The chapter presents a general overview on nuclear fission energy production. Chapter starts with... more The chapter presents a general overview on nuclear fission energy production. Chapter starts with the fundamentals of the fission process and fission energy production. Importance of the neutrons in nuclear reactor technology is highlighted. The Boltzmann neutron transport equation is explicitly evaluated. Conventional and advanced nuclear reactors are addressed. Major Gen-IV nuclear reactor types are reported. Finally, major space nuclear reactor systems are investigated. The importance of the space nuclear reactors for the commercial exploitation of the solar system is pointed out.
In a source-driven fissionable blanket, a flat fission power density (FPD) is achieved by using a... more In a source-driven fissionable blanket, a flat fission power density (FPD) is achieved by using a mixed fuel (Th0 2 and natural U0 2) with the thorium/uranium ratio changing from front to back in the ten fuel rows along the radial direction. A straightforward graphic method is used. The temporal behavior of the FPD has been observed for an operation period of 6 months and for a plant load factor of 75% by applying a fusion driver neutron flux of 10 14 14-MeV neutrons/(cm 2-s) at the first wall, corresponding to ~2.25 MW/m 2. To keep the power density flat, it is necessary to replace the fuel in rows 1, 2, and 3, close to the first wall. The time intervals for this operation increase, counting from initial start-up, typically, 2 months, 6 months, etc. One result of this study is that plutonium produced in such a hybrid blanket contains very low amounts of even isotopic components even over very long operation times of ~3 yr. Hence, if fusion reactors are introduced into the energy market, special regulations are needed for international safeguarding.
Preliminary study on influence of changing void fraction (VF) on SUPEL (Straight Utilization of s... more Preliminary study on influence of changing void fraction (VF) on SUPEL (Straight Utilization of sPEnt LWR fuel in LWR system) scenario for boiling water reactor (BWR) spent fuel direct recycling scheme has been carried out. Several VF values of BWR have been investigated to determine the criticality of reactor. The VF values range from 20% to 60%. The fraction of spent fuel to the total loaded fuel was changed from 5% to 20%. The required uranium enrichment for criticality becomes higher with the increasing of VF as well as the enlarging of the fraction of spent fuel in loaded fuel. The neutron spectra become harder with the augmenting of VF. The plutonium and minor actinides isotopes are produced more in the reactor.
Energy Citations Database. ADOPT-A-DOC? ... Title, Control mechanism for spacecraft thermionic re... more Energy Citations Database. ADOPT-A-DOC? ... Title, Control mechanism for spacecraft thermionic reactors leading to lower radiation shielding mass. Creator/Author, Sahin, S. Publication Date, 1979 Jan 01. OSTI Identifier, OSTI ID: 5351674. Report Number(s), CONF-791103-. ...
... King Saud University, POBox 800, Riyadh 11421, Saudi Arabia On leave: University of Erciyes, ... more ... King Saud University, POBox 800, Riyadh 11421, Saudi Arabia On leave: University of Erciyes, Kayseri, Turkey CHIEN CHUNG National Tsing Hua ... foils of 93.15% enriched 235U (3mg each)ve been irradiated with thermal neutrons at a thermal pow r of 500 watts, to measure ...
Abstract Criticality and burn up behavior of the Fixed Bed Nuclear Reactor (FBNR) are investigate... more Abstract Criticality and burn up behavior of the Fixed Bed Nuclear Reactor (FBNR) are investigated for the mixed fuel 233UO2/ThO2 as an alternative to low enriched 235UO2 fuel. CERMET fuel with a zirconium matrix and cladding has been used throughout the study. The main results of the study can be summarized as follows: Reactor criticality is already achieved by ∼2% 233UO2 with the mixed 233UO2/ThO2 fuel. At higher 233U fractions, reactor criticality rises rapidly and exceeds keff > 1.5 already by 9% 233UO2. With 100% 233UO2, start up criticality can reach keff = 2.0975. Time dependent reactor criticality keff and fuel burn up have been investigated for two different mixed fuel 233UO2/ThO2 compositions, namely: 4% 233UO2 + 96% ThO2 for a reactor power of 40 MWel (120 MWth) and 9% 233UO2 + 91% ThO2 for a reactor power of 70 MWel (210 MWth). Sufficient reactor criticality (keff > 1.06) for continuous operation without fuel change can be sustained during ∼5 and 12 years with 4% and 9% 233UO2 fractions in the mixed fuel, leading to burn ups of ∼36000 and >105000MWD/t, respectively. Thorium based fuel produces no prolific uranium. Plutonium production remains negligible.
