The integration of fast reactor to the fuel cycle in Slovakia

International Nuclear Information System (INIS)

Zajac, R.; Darilek, P.; Necas, V.

2009-01-01

A very topical problem of nuclear power is the fuel cycle back-end. One of the options is a LWR spent fuel reprocessing and a fissile nuclides re-use in the fast reactor. A large amount of spent fuel has been stored in the power plant intermediate storage during the operation of WWER-440 reactors in Slovakia. This paper is based on an analysis of Pu and minor actinides content in actual WWER-440 spent fuel stored in Slovakia. The next part presents the possibilities of reprocessing and Pu re-use in fast reactor under Slovak conditions. The fuel cycle consisting of the WWER-440 reactor, PUREX reprocessing plant and a sodium fast reactor was designed. The last section compares two parts of this fuel cycle: one is UOX cycle in WWER-440 reactor and the other is cycle in the fast reactor - SUPER PHENIX loaded with MOX fuel (Pu + Minor Actinides). The starting point is a single recycling of Pu from WWER-440 in the fission products. The next step is multi recycling of Pu in the fission products to obtain equilibrium cycle. This article is dealing with the solution of power production and fuel cycle indicators. All kinds of calculations were performed by computer code HELIOS 1.10. (Authors)

The integration of fast reactor to the fuel cycle in Slovakia

International Nuclear Information System (INIS)

Zajac, R.; Darilek, P.; Necas, V.

2009-01-01

A very topical problem of nuclear power is the fuel cycle back-end. One of the options is a LWR spent fuel reprocessing and a fissile nuclides re-use in the fast reactor. A large amount of spent fuel has been stored in the power plant intermediate storage during the operation of VVER-440 reactors in Slovakia. This paper is based on an analysis of Pu and minor actinides content in actual VVER-440 spent fuel stored in Slovakia. The next part presents the possibilities of reprocessing and Pu re-use in fast reactor under Slovak conditions. The fuel cycle consisting of the VVER-440 reactor, PUREX reprocessing plant and a sodium fast reactor was designed. The last section compares two parts of this fuel cycle: one is UOX cycle in VVER-440 reactor and the other is cycle in the fast reactor - SUPER PHENIX loaded with MOX fuel (Pu + Minor Actinides). The starting point is a single recycling of Pu from VVER-440 in the FR. The next step is multirecycling of Pu in the FR to obtain equilibrium cycle. This article is dealing with the solution of power production and fuel cycle indicators. All kinds of calculations were performed by computer code HELIOS 1.10. (authors)

Modified-open fuel cycle performance with breed-and-burn advanced reactor concepts

International Nuclear Information System (INIS)

Heidet, Florent; Kim, Taek K.; Taiwo, Temitope A.

2011-01-01

Recent advances in fast reactor designs enable significant increase in the uranium utilization in an advanced fuel cycle. The category of fast reactors, collectively termed breed-and-burn reactor concepts, can use a large amount of depleted uranium as fuel without requiring enrichment with the exception of the initial core critical loading. Among those advanced concepts, some are foreseen to operate within a once-through fuel cycle such as the Traveling Wave Reactor, CANDLE reactor or Ultra-Long Life Fast Reactor, while others are intended to operate within a modified-open fuel cycle, such as the Breed-and-Burn reactor and the Energy Multiplier Module. This study assesses and compares the performance of the latter category of breed-and-burn reactors at equilibrium state. It is found that the two reactor concepts operating within a modified-open fuel cycle can significantly improve the sustainability and security of the nuclear fuel cycle by decreasing the uranium resources and enrichment requirements even further than the breed-and-burn core concepts operating within the once-through fuel cycle. Their waste characteristics per unit of energy are also found to be favorable, compared to that of currently operating PWRs. However, a number of feasibility issues need to be addressed in order to enable deployment of these breed-and-burn reactor concepts. (author)

Core-state models for fuel management of equilibrium and transition cycles in pressurized water reactors

International Nuclear Information System (INIS)

Aragones, J.M.; Martinez-Val, J.M.; Corella, M.R.

1977-01-01

Fuel management requires that mass, energy, and reactivity balance be satisfied in each reload cycle. Procedures for selection of alternatives, core-state models, and fuel cost calculations have been developed for both equilibrium and transition cycles. Effective cycle lengths and fuel cycle variables--namely, reload batch size, schedule of incore residence for the fuel, feed enrichments, energy sharing cycle by cycle, and discharge burnup and isotopics--are the variables being considered for fuel management planning with a given energy generation plan, fuel design, recycling strategy, and financial assumptions

Neutronic investigations of an equilibrium core for a tight-lattice light water reactor

International Nuclear Information System (INIS)

Broeders, C.H.M.

