INTERNATIONAL CONFERENCE ЩON NUCLEAR POWER AND ITS FUEL CYCLE jjw

SALZBURG. AUSTRIA • 2-13 MAY 1977

IAEA-GN-36/342

SAFETY OP SODIUM-COOLED FAST REACTORS

Yu.E.Bagdasarov, L.A.Kochetkov, V.I.Matveev,

M.Ya.Kulakovsky, I.A.Kuznetsov,

G.B.Pomerantsev, H.V.Krasnoyarov

USSR

Abstract

In the paper some physical and technical aspects of so-

dium-cooled fast reactor safety are discussed. An approach

to the solution of safety problems used in the designs of the

USSR first fast-reactor generation was based (as a result of

thermal-reactor operating experience) on a logical analysis

of the tree of failures, on careful development and testing

of safety systems and devices, and on duplication and eche-

lonment of the most important of them. A review of main works

carried out for the last two decades is given and the most

important results of design studies, research and operating

experience obtained on fast reactors as to accumulation of

data and characteristics in the area of safety are presented.

The most valuable of the above results have been obtained in

the process of operation and investigations on the BR-1O and

BOR-60 reactors, from design studies of the BN-350 and ВЫ-600,

and from the BH-350 experimental work and operation. A con-

templated extensive fast reactor development program calls

for a consideration, in the future, of a wider range of prob-

lems including definite attention to very low probability ac-

cidents. In the paper some advantages and disadvantages of

the concepts of nuclear power plant safety evaluation using

the maximum credible accident analysis and quantitative pro-

bability methods are discussed.

damaged subassemblies at the shut-down reactor both in the in-

1. IHTRODUCTIOH

The necessity for the extensive development of fast reac-

tor nuclear power plants does not give rise to any doubt now.

Opinions of various scientists differ only as regards the time

when particularly the plants serial construction should be

started. A considerable part in the determination of this time

is assigned, among other factors, to the problems of ensuring

the nuclear power plant personnel and environment safety.

As any new trend of power engineering the sodium-cooled

fast reactor power plants should pass through stages of expe-

rimental verification of a technical idea, of experimental

operation and of serial production.

It is obvious that in the process the attitude to the

problems of plants safety, the means and methods of its ensu-

ring should be considerably transformed.

Reactor development in the first and second stages is

based upon the available experience in other areas of techno-

logy and on the successively obtained results from special

experiments and operation of previous reactors. Therefore,

each new plant at these stages is a unique one and calls for

specific technical solutions of safety problems. Possibilities

for development of standards and their application in designs

open up directly in the period of preparation to the third

stage and in the period of nuclear power plants serial const-

ruction. This is connected also with the scale factor due to

additional technical and technological problems arising with

the increase of plant dimensions.

The program of sodium-cooled fast reactors has been de-

veloped in the USSR from 1950. The period from 1950 to 1970

may be considered as an experimental verification stage of a

technical idea that was concluded with putting into operation

and successful operation of the BR-5 (Ш-10) and BOR-60 re-

search reactors. ,

The second stage of fast-reactor problem development in

the USSR connected with the creation of pilot commercial and

heading serial nuclear power plants was begun from develop-

ment and start-up in 1973 of the ВЫ-35О plant which was

reactors with different fuel types and of various volume.

Theae ev " • • _

operated a t a power of 650 MW(t) producing e l e c t r i c a l powerand deealted water. At present mounting work at the BN-6Q0plant of 600 MWe power i s being completed, design developmentof a heading s e r i a l nuclear power plant of about 1500 MW elect -r i c a l power has been s tar ted [i • 5J»

2 . GENERAL APPHOAOH TO SAFETYOF FIRST-GEHERATION SODIDM-OOOLED FAST REACTORS

When analyzing and evaluating the safe ty of the USSRf i r s t - g e n e r a t i o n f a s t reactors which include also the BN-600,there were primarily taken into account requirements uponnuclear p lants and power plants with thermal reactors whichreceived f o r e s t a l l i n g development. As noted above, each sue-ceding p lant was considered as a unique i n s t a l l a t i o n with a l lthe r e s u l t i n g requirements upon experimental and c a l c u l a t i o -nal substantion of construct ions and operating cond i t ions ,upon the equipment fabr i ca t ion qual i ty and i t s control duringthe whole operation per iod . The unique character of the f i r s tresearch f a s t reactors was determined by such well known fun-damental features of sodium as i t s chemical a c t i v i t y towardsoxygen and water as we l l as a high melt ing temperature andinduced a c t i v i t y of coo lant , high s p e c i f i c power of the coreand r i s e of coolant temperature i n the reactor , i n combina-tion with a relatively low heat conduction of stainless steelused as a structural material, lack of technological opera-ting experience with sodium, small prompt-neutron lifetimeand small delayed-neutron fraction for plutonium fuel, etc.

