International Journal of Nuclear Energy Science and Technology (9 papers in press)
Burn/breed: a wavelet-based nuclear fuel burnup and decay heat code
by Hesham Nasif
Abstract: Burn/breed is a code designed to aid in the analysis, prediction and optimisation of fuel burnup performance and decay heat calculation in a nuclear reactor. The code uses the output parameters generated by the Monte Carlo neutronics codes to determine the isotopic inventory as a function of time and power density. Burn/breed directly uses the neutron absorption tally/reaction information generated by Monte Carlo code for each nuclide of interest to determine the nuclide inventory. When the isotope inventories have been calculated for a specific reactor operation and cooling period, the decay heat can be derived. Beside the U-235 and Pu-239 decay chains, the code includes the U-233 decay chain to manage the possible scenarios for using thorium in a nuclear fuel cycle. This paper describes the theoretical basis of this code, and shows the results of the code for two test cases. The results show good agreement with other codes for the first test case and with the experimental results for the second test case.
Keywords: burn/breed; burnup; decay heat; nuclear reactor; Monte Carlo; isotopic inventory; wavelet; U-233 decay chain.
Estimation of radioactivity released from CHASNUPP-1 nuclear power plant during loss of coolant accident
by Khurram Mehboob, Mohammad S. Aljohani
Abstract: The CHASNUPP unit 1 is a 996 MWth intermediate type pressurised water reactor that began commercial operation in May 2000 in Pakistan. The CHNUPP-1 is a conventional two-loop PWR operated by the Pakistan Atomic Energy Commission (PAEC). The expected amount of radiation exposure from the CHNUPP-1 R CHNUPP unit 1 is simulated for hypothetical severe accidents. For this purpose, modelling and simulation has been carried out in MATLAB, and a kinetic model has been developed and implemented in MATLAB to carry out the simulation of the release of radionuclides. The core and coolant activity of CHNUPP-1 is compared with the similar type reactor KORI-1. The developed model uses the Origen 2.2 (A.G. craft, 1983) core inventory as a subroutine. The coolant inventory has been evaluated with 0.25% fuel damage and compared with the KORI-1 reactor.
Keywords: CHASNUPP; pressurised water reactor; loss of coolant accident.
The enhancement of energy gain in a p6Li inertial fusion reactor by laser-driven protons
by Javad Bahmani, Baharak Eslami, Farhad Mohammad Jafari
Abstract: For the ignition of proton-lithium-6 (p6Li) fuel in a fusion reactor, two important challenges have emerged: the losses of energy and the need for high-temperature electrons and ions. To provide favorable conditions for this fuel these problems should be reduced as much as possible. In this paper, one of the considerable approaches for reducing p6Li ignition energy is proposed, which is the use of a proton beam in a fast ignition method with inertial confinement fusion (ICF). The proton particles in the beam can fuse with ions into p6Li plasma (protons and lithium-6) and in this way produce additional energy. The determination of this energy enhancement and the investigation of influential parameters can have a significant role in increasing energy gain in p6Li fusion reactors. In this study, the stopping power of the proton beam in the p6Li fuel is examined. The amount of the produced excess energy and the total energy are estimated at different proton beam energies and electron temperatures. The investigations indicate that total stopping power of proton particles in p6Li fuel at different temperatures is a function of proton energy and it decreases with rising temperature. The conclusions show that the amount of additional energy depends on the proton beam energy and the electron temperature. The amount of extra energy gain is considerable at low proton energy and high electron temperature.
Keywords: stopping power; fusion; plasma; proton beam.
Estimated radon concentration in drinking water samples for different regions of Hilla City
by Khalid Jassim, Inaam H. Kadhim, Nawras T. Shihab
Abstract: In this study, the concentration of radon in water has been measured for 17 samples from different regions of AL-Kifl in the province of Babylon, Iraq. This study performed using the electronic radon detector RADH2O, where the highest value was 0.199 Bq L-1 and the lowest value was 0.0 Bq. L-1, and effective dose for human exposure to radon average was 5.09*10-7 Sv.y-1. It may be concluded that the results of radon concentrations and annual effective dose in all samples show no significant radiological risk for the inhabitants in the study region. We have chosen this subject of the current study of the importance of water in human life and living because of the lack of previous studies in the study area.
Keywords: radon concentrations; water; annual effective dose; Babylon; RAD7; UNSCEAR.
Uncertainty calculation in small break LOCA in the emergency core cooling system connected to the hot leg of Angra 2 nuclear power plant
by Eduardo Madeira Borges, Gaiane Sabundjian, Francesco D'Auria, Alessandro Petruzzi
Abstract: Owing to the occurrence of nuclear accidents, worldwide nuclear regulatory organisations included the analysis of accidents considered as design basis accidents Loss of Coolant Accident (large- and small-break LOCA) in the safety analysis reports of nuclear facilities. In Brazil, the tool selected by the licensing authority, Comiss
Keywords: CIAU; RELAP5; SBLOCA; nuclear safety analysis.
