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International Journal of Nuclear Energy Science and Technology

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Full-text available via subscription

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ISSN (Print) 1741-6361 - ISSN (Online) 1741-637X
Published by Inderscience Publishers

[367 journals]

Determination of K shell parameters of silver using high resolution HPGe detector spectrometer
- Authors: K.M. Niranjana; Krishnananda; N.M. Badiger; Daisy Joseph; S. Kailas
  Pages: 179 - 190
  Abstract: The K shell parameters such as K shell photoelectric cross-section at K shell binding energy, K shell jump ratio, K shell jump factor, the Davisson-Kirchner ratio and K shell oscillator strength of silver have been determined using high resolution HPGe detector coupled to 16K multichannel analyser. In this method the K X-ray photons in the energy region from 8 to 52 keV are generated by sending 59.56 keV gamma photons from Am-241 source on various targets. These X-ray photons are allowed to pass through a silver foil in good geometry arrangement. By measuring the intensities of incident and transmitted photons, the mass attenuation coefficients are determined at various X-ray energies. Measured mass attenuation coefficient around K shell binding energy and K shell parameters have been compared with values predicted by XCOM and FFAST and with other experimental values.
  Keywords: silver|HPGe detector|K shell parameters|absorption edge|jump ratio|jump factor|EDXRF|mass attenuation coefficient|photon absorption|photon scattering
  Citation: International Journal of Nuclear Energy Science and Technology, Vol. 7, No. 3 (2013) pp. 179 - 190
  PubDate: 2013-02-19T23:20:50-05:00
  DOI: 10.1504/IJNEST.2013.052162
  Issue No: Vol. 7, No. 3 (2013)
Mitigation of PWR in-containment iodine source term under LOCA
- Authors: Khurram Mehboob; Cao Xinrong
  Pages: 191 - 208
  Abstract: Iodine is one of the most hazardous fission products from radiological consequences of LWR. In this paper, the iodine source term has been evaluated for two-loop PWR under severe accident initiated due LOCA. The TMI-2 reactor is considered as the reference reactor. The modelling and simulations of release of iodine have been carried out by developing a MATLAB code that uses the post-accident conditions and core inventory as its input. The containment response, in order to mitigate the environmental and in-containment iodine source term is studied in normal, emergency, and isolation states of containment. The in-containment iodine source term has been evaluated with, and without operation of Engineering Safety features (ESFs). The environmental iodine source term is calculated as the function of containment response. The containment retention factor (CRF) for iodine has also been studied in all confined states of containment. Results indicate the weak sensitivity of iodine towards exhaust rate with ESFs operation.
  Keywords: iodine source term|containment spray system|containment performance|mitigation|light water reactors|LWR accidents|nuclear reactors|nuclear energy|nuclear power|modelling|simulation|iodine release|nuclear safety|containment retention factor|hazardous
  Citation: International Journal of Nuclear Energy Science and Technology, Vol. 7, No. 3 (2013) pp. 191 - 208
  PubDate: 2013-02-19T23:20:50-05:00
  DOI: 10.1504/IJNEST.2013.052163
  Issue No: Vol. 7, No. 3 (2013)
Response time testing of temperature sensors using loop current step response method
- Authors: H.M. Hashemian
  Pages: 209 - 230
  Abstract: The Loop Current Step Response (LCSR) method is used for in-situ response time testing of temperature sensors as installed in an industrial process. This paper presents the theoretical foundation of the LCSR (test theory, LCSR equation and test procedures) and the author's research to validate its use in nuclear power industry applications. The validity of the LCSR method depends on how well the temperature sensor design satisfies LCSR test assumptions, the quality of LCSR data in terms of signal-to-noise ratio, the application of proper sampling parameters for the LCSR data acquisition phase, and the suitability of the algorithms and fitting processes used in the LCSR data analysis stage. In multiple tests the author demonstrated a ±10% agreement between the median values of the response time produced by the plunge and LCSR methods. The plunge test is the standard method for laboratory measurement of the response time of temperature sensors.
  Keywords: process instrumentation|dynamic measurements|response time testing|in-situ testing|response time degradation|resistance temperature detectors|loop current step response test|nuclear power plants|NPP: nuclear energy|temperature sensors|sensor
