The National Laboratories are run under contract by private bodies: Argonne National Laboratory (ANL), run by University of Chicago; Los

Alamos Scientific Laboratory (LASL), run by University of California; Oak Ridge National Laboratory (ORNL), run by Union Carbide; Brook — haven National Laboratory; and Savannah River National Laboratory.

They are almost wholly financed by the AEC and indirectly organized by the AEC. Their main function is to engage in basic research and code de­velopment, using facilities which are too large and expensive for private industry to operate.

Argonne National Laboratory and its facilities in the National Reactor Testing Station (NRTS), Idaho, forms the major force in fast reactor development. A good deal of experimental work including EBR-II and the TREAT facility comes within the ANL orbit. The ANL code center acts as a clearing house for ANL and industry codes developed under AEC contract, available for the good of fast reactor development.

Oak Ridge National Laboratory, besides having an experimental fast reactor program devoted to both sodium facilities and gas-cooled work, runs the National Safety Information Center (NSIC). The NSIC distributes abstracts on the user’s selected safety topic to the user on a regular basis. The service, operated as a totally computerized system, is free to any

Fig. 6.3. National Safety Information Center abstract card.

engineer engaged in safety work. It is very comprehensive. Figure 6.3 shows an example of an abstract card produced by the NSIC and Fig. 6.4 shows an example of a page from an NSIC-produced listing of articles which have appeared in the journal Nuclear Safety, also issued by ORNL [see general references, Chapter 1].

08-5-2-455 CABRI — A TEST REACTOR FOR SAFETY STUDIES MILLOT, J. P.

CADARACHE NUCLEAR RESEARCH CENTER, FRANCE

THE SAFETY OF SWIMMING-POOL REACTORS HAS BEEN THOROUGHLY STUD IEO IN FRANCE. THE CABRI REACTOR, DESIGNED FOR THAT PURPOSE, WAS INITIALLY USED TO INVESTIGATE REACTIVITY ACCIDENTS. LOSS-OF-COOLANT-FLOW ACCIDENTS WERE ALSO STUDIED,

AND THIS WORK IS BEING EXTENDED WITH A NEW FACILITY THAT BECAME OPERATIONAL THIS YEAR, CABRI PUISSANCE. A PROGRAM OF FUEL TESTING FOR BOTH WATER AND FAST REACTORS, INVOLVING CABRI AS A CAPSULE-DRIVER CORE, IS BEING ESTABLISHED.

08-5-2-461 RADIATION DAMAGE TO PRESSURE VESSEL STEELS WECHSLER, M. S.

OAK RIDGE NATIONAL LABORATORY, OAK RIDGEt TENNESSEE

THE EFFECTS OF NEUTRON IRRADIATION ON THE PROPERTIES OF STRUCTURAL MATERIALS ARE OF SIGNIFICANT INTEREST TO THE REACTOR INDUSTRY BECAUSE OF THEIR POSSIBLE RELATION TO THE INTEGRITY OF REACTOR PRESSURE VESSELS. IRRADIATION EFFECTS ON PRESSURE-VESSEL STEELS ARE CONSIDERED. THE VARIABLES THAT AFFECT THE NIL CONDUCTIVITY TRANSITION TEMPERATURE ARE SEPARATED INTO TWO CATEGORIES — MATERIALS VARIABLES AND RADIATION VARIABLES. AMONG THE MATERIALS VARIABLES CONSIDERED ARE CHEMICAL IMPURITIES (SUCH AS BORON,. CARBON, AND NITROGEN), GRAIN SIZEt AND METALLURGICAL STRUCTURE ASSOCIATED WITH WELDING. THE RADIATION VARIABLES INCLUDE DOSE, DOSE RATE, NEUTRON SPECTRA, AND IRRADIATION TEMPERATURE. OTHER AREAS REVIEWED INCLUDE POSSIBLE APPLICATIONS (AND LIMITATIONS) OF FRACTURE MECHANICS TO THE PROBLEM OF RAOIATION EMBRITTLEMENT IN PRESSURE-VESSEL STEELS AND THE EFFECTS OF IRRADIATION ON THE FATIGUE STRENGTH OF STEELS.

08-5-3-470 RELIABILITY ANALYSIS OF ENGINEERED SAFEGUARDS GARRICK, B. J. ♦ GEKLER, W. C.

