S. P. Belov

Spatial and energy distributions of neutrons in boron carbide

Studies were made of the space and energy distributions of 3- to 15-Mev neutrons in boron carbide. Reactions H/sup 2/(H/sup 2/, n) He/sup 3/ and H/sup 2/ (H/sup 3/,n)He/sup 4/ were use d as sources for 3-Mev and 15-Mev neutrons, respectively. A proportional counter with BF/sub 3/ enriched up to 88%, a fission chamber with U/sub 235/ (enriched up to 75%) and Th/sup 232/, and threshold indicators of P/sup 31/(n,p)Si/sup 31/, Al/sup 27/(n,p)Mg/sup 27/ , 3/ (n,2n) Cu/sup 62/, and In/sup 115/(n, gamma ) I/sup 116m/ were used as neutron detectors. The results are plotted and tabulated. The ratio of the U/sup 235/ fission cross section sigma /sup 235/ (E/sub eff/) to sigma /sub B/) in (n, alpha ) reactions on B/sup 10/-boron carbide was determined. The ratio of 3-Mev neutron weakening in the region of fixed spectra was 0.97 plus or minus 0.03. The obtained relation was used for finding the effective energy E/sub eff/ = 120 plus or minus 10 kev. (R.V.J.)

Calculational-experimental research models for a fast reactor with a heterogeneous core

The physical characteristics of heterogeneous metallic oxide cores were experimentally studied by physical tests of the critical assemblies BFS-46 and BFS-46AZ, which simulate a reactor of the BN-1600 type, into the core of which a fuel assembly with metallic uranium is inserted. A calculational model for the critical assemblies being investigated, showing the zones and their dimensions, is presented. The critical assembly BFS-46AZ is a modification of the basic critical assembly BFS-46 which adds plutonium to the IBZ to simulate its accumulation during reactor operation. The BFS-46 and BFS-46AZ assemblies have identical dimensions for the IBZ and LEZ, and have different HEZ dimensions, necessary to ensure the criticality of each assembly. Plutonium with a /sup 240/Pu content equal to 3.8% is used in the LEZ. The critically parameters are calculated using one-dimensional and two-dimensional models in a 26-group diffusion approximation based on the BNAP-78 system of group constants.

BFS critical assembly studies of the fast reactor safety

ConclusionsThe results of investigations of physical problems in the fast-reactor safety conducted in recent years on the BFS CA make it possible to draw the following conclusions:the calculation gives a more positive SVRE value, the absolute shift for central areas being about the same in the CA both with a mixed and uranium charge for equal volumes of the core; relative error of the SVRE calculation for the cores with mixed charges is about 20%;the measurements of the SVRE with the use of the BN-600 standard FA made it possible to estimate the accuracy of simulating this RE on the BFS CA to be about 20%;the RE due to the entering of hydrogenous substances into the fast-reactor core is estimated to be 15–20%;the RE due to the failure of a single FA is estimated by standard procedures (diffusion approximation, first-order perturbation theory) with an accuracy of about 20%.

Theoretical and experimental investigation of sodium void effect of reactivity

In analysis of the safety of fast reactors under design it is absolutely necessary to consider emergency situations due to partial or complete loss of coolant from the reactor core. For power reactors with a high unit power this problem becomes particularly acute since in this case the loss of sodium from the reactor core may result in a large positive reactivity effect. An emergency situation with escape of coolant is considered both for complete drainage of the reactor and within the framework of a local failure when the coolant escapes from one fuel assembly or molecule. Studies on the dynamics of reactivity variations during local failures are of particular interest since they provide the necessary information for diagnosis of an operating reactor. The methods used to calculate the sodium-void effect of reactivity were verified on uranium critical assemblies.

Experimental shield studies on the riz test-bed☆

Abstract A description is given of a test-bed with a zero-power reactor intended for studying the processes which take place in the part of the shield immediately adjoining the reactor core. A knowledge of the processes occurring in this part of the shield (shaping of the neutron spectrum and formation of powerful radiation) enables the dimensions and weight of the whole reactor shield to be correctly calculated. The paper gives the results of measuring neutron spectra and of studying shielding materials (iron, nickel and borated nickel) on the test-bed.

γ-Ray yield from radiative capture in iron

A technique is described for measuring the yield of capture γ radiation from iron by the use of a single crystal scintillation γ-spectrometer. Results are reported for the yield of capture γ radiation from iron, normalized to the number of neutrons either entering shielding or escaping from it. Measurements were made for the neutron spectrum from a Po-Be source. The dependence of capture y-radiation yield on shielding thickness is given. The effect of boron-containing impurities on the yield of capture γ radiation was investigated.