Experimental studies, equipment development, operatingexperience of the f i r s t research reactors (especially of theBR-5 and BOR-60) and passing to designing of large nuclearpower plants have substantially displaced the centre of at-tention as regards the fast reactor re l iabi l i ty and safetyproblems.

Heat-removal channel hydraulic characteristic ambiguityand a positive sodium void reactivity effect, structural ma-terial swelling, the effect of the scale factor upon the fab-rication quality and working ability of systems and devices,

- development and promotion of integral means and instru-

as well as upon the scale of radioactive contamination at re-

pair work and in case of an accident, etc., - all these fac-

tors have brought forth new problems which are important from

the nuclear-power-plant safety point of view. These circum-

stances have also called for an individual approach to the

Ш-35О and EN-600 projects. The uniqueness of each new plant

was determined, besides the above, by designers'desire to ve-

rify various fundamental technical ideas the realization of

which affected also the results of safety analysis. E.g., an

important distinction of the Ш-600 from the Ш-35О lies in

different principles of primary-circuit equipment arrangement,

steam-generator types and parameters, in the character of ge-

nerated heat utilization*

When evaluating the first fast-reactor generation safety

under emergency conditions there was used a method of logical

analysis of all really probable situations with calculated

studies of consequences at some attendant failures of safety

and control systems and devices. Particular attention was gi-

ven to safety systems on power value and its rise rate, as

well as to protection from situations causing sharp decrease

of coolant flow rate through a reactor. Characteristics of

the above processes call for a safety system action during

very short times. Therefore, in these cases safety systems

over parallel channels are provided, and detection of dange-

rous deviations and reactor shut-down over several parameters

are ensured. E.g., the alarm signalling at switching-off of

the Ш-600 primary pumps is formed over the channels of de-

energizing, of pumps revolution speed decrease and of flow

rate decrease through a reactor with appropriate correlation

of safety system limits. Besides, for forming an additional

alarm signal it is planned to use an analogue unit with spe-

cial processing of readings from thermocouples for tempera-

ture control at the reference subassembly outlet, inertial

characteristics of this analogue unit approaching those of

the above three channels* Such an approach and calculated

analysis were carried out for perturbations of reactivity and

coolant temperature at the reactor inlet, as well as for

other accidents including the technological ones. On the basis

avalanche-type release and passing through the core of gaseous

fission fra nts enetration of lar e as bubb e m

of logical and calculated analyses the requirements on the

development of safety systems and devices were formulated and

parameters of systems and devices for normal operation were

refined. After project completion, experimental testing and

determination of characteristics for mock-ups or prototypes

of equipment, instruments, systems, a repeated analysis of

all emergency cases was carried out with talcing into account

the output data obtained [6 * 8].

In fast reactor projects (up to the Ш-600) very low pro-

bability accidents (de-energizing of sodium circulating pumps

with simultaneous failure of all safety rods, postulated in-

put of positive reactivity, etc.) were not postulated and

considered.

In connection with the above circumstances, specific sec-

tions in the Government standards which would consider sodium-

cooled fast-reactor features are still under development now.

Designers and constructors are guided by accumulated expe-

rience and by requirements which are from time to time refin-

ed and formulated in project specifications foi designing of

each particular nuclear power plant.

3. Ш SAFETY REQUIREMENTS TO A PAST REACTOB PROJECT

Within the frames of the present paper there is no possi-

bility to cover in detail all safety requirements imposed

upon a fast reactor nuclear power plant project at present.