  Citation: International Journal of Nuclear Energy Science and Technology, Vol. 7, No. 3 (2013) pp. 209 - 230
  PubDate: 2013-02-19T23:20:50-05:00
  DOI: 10.1504/IJNEST.2013.052164
  Issue No: Vol. 7, No. 3 (2013)
Numerical solution for stochastic point-kinetics equations with sinusoidal reactivity in dynamical system of nuclear reactor
- Authors: S. Saha Ray; A. Patra
  Pages: 231 - 242
  Abstract: In this present analysis, the numerical simulation methods are applied to calculate the solution for stochastic point-kinetic equations with sinusoidal reactivity in dynamical system of nuclear reactor. The resulting system of differential equations is solved for each time step-size. Using experimental data, the methods are investigated over initial conditions and with sinusoidal reactivity. The computational results designate that these numerical approximation methods are straightforward, effective and easy for solving stochastic point-kinetic equations.
  Keywords: stochastic point-kinetic equations|Euler-Maruyama method|Taylor 1.5 strong order approximation|neutron density|delayed neutron precursor concentrations|sinusoidal reactivity|nuclear reactors|nuclear power|nuclear energy|numerical simulation|r
  Citation: International Journal of Nuclear Energy Science and Technology, Vol. 7, No. 3 (2013) pp. 231 - 242
  PubDate: 2013-02-19T23:20:50-05:00
  DOI: 10.1504/IJNEST.2013.052165
  Issue No: Vol. 7, No. 3 (2013)
Thermal-hydraulic modelling of LOCA accident using CONTAIN 2.0 code
- Authors: Molood Abbasi; Mohammad Rahgoshay
  Pages: 243 - 257
  Abstract: This paper has modelled one of the accidents in the Pressurised Water Reactor (PWR) reactor containment that is known as Loss of Coolant Accident (LOCA) in its worst condition that is called Large Break LOCA (LB LOCA) with the CONTAIN 2.0 code. Specific type of the LB LOCA is Double-Ended Cold Leg (DECL) break that means guillotine break in the cold leg pipe. The coolant is lost when a LOCA occurs and the fluid of primary loop leaks to the containment and increases pressure and temperature in the containment. This modelling is performed in the VVER-1000 reactor containment. The analysis includes average pressure in the containment and temperature distribution in the sample cells in a short time (during 200 seconds). The results are compared with the existing reports that used the ANGAR code. Results show that the CONTAIN 2.0 code is an adaptable tool for analysis of nuclear events such as LOCA.
  Keywords: CONTAIN 2.0 code|LB LOCA|DECL|double-ended cold leg|containment|VVER-1000 reactors|nuclear reactors|nuclear energy|nuclear power|modelling|nuclear accidents|pressurised water reactors|PWR|loss of coolant|temperature distribution|nuclear safet
  Citation: International Journal of Nuclear Energy Science and Technology, Vol. 7, No. 3 (2013) pp. 243 - 257
  PubDate: 2013-02-19T23:20:50-05:00
  DOI: 10.1504/IJNEST.2013.052168
  Issue No: Vol. 7, No. 3 (2013)
Chemical kinetics parameters of nuclear graphite gasification
- Authors: Mohamed S. El-Genk; Jean-Michel P. Tournier
  Pages: 258 - 284
  Abstract: This paper provides chemical kinetics parameters for the gasification of nuclear graphite grades of IG-110, IG-430, NBG-18 and NBG-25 and presents empirical correlations for their surface areas of free active sites as a function of mass. The kinetics parameters for the four elementary chemical reactions of gasification of these grades of nuclear graphite include the values and Gaussian distributions of the specific activation energies and the values of the pre-exponential rate coefficients for the adsorption of oxygen and desorption of CO and CO<SUB align="right">2 gases. The values of these parameters and the surface area of free active sites for IG-110 and NB-25, with fine and medium petroleum coke filler particles, are nearly the same, but slightly different from those of NBG-18 and IG-430, with medium and fine coal tar pitch coke filler particles. Recommended parameters are applicable to future safety analysis of high and very high temperature gas cooled reactors in the unlikely event of a massive air ingress accident.
  Keywords: nuclear graphite gasification|chemical reaction kinetics|high temperature nuclear reactors|very high temperature nuclear reactors|air ingress accidents|safety analysis|nuclear safety|nuclear energy|nuclear power|nuclear accidents|gas cooled reactors
  Citation: International Journal of Nuclear Energy Science and Technology, Vol. 7, No. 3 (2013) pp. 258 - 284
  PubDate: 2013-02-19T23:20:50-05:00
  DOI: 10.1504/IJNEST.2013.052169
  Issue No: Vol. 7, No. 3 (2013)

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