef f = 6000 h/yr; N~ = 0.27; D = 0.33. For the assumed conditions,

813

TABLE i. Correlation of Factors Defining the Probability of Damaging the Core of Prospective Reactors, %

i /

Event VVER VVER-H

Transient conditions Disconnecting the plant from the electrical system Steam generator nozzle rupture Primary coolant leak large small Steam line rupture Loss of cooling/de-energizing the coolant pumps

Loss of pressure in shielding jacket due to a primary leak

6,25. l 0-:'/25,4 I ,15-I()-q41;,l;

3,75.106/1,5

?, 50.10-~f3, 1 f;,25 . l i t -e/2,G 4, Of I. 1 (I-~/0, OZ 5. n(l. I 0-.~/20, fl

l , 2f~. 1 . - % , .5

3, 12.10-;>/9G, |~

4,0IJ, |0-~I0, 1,5

-! ,20.1( r 6/3,7

Total probability 2,4"|O-411Reac'3,2"|0-~I/Reac tor-year tor-year

the calculation using formula (i) demonstrated that the excess nuclear fuel costs for a sys- tem with natural coolant circulation comes to 3.3 million rubles/1000 MW-yr.

However, it is necessary to note that the optimum geometry of the fuel grid, taking account of changes in the WER-B of the coolant parameters as well as a reduction of its average temperature, can give a certain advantage in fuel consumption. Therefore, the deter- mination of the total effectiveness of fuel utilization requires conducting detailed neutron- physical calculations on the core.

The correlation of capital component expenses of the alternative variations is deter- mined, on the one hand, by "scaling" factors (a reduction in unit power of the safety ele- ment compared to a W~R-1000 with another design, by the circuit arrangement, and by the specifications of the W~R-B assembly. The effect of the "scaling" factors studied are in agreement with the publications of various authors, including foreign ones [6]. Thus, for a i000 MW increase, capital expenditures total 30 and 20% for WERs of 350 and 500 MW power, respectively (with regard to the several additional increases in price because of the reduc- tion in parameters). At the same time, the exclusion from the layout of a series of elements (coolant pumps, steam generator vessels, and so on) leads to a decrease in expenses because of the avoided cost of their manufacture and assembly. Thus for a i000 MW WER unit, the body of the main circulation system totals a length of 200 m of piping 850 mm in diameter and weighing about 400 tons. The mass of the four steam generator vessels (PGV-250) comes to no less than i000 tons. The cost of four coolant pumps is estimated at four million rubles. A high temperature differential in the steam generators leads to a reduction in heat transfer surface (around i0,000 m 2 at a cost of about 500 rubles/mZ). Calculations show that savings, associated with the considerable reduction in branching of the reactor loop, amount to no less than 25 million rubles. The main part of the reduction in expenses makes possible factory assembly of the main parts of the primary system (according to various assessments, up to 20% of the cost of the reactor portion of the assembly). It is possible to obtain a certain reduction in expenses by an optimum choice of layout of the safety assem- bly (Fig. 2). Two reactors with a 350 MW electrical power output each, ensure operation of one turbine as the selected part of the high and low pressure turbine K-i000-60/1500 with dual intermediate moisture separators.

A complex calculation of all the enumerated factors reveals that the overexpenditure of capital component expenses for a W~R-B compared to a W~B-1000 comes to no more than 0.5-7 million rubles per I000 MW-yr (the higher limit for 350 MW, the lower limit for 500 MW).

From all the variety of failures, special significance is given to those which cause accidents of a definite probability leading to serious ecological incidents. Today's world- wide experience knows of at least two serious incidents: the accidents leading to the melt- ing of the cores of the Three Mile Island and Chernobyl' reactors. Significant differences in both incidents from the ordinary accidents involving the power plant are: significant cost overruns in cleaning up the consequences in comparison to downtime compensation of loss from underrutilization of electrical power in the system. Thus, it was assumed that only by

,= 0,3

0

0 J, S v -t 9

Fig. 3

"~ l ° l j

i

..T I I

3,~ J,Z - t,g,~

Fig. 4

Fig. 3. Specific insurance expenses Es~ for compensation following possible accidents as a functlon of operating time for an accident with a probable loss of 30 million rubles. (i000 MW unit power).