We shall briefly dwell only on main principal points. [4,6*8]

3.1. Sodium Void Effect of Reactivity

With increase of reactor sizes the sodium void effect of

reactivity becomes positive not only in the limited region of

the core but also over th. reactor as a whole. Unless any

special steps will be taken, then in the future serial nuclear

power plants of about 1000 We and higher power the integral

positive sodium void effect of reactivity шау reach 4*5 dol-

lars. For the first generation of fast reactors in the'USSR

the integral effect was considerably negative. The calculated

analysis of experiments on critical assemblies which was car-

ried out with the use of the BKAB-70 system of constants

are introduced. By an intermediate safety criterion is under-

shows that the calculation accuracy of the considered effect

for large reactors may be within 50*100%. For medium-size

reactors there is an obviously pronounced tendency toward in-

creasing of the positive component of the effect. Рог example,

measurements performed on the Ш-35О have shown the absence

of any pronounced region of the positive effect, whereas at

the development stage a region of the positive effect with a

radius of 30 cm was predicted. The presence of expected limi-

ted regions of the positive sodium void effect in the EH-600-

type reactors does not play any decisive part from the safety

standpoint in view of a relatively large negative temperature

feedback with the Doppler effect playing a determining part

(70+80% of the total effect). Experimental and calculated

studies and testing of temperature and power reactivity effects

at the BN-350, an analysis of results using the ШАВ-70 Con-

stants give ground to believe that the above reactivity ef-

fects can be calculated for large reactors to an accuracy of

about 15%. Despite the dangerous role of the sodium void ef-

fect of reactivity under hypothetical emergency conditions of

fast discharge of coolant from large reactor cores, a requi-

red degree of talcing into account this factor at determining

optimum physical and design characteristics of a core for fu-

ture nuclear power plants is still to be determined.

3.2. Protection against Core Drainage Accident

In addition to dangers connected with the sodium void ef-

fect of reactivity, the core drainage inevitably results in

fuel melting and draining with all the ensuing consequences.

To eliminate such a possibility, the reactor vessel is enclo-

sed into a protective housing which should take loads after

its communication with the in-vessel space and whose volume

should be confined so that sodium level in the reactor would

not fall below the limiting permissible one. In case of a

loop arrangement (the Ш-35О) the housing encloses also some

of the main pipings up to the cut-off valves. When consider-

ing this problem, the philieophy of an approach to the acci-

dents with a prompt rupture of the primary piping over the

practically atmospheric pressure, a rupture in piping is con-

siderably less probable, and at the inte ral-t e arran ement

whole cross-section iB of great importance. As is known, for

water reactors it is postulated as a maximum credible accident

that results in providung appropriate safety and localizing

devices. We believe that, as applied to sodium-cooled reactors

(low pressure in the primary circuit in combination with the

use of highly plastic structural materials), this requirement

is hardly justified. In the Ш-35О reactor project such an

accident was not considered.

In this connection, the integral-type arrangement of the

primary circuit (the BH-6OO) has significant advantages. If

operating experience of loop- and integral-type reactors shows

that the latter do not cause any additional fundamental diffi-

culties (in assembling, repair, inspection of equipment and

systems), then, in future, integral-type arrangement should

be preferred as a more safe and compact one.

3.3. Emergency Heat-Removal Systems

High specific heat releases in a fast reactor cause a t

.a 'i

need for ensuring large flow rates of coolant. Even after a

safety-rod trip, the specific heat loads in a fast reactor

prove still to be at a nominal-load level of water-moderated

water-cooled power reactors. This imposes corresponding re-

quirements upon the laws of decreasing coolant flow-rate

through a reactor and then to the natural circulation flow-rate

level. From this point of view, the mechanical pumps have de-

finite advantages over the electromagnetic ones. The mechani-

cal rundown laws of the BN-350 and BN-6OO primary and seconda-

ry pumps satisfy this requirement: the relative sodium flow

rate, practically over the whole rundown range of interest,

remains higher than the relative value of heat release. Other

important requirements to safety heat-removal systems are:

the provision of such an arrangement of equipment that allows

under extreme emergency conditions to realize heat removal

througb natural circulation of primary and secondary coolants

and to ensure heat removal through several parallel loops., It

should be noted that, from the viewpoint of emergency heat re-

moval, sodium is the best of the coolants developed, barge

8. Kuznetsov I.A., et al. "Past Reactor Transient Operating

Conditions" The USSR-USA Semina P "

margin up to the boiling temperature at atmospheric pressure(400 + 500°C) and high heat-tranafer coefficients under natu-ral circulation conditions allow to remove heat out of thecore and to accumulate it in design components and coolantwithout core destruction even in case of prolonged cutting-off the feed of water to steam generators. In this case therate of temperature increase in the primary circuit of largereactors is 20 • 50°C/hr. The above eircumstanos gives somescope for optimization of technical solutions over additionalmeans for emergency shut-do\sn cooling.

