There is a sequence of events in the licensing process that must take place before a license is granted.

(a) The application for the license to build is submitted by the utility as applicant, using technical information supplied by the industrial designer. This technical information includes a preliminary safety analysis report, the PSAR.

(b) The DRL receives the application and gives a copy to the ACRS. It puts a copy on view in the public room to give objectors access to all the information. It considers the application and the evaluation report and then submits its report, also to the ACRS.

(c) The ACRS considers the application, the safety evaluation, and the DRL report. It then reports in turn to the AEC commissioners who make the official ruling. However, the ACRS by this time has made a recommen­dation which carries considerable weight. The AEC Commissioners are

TABLE 6.1

AEC Question 14.1 on Indian Point 2

unlikely to oppose this recommendation. During the ACRS review there is considerable communication between the committee and the utility — industrial designer partnership. The ACRS is likely to ask a series of search­ing questions before it is satisfied to make a recommendation to the com­missioners. Table 6.1 details a typical question asked by the AEC, in this case during a PWR licensing application.

(d) The commissioners call a public meeting for public review and objection by groups or individuals if required. The licensing board at this meeting again reports to the commission on objections received.

(e) The commission then decides on the application and issues a con­struction license if they are satisfied that the plant is safe, and that public objections have been met, where these objections are relevant.

(f) After construction of the plant, the utility must again make an appli­cation, this time for an operating license. The procedure to obtain it is identical with steps (a)-(e) and again occupies many months. This second application is accompanied by another safety analysis document, a final one (FSAR), which incorporates all the latest work and includes answers to all the ACRS questions.

Licensing of power reactors [2] [3] [4]

To summarize the procedure (Fig. 6.5):

1. application by utility (industrial vendor acts as an advisor);

2. DRL distribution of application and consideration of PSAR;

3. report by DRL to ACRS;

4. report by ACRS to AEC commissioners;

5. public review meeting;

6. report by licensing board to AEC commissioners;

7. construction license granted or refused.

Then after construction of the plant, there is a repetition of steps 1-6, following which the operating license may be granted or refused or simply delayed until all outstanding queries are settled.

Steps 1-7 occupy approximately eighteen months and there is approx­imately a two-year construction delay before the next application is made. During this time, the safety evaluation will be strengthened and made more specific to the design which will become fixed during the early construction period.

Construction license

3

Fig. 6.6. Licensing time scale for a nuclear power plant.

4

<F Criticality

<F Fu І power operation 6 ■

If q is very large then a>0 is large and greater than the largest Thus:

Q = (/*со<Де<r) + £ Pi so "о = (б — P) keaH* (1.16)

І=1

Now the neutron lifetime is all important and the delayed neutrons have no effect beyond reducing the reactivity q by /3 to (q — /fJ).

1.2.1 Delayed Neutron Data

The decay constants of the delayed neutron groups do not differ markedly among fissionable isotopes, but the yields (l do (4). A comparison of /3 for different fuels is shown in the accompanying tabulation. Thus an

effective /3 must be calculated before reactor kinetics calculations can pro­ceed. The value of /3 depends not only on the fuel in the system but on a large number of other system variables:

(a) The isotopic concentrations in the system and the fission rate in each isotope affect /3. In a fast reactor 238U may have 15% of the fissions.

(b) Delayed neutrons are slower than prompt neutrons and thus their importance with respect to leakage and fission is different (5). Their slow­ness reduces the effective /3 value in the fast core.

(c) The geometry of the system and the blanket fission rate effect /3 (6). In a fast system 10% of the fissions can occur in the blanket.

(d) Neutron hold-up in a reflector is sometimes represented by another delayed neutron group, although it is better represented by a modified neutron lifetime. This effect is particularly important in heavy water thermal systems but does not apply in a fast reactor system.

(e) Variation with burn-up also changes the balance as far as /9 is con­cerned. The plutonium is enhanced; thus the effective /9 decreases in a thermal system that may start with /9 ~ 0.0075, but in fast cores where 239Pu gives way to 241Pu the effective /9 increases. A fast reactor typically has a /9 of about 0.0035.

This is another form of fault tree most commonly used in the analysis of the course of the accident. It proceeds in the direction opposite to the single-failure tree, from cause to a number of consequences rather from a consequence to its various causes.

Starting from the initiating fault, an attempt is made to see where the events may lead, just as the accident might progress. Therefore, conse­quences may be both safe and unsafe.

To illustrate a flow-process tree, consider Fig. 1.28. This follows the

FUEL EJECTION —m FISS PRO ___ Ш ASEOF ION DUCT Л_____

f*| ( CRITICAL| AMOUNW

behavior of the particular system in the face of a reasonably probable fault: the installation of an oversized fuse. If an oversized fuse were installed, then the fuse would be unable to open in the face of excessive power. This is not a problem, unless the motor fails, although the failure is always a potential one while the oversized fuse is in position.

