Volume 80, Issue 3

A number of improvements of the data evaluation concerning nuclear model calculations and evaluation procedures are considered. A promising combination of the hybrid Monte Carlo simulation model of M. Blann for the modelling of non-equilibrium particle emission and the Hauser-Feshbach model, and a possible correction of simulations using intranuclear cascade evaporation model are discussed. In the last case the modelling of particle emission is improved by the consideration of nucleon-cluster interactions. The approach for reliable predictions of cross-sections using “optimal” nuclear model parameters and the method of their generation is discussed.

The reactivity monitoring is a key aspect for the safe operation of nuclear reactors, especially for subcritical source-driven systems. Various methods are available for both off-line and on-line reactivity determination from direct measurements carried out on the reactor. Usually the methods are based on the inverse point kinetic model applied to signals from neutron detectors and results may be severely affected by spatial and spectral effects. Such effects need to be compensated and correction procedures have to be applied. In this work, a new approach is proposed, by using the full information from different local measurements to generate a global signal through a proper weighting of the signals provided by single neutron detectors. A weighting techique based on the use of the adjoint flux proves to be efficient in improving the prediction capability of inverse techniques. The idea is applied to the recently developed algorithm, named MAρTA, that can be used in both off-line and on-line modes.

The feedback from the accident at the Fukushima-Daiichi nuclear power plant is crucial for defining and implementing measures for preventing accidents involving large releases of radioactive material at nuclear installations, including nuclear fuel cycle facilities. Following the lessons learned from this accident, assessment of the safety of nuclear fuel cycle facilities is essential to evaluate the robustness of the facilities' protection systems and components against the impact of extreme external events. A methodology to perform this safety assessment is presented, with discussions on possible preventive measures to be applied and mitigatory actions to be taken for further improvement of the robustness of nuclear fuel cycle facilities when subjected to extreme external events. Considerations in the assessment of multi-facility sites and use of a graded approach, commensurate with the facility's potential hazard, in application of the safety assessment methodology are also discussed.

In this research, the TRACE/SNAP model of Lungmen ABWR nuclear power plant (NPP) has been established for the simulation and analysis of ultimate response guideline (URG). The main actions of URG are depressurization and low pressure water injection of reactor and containment venting. This research focuses to assess the URG utility of Lungmen NPP under Fukushima-like conditions. This study consists of three steps. The first step is the establishment of Lungmen NPP TRACE/SNAP model. In order to evaluate the system response of TRACE/SNAP model, FSAR data (MSIV closure and loss of feedwater flow transient) were used to compare with the results of TRACE. The second step is the URG simulation and analysis under Fukushima-like conditions by using Lungmen NPP TRACE/SNAP model. In this step, the no URG case was also performed in order to evaluate the URG effectiveness of Lungmen NPP. In order to confirm the mechanical property and integrity of fuel rods, the final step is FRAPTRAN analysis. According to TRACE analysis results, the URG can keep the peak cladding temperature (PCT) below the criteria 1 088.7 K under Fukushima-like conditions which indicates that Lungmen NPP can be controlled in a safe situation. Nevertheless, if Lungmen NPP does not perform the URG under Fukushima-like conditions, the water level may drop lower than TAF after 1 100 s which means a safety issue about the fuel rods may be generated. The analysis results of FRAPTRAN also indicate the integrity of fuel rods cannot be kept under the above conditions.

The accident that occurred at the Fukushima Daiichi Nuclear Power Plant is a reminder of the danger of hydrogen explosion within a reactor building. Sufficiently high hydrogen concentration may cause an explosion that could damage the structure, resulting in the release of radioisotopes into the environment. In the first part of this study, a gas diffusion experiment was performed, in which helium was used as the working fluid. An analytical model was also developed using the GOTHIC code and the model predictions of the helium distribution were found to be in good agreement with the experimentally measured data. In the second part of the study, a model of the Mark III containment of the Kuosheng Plant in Taiwan was developed, and was applied to a long-term station blackout (SBO) accident similar to that of the Fukushima plant. The hydrogen generation was calculated using the Modular Accident Analysis Program and was used as the boundary condition for the GOTHIC containment model. The simulation results revealed that the hydrogen concentration at the first floor of the wetwell in the containment reached 4% 9.7 h after the accident. This indicated the possibility of dangerous conditions inside the containment. Although active hydrogen ignitors are already installed in the Kuosheng plant, the findings of this study indicate that it may be necessary to add passive recombiners to prolong an SBO event.

The present study aims to show the effect of different cross section libraries on MINERVE reactor safety parameters. The MCNP5 calculation model of the MINERVE reactor facility is used to determine core and safety parameters such as axial and radial fission rate distributions, control rod worth and spectral indices. Different neutron spectra were achieved by changing the experimental lattice within the MINERVE reactor. MINERVE provides a large experimental basis for the improvement of the cross section databases. The current study calculates these parameters by MCNP5 code using three different cross section libraries (the continuous energy cross sections of the ENDFB-VI, T16-2003, and ENDF/B-VII.1 libraries). Neutronic calculations were performed for R1UO 2 , R1MOX, R2UO 2 , and R2MOX core configurations, representative of a LWR loaded with UO 2 , mixed oxide matrix, over moderated UO 2 and over moderated MOX, respectively. The study aims to determine the most suitable cross section library to be used with MCNP5 reactor calculation code for light water reactor fuel lattice. The MCNP5 results were compared with the experimental results.