HOLMES AND NARVER, INC., LOS ANGELES, CALIFORNIA THE REQUIREMENTS FOR ASSESSING RELIABILITY IN EN SAFEGUARDS ARE REVIEWED, AN EXAMPLE OF RELIABILI A TYPICAL EMERGENCY CORE-COOLING SYSTEM IS GIVEN PRESENTED FOR USE OF RELIABILITY TECHNIQUES IN N SAFETY ANALYSIS. RELIABILITY CAN NOW BE USEO IN NUCLEAR SAFETY, BUT, BEFORE IT CAN BE A TRULY EF CERTAIN ADDITIONAL REQUIREMENTS MUST BE MET. REL PROBABILISTIC MODELS MUST BE DEVELOPED, AND REAL STATISTICAL DATA ARE NEEDED. TECHNIQUES EXIST FO RELEVANT MODELS. STATISTICAL DATA NOW DERIVE LAR OPERATING EXPERIENCE. THESE DATA, ALTHOUGH OF SO IMPROVED TO PERMIT CONFIDENCE IN MATHMATICAL EVA RELIABILITY. THE IMPORTANCE OF RELIABILITY LIES NUMBERS GENERATED AS IT DOES IN REQUIRING SYSTEM OF SAFETY ANALYSIS AND TESTING AND ACCIDENT-PREV PROGRAMS.

08-5-3-479 RELIABILITY ESTIMATES AND REACTOR SAFETY SYSTEM OPERATION SCOTT, R. L.

OAK RIDGE NATIONAL LABORATORY, OAK RIDGEt TENNESSEE

TWO PAPERS PRESENTED AT THE WINTER 1966 MEETING OF THE AMERICAN NUCLEAR SOCIETY ARE SUMMARIZED. ONE PAPER DISCUSSES DATA OBTAINED FROM FIVE POWER-REACTOR OPERATING AGENCIES IN ORDER TO MAKE RELIABILITY ESTIMATES. SINCE THE DATA WERE OBTAINEO FOR PURPOSES OTHER THAN RELIABILITY PREDICTIONt IT HAS LIMITED APPLICABILITY. RELIABILITY ESTIMATES ARE PRESENTED FOR EMERGENCY CORE-COOLING SYSTEMSt EMERGENCY POWER SYSTEMSt AND SECONDARY SHUTDOWN SYSTEMS. THE OTHER PAPER DISCUSSES REACTOR-SAFETY SYSTEM OPERATION DATA OBTAINED FROM THE SAME FIVE POWER-GENERATING PLANTS. THE COMPOSITE DATA INDICATE THAT REAL AND SPURIOUS SCRAM RATES ARE EACH BETWEEN FIVE AND SIX PER YEAR. MOST OF THESE SCRAMS OCCURRED AT LOW REACTOR POWER AND PROOUCED ONLY ONE OR TWO POWER-GENE RAT I ON OUTAGES PER YEAR.

Fig. 6.4. Page of National Safety Information Center listing of articles which have appeared in Nuclear Safety.

References and abstracts are located by certain key words. Thus, by specifying a number of key words such as reactor, fast, melt-down, accident — analysis, the user is able to specify a very well defined topic of his interest on which he would like information.

In addition, ORNL produces survey documents on selected technical topics such as missile protection, nuclear reactors and earthquakes, etc. This service is also an AEC-financed information service.

Equation (1.3) is still space and time dependent even though the energy dependence can be averaged out by a judicious choice of the diffusion coefficient D. The main space dependent term is eliminated if the flux shape is assumed to remain constant as far as leakage is concerned during the kinetic calculation. The critical distribution derived from Fermi age theory

(3) may be assumed:

ГЦ = — ВЦ (1.9)

where В2 is the geometric buckling.

Then, with the following mathematical transformations

L2 = 2>/Га, /„ = 1/t;Г„ l* = /„/(1 + L2B% фі=р ехр(-£2т)Сі//0Га,

and assuming that the Fermi ages r and rD are equal and that the average neutron velocity is constant in time, we can further simplify Eq. (1.3).

In graphite-moderated systems typically r = 350 cm2 and rD = 333 cm2 so that the above assumption is reasonable. In water-moderated systems where r = 25 cm2 and rD = 14 cm2 the assumption is not so good. How­ever in the fast reactor r and rD are both very large and the assumption is excellent. Thus Eqs. (1.3) and (1.4) reduce to

дф/dt = [(<5& — /%efr) ф/П + £ Ліфі

,_1 (1.10)

дфі/dt = №<*Ф1П — Чі, і = 1,2,…, n v ‘

These equations are time dependent and ф may represent the total neutron flux in the system. In a fast system, which is relatively small, the approxima­tion is good.

This is the most usually reported version of the fault trees.