However if the motor does fail shorted, then an overcurrent will result through the system wiring and through the relay contacts. So the tree is further extended into two further possible terminations—one in which there is transient overheating but the wire is safe and the other in which the wire does fail.

Such a" fault tree answers the question: What happens as a consequence of this event? It is the tree that one would follow in any calculation of the ensuing accident. There will be further examples throughout this book; a particularly good one lies in the analysis of fuel failure and its possible extension to further failures.

Figure 1.29 shows this tree in its accident process configuration while Fig. 1.30 shows the single-failure fault tree for the pin-failure propagation. They are equivalent, and they should be referred to again after reading Chapter 4. Because such trees require an intimate knowledge of the proc­esses involved before they can be made unique, there is in fact great variety in how such trees may be laid out.

All the foregoing text has referred to linear stability analysis and any system with nonlinearities would have to be linearized before being treated in this way. [Witness Eq. (2.28), which linearizes the дк P term of Eq. (2.25).] However, many systems are significantly nonlinear.

Fig. 2.37. (a) Nonlinear system: spring displacement for a given force, (b) Nonlinear system: saturated system behavior.

After a core dispersion accident, it would still be necessary to contain the fission gas and any plutonium in the system as far as possible (see also Sections 5.5 and 5.6.2). The following methods might be used: radial blast shielding, shock attenuation volumes, sodium hammer suppressor plate, head hold-down system, missile barriers, splash plates to agglomerate so­dium sprays, aerosol settling devices, filtration, pressure suppression systems, and containment barriers.

This section has so far treated failure arising from a single pin; however, it is possible that due to an overall reactivity accident or an overall flow failure, the whole core might almost simultaneously fail. The characteristics of the widespread failure would be very different. To obtain a widespread failure one must start with a hypothetical accident.

Section 5.4.4 treats suggested design bases and outlines the behavior following a loss of electrical power to the pumps associated with a failure to scram the reactor. The sequence of events in that case is as follows (remembering always that this is the supposed chronology of a hypothetical event):

(a) Widespread voiding due to either sodium boiling or fission gas release or both effects and consequent reactivity feedback.

(b) Prompt criticality resulting in widespread fuel failure and melting. A slight dispersion shuts the system down.

(c) Slumping of the core under gravity into a more reactive position.

(d) A second more violent superprompt criticality.

(e) Violent dispersion of the core.

The calculations leading to estimates of voiding times are the same as those previously described in Section 1.3. However the main difficulty in this analysis enters into the slumping calculation. Just how does molten fuel collapse? The time of collapse is all-important in defining the time scale in

which the slumping-induced reactivity feedback is added. This in turn de­fines the subsequent energy release.

Several fuel slumping models have been suggested in the past. One code, MELT II, allows the user to specify his own mode of fuel collapse for a single representative pin (31).

Another method used in the Fermi hazards report calculated the molten patterns in a homogeneous core following a loss of cooling. These molten regions, shown in Figs. 4.19-4.24, were then assumed to slump under gravity (32a).

In several cases, EBR-II and Fermi, the hypothetical case of a gravita-

tional slumping of the top third of the core onto a previously slumped and compacted bottom two-thirds was used as a design basis (32b).

Another suggested method is shown in Fig. 4.25 in which a representative fuel pin is modeled. The molten fuel distribution is assumed to run down as a collar in the equivalent channel of the pin and to successively grow as

more molten fuel is produced from the transient. The assumptions in this approach are that the core is voided and the void does not reenter while slumping is proceeding; there is no separated dripping of fuel; one rod is representative; the top of slumped collar is level with the top of the un­melted fuel; and there is apparently no hold-up of molten fuel by the surrounding unmolten material. The equations that define the configuration are those for conservation of mass and gravitational fall. The assumptions of this model are at least no worse than those of the others. However, they all suffer from one considerable defect: None of the core collapse models to date treat the combination of core voiding and fuel collapse and the inter­action between the two phenomena, especially inasmuch as sodium liquid reentry is liable to have a compacting effect on the molten fuel. In some cases (23b) this interaction is not present, since the void exists for a long enough time for the fuel to slump while the liquid sodium is out of the core, but generally this may not be true.

Another hypothetical event which can lead mechanically to a core disrup­tive accident is an uncontrolled core reactivity increase. In this case the sequence is as follows:

(a) Power increase leading to fuel melting and eventual melting of the cladding from the high fuel heat fluxes.

(b) Molten fuel and fission gas ejection and fuel slumping into the coolant channel.

(c) Widespread voiding throughout the core (notice that in this case the voiding comes after rather than before the cladding failure).

(d) Widespread fuel slumping in combination with voiding mechanics.