3.4. An Approach to Local Accidents Detection

Loss-of-power and reactivity-perturbation accidents arethe most quick-proceeding and dangerous ones from the pointof view of their timely detection and reactor shutdown.

The arrangement of the alarm signal system in these caseswas described above. Of other quick-proceeding accidents, thelocal accident with prompt plugging of a subassembly cross-section is usually considered. A detailed analysis of condi- .; 'tions for coolant passage from the overall circuit into sub- ^assemblies has shown that such an accident is eliminated bymeans of adopted design solutions for the head collector andsubassembly tails. Partial plugging (even up to 50% inletcross-section) does not result in dangerous and fast deterio-ration of fuel-elements temperature conditions. On the otherhand, up to now there are no reliable instruments and systemswhich would allow to detect in time an accident with promptand complete subassembly plugging and to shut down a reactor.This also holds true for the most developed method, that oftemperature measurement at each subassembly outlet. In con-nection with the above circumstances, for the first genera-tion of fast reactors no indispensable requirement for con-tinuous individual monitoring of each subassembly was putforward. As to slow processes of heat-removal deterioration(with the time constant of above 1 min.) that result in agradual loss of tightness of fuel elements, they can be easilydetected during reactor operation by the delayed-neutron

monitoring system and during shutdown and refuelling periods -by sodium flow rate measuring systems through, each subassembly,as well as by the in-reactor system for unsealed fuel subas-semblies detection.

3.5. Emergency Shut-Down System

A great deal of parameters which should be monitored atan operating reactor and whose deviations from standard valuesshould result in a reactor shut-down pose complicated problems.On the one hand, from the viewpoint of maximum plant safety,any really dangerous deviation should lead to dropping of thesafety rods. On the other hand, false signals from instrumentsand devices should be revealed to a maximum degree, and inthese cases forming of an alarm signal must be prevented.Therefore, shaping of these signals is realized according tologic schemes (2 of 2, 2 of 3» etc.). On a logical block arequirement is imposed that no unit failure within a block orno personnel fault would result in a failure of shaping ofan alarm signal should it be required under these conditions. |

All the safety rod drives must be separated from one another |each of them being provided with an electromagnet. The choiceof the number and total efficiency of safety rods so far didnot cause any difficulties. The analysis shows that theseproblems will also cause no serious difficulties in case ofserial nuclear power plants. The requirements imposed on thetotal safety rode efficiency are based on the assumption of apossible failure of one most effective rod; the rest of themmust ensure the reactor transition into the subcritical stateunder any normal and emergency conditions (hypothetical accidentsof fuel redistribution being not considered) and maintainingof this state up to temperatures of about 100°C.

3.6. Reactor Protection against Penetrationof Hazardous Substances

One of the main requirements imposed on a fast reactorplant project is elimination or limitation of penetration intothe primary circuit and in the reactor of moderating substances

and materials which actively react with sodium or cause sub-

stantial deterioration of coolant properties. This has resul-

ted in incorporation of the intermediate sodium circuit into

the plant scheme and poses some design limitations on auxilia-

ry cooling and lubricating systems for the mechanisms used in

the primary circuit. But it should be noted that incorporation

of the intermediate sodium circuit by itself does not eliminate

the danger of water and water-sodium interaction products pe-

netration into the primary circuit: we mean a possibility of

intermediate heat exchanger damage'due to static and impact

loads which may arise in the secondary circuit at a large ac-

cident in the steam generator. That is why the project require-

ments on safety and containment systems and devices for the

secondary circuit and steam generators are important from

the viewpoint of safety. Note, that they are dictated not only

by conditions of assured retention of intermediate heat ex-

changer integrity, but also by conditions of preserving secon-

dary circuit tightness.