The possibility of spent nuclear fuel rejuvenation in fusion reactors is investigated for both (D... more The possibility of spent nuclear fuel rejuvenation in fusion reactors is investigated for both (D,T) and catalysed (D,D) modes. The analysis is conducted for light water reactor (LWR) spent nuclear fuel which was used up to a total enrichment grade (fissionable plutonium plus 235U) of 2.174% The behaviour of the spent fuel is observed during 48 months for discrete time intervals of At = 6 months. The cooling of the fissile fuel zone is considered with three different coolants, namely pressurized helium ('gas'), Flibe (Li2BeF4) ('molten salt') and natural lithium ('liquid metal'). The best rejuvenation performance is obtained with Flibe, followed by helium and natural lithium. The catalysed (D,D) mode has a better rejuvenation capability than the conventional (D,T) mode. A rejuvenation time of 12 months is evaluated for a final fissile fuel enrichment grade of around 3.5% with Flibe coolant in the fissile zone under a first-wall fusion neutron current load of 1014 (2.45 MeVncm-2s) and 1014 (14.1 MeV ncm-2s), corresponding to 2.64 MWm-2 by a plant factor of 75% for the catalysed (D,D) fusion reactor mode. The rejuvenation time increases to 24 months for the same fissile fuel enrichment grade using the (D,T) fusion reactor mode under a first-wall fusion neutron current load of 1014 (2.45 MeV ncm-2 s), corresponding to 2.25 MWm-2 by a
... REDUCED SHIELDING MASS FOR THE VISTA SPACECRAFT Sümer Şahin* and Hacı MehmetŞahin Gazi Üniver... more ... REDUCED SHIELDING MASS FOR THE VISTA SPACECRAFT Sümer Şahin* and Hacı MehmetŞahin Gazi Üniversitesi, Teknik Eğitim Fakültesi Beşevler, Ankara 06503, Turkey and ... Page 2. Sümer Şahin, H. Mehmet Şahin, and Abdulmuttalip Şahinaslan ...
A thermo-hydrodynamic-neutronic analysis is performed for a fast, uranium carbide (UC) fueled spa... more A thermo-hydrodynamic-neutronic analysis is performed for a fast, uranium carbide (UC) fueled space-craft nuclear in-core thermionic reactor. The thermo-hydrodynamic analysis shows that a hybrid thermionic spacecraft nuclear reactor can be designed for both electricity generation and nuclear thermal propulsion purposes. The neutronic analysis has been conducted in S[sub 8]-P[sub 3] approximation with the help of one- and two-dimensional neutron transport codes ANISN and DORT, respectively. The calculations have shown that a UC fueled electricity generating single mode thermionic nuclear reactor can be designed to be extremely compact because of the high atomic density of the nuclear fuel (by 95% sintering density), namely, with a core radius of 8.7 cm and core height of 25 cm, leading to power levels as low as 5 kW (electric) by an electrical output on an emitter surface of 1.243 W/cm[sup 2]. A reactor control with boronated reflector drums at the outer periphery of the radial reflector of 16-cm thickness would make possible reactivity changes of [Delta]k[sub eff] > 10% -- amply sufficient for a fast reactor -- without a significant distortion of the fission power profile during all phases of the space mission. The hybrid thermionic spacecraft nuclear reactor mode contains cooling channels inmore » the nuclear fuel for the hydrogen propellant. This increase the critical reactor size because of the lower uranium atomic density in this design concept. Calculations have lead to a reactor with a core radius of 22 cm and core height of 35 cm leading to power levels [approximately] 50 kW(electric) under the aforementioned thermionic conversion conditions.« less