This case is very different from the loss of flow combined with loss of scram, since in the previous case the voiding initiated the cladding failure, whereas here the fuel fails into a subchannel which is still filled with sodium. This latter event is likely to be more chaotic, although it is much less likely because of the design of the reactivity control system and the plant pro­tective system.

A description of the effects of acute radiation sickness and its treatment covers the whole range of somatic effects of radiation; in less severe cases only the earlier symptoms are experienced, in more severe cases the symp­toms are intensified and are usually also associated with external signs of radiation damage.

In the range of 100-500 rem the average effects may be described as follows.

Sickness:

(a) Within a dozen hours or so the patient experiences nausea, vomiting, and fatigue.

(b) After a day or two the patient begins to feel normal for a few weeks, although the blood cells (white and red corpuscles and platelets) are dimin­ishing.

(c) The drop in red and white cells becomes obvious along with a drop in red platelets. There are two consequences: a feeling of weakness due to the anemia, and a proneness to bacterial invasion from other sicknesses as a result of the drop in the protective white corpuscles. The loss of red plate­lets leads to various forms of hemorrhage—from the nose, gums, or even intestines.

(d) Either the patient overcomes the blood deficiency in time or other infections, anemia, or hemorrhage lead to death.

However such sickness can be treated and the blood deficiency can be monitored and overcome in all but the most severe cases.

Treatment:

(a) It is important to first thoroughly wash and decontaminate the patient of all external and orifice radiation contamination.

(b) Then a bacterially clean environment needs to be established before stage (c) above of the sickness, to avoid infection due to external bacteria while the patient’s blood count is low.

(c) It may be necessary to transfuse the constituents of blood (white and red corpuscles and platelets), and then to allow the patient to recover his normal blood-making processes after the critical period when his blood count is at a minimum.

(d) In extreme cases it may also be necessary to transfuse bone marrow which makes its way to the bone to begin its blood-forming function. The new blood-forming cells must come from a close match individual such as a twin, even though the radiation damage has also destroyed the normal rejection mechanism to strange body tissue. The rejection response will be restored as the patient recovers and it is necessary to ensure that the injected bone marrow is not subsequently rejected (5).

It is emphasized that such a sickness as is described here is the worst consequence of a radiation accident. No such accident has ever been asso­ciated with a power reactor of any type, thermal or fast. It is the intention of this book to help to ensure in some small way that such an accident will never occur in connection with a fast power reactor.

An alternative method of assessment of vessel strain is simply to integrate the standard stress-strain curve for the work energy absorbed. This assumes uniform energy absorption, no effect from the internal structure within the vessel, and of course, that the classic stress-strain curve applies. These are the same assumptions as were made in the Proctor method. Such an integra­tion method is’.

3{?(5o-y + <ти)є(Яе2 — /?;2)2
3.53-108

where the units are those of Eq. (5.10). This equation also assumes, as did Proctor’s, that there is a 67% efficiency in transforming blast into impulse to the vessel wall.

A nuclear plant has to conform to strict release limitations set by the AEC in its Code of Federal Regulations outlined in the previous chapters. Moreover, this compliance is not measured against actual releases, as in the case of the fossil-fueled plants but against hypothetically bad conditions described as the design basis for the plant.

Thus Table 6.5 compares the annual doses which are recommended as maxima by the Federal Radiation Council against those that are calculated to occur if 1% of the fuel were leaking and also against those that are actually expected on the basis of present operating experience in PWR plants.

The table shows that the design basis conditions are about a hundredth of the recommended maxima or better at the various exclusion zone bound­aries, while the actual expected values are again a further factor of a hundred lower {8).

The assumptions in this table are that a 45-day hold-up system is used to eliminate all short-lived isotopes and that the gaseous releases are those resulting from pessimistic wind dispersal conditions. The liquid releases here are calculated from the intake of aquatic food as well as drinking water. The fish and mollusks account for about 5% of the total.

If multiple plants in one location are considered, then sample calculations showed that, for pessimistic conditions downstream, a set of three plants resulted in about a 50% increase in dose (8). Combined doses are not ad­ditive, unless the plants are in identical positions in relation to winds and ground water flows.

It is worth noting that the FRC limits apply only at the boundary of each zone, and the doses will be significantly less elsewhere farther away. Figure 6.7 shows that while the design basis release is much lower than the FRC exposure limit at the fence line, as one moves away the dose becomes exponentially less. In fact, even with a complete failure of the fuel in this example, where the FRC exposure limit would be exceeded at the fence line, doses drop below the limitation very rapidly as the distance is increased.

It has been shown in preceding sections that one can predict the power generation level from the reactivity, and that temperatures can be predicted from this power level. It has been noted that the temperatures affect the reactivity through variations in the cross sections which determine the neutron behavior. Thus a complete feedback path exists (Fig. 1.11). It is necessary to know how reactivity depends on temperatures and how this dependence may be affected by the reactor and plant design.