**8.1. Possibility of neutron loss decrease**

Obtaining of high fission materials reproduction is possible only in reactors, which have the minimal neutron loss in construction materials and leakage. This loss must be about 1-2% by the preliminary analysis of previous materials. It is not possible to realize on practice all of the discussed fuel cycles.

Let us consider in this chapter could be achieved good results with bigger neutron loss level and what must be considered as good results.

CANDU (and its many versions) with heavy water as coolant and moderator, with zirconium shells and channel walls is the nearest reactor to high fission materials reproduction reactors. Let's look at the characteristics of this reactor and its possible modifications, which are directed to neutron loss decrease. Fuel assembly of CANDU reactor with 37 fuel element (on left) and modification of this assembly with replacement of 7 central fuel element by beryllium block are presented in the figure 7.

Characteristics of initial reactor and its 6 modifications are presented at the table 3:


6. Reactor by the i.5, with fuel assembly shown at the figure 19 (on the right) with beryllium insertion.

196 Nuclear Power – Practical Aspects

pike lowering.

**7.2. Fuel is natural uranium** 

of such campaign are high enough.

the discussed fuel cycles.

**8.1. Possibility of neutron loss decrease** 

and what must be considered as good results.

assembly casing zirconium and tin.

bismuth (20% of volume).

Let's examine how neutron loss in control elements used for generation of 235U in reactor with initial fuel 235U + 238U influence on campaign characteristics (figures 10, 12 - 14). Characteristics of detailed campaign with 1014 sm-2sec-1 neutron flux, 233U production and 1,7

Maximum fuel burn-up (5.27%) is reached in this campaign in comparison with previous campaigns; duration of detailed campaign is 34000 hours. Reactivity change during work is close to optimum results (K-1 from 1 % up to 3.2 %). Reactor power is stable. Reactor power change in constant neutron flux is not exceeding 2.5% from the average. 233U generation increase in the campaign with this fuel type is possible at the expense of operating reactivity

Enough good indexes by this technology can be received into the reactor with 5.2% neutron loss. Description of such campaign is presented in string 24 of Attachment Table 1. Indexes

Obtaining of high fission materials reproduction is possible only in reactors, which have the minimal neutron loss in construction materials and leakage. This loss must be about 1-2% by the preliminary analysis of previous materials. It is not possible to realize on practice all of

Let us consider in this chapter could be achieved good results with bigger neutron loss level

CANDU (and its many versions) with heavy water as coolant and moderator, with zirconium shells and channel walls is the nearest reactor to high fission materials reproduction reactors. Let's look at the characteristics of this reactor and its possible modifications, which are directed to neutron loss decrease. Fuel assembly of CANDU reactor with 37 fuel element (on left) and modification of this assembly with replacement of

1. Reactor with fuel assembly at the figure 19 (on the left) with standard reflector,

2. Reactor by the i.1, with enriched by isotopes 90Zr and 120Sn in fuel rod shells and fuel

5. Reactor by the i.4, with fuel rods made of metallic uranium (80% of volume) and

7 central fuel element by beryllium block are presented in the figure 7.

3. Reactor by the i.2, with bigger thickness of heavy-water reflector.

standard fuel on the basis of natural uranium dioxide;

Characteristics of initial reactor and its 6 modifications are presented at the table 3:

4. Reactor by the i.2, with addition of graphite reflector to heavy-water reflector.

**8. Characteristics of reactors with high fission materials reproduction** 

% neutron loss are shown at the string 15 of Attachment Table 1.



**Table 3.** Neutron loss (string 1-11, %), multiplication factor and reproduction coefficient (string 12, 13, relative units) and fuel types and reflector types in variants of CANDU reactor models.

**Figure 7.** Fuel assembly of CANDU reactor (on the left) and its modification with beryllium insertion (on the right). 1 – fuel assembly casing; 2 – fuel element shell; 3 – fuel rods; 4 – coolant; 5 – beryllium insertion.

Analysis of calculation results allows saying:


Thermal Reactors with High Reproduction of Fission Materials 199

enough high absorption cross-section) is possible to delete by using tin enriched by 120Sn

It is possible, that alloy of tin with bismuth and lead also will form the tin stannide on the surface of zirconium, which prevents the interaction of zirconium with bismuth and lead.

We should specify requests to whole of possible varieties of core composite elements, which

In this case the statement is follows: space between particles must create stable conditions for separate particles location under fuel element cover when increasing volume in specified



It should be marked that important factor in such construction of fuel element, which decrease the metallic fuel elements swelling, is their small thickness. The small thickness increases the migration of fission gaseous splinters over the external surface with decreasing

By the estimations, forms changing of metallic fuel rods under the execution of given claims and initial content of liquid-metal filler in 20% by the volume is not leading to trespassing of


isotope. 120Sn has the best properties among tin isotopes with even atomic weight.

Rigorous research in this direction is necessary.

1 – fuel element cover; 2 and 3 – fuel elements; 4 – filler.

**Figure 8.** Variants of fuel element construction with composite metallic core.

limits without appearing of additional effort between particles;

between them for realization of increasing of particle sizes;

This claim dissects for some independent claims:

surface of particles;

of inner mechanical stresses.

negligible.

must have high workability of fuel element with maximal level of burn-up.


1.7 %, 2.8 % and 5.2 % neutron loss are used in Table 1 as results of CANDU reactor calculation.