З Л . Fire Safety

A distinguishing feature of sodium-cooled fast reactors

is a need for the development of specific systems, devices

and means to prevent hazardous consequences of sodium leakage

from tight circuits, especially from the primary circuit. The

efficient means used to localize the consequences in the lat-

ter case are sealed shells for equipment and air-tight com-

partments of the primary circuit filled with inert gas and

having a closed system of technological ventilation. In some

cases, particularly as applied to auxiliary systems of the

primary circuit, as well as for all secondary circuit com-

partments, the requirement for filling with inert gas has

not been brought forward. Steel lining of compartments, pro-

vision of special drains and pans of different types with

hydroseals, and sometimes the volumetric-fire extinguishing

system, are used as necessary additional fire-prevention means*

Operational experience of the B0R-6O and Ш-35О plants has

shown that in order to achieve acceptable tightness of

10

primary-circuit compartment it is песеввагу to carry out se-

rious design studies and to perform carefully the assembling

operations.

3.8. Radiation Safety

When developing the requirements and means on sodium-

cooled fast reactor radiation safety, first of all, the All-

Union well known health regulations for nuclear power plant

designing and radiation safety standards are taken into ac-

count. Therefore,in this section we shall briefly dwell only

on those requirements and means which are connected with de-

sign features of the considered reactor type. One of the main

requirements is a limitation on a list or conditions of appli-

cation for shielding materials: there should be eliminated a

possibility of penetrating into sodium coolant of carbon,

hydrogen, materials which actively react with sodium. She

main source of radioactive radiation of coolant at an opera-

ting reactor is the inherent activity of sodium. The calcula-

ted value of sodium-24 activity for the Ш-600 reactor is

20 curie/1, that of sodium-22 at saturation - 5 :millicuri/l.

A significant factor affecting the radiation situation at

performing repair work is the escape of fission fragments

from unsealed fuel elements and corrosion of structural mate-

rial activated components* Escape of fission products from

faulty fuel elements and their distribution over the circuit

were intensively studied on the BR-5 and BOR-60 reactors.

The experimental data analysis leads to the following

maximum estimates of the isotope yield fractions from faulty

fuel elements: X e1 3 3

- 30%, C s1 3 7

- 30%, I1 3 1

- 2%,

B a1 4 0

+ b aU G - 0.3%, Zr

9 5 + Nb

9 5 - 0.3%. The major fraction

of I1 3 1

is contained in the cold trap (-90%).

The calculated values of the primary-circuit total acti-

vity for corrosion products for the Ш-600 reactors are:

Co5 8 - 6.8 x 10

3 curie, Go

6 0 - 3.8 x 10

3 curie, M n

5 4 - 5 Л П О

3

curie.

The allowable design loss of tightness for fuel elementsis 0.1% of their total number.

11

The total activity of noble gases in the Ш-600 reactor

gas system (2500 curie is below the allowable daily discharge

in case of emergency discharge of the whole gas out of the

gas blanket.

During repair work some special precautions are taken to

ensure the primary circuit tightness: inert gas (argon) supply,

sealing of withdrawn and transported* removable components

within special containers, etc.

Means taken for the restriction of radiation effects due

to sodium burning and aerosol activity spread are closely con-

nected with the fire-safety means (see З Л ) . Besides, there

are imposed some additional requirements of separation of

technological and special ventilation systems, of trapping

and directed discharge of radioactive sodium aerosols, of

post-accident cleaning and repair technology for rooms and

equipment, etc.

4* REVIEW OP RESULTS FROM Ж Е Н STUDIES

FOR SUBSTAHTIATIOH OF FAST REACTOR SAFETY IK THE USSR

4.1. Work on Substantiation of Safety and Reliability

of Systems and Devices for Hormal Operation [1 • 12]

The questions of plants reliability in operation and sa-

fety are especially closely connected in initial stages of de-

velopment of this problem and become appreciably separated

only when passing to mass introduction of serial plants. That

is why in the last few decades primary emphasis was placed on

studies and development of systems and devices for the normal

operation.

Past reactor core physical characteristics were substan-

tiated with an extensive complex of calculational and theore-

tical work and experimental studies on the rigs including

those on the BPSzero-power reactors with full-scale assemblies

of power reactor cores being designed. Por substantiation of

fuel elements design and operational characteristics there

were carried out extensive studies on test specimens, and mass

experience was obtained on fuel elements of the BR-5 and

B0E-6O oxide-fuelled cores up to burn-up of above 135S heavy

12

atoms; fuel elements of the Ш-35О plant reached tne

rated burn-up.