A thermo-hydrodynamic-neutronic analysis is performed for a fast, uranium carbide (UC) fueled spa... more A thermo-hydrodynamic-neutronic analysis is performed for a fast, uranium carbide (UC) fueled space-craft nuclear in-core thermionic reactor. The thermo-hydrodynamic analysis shows that a hybrid thermionic spacecraft nuclear reactor can be designed for both electricity generation and nuclear thermal propulsion purposes. The neutronic analysis has been conducted in S[sub 8]-P[sub 3] approximation with the help of one- and two-dimensional neutron transport codes ANISN and DORT, respectively. The calculations have shown that a UC fueled electricity generating single mode thermionic nuclear reactor can be designed to be extremely compact because of the high atomic density of the nuclear fuel (by 95% sintering density), namely, with a core radius of 8.7 cm and core height of 25 cm, leading to power levels as low as 5 kW (electric) by an electrical output on an emitter surface of 1.243 W/cm[sup 2]. A reactor control with boronated reflector drums at the outer periphery of the radial reflector of 16-cm thickness would make possible reactivity changes of [Delta]k[sub eff] > 10% -- amply sufficient for a fast reactor -- without a significant distortion of the fission power profile during all phases of the space mission. The hybrid thermionic spacecraft nuclear reactor mode contains cooling channels inmore » the nuclear fuel for the hydrogen propellant. This increase the critical reactor size because of the lower uranium atomic density in this design concept. Calculations have lead to a reactor with a core radius of 22 cm and core height of 35 cm leading to power levels [approximately] 50 kW(electric) under the aforementioned thermionic conversion conditions.« less
Abstract The aim of this paper is to determine neutronic performances of the light water reactor ... more Abstract The aim of this paper is to determine neutronic performances of the light water reactor (LWR) spent fuel mixed with fertile thorium fuel in a FFHR. Time dependent three dimensional calculations for major technical data, such as blanket energy multiplication, tritium breeding ratio, cumulative fissile fuel enrichment and burnup have been performed by using Monte Carlo Neutron-Particle Transport code MCNP5 1.4, coupled with a novel interface code MCNPAS, which is developed by our research group. A self-sustaining tritium breeding ratio (TBR>1.05) has been kept throughout the calculations. The study has shown that the fissile fuel quality will be improved in the course of the transmutation of the LWR spent in the FFHR. The latter has gained the reusable fuel enrichment level conventional LWRs between one and two years. Furthermore, LWR spent fuel - thorium mixture provides higher burn-up values than in light water reactors.
Science and Technology of Nuclear Installations, 2016
Emergency planning zones (PAZ and UPZ) around the Karachi-2 and Karachi-3 nuclear power plants (K... more Emergency planning zones (PAZ and UPZ) around the Karachi-2 and Karachi-3 nuclear power plants (K-2/K-3 NPPs) have been realistically determined by employing Gaussian puff model and Gaussian plume model together for atmospheric transport, diffusion, and deposition of radioactive material using onsite and regional data related to meteorology, topography, and land-use along with latest IAEA Post-Fukushima Guidelines. The analysis work has been carried out using U.S.NRC computer code RASCAL 4.2. The assumed environmental radioactive releases provide the sound theoretical and practical bases for the estimation of emergency planning zones covering most expected scenario of severe accident and most recent multiunit Fukushima Accident. Sheltering could be used as protective action for longer period of about 04 days. The area about 3 km of K-2/K-3 NPPs site should be evacuated and an iodine thyroid blocking agent should be taken before release up to about 14 km to prevent severe determinist...
