<p>A few years ago Westinghouse started the development of a new method for criticality calculations for spent nuclear fuel storage pools called “PHOENIX-to–MCNP” (PHX2MCNP). PHX2MCNP transfers burn-up data from the code PHOENIX to use in MCNP in order to calculate the criticality. This thesis describes a work with the purpose to further validate the new method first by validating the software MCNP5 at higher water temperatures than room temperature and, in a second step, continue the development of the method by adding a new feature to the old script. Finally two studies were made to examine the effect from decay time on criticality and to study the possibility to limit the number of transferred isotopes used in the calculations.</p><p>MCNP was validated against 31 experiments and a statistical evaluation of the results was done. The evaluation showed no correlation between the water temperature of the pool and the criticality. This proved that MCNP5 can be used in criticality calculations in storage pools at higher water temperature.</p><p>The new version of the PHX2MCNP script is called PHX2MCNP version 2 and has the capability to distribute the burnable absorber gadolinium into several radial zones in one pin. The decay time study showed that the maximum criticality occurs immediately after the takeout from the reactor as expected.</p><p>The last study, done to evaluate the possibility to limit the isotopes transferred from PHOENIX to MCNP showed that Case A, a case with the smallest number of isotopes, is conservative for all sections of the fuel element. Case A, which contains only some of the actinides and the strongest absorber of the burnable absorbers gadolinium 155, could therefore be used in future calculations.</p><p>Finally, the need for further validation of the method is discussed.</p>

A few years ago Westinghouse started the development of a new method for criticality calculations for spent nuclear fuel storage pools called “PHOENIX-to–MCNP” (PHX2MCNP). PHX2MCNP transfers burn-up data from the code PHOENIX to use in MCNP in order to calculate the criticality. This thesis describes a work with the purpose to further validate the new method first by validating the software MCNP5 at higher water temperatures than room temperature and, in a second step, continue the development of the method by adding a new feature to the old script. Finally two studies were made to examine the effect from decay time on criticality and to study the possibility to limit the number of transferred isotopes used in the calculations. MCNP was validated against 31 experiments and a statistical evaluation of the results was done. The evaluation showed no correlation between the water temperature of the pool and the criticality. This proved that MCNP5 can be used in criticality calculations in storage pools at higher water temperature. The new version of the PHX2MCNP script is called PHX2MCNP version 2 and has the capability to distribute the burnable absorber gadolinium into several radial zones in one pin. The decay time study showed that the maximum criticality occurs immediately after the takeout from the reactor as expected. The last study, done to evaluate the possibility to limit the isotopes transferred from PHOENIX to MCNP showed that Case A, a case with the smallest number of isotopes, is conservative for all sections of the fuel element. Case A, which contains only some of the actinides and the strongest absorber of the burnable absorbers gadolinium 155, could therefore be used in future calculations. Finally, the need for further validation of the method is discussed.

Canada has participated in the Generation IV International Forum (GIF) collaboration in the area of Super Critical Water-cooled Reactors (SCWR). Similar to the current CANDU technologies, in the Canadian SCWR design the low pressure heavy water moderator system is separated from the supercritical coolant system (25MPa). The High Efficiency Re-entrant Channel (HERC) design in the Canadian SCWR has multiple coolant regions (i.e. coolant in the downward center flow tube and coolant in the upward outer fuel region) and provides thermal isolation between the moderator and heat transport system fluid. Although the overall reactivity feedback in the channel is negative for equilibrium density decrease transients, a temporary positive reactivity may be induced during non-equilibrium conditions such as cold-leg Loss of Coolant Accidents (LOCA). The primary objective of this study is to investigate the fuel and coolant behaviors under postulated LOCA transients, in particular those caused by cold-leg breaks, and to demonstrate the effectiveness of several proposed safety systems in the Canadian SCWR. The one-dimensional thermalhydraulic system code RELAP5 has been used for the safety analysis in many LWRs. The latest version RELAP5/MOD4 has been improved to accommodate supercritical water and is used in this study. A RELAP5 model is constructed based on the most recent Canadian SCWR design. The 336 fuel channels are split into two representative groups each with a series of hydraulic and heat structure models. A benchmark study is conducted by comparing RELAP5 to CATHENA simulations and shows good agreement for both steady-state and transient predictions. The RELAP5 model is then used to predict the system response to several postulated LOCA transients. For a 100% (single-ended) cold-leg break located in between the feedwater pump and the inlet plenum, the system pressure immediately drops followed closely by a flow reversal with rapid discharge from the break. A brief power pulse (178%FP) is observed under this non-equilibrium depressurization scenario. The transient simulations show the potential for two sheath temperature maxima, one early in the transient as a result of the power pulse and the subsequent flow-power mismatch, and another later peak resulting from the fuel heat-up under near stagnant channel flow conditions (such as in the failure of the Emergency Core Cooling Systems) as the heat transfer regime changes to radiation dominated. The Automatic Depressurization System (ADS) located on the hot-leg side mitigates the later fuel heat-up by introducing forward channel flows. This effect is enhanced by additional coolant supplied from Low Pressure Coolant Injection (LPCI) which is part of the Emergency Core Cooling System (ECCS). Under the 100% break LOCA/LOECC transient, the core inventory is depleted rapidly after the break and thermal radiation becomes the dominant heat removal mechanism. The highest MCST, 1331 K, is achieved approximately 136s after the break and meets the safety criterion (1533 K). Beyond this time the sheath temperatures gradually decrease either by the continuous LPCI from the reactor sumps, gravity driven core cooling, or in the event of a failure of those systems by the Passive Moderator Cooling System (PMCS). LOCAs initiated by break sizes varying from 5% to 100% of the cold-leg cross-section area are simulated under loss of ECCS. In this specific design, break sizes less than 15% are defined as SBLOCAs and show an early pressure increase up to the Safety Relief Valve (SRV) setpoint. During SBLOCAs, the first MCST peak is more limiting than the large LOCA case because of insufficient fuel cooling caused by relatively low reverse flow. However, these lower reverse flows prolong the period of blowdown cooling and hence help to mitigate the secondary MCST peak. The worst LOCA case occurs in the 15% break case with a maximum cladding temperature of 1450 K. The results showed the most sensitive parameters are delays associated with SDS action, emissivity and ADS actuation parameters. / Thesis / Master of Applied Science (MASc)

No description available.