Fig. 4. Savings in insurance expenses Ein with a compari- son of alternative nuclear units with natural circulation and WER-1000 units, with a five-fold reduction in the operating time parameter, for an accident in a W~R-B.

the end of 1988 was decontamination of the reactor containment building to be completed. From published data, admittedly somewhat contradictory, the loss from the accident, which includes direct and indirect expenditures associated with the accident, is enormous. What has been said allows a conclusion, that in system technical/economical comparison and opti- mization calculations regarding nuclear power plants, one of the necessary demands is safety in all the variations. Among those variations with different reliabilities, not only is a constant output of electrical power important, but in first place is the ecological cost of the.variations, being distinguished by the probability of the emergence of dangerous situa- tions. The most dangerous accident involves destruction of the core and especially the dis- charge of radioactivity into the ambient atmosphere. Its ecological consequences are the most serious and they can lead to catastrophic results. The probability of destruction of the core is made up of various factors, a tentative relationship which is estimated for prospective WER units in a chart (data from the I. V. Kurchtov IAE). Orienting oneself in these data, one tries to assess the reduction in probability of core destruction on the basis of the safer structure of a reactor with natural circulation (Table i).

From an examination of such a main coolant loop structure, there emerges a class of accidents associated with abnormalities in its operation and some failures of the steam gener- ators. The influence of transient conditions is significantly limited. This is inherent to a reactor plant with self-regulating natural circulation which compensates both for long term effects connected with the burning of the fuel and for short term ones dictated by changes in power, without having to inject boron into the coolant or move the control rods. These prop- erties depend on the close coupling of negative reactivity with a change in coolant density. An analysis of the table shows that the more secure structure of the main coolant system in a VVER-B allows one to reduce the number of standard possible accident situations and thereby lower the probability of an accident involving core damage.

To take into account the ecological component of expenditures, one can use the insurance charges on electrical power generation by nuclear power plants given by the formula

Eec = Eac%/(Nh) rubles/kWt.h ,

where Eac - expected loss from an accident with core damage, rubles; ~ - probability of a large-scale accident with core damage, per reactor-year; N - unit power, kW; h - number of h of utilization of the installed rated power, h/yr.

The dependence of Eec on the parameter ~ in a logarithmic scale is shown in Fig. 3. From this it can be seen that with the attainment of X = (2-3) × 10 -5 per reactor-year, the need for insurance payments is significantly reduced. In Fig. 4, showing the savings in ex- penditures with the alternative nuclear unit, the loss from supposed accidents was assumed to be proportional to the unit's rated power. Complex analysis of all the factors considered testifies that with the attainment by the traditional V~ units of high safety parameters

T,"C

~ °

1 7 0 ~ I t 18 ~ 2 Pc, MPa

F i g . 5. Optimum f e e d w a t e r t e m p e r a t u r e o f a WER-B u n i t f o r s t eam g e n - e r a t o r s u r f a c e h e a t t r a n s f e r modulus c o s t s (CSG) of 300 ( ) , 500 ( .... ), and 700 rubles/m 2 ( ..... )and for fuel at 400 (I), 600 (2), and 800 (3) rubles/kg UO 2 respectively, with a change in pressure in the intermediate steam separator.

(~ ~ 2 × 10 -5 per reactor-year), their technical/economical effectiveness would be high. Low thermodynamic effectiveness and possible overexpenditure of capital investiments in units with natural circulation is compensated by the higher structural reliability and ecological safety with ~ = I03-I0 -~ per reactor-year, typical for current power plants. Under these conditions, the plants considered would be effective.

According to the evidence, the efficacy of a WER-B is the most important step in the technical/economical optimization of some subassembly parameters of the fuel arrangement, such as the temperature of the feed water and the pressure of the intermediate separator, given the predetermined design flow characteristics of the reactor and turbine. Low initial operating fluid parameters (minus the steam drum of the high pressure turbine K-1000-60/ 1500) determine the regenerative scheme for pre-heating the feed water, limited by the deaerating heater (see Fig. 2). The stipulation on selection of the optimum feed water tem- perature is written in for form

(SE/tFW)G0= (SEF/StFW + 8Ec/StFW)G0 - (SN/ 8tFW)G ° (SEFISN + 8Ec/ 8NtFw) = 0,

where E, E F - derived and fuel costs, rubles/year; E c - capital equipment costs, rubles/year; N - electrical power rating of the plant, kW; tFW - feed water temperature, °C; C o - steam flow rate to the turbine, kg/second.

With an increase of E F from 400 to 800 rubles/kg UO 2 there occurs an increase in the optimum feed water temperature of 10°C (Fig. 5). This is due to the fact that the thermody- namic optimum feed water temperature is higher than the technical/economical optimum tempera- ture and is ~150°C. A reduction in tFW with an increase in CSG from 350 to 700 rubles/m 2 is due to the stronger influence on the surface of the steam generators of the &T in temperature than heat being supplied during a change in tFW because of a significant difference in the total thermal equivalent of the coolant and feed water.

Thus, the principal opportunity for a significant increase in ecological safety through the generation of power in nuclear power plants having reactors without external coolant loops and having naturally circulating coolant predisposes interest in the plan for the gen- eration not only of heat (AST and ASPT projects), but also of the production, at its site, of electrical energy. A comprehensive analysis is necessary of the possible characteristics of the fuel cycle, to define more precisely the cost characteristics, and to carry out the complex optimization of the parameters of the thermal schemes of such plants.