Abstract Neutron flux and energy spectrum were determined at the surface of three in-vessel Surve... more Abstract Neutron flux and energy spectrum were determined at the surface of three in-vessel Surveillance Capsule Assemblies (SCAs) removed from CHASNUPP Unit-1 after 2nd, 4th, and 9th fuel cycles for the life assessment of reactor pressure vessel belt line region. Dosimetry data were measured from radiometric sensors irradiated in base material section of SCAs. Fast neutron flux (E > 1.0 MeV) was best estimated at the surface of three SCAs corresponding to the center of C-1 core using the least square method by employing LSL-M2 package. These results were compared with fast neutron flux calculated using DOT3.5 code and both results are within good agreement of ±20% acceptance criteria as described in Regulatory Guide 1.190. Therefore, calculational model was validated by dosimetry evaluation and these results can be used in the life assessment of CHASNUPP Unit-1 pressure vessel belt line region.
Large quantities of weapon grade (WG) plutonium have been accumulated in the nuclear warheads. Pl... more Large quantities of weapon grade (WG) plutonium have been accumulated in the nuclear warheads. Plutonium and heavy water moderator can give a good combination with respect to neutron economy. TRISO type fuel can withstand very high fuel burn up levels. The paper investigates the prospects of utilization of TRISO fuel made of WG-plutonium in CANDU reactors. Three different fuel compositions have been investigated: 1: 90 % ThC + 10 % PuC, 2: 70 % ThC + 30 % PuC and 3: 50 % ThC + 50 % PuC. The temporal variation of the criticality k∞ and the burn-up values of the reactor have been calculated by full power operation up to 17 years. Calculated startup criticalities for these fuel modes are k∞,0 = 1.6403, 1.7228, 1.7662, respectively. Attainable burn up values and reactor operation times with the same fuel charge will be 94 700, 265 000, 425 000 MW.D/MT and ~ 3.5, 10, 17 years, respectively. These high burn ups would reduce fuel fabrication costs and nuclear waste mass for final disposal per unit energy drastically.
The chapter presents a general overview on nuclear fission energy production. Chapter starts with... more The chapter presents a general overview on nuclear fission energy production. Chapter starts with the fundamentals of the fission process and fission energy production. Importance of the neutrons in nuclear reactor technology is highlighted. The Boltzmann neutron transport equation is explicitly evaluated. Conventional and advanced nuclear reactors are addressed. Major Gen-IV nuclear reactor types are reported. Finally, major space nuclear reactor systems are investigated. The importance of the space nuclear reactors for the commercial exploitation of the solar system is pointed out.
In a source-driven fissionable blanket, a flat fission power density (FPD) is achieved by using a... more In a source-driven fissionable blanket, a flat fission power density (FPD) is achieved by using a mixed fuel (Th0 2 and natural U0 2) with the thorium/uranium ratio changing from front to back in the ten fuel rows along the radial direction. A straightforward graphic method is used. The temporal behavior of the FPD has been observed for an operation period of 6 months and for a plant load factor of 75% by applying a fusion driver neutron flux of 10 14 14-MeV neutrons/(cm 2-s) at the first wall, corresponding to ~2.25 MW/m 2. To keep the power density flat, it is necessary to replace the fuel in rows 1, 2, and 3, close to the first wall. The time intervals for this operation increase, counting from initial start-up, typically, 2 months, 6 months, etc. One result of this study is that plutonium produced in such a hybrid blanket contains very low amounts of even isotopic components even over very long operation times of ~3 yr. Hence, if fusion reactors are introduced into the energy market, special regulations are needed for international safeguarding.
Preliminary study on influence of changing void fraction (VF) on SUPEL (Straight Utilization of s... more Preliminary study on influence of changing void fraction (VF) on SUPEL (Straight Utilization of sPEnt LWR fuel in LWR system) scenario for boiling water reactor (BWR) spent fuel direct recycling scheme has been carried out. Several VF values of BWR have been investigated to determine the criticality of reactor. The VF values range from 20% to 60%. The fraction of spent fuel to the total loaded fuel was changed from 5% to 20%. The required uranium enrichment for criticality becomes higher with the increasing of VF as well as the enlarging of the fraction of spent fuel in loaded fuel. The neutron spectra become harder with the augmenting of VF. The plutonium and minor actinides isotopes are produced more in the reactor.