This report accounts for investigations of ways to reduce the calculation times forsimulations of falling control rods in boiling water reactors done prior to everyreactor startup, known as RIA-calculations. Two methodologies to lower thecalculation times have been proposed, developed and implemented in a set ofmatlab-scripts, which are fully compatible with the previously used methodology.The new methodologies have been applied on 17 authentic power cycles at the threeForsmark reactors, whereby a reduction in calculation times by 70 to 90 % could bedemonstrated while still confidently maintaining the analysis performance. Thesimulations made and the basis of the new methodologies are described in detail inthis report, and possible steps to further lower the calculation times are alsoproposed.

Jatropha curcas L. is a drought-resistant plant which can be grown in poor soil and marginal lands. The use of Jatropha seed oil to produce biodiesel has been widely studied in recent years. Results showed that it is one of the most promising alternatives for conventional petro-diesel. Currently, hexane is still the most commonly used solvent for commercial oil extraction. However, the increasing price and flammability properties of hexane are motivating industry to seek alternative solvents. The objectives of this study are to design and analyze the Jatropha seed oil extraction for use in biodiesel production, and to provide a systematic safety-economic analysis of alternative solvents that can be used in Jatropha seed oil extraction. First, a base-case flowsheet is synthesized for oil extraction. Then, the base-case extraction process and each solvent Fire and Explosion Index (F & EI) and the Solvent Safety Index (SSI). Eight solvents, including n-heptane, toluene, xylene, dichloromethane, chloroform, 1,2-dichloroethane, methanol and ethanol are selected for candidates by comparing these results to those for hexane. Two cases are developed to evaluate the economic potential of these candidates. Furthermore, heat integration is applied to the process to minimize energy usage. Finally, a comprehensive solvent comparison is developed based on F & EI, SSI, solvent makeup cost, utilities cost, and capital investment. The results show that chloroform is the optimal solvent, while dichloromethane is the next best.

Made available in DSpace on 2014-10-09T12:28:22Z (GMT). No. of bitstreams: 0 / Made available in DSpace on 2014-10-09T13:56:28Z (GMT). No. of bitstreams: 0 / Tese (Doutoramento) / IPEN/T / Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP

Made available in DSpace on 2014-10-09T12:32:55Z (GMT). No. of bitstreams: 0 / Made available in DSpace on 2014-10-09T14:08:33Z (GMT). No. of bitstreams: 1 17896.pdf: 536541 bytes, checksum: 914e67d2a1b839fa60af631909047c5a (MD5) / Dissertacao (Mestrado Profissionalizante em Lasers em Odontologia) / IPEN/D-MPLO / Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP; Faculdade de Odontologia, Universidade de São Paulo, São Paulo

Made available in DSpace on 2014-10-09T12:28:22Z (GMT). No. of bitstreams: 0 / Made available in DSpace on 2014-10-09T13:56:28Z (GMT). No. of bitstreams: 0 / O presente trabalho tem o objetivo de avaliar os riscos potenciais apresentados pelos incidentes em reatores nucleares de pesquisa. Para o desenvolvimento do trabalho, duas bases de dados do Organismo Internacional de Energia Atômica, OIEA, foram usadas, o Incident Report System for Research Reactor e Research Reactor Data Base. Para este tipo de avaliação fez-se uso de Análise Probabilística de Segurança (APS), dentro de um limite de confiança de 90% e, Análise de Probabilidades determinística (APD). Para obtenção dos resultados dos cálculos de probabilidades por APS, utilizou-se a teoria e as equações sugeridas em documento da IAEA TECDOC - 636. O desenvolvimento dos cálculos das probabilidades por APS utilizou-se o Programa Scilab versão 5.1.1, de livre acesso, executável nas plataformas do Windows, Linux. Um programa específico para obter os resultados das probabilidades foi desenvolvido dentro do programa principal Scilab 5.1.1., para duas distribuições Fischer e Chi-quadrado, ambas no limite de confiança de 90%. Fazendo uso das equações de Sordi e do programa Origin 6.0, foram obtidas as doses máximas admissíveis relacionadas às probabilidades que satisfazem os limites de riscos estabelecidos pela Comissão Internacional de Proteção Radiológica, CIPR e, pode-se também obter estas doses máximas graficamente com a figura 1 resultante dos cálculos de probabilidades x doses máximas admissíveis. Verificou-se que a confiabilidade nos resultados das probabilidades está relacionada com a experiência operacional (reator x ano e fração) e, que quanto maior ela for, maior é a confiabilidade no resultado. Finalizando, sugere-se uma lista de futuros trabalhos que complementam este. / Tese (Doutoramento) / IPEN/T / Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP

Made available in DSpace on 2014-10-09T12:32:55Z (GMT). No. of bitstreams: 0 / Made available in DSpace on 2014-10-09T14:08:33Z (GMT). No. of bitstreams: 1 17896.pdf: 536541 bytes, checksum: 914e67d2a1b839fa60af631909047c5a (MD5) / Dissertacao (Mestrado Profissionalizante em Lasers em Odontologia) / IPEN/D-MPLO / Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP; Faculdade de Odontologia, Universidade de São Paulo, São Paulo