LITERATURE CITED

i. B. Binghman, Transactions of the American Nuclear Society, No. 50, pp. 430-431 (1985). 2. R. Harris, R. Letendre, and M. Grump, "Physics and safety aspects of the minimum atten-

tion plant," Topical Meeting, N.Y., September 17-19, 1986. 3. V.P. Girshfel'd and E. G. Sklovskaya, "Definition of emergency underproduction via a

technical/economical comparison of variations in steam turbine installations," Thermal Power, No. i0, pp. 9-12 (1970).

4.

5.

6.

A. I. Andryushchenko and R. Z. Aminov, Optimization of the Operating Regime and Param- eters of a Thermal Electric Power Plant, Vysshaya Shkola, Moscow (1983). N. E. Buinov, S. M. Kaplun, and L. S. Popyrin, "A calculation of reliability by the optimization of the arrangements of the power units of W~R power plants," At. ~nerg., 57, No. 3, 157-161 (1984). R. Sullivan, Design Evolution of the Electric Power System [Russian translation], ~nergoatomizdat, Moscow (1982).

TESTS ON A SETUP WITH NATURAL CIRCULATION OF LiF-BeF2-UF 4

LIQUID-SALT FUEL

V. D. Braiko, V. V. Ignat'ev, V. M. Novikov, A. I. Surenkov, I. B. Tikhomirov, V. I. Prusakov, V. I. Fedulov and V. N. Cherednikov

UDC 621.039.52.034.6+621.039.519

Efforts are currently being devoted to the development of new nuclear reactor designs whose physicochemical characteristics provide the means of radically improving safety. A design of this type is a circulating-fuel reactor using light- or heavy-metal fluorides. Liquid-salt reactors (LSR) offer enhanced core-control capabilities, and hence improved safety [i] - dependent, of course, on the efficient operation of process equipment (fuel circulation,

full-scale fuel cleanup system, etc.).

The performance of new process systems can be tested on research-reactor stands under conditions most closely approximating the operation of an LSR fuel circuit. A MURS-2 reactor setup with natural circulation of liquid-salt fuel was built for this purpose at the I. ¥. Kurchatov Institute of Atomic Physics [2]. The setup is an integrated unit, designed for testing the performance of LSR components, studying the thermal characteristics of natural circulation, and assessing the radiation-chemical stability of liquid-salt fuel compositions.

Tests were performed in a WR-SM reactor with a neutron flux up to 0.76"10 l~ cm-2"sec -l (Fig. i). The test compartment i is fixed with its lower section in the center channel 2 of the core 3 by means bracket 4, and its bottom resting on the base plate 4. The outer surface of the test compartment is cooled by reactor water. The reactor tank 6 is encased in radia- tion shielding 7, and water 8 fed into it through pipes 9 passes through a mixing grille i0 and runs down the sides of the test compartment. The assembly of the test compartment is shown in Fig. 2. The main component I is a sealed-in cylindrical thermosiphon filled with a 7LiF-BeF~-23sUF~ composition (melting point 458°C). The thermosiphon is made of 12KhlBNIOT steel and enclosed in two security jackets 2. Heating coils 3, 4 are wrapped around the length of the thermosiphon for salt premelting. The thermosiphon section located in the core is a heating zone, and the section above the core is a cooling zone. The heat released by nuclear reactions directly in the melt and thermosiphon wall is transferred by free convection along the center tube 5, through the overflow holes 6, into the cooling zone, where it is removed by reactor water. The cooled melt then proceeds to the core, where it heats up. Heat is removed from the thermosiphon surface primarily in the upper section, since the annular gas gaps separating the melt and cooling water are five times narrower there than in the core section. The thermosiphon and security jackets are connected to a gas-evacuation system 7. The setup provides for fluorine prepassibation of the gas lines, mass-spectrometric analysis of gas samples during irradiation, and continuous pressure monitoring in the gas spaces. The test-compartment design ensures that the space heat release in the melt and am- pul wall at a maximum reactor output of i0 MW should not raise the thermosiphon temperature to more than 750-800°C. Heat is removed from the thermosiphon surface by radiation and con- duction, without free gas convection in the gaps. The thermal output is calculated from working formulas and calibration readings of 15 thermocouples spaced out along the surface of the thermosiphon and security jacket. Heat removal is controlled by varying the pressure

Translated from Atomnaya ~nergiya, ¥oi. 69, No. 4, pp. 211-215, October, 1990.

article submitted April 18, 1989.

Origindl

nn~_ERixlgO16904-0817512.50 © 1991 Plenum Publishing Corporation 817