Energy Citations Database. ADOPT-A-DOC? ... Title, Control mechanism for spacecraft thermionic re... more Energy Citations Database. ADOPT-A-DOC? ... Title, Control mechanism for spacecraft thermionic reactors leading to lower radiation shielding mass. Creator/Author, Sahin, S. Publication Date, 1979 Jan 01. OSTI Identifier, OSTI ID: 5351674. Report Number(s), CONF-791103-. ...
... King Saud University, POBox 800, Riyadh 11421, Saudi Arabia On leave: University of Erciyes, ... more ... King Saud University, POBox 800, Riyadh 11421, Saudi Arabia On leave: University of Erciyes, Kayseri, Turkey CHIEN CHUNG National Tsing Hua ... foils of 93.15% enriched 235U (3mg each)ve been irradiated with thermal neutrons at a thermal pow r of 500 watts, to measure ...
Abstract Criticality and burn up behavior of the Fixed Bed Nuclear Reactor (FBNR) are investigate... more Abstract Criticality and burn up behavior of the Fixed Bed Nuclear Reactor (FBNR) are investigated for the mixed fuel 233UO2/ThO2 as an alternative to low enriched 235UO2 fuel. CERMET fuel with a zirconium matrix and cladding has been used throughout the study. The main results of the study can be summarized as follows: Reactor criticality is already achieved by ∼2% 233UO2 with the mixed 233UO2/ThO2 fuel. At higher 233U fractions, reactor criticality rises rapidly and exceeds keff > 1.5 already by 9% 233UO2. With 100% 233UO2, start up criticality can reach keff = 2.0975. Time dependent reactor criticality keff and fuel burn up have been investigated for two different mixed fuel 233UO2/ThO2 compositions, namely: 4% 233UO2 + 96% ThO2 for a reactor power of 40 MWel (120 MWth) and 9% 233UO2 + 91% ThO2 for a reactor power of 70 MWel (210 MWth). Sufficient reactor criticality (keff > 1.06) for continuous operation without fuel change can be sustained during ∼5 and 12 years with 4% and 9% 233UO2 fractions in the mixed fuel, leading to burn ups of ∼36000 and >105000MWD/t, respectively. Thorium based fuel produces no prolific uranium. Plutonium production remains negligible.
The possibility of spent nuclear fuel rejuvenation in fusion reactors is investigated for both (D... more The possibility of spent nuclear fuel rejuvenation in fusion reactors is investigated for both (D,T) and catalysed (D,D) modes. The analysis is conducted for light water reactor (LWR) spent nuclear fuel which was used up to a total enrichment grade (fissionable plutonium plus 235U) of 2.174% The behaviour of the spent fuel is observed during 48 months for discrete time intervals of At = 6 months. The cooling of the fissile fuel zone is considered with three different coolants, namely pressurized helium ('gas'), Flibe (Li2BeF4) ('molten salt') and natural lithium ('liquid metal'). The best rejuvenation performance is obtained with Flibe, followed by helium and natural lithium. The catalysed (D,D) mode has a better rejuvenation capability than the conventional (D,T) mode. A rejuvenation time of 12 months is evaluated for a final fissile fuel enrichment grade of around 3.5% with Flibe coolant in the fissile zone under a first-wall fusion neutron current load of 1014 (2.45 MeVncm-2s) and 1014 (14.1 MeV ncm-2s), corresponding to 2.64 MWm-2 by a plant factor of 75% for the catalysed (D,D) fusion reactor mode. The rejuvenation time increases to 24 months for the same fissile fuel enrichment grade using the (D,T) fusion reactor mode under a first-wall fusion neutron current load of 1014 (2.45 MeV ncm-2 s), corresponding to 2.25 MWm-2 by a
... REDUCED SHIELDING MASS FOR THE VISTA SPACECRAFT Sümer Şahin* and Hacı MehmetŞahin Gazi Üniver... more ... REDUCED SHIELDING MASS FOR THE VISTA SPACECRAFT Sümer Şahin* and Hacı MehmetŞahin Gazi Üniversitesi, Teknik Eğitim Fakültesi Beşevler, Ankara 06503, Turkey and ... Page 2. Sümer Şahin, H. Mehmet Şahin, and Abdulmuttalip Şahinaslan ...
