Simulation of pulsed accidental energy release in a reactor core

All-Union Scientific-Research Institute of Engineering Physics. Translated from Atomnaya énergiya, Vol. 77, No. 2, pp. 112–118, August, 1994.     

Testing a system prototype for out-of-reactor monitoring of reactor power and its distribution in the core volume

Conclusion Prototype tests were performed for a representative set of reactor conditions: the power distribution was performed by a wide range of changes in the position of three groups of control and protection system regulatory devices, the A0 values were varied in the range from –0.40 to 0.16, and the coefficient of non-uniformity k_v in the range from 1.7 to 2.6. The results showed the high accuracy and effectiveness of out-of-reactor monitoring of the power and its distribution throughout the core volume.For an LWR, the out-of-reactor monitoring system assemblies can be placed in radiation shielding channels; for this three assemblies, each having three detectors, is sufficient. Chambers having an energy range of standard neutron flux monitoring equipment can be used as detectors.Determining the thermal power and the coefficients of nonuniformity of its distribution in the core does not begin to exhaust the possibilities for out-of-reactor monitoring. Algorithms already exist or are being developed which would allow increased accuracy in the monitoring of power and its distribution, localization of the region or fuel element with the greatest energy loading, detection of stuck control rods and nonfunctional thermal monitoring sensors, and diagnosis of fluctuations and position shifts of internal reactor vessel structures. Thus a reliable, cheap, rapdily responding core condition diagnostics systems can be constructed on the base of out-of-reactor detectors.Translated from Atomnaya Énergiya, Vol. 64, No. 3, pp. 174–180, March, 1988.     

The study of aeroball system for measuring 3D neutron flux distribution in reactor core

Aeroball system is attractive in several aspects because it can easily transport the map of neutron flux distribution to be measured from incore to outside of a reactor vessel.However,before the aeroball system is put to practical use in the heating reactor.there are four topics that have to be further studied.They are the stability of the activated positions,enhancement of signal/noise(S/N)ratio,distributed control and data-acquisition system and on-lin nbeutron flux distribution reconstruction.Besides describing the rasons for them,this paper gives out the theory,concept and solution about the first two topics and it is helptul to give the possibility to enhance the reactor-power.     

Testing of compton-emission type detectors of neutrons and of triaxial fission chambers in a pulsed reactor

Translated from Atomnaya Énergiya, Vol. 65, No. 1, pp. 59–60, July, 1988     

Applicability evaluation to a MOX fueled fast breeder reactor for a self-consistent nuclear energy system

The potential for a MOX fueled fast breeder reactor (FBR) is evaluated with regard to its ability to transmute radioactive nuclides and its safety when incorporated in a self-consistent nuclear energy system (SCNES). The FBR's annual production amounts of selected long-lived fission products (LLFPs), Se-79, Tc-99, Pd-107, I-129, Cs-135 and Sm-151, can be transmuted by using a two layer radial blanket region without a significant impact on core nuclear and safety characteristics. The other LLFPs are confined in the system. The hazard index level of the LLFPs per one ton of spent fuel from the system after 10² years is as small as that of a typical uranium ore. Regarding self-controllability in the system's safety, the proposed FBR core concept has an inherent negative reactivity feedback with a gas expansion module, sodium plenum above the core and burnup reactivity compensation module. So sodium boiling and fuel melting will be avoided in anticipated transient without scram events. Regarding self-terminability, even if the MOX fuel melting should cause a core compaction process, re-criticality of the core can be avoided by a fuel dilution and relocation module.     

A point kernel model for the energy deposited on samples from gamma radiation in a research reactor core

A basic safety requirement for a research reactor is the reliable estimation of the gamma heating of samples irradiated in the reactor core. A three-dimensional numerical code of gamma heating using a point kernel parameterization is developed. The heating due to γ-rays, produced from U235 fission and from (n, γ) reactions with the core materials is considered. The dose build-up due to photons scattering on the core materials as well as the energy absorption build-up in the sample are also included, based on empirical relationships. The developed code (GHRRC: Gamma Heating in Research Reactor Cores) is applied for the Greek Research Reactor (GRR-1) core. The required microscopic cross-sections and the three-dimensional neutron flux are obtained with the neutronics code system XSDRNPM and CITATION. The macroscopic cross-sections of the U235 fission and the (n, γ) reactions in the core materials are determined assuming a homogenized core. Comparisons of the computed gamma heating power deposited on a Fe sample with in-pile and out of pile measurements of the sample temperature show that GHRRC gives reasonable estimations. GHRRC may easily be handled even by poorly experienced users.     

Methods for the seismic verification of a fast reactor core

This paper presents the main features of the methods applied by the Fast Reactor Department of the Italian Energy Authority (ENEA) to the seismic verification of the core of the Italian PEC fast reactor test facility, which is in advanced construction at the Brasimone site. It also points out the problems which in general remain open in the fast reactor core seismic analysis. The aim is to contribute towards a generally acceptable philosophy on core seismic verification techniques and suggest items of possible co-operation for future developments.     

Using a measuring fuel assembly in the local thermal regulation system of an IRT reactor

Translated from Atomnaya Énergiya, Vol. 70, No. 5, pp. 347–349, May, 1991.