A complex of calculational and experimental hydraulic

and heat-transfer studies and developments of the core tempe-

rature calculation methods with taking into account non-uni-

formities of neutron and pressure fields under the conditions

of central and peripheral heat-removal cells for different re-

lative fuel-elemert spacinge, various diameter and pitch va-

lues of spacing wire winding, etc., were carried outс Develo-

ped are calculational methods for evaluation of fuel elements

working capacity and permissible fuel burn-up limits with

account for non-uniform neutron and temperature fields, swel-

ling, interaction between fuel and canning, and other factors.

Much attention w§s given to technological equipment develop-

ment. The main BN-350 and BH-6OO sodium circulating pumps

have undergone prolonged testing on full-scale rigs. Mock-ups

of intermediate heat exchangers and steam generators were

tested on air- and water-filled hydraulic rigs; a 3 Mw mock-up

of the BH-350 steam generator stood up to a resource of

20 000 hours, a 30 Mw mock-up of the direct-flow coil-type

steam generator was successfully operated at the BOR-60 for

more than 20 000 hours. At present, at the BOR-60 a 30 MW(t)

modular type steam generator of Gzechoslovakian design has

been operating for more than 15000 hours, and a EN-600 steam-

generator module of the same power is being prepared for

testing.

Methods and equipment for preparation of specified-purity

sodium, methods and means of sodium purification from impuri-

ties during reactor operation have been developed on rigs with

subsequent testing in industrial conditions.

Technological studies on the choice of structural mate-

rial and development of equipment fabrication technology were

carried out; methods of quality monitoring during fabrication

aiid assembling were developed.

The most important units of large-sized equipment were

tested on the mock-ups of various scale. Characteristic

examples of these tests are: testing of the outlet piping

joint unit at the BE-350 reactor vessel for heat changes,

13

investigation of hydraulic characteristics of the joint unit

at the coolant inlet into the vessel; tensometric and vibra-

tion investigation of the Ш-35О vessel and in-vessel compo-

nents, structural strength tests of the mock-up of the BN-600

vessel upper part, full-size testing on a rig of vibration

characteristics of the ЕИ-6ОО primary circuit pressure piping

together with the standard pump, mock-up investigation of the

Ш-600 primary circuit hydraulic characteristics, development

and testing of standard mechanismes for the in-vessel and

extra-vessel handling systems of the Ш-35О and Ш-600 plants

at sodium rigs, etc.

4.2. Investigations and Tests on Substantiation

of Safety and Localization Systems

and Devices [4» 6, 7, 8, 10, 13J

The most important safety system is that of reactor shut-

down at emergency situations appearance. The standard EN-350

and Ш-600 safety rods together with the drives underwent re-

source testing and full-sized tests for thermal shock on so-

dium rigs. Subsequent experience of the EN-350 operation has

confirmed high reliability of the adopted designs* Electronic,

electric and logic components of the safety system have under-

gone v ~ification tests at rigs and under reactor conditions.

Much attention was given to an increase in reliability and to

a decrease in inertia of instruments and devices displaying

dangerous deviations in parameters. In this connection one

should mention development and testing in BH-350 plant condi-

tions of analogue safety devices based on sodium flow rate

and temperature monitoring.

An important complex of safety problems is related to

the problems of fast-reactor operation with unsealed fuel

elements. At the ER-5 and BOR-60, extensive studies were car-

ried out on the development of technological operating methods

for oxide- and earbide-fuslled reactors with unsealed fuel

elements. Developed and tested were a system of unsealed-

subassembly detection at the operating reactor from delayed-

neutrons monitoring systems and devices for detection of

14

damaged subassemblies at the shut-down reactor both in the in-

pile and out-of-pile conditions.