A thermo-hydrodynamic-neutronic analysis is performed for a fast, uranium carbide (UC) fueled spa... more A thermo-hydrodynamic-neutronic analysis is performed for a fast, uranium carbide (UC) fueled space-craft nuclear in-core thermionic reactor. The thermo-hydrodynamic analysis shows that a hybrid thermionic spacecraft nuclear reactor can be designed for both electricity generation and nuclear thermal propulsion purposes. The neutronic analysis has been conducted in S[sub 8]-P[sub 3] approximation with the help of one- and two-dimensional neutron transport codes ANISN and DORT, respectively. The calculations have shown that a UC fueled electricity generating single mode thermionic nuclear reactor can be designed to be extremely compact because of the high atomic density of the nuclear fuel (by 95% sintering density), namely, with a core radius of 8.7 cm and core height of 25 cm, leading to power levels as low as 5 kW (electric) by an electrical output on an emitter surface of 1.243 W/cm[sup 2]. A reactor control with boronated reflector drums at the outer periphery of the radial reflector of 16-cm thickness would make possible reactivity changes of [Delta]k[sub eff] > 10% -- amply sufficient for a fast reactor -- without a significant distortion of the fission power profile during all phases of the space mission. The hybrid thermionic spacecraft nuclear reactor mode contains cooling channels inmore » the nuclear fuel for the hydrogen propellant. This increase the critical reactor size because of the lower uranium atomic density in this design concept. Calculations have lead to a reactor with a core radius of 22 cm and core height of 35 cm leading to power levels [approximately] 50 kW(electric) under the aforementioned thermionic conversion conditions.« less
A thermo-hydrodynamic-neutronic analysis is performed for a fast, uranium carbide (UC) fueled spa... more A thermo-hydrodynamic-neutronic analysis is performed for a fast, uranium carbide (UC) fueled space-craft nuclear in-core thermionic reactor. The thermo-hydrodynamic analysis shows that a hybrid thermionic spacecraft nuclear reactor can be designed for both electricity generation and nuclear thermal propulsion purposes. The neutronic analysis has been conducted in S[sub 8]-P[sub 3] approximation with the help of one- and two-dimensional neutron transport codes ANISN and DORT, respectively. The calculations have shown that a UC fueled electricity generating single mode thermionic nuclear reactor can be designed to be extremely compact because of the high atomic density of the nuclear fuel (by 95% sintering density), namely, with a core radius of 8.7 cm and core height of 25 cm, leading to power levels as low as 5 kW (electric) by an electrical output on an emitter surface of 1.243 W/cm[sup 2]. A reactor control with boronated reflector drums at the outer periphery of the radial reflector of 16-cm thickness would make possible reactivity changes of [Delta]k[sub eff] > 10% -- amply sufficient for a fast reactor -- without a significant distortion of the fission power profile during all phases of the space mission. The hybrid thermionic spacecraft nuclear reactor mode contains cooling channels inmore » the nuclear fuel for the hydrogen propellant. This increase the critical reactor size because of the lower uranium atomic density in this design concept. Calculations have lead to a reactor with a core radius of 22 cm and core height of 35 cm leading to power levels [approximately] 50 kW(electric) under the aforementioned thermionic conversion conditions.« less
Abstract The aim of this paper is to determine neutronic performances of the light water reactor ... more Abstract The aim of this paper is to determine neutronic performances of the light water reactor (LWR) spent fuel mixed with fertile thorium fuel in a FFHR. Time dependent three dimensional calculations for major technical data, such as blanket energy multiplication, tritium breeding ratio, cumulative fissile fuel enrichment and burnup have been performed by using Monte Carlo Neutron-Particle Transport code MCNP5 1.4, coupled with a novel interface code MCNPAS, which is developed by our research group. A self-sustaining tritium breeding ratio (TBR>1.05) has been kept throughout the calculations. The study has shown that the fissile fuel quality will be improved in the course of the transmutation of the LWR spent in the FFHR. The latter has gained the reusable fuel enrichment level conventional LWRs between one and two years. Furthermore, LWR spent fuel - thorium mixture provides higher burn-up values than in light water reactors.
