Volume 84, Issue 2

In this research, TRACE code with the interface SNAP was used to evaluate and simulate the postulated Extended Loss of AC Power (ELAP) accident in Maanshan Nuclear Power Plant (NPP). This study is to determine whether Reactor Coolant System (RCS) water level will be below Top of Active Fuel (TAF) while the 5th diesel generator and gas turbines are all disabled when the accident occurred. The scenario and assumptions of postulated ELAP in this research are referred to the WCAP-17601-Pt and NUREG-1953 reports. Additionally, this study will run a base case without any mitigation strategy and four cases with multiple mitigation strategies to analyze the effectiveness of URG and FLEX strategies. According to the analysis results of TRACE, it can be found that all four cases with mitigation strategies can keep RCS water level above TAF. This indicates that URG and FLEX strategies can ensure the safety function of reactor.

The direct contact condensation is an effective way to rapid depressurization for light water reactor, such as pressure suppression pool, for which the characteristic of steam plume is a key parameter to evaluate the efficiency of depressurization. In this paper, a series of visualization experiments are described which investigate the characteristic of steam plume, the influence of jetting direction, air mass fraction, water temperature, diameter of jet on the steam plume were analyzed. The results show that, the dimensionless steam plume length increased with the increase of inlet pressure and water temperature for pure steam submerged jet, a correlation was set up to predict the dimensionless penetration length and the predicted errors were within the band of ±15%. Furthermore, the penetration length decreased with the rise of air mass fraction due to buoyancy force, the penetration length increased with the increase of inlet pressure, and the jet penetration length for vertical jet is longer than horizontal jet about 25% at the same inlet pressure and air mass fraction.

Fluid transients occur mainly during fluid transfer in the networked pipe systems in nuclear power plants. Research on pipe network simulation is presented in literature. However, there are still some open technical gaps. In this paper, to satisfy the different time step requirements in different stages of nuclear power plant operation, and to improve the accuracy of simulation results, a multi-scale simulation model of single-phase water pipe networks is proposed. Therefore, the pipe networks are divided into control nodes and flow lines in a complete simulation map. The multi-scale simulation is obtained through the combination of simplified equations, time step selection and refined nodes. The accuracy and stability of the simulator are verified by modeling the feed water system of the Chinese experimental fast reactor. In addition, the output is compared with output of the commercial software, Jtopmeret. As a conclusion, the new model calculates results for this feed water system with a higher accuracy under dynamic operations which consistent with theoretical analysis.

The CHASHMA Nuclear Power Plant unit 1 is known as CHASNUPP-1. The CHASNUPP-unit 1 is a 996 MWth intermediate type pressurised water reactor that began commercial operation in June 2000 in Pakistan. The CHASNUPP unit 1 is a conventional two loop PWR operated by the Pakistan Atomic Energy Commission (PAEC). The radiological source term of CHASNUPP-1 has been evaluated and compared with the advanced modular reactor (SMART) and KORI-1 reactor. For this purpose, modelling and simulation has been carried out in MATLAB. a kinetic model has been developed to carry out the simulation of the release of radionuclides. The core and coolant activity of CHASNUPP-1 is compared with the similar type reactor KORI-1. The developed model uses the ORIGEN 2.2 core inventory as a subroutine. The coolant inventory has been evaluated with 0.25% fuel damage and compared with SMART and KORI-1 reactor.

The health impacts and corresponding damage costs of radioactive emission of Muğla province lignite-fired power plants have been assessed by using the simplified impact pathway approach. Radiation dose and risk calculations have been carried out by the code CAP88-PC around the power plants. Specific isotopes, 238 U, 232 Th, 226 Ra and 40 K in the flying ash samples are considered as radioactive sources. Power plants intersecting regions are also considered in the collective dose calculation. The estimated total collective dose around Muğla province due to Yatağan, Yeniköy and Kemerköy power plants is 5.25 × 10 −2 man.Sv/year. The total health effects damage cost around the Muğla province due to the radioactive emissions from the power plants is 2504.3 US $ 2000/year. The results obtained from the calculations are under the dose limits of International Commission of Radiation Protection and it does not any risk for public health around the plant environment.

The Minim and Martap villages are famous for their richness in bauxite mineralization. Acute respiratory infection, known to stem from 238 U and 232 Th activities, is one of the main reasons to the mortality and morbidity of the over 40,000 local inhabitants. Radioactive elements in top of the soils of the area are potential causes to this infection. Therefore, the natural activity concentration of radioactive elements in the area was investigated. Top soil samples collected from the village settlement area and some plateaus were analyzed using methods LF700 (for 40 K) and LF100 (for 238 U and 232 Th). The measured concentration of the elements was higher than the world mean value. Samples of felsic volcanic soils and rocks were identified to have the highest concentration of the elements. 232 Th and 238 U revealed relatively higher activity concentration while that of 40 K was lower. Absorbed dose rate are higher in all the samples by a factor range of 1.2–2.1 times greater when compared to world average. Annual effective indoor and outdoor doses, radi-um equivalence (Raeq), internal (Hin) and external (Hex) hazard indexes, and the Excess lifetime cancer risks of samples, are all within the world averages by UNSCEAR .

The analysis of accident scenarios in a nuclear power plant and consequently a review of the proper functioning of safety systems is very importance. In this paper, an analysis of a anticipated transient without scram has been done in the Loss of flow scenario with modeling the VVER-1000 reactor of Bushehr by MELCOR1.8.6 code. The chronological sequence of the different steps and the history of the parameters under this accident calculated with the code are evaluated. Parameters such as temperature and pressure changes in the primary and secondary circuit, temperature changes of fuel and fuel clad are calculated as well as a comparison with the melting temperature. Also, the rate of hydrogen production and pouring molten material into the reactor cavity is calculated. This information is very useful for managing the ATWS accident and controlling the reactor. Also, this results to assess Bushehr nuclear power plant. The results show that thermodynamic parameters of pressure and temperature passes the limit without any accident management.

As a perspective plan to transport the spent fuel of material testing of research reactor (MTR) from the temporary storage to permanent storage, choosing an adequate cask is very important to ensure the safety precautions during the transport process. Latin America cask is one of the transportation cask types may be chosen to transport the spent fuel elements where it was designed to transport the irradiated fuel for MTR and the TRIGA research reactors. Therefore, it must be evaluated from the neutronic, radiological, and thermal points of view. The cask has internal diameter of 60 cm which make it possible to content 21 of spent fuel elements of MTR. This study aims to evaluate dose rate distribution around the cask after loading with 21 of MTR spent fuel elements which have been stored for 5-years as a minimal decay time in the temporary storage. For this, MCNP5 code was used to determine the dose rate in the radial and axial directions around the cask. The results show that the dose rates at the cask surface and at 200 cm from the surface are lower than the permissible transportation limits.