The problems of sodium burning and fire extinguishing

were investigated as in the process of experimental and reac-

tor plant operation, so en special rigs with cell volumes up

to 390 m^ and poured-out sodium quantities up to 130 kg. On

these latter there were investigated effects of sodium tempe-

rature and pressure, initial-fault size, thermal insulation

type, environmental conditions in a cell where sodium was

poured-out upon the rate of leakage development, time-to-rup-

ture of thermal insulation layer and jacket, rate of sodium

burning, maximum aerosol concentration, fractions of sodium

taken away into aerosols, and others. As pointed out above,

fast reactor safety problems are closely connected with the

processes taking place in the secondary circuit at large rup-

tures in steam-generator heat-transfer tubes. Here a complex

of investigations covered mock-up studies of physical patterns

of small water-into-sodium leak initiation and development

processes and of the effects from the insantaneous heat-trans-

fer tube failure over the total cross section, development and

experimental testing of safety and localizing systems and de-

vices.

Transient process studies take much space in fast-reactor

safety studies. Over the past years there were developed and

experimentally tested models: of calculation of transient ope-

rating conditions for fuel elements, heat-exchange devices,

structural elements; of reactor dynamics with taking into ac-

count of temperature, flow rate, pressure and power feedbacks

and with account for safety systems and devices operation; of

calculation of transient thermal stresses in plant structural

elements; of transient hydraulic processes at various equip-

ment failures; of reactor eiaergency heat removal calculation

due to forced and natural circulation of coolants, etc. An

extensive complex of experimental investigations on the Ш - 5 ,

BOR-60, Ш-35О was carried out to confirm and refine the deve-

loped models.

Studies of the "reactivity-power11 channel stability al-

lowed to formulate criteria and to evaluate stability of fast

15

reactors with different fuel types and of various volume.

These evaluations convincingly reveal stability of oxide-

fuelled, liquid-metal cooled fast reactors with core volumes

up to 104l and more. In contrast to thermal reactors, in

fast reactors there is no xenon poisoning, and in the process

of their development no problem of spatial instability was

met yet. Therefore, instability as a physical phenomenon pre-

sents no serious hazard to fast reactor safety, and some

cases of instability were caused by design errors in reactors

of very small volumes.

5. FURTHER DEVELOPMENT AND RESEARCH PROBLEMS

Ш THE REGION OF FAST REACTOR SAFETY

Further development and research problems in the region

of fast reactor safety can be divided into two trends which

should be developed in parallel.

The first trend covers the problems of advancement and

reliability improvement of safety and control systems and de-

vices intended to prevent or decrease the effects of really pre-

dictable emergency conditions. The most important of them are:

- a transition from parameters measured on a real time

scale in safety systems to an early prediction of a future

failure forecasted by computers;

- accumulation of statistical data; development and em-

ployment in the process of design of practically confirmed

methods for calculating the safety of individual components

related to the reactor safety; optimization of technical so-

lutions on the basis of reliability analysis;

- development of well-founded core-damage limits permis-

sible in various stages of accident development (ideally5 ob-

taining of a relationship between a core state and the allow-

able time of stay in this state);

- calculated and experimental studies of the sodium void

effect of reactivity for various compositions, states and

sizes of fast reactors; increase in accuracy of methods for

calculating the local and integral values of the effect;

16

- development and promotion of integral means and instru-

ments for timely detection of the onset of sodium boiling in

the reactor;

- refinement of methods and means for primary-circuit

room sealing, for early detection of sodium leaks out of the

circuits, for sodium-fire extinguishing and aerosol trapping;

- increase of steam-generator reliability, refinement of

their safity and localizing systems and devices;

- development and introduction into practice of standards

for sodium-cooled fast reactor plants design and operation.

The second trend of further work in the area of fast-

reactor safety includes work on revealing the mechanism and

probability of process development at very low probability ac-

cidents, as well as development of nuclear power plant safety

evaluation methods п Д

Very low probability accidents in a sodium-cooled fast

reactor may include a rupture over the total cross section of

the primary-circuit main piping and an accident with fast rea-

lization of the positive sodium void effect of reactivity.

The first of them was considered in a common way because it

had been postulated as a maximum credible accident for water

reactors. A possibility of realizing the sodium void coeffi-

cient of reactivity is mainly connected with two schemes of

accident development: propagation of a local accident in an

individual subassembly onto the whole core and complete de-

energing of sodium pumps followed by a failure of all safety

rods. Hote, that a probability of both processes is conside-

rably less than that of a high-pressure piping rupture: if a

possibility of sodium void effect realization is connected

with a failure of many safety-system operating devices, then

a pipe rupture may be caused by an undetected manufacturing

fault. Calculation and experimental studies performed to date

show that a possibility of rapid avalanche-type propagation

of an accident at sodium local-boiling initiation within

subassembly inter-rod space is practically excluded. The me-

chanism of accident development at a simultaneous failure of

pumps and safety rods or under other predictable conditions

of rapid positive reactivity insertion (for example,

17

avalanche-type release and passing through the core of gaseousfission fragments, penetration of large gas bubbles, modera-ting materials, at the reactor inlet, etc.) is still ratherobscure.