The present paper demonstrates nondestructive application of the 131 I and 82 Br radioisotopes in performance evaluation of organic base anion exchange resins Purolite NRW-4000 and Duolite A-378. The evaluation was based on kinetics of iodide and bromide ion-isotopic exchange reactions taking place between the external ionic solution and the resin surface. It was observed that at a constant temperature of 35.0 °C when the concentration of bromide ions in solution increases from 0.001 mol/L to 0.004 mol/L, the specific reaction rate values for bromide ion-isotopic exchange increases from 0.228 to 0.266 min −1 using Purolite NRW-4000 resins and from 0.160 to 0.185 min −1 using Duolite A-378 resins. Also it was observed that under the same experimental conditions, the percentage of bromide ions exchanged increases from 60.67 % to 67.00 % using Purolite NRW-4000 resin, while for Duolite A-378 resin it increases from 49.30 % to 52.00%. The identical trend was observed for the two resins during iodide ion-isotopic exchange reaction. The overall results indicate that under identical experimental conditions, as compared to Duolite A-378 resins, Purolite NRW-4000 resins shows superior performance. It is expected that the present nondestructive radiotracer technique can be extended further for characterization of series of nuclear as well as non-nuclear grade ion exchange resins. The results of such study will be immensely useful in proper selection of the ion exchange resins in various industrial applications.

Utilization of thorium-based fuel in different reactors has been under investigation for several decades. In fact, excellent breeding features, rather flattened distribution of power as well as proliferation resistance of such fuel cycle draws the attention towards utilization of this type of fuel in nuclear power technology. In the present study, the neutronic performance of a typical thorium core loading is addressed. In this configuration, a mixed uranium and thorium oxide is loaded in CANDU 6 reactor. The obtained results determine a total peaking factor of 2.73 for the proposed configuration. The values obtained for the β and the β eff are 332 and 303 pcm respectively. The core reactivity coefficients were more negative comparing the CANDU 6 loaded with nat UO 2 . The initial fissile material loaded in the core increased by a factor of 1.5 after 730 GWd burn-up. The obtained burn-up results show the core reactivity variations were highly positive after 6 and 12 h shut down because of considerably high buildup of 233 Pa after 1-year core operation at 2000 MW power.

Radiopharmaceuticals are radioactive drugs used for diagnosis or treatment in a tracer quantity with no pharmacological action. The production of radiopharmaceuticals is carried out in the special research centers generally using by the cyclotron systems. Indium-111 is one of the most useful radioisotopes used in nuclear medicine. In this paper, we calculated the production cross sections of 111 In radioisotope via 111–114 Cd(p,xn) nuclear reactions up to 60 MeV energy. In the model calculations, ALICE/ASH, TALYS 1.6 and EMPIRE 3.2 Malta nuclear reaction code systems were used. The model calculation results were compared to the experimental literature data and TENDL-2014 (TALYS-based) data.

In this work the performance of backscatter gamma densitometer is studied using 60 Co as gamma source. The ability of the densitometer to determine the liquid's interfaces in multi-phase flows was assessed. The aim mainly was to investigate the possibility of using 60 Co in this system. Furthermore, a comparison between 60 Co and was done. The obtained results indicate that 60 Co can be advantageous in this regard.

In this paper, we attempt to determine the activity concentration of natural and artificial radionuclides in 37 sand samples from the Mediterranean coast of Antalya in Turkey by using a high purity Germanium (HpGe) detector. 238 U, 232 Th, 40 K and 137 Cs activity concentrations, absorbed dose rate, annual effective dose equivalent, radium equivalent activity, external and internal hazard index of sand samples are determined res-pectively. The average values are 13.43 ± 0.21 Bq/kg, 6.96 ± 0.06 Bq/kg and, 122.46 ± 18.58 Bq/kg, for 238 U, 232 Th and 40 K respectively. Most of the activity concentration values are less than below minimum detection limit for 137 Cs. The average values of the absorbed dose rate D (nGy/h), annual effective dose equivalent AEDE(μSv/y), radium equivalent activity Ra eq (Bq/kg), external hazard index H ex and internal hazard index H in are 15.52 nGy/h, 19.03 μSv/y, 32.81 Bq/kg, 0.09 and 0.12 respectively. It is observed that 238 U, 232 Th, 40 K and 137 Cs activity concentrations, absorbed dose rate, annual effective dose equivalent, radium equivalent activity are in the limit of the published values, external and internal hazard index values are less than unity.

In this symbolic investigation, a cylindrical cell in a LWR, which consists of one fuel pin and moderator (water), is considered. The width of this cylindrical cell is divided into 100 equal units. Since the neutron flux in a cylindrical fuel pin is resulting from the diffusion equation: -1rddrφ(r)+∑aφ(r)=S(r)$ - \frac{1}{r}\frac{d}{{dr}}\varphi \left( r \right) + \mathop \sum \nolimits_a \varphi \left( r \right) = S\left( r \right)$, the amount of fast neutron fluxes are obtained on the basis of the numeric solution of this equation, and the applied boundary conditions are considered: φ′(0)=φ′(1)=0$\varphi '\left( 0 \right) = \varphi '\left( 1 \right) = 0$. This differential equation is solved by the tridiagonal method for variant enrichments of uranium. Neutron fluxes are obtained in variant radii of fuel pin and moderator and are finally compared with each other. There are some interesting outcomes resulting from this investigation. It can be inferred that because of the fuel enrichment increment, the fast neutron flux increases significantly at the centre of core, while many of the fast neutrons produced are absorbed after entering the water region, moderation of lots of them causes the reduced neutron flux to get improved in this region.