Validating COMSOL multiphysics for VVER-1000 whole-core-steady-state via AER benchmark problem
by Ned Xoubi, Abdelfattah Soliman
Abstract: In this work, a full core, three-dimensional, multi-group model of the VVER-1000 reactor core is developed using specifications of the Schulz benchmark. This paper presents a new 3-D full core solution, and simulates the neutronic behaviour of the VVER-1000 reactor by predicting the system criticality, power distribution, and the neutron flux distribution. Multi-group constants are applied to the COMSOL model to perform neutronics calculations using finite element method with adaptive mesh refinement. The study found that the calculated effective multiplication factor (keff) compares well with the reference value. Furthermore, the fission rates, 3-D power distributions and axially averaged 2-D power distribution are in good agreement with reported reference results. The thermal neutron spectrum is also calculated by the COMSOL model and presented in this paper. This study allows us to validate the COMSOL calculation schemes for VVER-type reactors and to compare our solutions with reference solutions at steady state
Keywords: VVER; full core; benchmark; neutronics calculation; power distribution; COMSOL.
An R-package for water and steam properties for scientific and general use
by Benedito Baptista, Eduardo Cabral, Antonio Barroso
Abstract: The International Association for the Properties of Water and Steam (IAPWS) develops formulations for the calculation of thermophysical properties of water as a function of different combinations of temperature, density, pressure, enthalpy, and entropy. These properties are useful for scientists and nuclear, chemical and mechanical engineers who analyse experimental data or are involved with projects and equipment development, such as heat exchangers, turbines or nuclear power reactors. The IAPWS-95 formulation solves the fundamental equation of Helmholtz free energy as a function of temperature and density. This paper gives a description of how these equations are solved and exemplifies the use of a package developed for the free platform R. The IAPWS95 package was developed to help users to get access to the IAPWS-95 formulation in a free software environment that is growing exponentially. Transport properties were programmed using other IAPWS releases. The examples consider the uncertainty analysis of thermal parameters of a nuclear power reactor and the preparation of tables and graphs of water properties.
Keywords: steam-water properties; thermophysical water properties; R software environment; IAPWS-95 water formulation; Helmholtz free energy equation; uncertainty analysis in PWR.
Prediction of peak cladding temperature in a three-loop pressurised water reactor with accident-tolerant fuel during loss-of-coolant accident
by Alexander Agung
Abstract: Safety analysis to a pressurised water reactor fuelled with ATF (Accident-Tolerant Fuel) has been performed at LB-LOCA condition. The ATF being used is uranium silicide (U3Si2) and FCMF (Fully Ceramic Microencapsulated Fuel) with silicon carbide (SiC) and FeCrAl alloy as a cladding material. The objective of this research is to obtain dynamic characteristics of ATF-fuelled PWR at LB-LOCA condition as compared to PWR-fuelled with monolithic UO2. The research activities were performed by modelling the PWR core as well as other related components in both primary and secondary systems. Modelling and simulation were performed by using RELAP5-3D system code, incorporating hydrodynamic structures, heat structures, trips and control variables to the reactor as well as its balance of plant. In addition, thermophysical properties of monolithic UO2 and ATF were implemented to the RELAP5-3D model. The LB-LOCA analysis was performed by assuming a safe shutdown of the reactor after a depressurisation following a double-ended guillotine breach in the main pipe. Neutronic feedback was therefore irrelevant and consequently the heat source originating from decay heat was estimated from ANS-5.1-2014 decay heat curve. Simulations were performed by assuming one out of three high-pressure safety injection (HPSI) system and one out of two low-pressure safety injection (LPSI) system are not functioning. The results of simulations show that during LB-LOCA with functioning ECCS, the transient PCTs were far below the maximum allowable limit (1204
Keywords: nuclear safety; peak cladding temperature; accident-tolerant fuel; loss-of-coolant accident; RELAP5-3D.
Core conversion design study of TRIGA Mark 2000 Bandung using MTR plate type fuel element
by Surian Pinem, Tagor Malem Sembiring, Tukiran Surbakti
Abstract: The TRIGA Mark 2000 Bandung (TRIGA 2000) is a TRIGA Mark II type reactor with the nominal thermal power of 2 MW. It uses the reactor back-up reactor for irradiation radioisotope targets to the local market. Some standard and control fuel elements reached the limit value of burn-up level after 53 years operations. Therefore, the fresh fuel elements are needed to keep the routine reactor operation while the fuel fabrication does not supply the fuel. This present study describes the fuel type conversion, from TRIGA fuel rod type to the MTR fuel plate type, as a solution to the lack of supply. The study obtained an optimum equilibrium core of the TRIGA 2000 reactor using the fuel plate type of silicide fuel (U3Si2Al) with the low uranium enrichment. The reactor uses graphite as the reflector. The WIMSD/5 code has been used for the cell calculations to generate the neutron diffusion constants as a function of temperature and xenon condition. Those constants are using for the two-dimension and three-dimension core calculations by using multigroup neutron diffusion method to determine the static and kinetic parameters of the equilibrium cores. The core calculation results showed that the optimum equilibrium core configuration is 5
Keywords: TRIGA Mark II 2000 Bandung; MTR fuel plate type; core conversion; silicide fuel; TRIGA fuel rod type.