1.6.2.1 Logic Symbols

The shorthand symbols which are used to define a fault tree are very few. They consist of a set of logic symbols which interconnect an undesirable event with its causes and at the same time define the relationship between the event and its causal events. As we shall see they are Boolean operators.

Figure 1.23 shows the three operators:

(a) AND gate: This describes the connection by which an event takes place only if all the input events occur.

Basic component fault which needs no further development

EVENT Fault event not developed to its causes

I Fault event of probability one

xransfer symbol

Fig. 1.23. Fault tree symbols.

(b) OR gate: This describes the connection by which an event takes place even if only one of the input events occurs.

(c) INHIBIT gate: This is a causal relationship which states that the output event is caused by the input event if the indicated condition is satisfied. It is a conditional connection.

In addition a set of symbols defines events in terms of their state of analysis within the fault tree. Figure 1.23 shows the fault event which is expanded from its causes to its consequences, the fault event which, as a basic component fault, requires no further analysis, and the fault event which, although it could be developed further, is not.

Two special symbols define a transfer of one part of the tree to another branch and an event which has unity probability of occurring (that is, a certainty).

These 8 symbols form the basic blocks from which a fault tree is construc­ted. All that remains is to connect the different events in a particular system.

This method does exactly what it states, it locates the poles by solving the characteristic equation from the closed-loop transfer function:

1 + G(s)H(s) = 0 (2.12)

In a reactor system, these poles will be functions of core parameters, the power level, and plant time constants.

Depending on the degree of s in G(s)H(s), various criteria can be esta­blished for defining the stability of the system. These are algebraic condi­tions on the coefficients of Eq. (2.12) to ensure that there are no poles in the right half of the complex s plane.

a. Routh-Hurwitz method. This method (7) starts from the closed-loop characteristic [Eq. (2.12)], written as

ansn + an_i5n_1 + • • • + a0 = 0 (2.13)

Then an array of n + 1 rows is prepared as follows:

(2.14)

Ci = (Vn-з — I bi, etc.

The number of roots is the number of sign changes in the first column of this array (2.14). Thus the stability criterion is that there should be no changes of sign in this column and thus no poles in Eq. (2.11).

b. Root-Locus method. This method also ensures no poles in the right half of the plane by drawing a locus of values of s which satisfies the char­acteristic Eq. (2.12). The locus is drawn by a graphical method and then the stability criterion is stated in terms of the points at which the locus passes into the right half-plane. A knowledge of G(s)H(s) is required as well as its zeros and poles and the locus is drawn from these points such that a value of s on the locus satisfies the two equations

I G(s)H(s) I = 1 arg G(s)H(s) = nn

which is Eq. (2.12) in its gain and phase components. The rules for drawing the root-locus and the statement of the criteria for stability are summarized in Table 2.2. The subject is treated in excellent fashion by Weaver (7).

c. Solution of the characteristic equation. If the system is simple, then one final method is available. The characteristic equation could be solved for its roots [that is, for the poles of Eq. (2.11)], given all the values of the relevant coefficients, which would include heat-transfer coefficients, feed­back coefficients, and time constants. In even a simple system, this method can be, at best, time-consuming; in a more complex system it is generally impossible.

Some reactor conditions require a reduction in power. This can be ac­complished safely through the scram function but it may be not necessary or desirable to scram in every case. A series of partial power reductions is sometimes more appropriate. In sequence: (a) warning, (b) setback in power, (c) controlled shut-down, and (d) scram. Such a scheme would avoid thermal shocks to the reactor components in many instances.

Table 3.9 shows the standard earthquake scale (11,12). Because of the possibility of structural movement bringing about core reactivity changes, the reactor may require power reductions when a serious earthquake occurs. Generally a scram is initiated at a modified Mercalli intensity of V. However, it is more consistent with modern operating practice and modern knowl­edge of seismology to provide a power setback system on the following scheme (11).

At a modified Mercalli intensity of V a warning is provided, at an inten­sity of VII a controlled shut-down would be instituted and then at an inten­sity of VIII a scram would be initiated.

This philosophy can be applied to many secondary warning indications in the plant without sacrificing any of the overall safety of the system. Usually, however, the warning and setback functions are part of the control system rather than the protective system, and it is necessary to coordinate the two systems most carefully to ensure overall safety under this scheme.

3.4.3 Other Safety Features

It is appropriate to discuss safety features in relation to the accidents which they were designed to mitigate.