Therefore, one of the main problems of the second trendin the safety area is verification of the mechanism and pro-babilities of proceeding of local accidents, accidents withpipe ruptures and of hypothetical accidents with rapid posi-tive reactivity insertion.

A current approach to nuclear power plant safety evalua-tion by means of the postulated maximum credible accident ana-lysis without taking into account its probability of initia-tion will evidently exhaust itself with transition to a stageof priority development of power engineering based on nuclearpower plants. At such an approach to nuclear power plant sa-fety evaluation it is assumed that the safety and localizingsystems provided to prevent hazardous consequences will ope-rate (for otherwise fuel would melt, burn through all the bar-riers and go into ground). In reality, there is always cer-tain probability of failure of these systems, and it may berather high since it is difficult to pose well-founded requi-rements on their reliability characteristics at the approachconsidered. Therefore, another important problem of the se-cond trend in development of the fast-reactor safety area isaccumulation and processing of statistical data required anddevelopment of quantitative methods for nuclear power plantevaluation. Main difficulties in solving this problem are in-sufficiently precise knowledge of reliability characteristicsof nuclear power plant equipment, instruments and systems andlack of any data on probability of realization of one oranother of maximum accident development schemes. In this con-nection, for a provisional use, a method which combines posi-tive qualities of postulated and quantitative approaches whenevaluating fast reactor safety may be recommended. The essenceof the method is that, along with the general safety criterion(limitation of radiation effect upon personnel or populationby some specified levels under any normal or emergency condi-tions at a nuclear power plant), some intermediate criteria

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are introduced. By an intermediate safety criterion is under-stood the one which refers to such limiting stages of emer-gency processes that can be sufficiently well described onthe basis of already known data and methods and the fulfilmentof which allows to prevent dangerous development of an acci-dent long before a stage of radioactive discharge occurence.

On the basis of the present state of knowledge, as oneof such limiting stages of emergency processes in the coremay be taken, e.g., a state characterized by loss of tightnessand by partial fuel melting without loss of fuel-element shape.Then an allowable probability of reaching any of the limitingstates is equated to an ultimate permissible probability (es-tablished by developed standards) of an accident with the ma-ximum radioactivity discharge. This allows to formulate quan-titative requirements to reliability of the normal operationsystems, to safety and localization systems and devices, todetermine the necessity of duplication or introduction of ad-ditional safety systems.

COHCLUSIOH

Accumulated experience of nuclear power plants operation,including that with fast reactors, shows that at normal opera-tion with taking into account of possible operating failuresand accidents they ensure a considerably lower level of riskfor personnel and population as compared with other fieldsof industry or natural disasters. Therefore, some fears con-nected with extensive development of nuclear power engineeringare dictated not by real risk but by that which, in principle,can be realized only at very low probability accidents. Fromthis point of view sodium-cooled fast reactors have certainadvantages. Indeed, probability of an ultimate accident witha piping rupture in pressurized reactors should be conside-rably higher. Here, realization of a single (and difficult-todetect) event, e.g., missing of a manufacturing defect, issufficient for arising of a very dangerous initial stage ofan accident. In the primary circuit of a fast reactor at

19

practically atmospheric pressure, a rupture in piping is con-

siderably less probable, and at the integral-type arrangement

it is quite safe. All other chains of ultimate accident deve-

lopment in a fast reactor are connected with the necessity of

realizing simultaneously a number of events in systems and de-

vices being permanently monitored (power supply systems, sa-

fety systems, etc.). The above considerations, along with such

important sodium properties as large margin up to the boiling

temperature and its capacity to transfer heat,practically

without inertia, out of the reactor to structural elements

and heat-exchanging devices under natural circulation condi-

tions, allow to hope that the level of risk for future serial

nuclear power plants with fast reactors will be at least no

higher than that for nuclear power plants with thermal reac-

tors.

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