Shutdown Reactors Research Articles

LASSO reconstruction scheme for radioactive source distributions inside reactor building rooms with spectral information and multi-radionuclide contaminated situations

Clarification of radioactive source distributions is one of the most important steps in initial decommissioning of not only normally shutdown reactors but also damaged ones by accidents like Fukushima Daiichi Nuclear Power Plants (FDNPP). Generally, since radioactive hot spots are restricted into specific areas in normal operating conditions, the clarification scheme can be mapped onto the inverse estimation in sparse source distributions. On the other hand, the fact that radioactive hot spots are largely spread in unknown manner as seen in FDNPP motivates to construct an inversion scheme in non-sparse source conditions. Thus, a reconstruction scheme applicable to both sparse and non-sparse radioactive distributions is highly in demand. In addition, a variety of radionuclides is produced in reactors. Thus, we also need a scheme to distinguish each source distribution in mixed multi radionuclides. In this paper, we confirm that the inverse estimation scheme using Least Absolute Shrinkage and Selection Operator (LASSO) method with spectral information commonly shows excellent performance in the above all situations. The proposed LASSO scheme with the spectral information enables to reduce the number of measurement points in sparse conditions, while information proliferation by sensing the spectrum makes it possible to directly reconstruct source distribution as almost solvable problems in non-sparse ones. Moreover, the LASSO scheme allows to reconstruct the source distribution of each potential radionuclide in multi-radionuclide coexisting situations. Consequently, we confirm that the LASSO scheme to reconstruct radioactive sources is promising for the future nuclear decommissioning projects widely from normally shutdown reactors to damaged ones like FDNPP.

On Measurement of Subcriticality of Reactors with Statistical Methods

The possibility of measurement of subcriticality of reactors of nuclear power plants by statistical methods in the performance of requirements of Regulations NP-082-07 is discussed. The statistical methods, in particular, the Feynman method, make it possible to measure the subcriticality of shutdown reactors and at initial start-ups within the required range of 0.01–0.02 at the level of neutron detector counting of about one pulse per second. The Feynman method was perfected at the critical assembly of the high-temperature gas-cooled nuclear reactor PROTEUS at the Paul Scherrer Institute (Switzerland). The measurement results of subcriticality are presented. The conditions which should be ensured to obtain an acceptable result of experiment are formulated.

Transportation operations model analysis for removal of stranded fuel from shutdown reactors

The transportation operations model was used to identify options for removing stranded fuel currently in dry storage at nine shutdown reactor sites to a hypothetical consolidated storage facility. The logistical variables included the campaign duration, fuel selection priority, consist size and location of the consolidated storage and maintenance facilities. The major factors affecting the logistics of fuel removal were identified. Recommendations for optimal strategies for the transport of stranded fuel from shutdown sites are made.

Dismantling design for the loop rooms on the MR reactor

The recently completed international co-operation project was aimed at planning for decommissioning the MR reactor identified as a pilot plant for the decommissioning of the other shutdown reactors on the site. The MR reactor was a pool-type, materials testing reactor with the total thermal power of 50 MW which incorporated pressure tubes containing fuel under test. The MR facility includes the reactor with its nine loop rig rooms containing pumps, heat exchangers and experimental equipment as well as systems and equipment located in other buildings in the complex. The objective of the MR reactor decommissioning project was to identify dismantling equipment and the decommissioning methodology for the reactor, loop rooms and redundant services to permit the refit and re-use of the building for a different nuclear related purpose. The dismantling design comprises two separate, but combined, tasks, namely, the dismantling of reactor installation itself and dismantling of experimental loops. The techniques proposed to undertake the dismantling operations within the loop rooms are described. Two options have been developed for removing contaminated equipment from the high radiation field loop rooms and packaging the waste into approved waste containers. The benefits and detriments of both methods have been identified, which allows implementing the safe, timely and cost-effective decommissioning.

Determination of the neutron flux and γ radiation in the core of operating and shut-down reactors using quartz glasses and element monitors

The neutron fluxes and the intensity of γ radiation are measured in 26 channels of a WR-SM reactor and its thermal column. The fast neutron fluxes in the channels are determined using Ni, Fe, Co, Au, and Mn element monitors with different threshold energies, together with a theoretical calculation using the MCNP-4C program. The energy distribution of the neutron flux inside the fuel assembly is obtained for selected channels around the core. The flux of neutrons with energies >1 MeV is in the range (0.5-43).10 12 cm -2 sec -1 , depending on the location of the channel. A linear correlation is discovered between the induced optical absorption at the 215 nm line (E' center) of SiO 2 -BaO glass and the fast neutron flux in the channels. The γ-ray intensity in the thermal channel is estimated for the reactor during operation (∼38.4 Gy/sec) and 24 hours after it is shut down (∼24. 7 Gy/sec ) using the E' centers induced in pure quartz glasses. The observed difference in the efficiency with which oxygen defects are formed during dry and wet irradiation of glass owing to the radiolysis of water must be taken into account when developing radiation technology and during the burial of radioactive waste.

Decommissioning of gas-cooled reactors

Abstract Gas-cooled reactors represent an important part of the world's nuclear programme. A wide spectrum of designs has been developed and tested. The majority of these test reactors and 1st generation of power reactors has been shut down and is now in different stages of decommissioning. This is also the case for the pilot power reactors of the types High Temperature Gas-cooled Reactor (HTGR) and Heavy Water Gas-cooled Reactor types (HWGCR). Gas-cooled reactors show major generic design characteristics different from Light Water Reactors (LWR), hence influencing the choice of decommissioning strategy, waste/material management and also to a certain extent the decommissioning techniques and procedures applied. The most important issue influencing the decommissioning is the presence of large amounts of graphite in most of these reactors. The majority of the shut-down reactors is located in the UK and in France and major decommissioning programmes are under way in these countries. Interesting decommissioning work has also been carried out in Germany and in the USA. Most reactors are in a stage of safe enclosure. However, decisions have been taken recently to perform direct complete dismantling, notably in France, Italy and, to a certain extent, in Germany.

A Revised ANS Standard for Decay Heat from Fission Products

Heating from the decay of radioactive nuclides in shutdown reactors plays an important role in the safety evaluation of nuclear power plants. Although there are many other important uses for this i...

Evaluation of decay heating in shutdown reactors

Heating from the decay of radioactive nuclides in shutdown reactors plays an important role in the safety evaluation of nuclear power plants. It also must be known in order to design spent fuel storage systems, shutdown reactor cooling systems and heat sink, reprocessing and nuclear waste disposal systems. Of these applications, the analysis of reactor accident scenarios has been the main impetus to develop more accurate methods of decay heat evaluation. It was recognized early in the 1970s that the knowledge was inadequate for safety requirements and that this placed an economic burden on nuclear power plants. Intensive research has been undertaken in the past few years and this has led to a much more precise knowledge of decay heat power in Light Water Reactors (LWR). With additional work this improvement can soon be extended to other reactors types. This paper reviews the background, recent research developments and the evolution of a major revision of the American Nuclear Society Standard for decay heat power in LWR.