While the failed subassembly is being subjected to the sequence of events just described, the adjacent assembly can be expected to respond to two effects in addition to that of power changes caused by feedbacks from the failed assembly. First, it can be expected to feel some effect of the high pressures produced by successive voiding explosions in the failed sub­assembly, and finally it may be subjected to high heat fluxes due to failed and molten fuel against the assembly duct.

Damage to body tissue can be conveniently divided into that produced by direct and indirect action: the first being the case in which a particle of biological interest is struck by either the electron or nucleus of the ionized tissue hydrogen atom, and the other is the case in which the fragments of the ionized tissue may drift over to damage the more critical biological

+ The International Commission on Radiological Protection now prefers the nomen­clature Quality Factor QF for RBE when applied to protection rather than radiobiology.

molecules. Molecules which are of special interest are the DNA of the cell nucleus because damage to this molecule could have genetic consequences.

Figure 5.1 shows diagrammatically the linear energy transfer process in tissue and differentiates between the direct and indirect damage modes. Approximately, the LET values may be shown to correspond to the RBE values as shown in Table 5.2 (2).

Biological effects of body tissue damage can be separated into genetic effects, which may be transmitted to progeny, and somatic effects, which are bodily effects to the recipient itself.

The ultimate objective of any safety evaluation is to be able to specify the eventual and safe condition of the plant following any malfunction. Thus in a core disruptive accident the ultimate objective is to be able to say where the damage debris comes to rest and how this will be contained and cooled.

Thus, taking the accident described in Table 5.10 as an example, the pre­sent need is to show what damage is done by the energy release of the postulated 500 MW-sec work energy. (Section 4.3.4 gives the meaning of this work energy.)

Salt water cooling has already been used for the large graphite-moderated thermal reactors of Britain which are situated mainly on coastal sites. Good condensing temperatures are obtained but the corrosion problems make the use of salt water more expensive than fresh water.

It has been suggested that the very slight temperature rise of the sea which results could be put to good use in keeping northern channels free from ice, in lengthening the lobster season in Maine, in allowing fish herding in Long Island Sound, or even in increasing the numbers of tropical fish in the south. Additional uses include warming recreational waters (10). However, adverse effects have been suggested that change existing fish populations. Thus limits are placed on temperature changes by the Federal Water Pollution Control Administration. In coastal or estuarine waters, the discharge of heated waste should raise the maximum daily temperature on a monthly mean basis no more than 4°F during fall, winter, and spring, and no more than 1.5°F during the summer.

Once-through cooling from rivers is the most economical cooling method and was the usual method applied in the past. For large plants, depending on the size of the river, temperature changes could be fairly large. The Vermont Yankee based on the Connecticut river might have raised the temperature at the river discharge by about 20°F (10). Such a change of river temperature would change the present ecological balance and result in a new balance with different fish and different plant groups. Such a change would most likely be an undesirable one although research is not complete on the exact effect that would be obtained (9a), and the undesirability has not yet been demonstrated. The limitations now set on rises in river tempe­ratures are set by each state and they range below 10°F depending on the value of the temperature at any time (11).

Fresh water circulated through a power plant’s condensers from cooling ponds or lakes with subsequent ejection to the atmosphere by radiation and convection is attractive, since the water is used over and over again in a self-contained system. The method appears to be economical.

The rejection of heat for other applications such as heating apartment buildings or even whole towns (10) has been practiced in local areas, but the method cannot be considered an overall solution for the rejection of excess heat, especially since many power plants are situated in relatively remote areas.

There are different types of cooling towers. The main division is between the wet evaporative kind, in which some of the cooling water is evaporated to remove heat, and the dry kind, in which a tubed radiator transfers heat. Both kinds can employ natural draft cooling or induced draft cooling. They are designed broadly to lower the temperature of the cooling water by about 15-17°F. Which kind is used in a particular application depends on the natural climate and its effect on the possibility of natural draft, the surround­ing countryside, and the effect of physically obtrusive structures, and whether or not it is necessary to avoid the use of noisy blowers for induced draft versions.

The advantage of cooling towers is mainly that they are a method of cooling recirculating water by direct contact with the environment rather than through the medium of a river. An additional advantage is that although they are very tall, cooling towers occupy relatively little site area when compared to cooling ponds which may occupy hundreds of acres of land area.

In siting any reactor plant, including a fast reactor installation, a com­prehensive assessment of the impact of cooling upon the environment is now a necessity and the AEC, as the licensing authority, now has the power to refer applications for plant licenses to agencies having legal jurisdiction in environmental matters and to require that the licensee observe certain applicable environmental limits. This extension of the AEC powers was enacted in 1969 (12).