Science and Technology of Nuclear Installations, 2016
Emergency planning zones (PAZ and UPZ) around the Karachi-2 and Karachi-3 nuclear power plants (K... more Emergency planning zones (PAZ and UPZ) around the Karachi-2 and Karachi-3 nuclear power plants (K-2/K-3 NPPs) have been realistically determined by employing Gaussian puff model and Gaussian plume model together for atmospheric transport, diffusion, and deposition of radioactive material using onsite and regional data related to meteorology, topography, and land-use along with latest IAEA Post-Fukushima Guidelines. The analysis work has been carried out using U.S.NRC computer code RASCAL 4.2. The assumed environmental radioactive releases provide the sound theoretical and practical bases for the estimation of emergency planning zones covering most expected scenario of severe accident and most recent multiunit Fukushima Accident. Sheltering could be used as protective action for longer period of about 04 days. The area about 3 km of K-2/K-3 NPPs site should be evacuated and an iodine thyroid blocking agent should be taken before release up to about 14 km to prevent severe determinist...
Abstract Neutron flux and energy spectrum were determined at the surface of three in-vessel Surve... more Abstract Neutron flux and energy spectrum were determined at the surface of three in-vessel Surveillance Capsule Assemblies (SCAs) removed from CHASNUPP Unit-1 after 2nd, 4th, and 9th fuel cycles for the life assessment of reactor pressure vessel belt line region. Dosimetry data were measured from radiometric sensors irradiated in base material section of SCAs. Fast neutron flux (E > 1.0 MeV) was best estimated at the surface of three SCAs corresponding to the center of C-1 core using the least square method by employing LSL-M2 package. These results were compared with fast neutron flux calculated using DOT3.5 code and both results are within good agreement of ±20% acceptance criteria as described in Regulatory Guide 1.190. Therefore, calculational model was validated by dosimetry evaluation and these results can be used in the life assessment of CHASNUPP Unit-1 pressure vessel belt line region.
Large quantities of weapon grade (WG) plutonium have been accumulated in the nuclear warheads. Pl... more Large quantities of weapon grade (WG) plutonium have been accumulated in the nuclear warheads. Plutonium and heavy water moderator can give a good combination with respect to neutron economy. TRISO type fuel can withstand very high fuel burn up levels. The paper investigates the prospects of utilization of TRISO fuel made of WG-plutonium in CANDU reactors. Three different fuel compositions have been investigated: 1: 90 % ThC + 10 % PuC, 2: 70 % ThC + 30 % PuC and 3: 50 % ThC + 50 % PuC. The temporal variation of the criticality k∞ and the burn-up values of the reactor have been calculated by full power operation up to 17 years. Calculated startup criticalities for these fuel modes are k∞,0 = 1.6403, 1.7228, 1.7662, respectively. Attainable burn up values and reactor operation times with the same fuel charge will be 94 700, 265 000, 425 000 MW.D/MT and ~ 3.5, 10, 17 years, respectively. These high burn ups would reduce fuel fabrication costs and nuclear waste mass for final disposal per unit energy drastically.
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