1992-01-01

Calculation procedures and first results concerning the neutronic design of an equilibrium core of an advanced pressurized water reactor (APWR) with mixed oxide fuel in a compact light water moderated triangular lattice are presented. Principle and qualification of the cell burnup calculations with the KARBUS program are briefly discussed. The fuel assembly design with single control rod positions filled with control rod material or coolant water requires special transport theory calculations, which are performed with a one-dimensional supercell model. The macroscopic fuel assembly cross section data is collected in a special library to be used in a new calculational procedure, ARCOSI, for multi-cycle reactor core simulations. Its first application for a reference design resulted in an equilibrium configuration with moderator density reactivity coefficients which are satisfactory as regards safety. (orig.) [de

Growth scenarios with thorium fuel cycles in pressurised heavy water reactors

International Nuclear Information System (INIS)

Balakrishnan, M.R.

1991-01-01

Since India has generous deposits of thorium, the availability of thorium will not be a limiting factor in any growth scenario. It is fairly well accepted that the best system for utilisation of thorium is the heavy water reactor. The growth scenarios possible using thorium in HWRs are considered. The base has been taken as 50,000 tons of natural uranium and practically unlimited thorium. The reference reactor has been assumed to be the PHWR, and all other growth scenarios are compared with the growth scenario provided by the once-through natural cycle in the PHWR. Two reactor types have been considered: the heavy water moderated, heavy water cooled, pressure tube reactor, known as the PHWR; and the heavy water moderated and cooled pressure vessel kind, similar to the ATUCHA reactor in Argentina. For each reactor, a number of different fuel cycles have been studied. All these cycles have been based on thorium. These are: the self-sustaining equilibrium thorium cycle (SSET); the high conversion ratio high burnup cycle; and the once through thorium cycle (OTT). The cycle have been initiated in two ways: one is by starting the cycle with natural uranium, reprocessing the spent fuel to obtain plutonium, and use that plutonium to initiate the thorium cycle; the other is to enrich the uranium to about 2-3% U-235 (the so-called Low Enriched Uranium or LEU), and use the LEU to initiate the thorium cycle. Both cases have been studied, and growth scenarios have been projected for every one of the possible combinations. (author). 1 tab

Prospects of subcritical molten salt reactor for minor actinides incineration in closed fuel cycle

Energy Technology Data Exchange (ETDEWEB)

Alekseev, Pavel N.; Balanin, Andrey L.; Dudnikov, Anatoly A.; Fomichenko, Petr A.; Nevinitsa, Vladimir A.; Frolov, Aleksey A.; Lubina, Anna S.; Sedov, Aleksey A.; Subbotin, Aleksey S.; Blandinsky, Viktor Yu. [Nuclear Research Centre ' ' Kurchatov Institute' ' , Moscow (Russian Federation)

2015-09-15

A subcritical molten salt reactor is proposed for minor actinides (separated from spent fuel VVER-1000 light water reactor) incineration and for {sup 233}U conversion from {sup 232}Th. Here the subcritical molten salt reactor with fuel composition of heavy nuclide fluorides in molten LiF - NaF - KF salt and with external neutron source, based on 1 GeV proton accelerator and molten salt cooled tungsten target is considered. The paper presents the results of parametrical analysis of equilibrium nuclide composition of molten salt reactor with minor actinides feed in dependence of core dimensions, average neutron flux and external neutron source intensity. Reactor design is defined; requirements to external neutron source are posed; heavy nuclides equilibrium and fuel cycle main parameters are calculated.

The procedure and results of calculations of the equilibrium isotopic composition of a demonstration subcritical molten salt reactor

Energy Technology Data Exchange (ETDEWEB)

Nevinitsa, V. A., E-mail: Neviniza-VA@nrcki.ru; Dudnikov, A. A.; Blandinskiy, V. Yu.; Balanin, A. L.; Alekseev, P. N. [National Research Centre Kurchatov Institute (Russian Federation); Titarenko, Yu. E.; Batyaev, V. F.; Pavlov, K. V.; Titarenko, A. Yu., E-mail: yuri.titarenko@itep.ru [Institute for Theoretical and Experimental Physics (Russian Federation)

2015-12-15

A subcritical molten salt reactor with an external neutron source is studied computationally as a facility for incineration and transmutation of minor actinides from spent nuclear fuel of reactors of VVER-1000 type and for producing {sup 233}U from {sup 232}Th. The reactor configuration is chosen, the requirements to be imposed on the external neutron source are formulated, and the equilibrium isotopic composition of heavy nuclides and the key parameters of the fuel cycle are calculated.

Design of an equilibrium nucleus of a BWR type reactor based in a Thorium-Uranium fuel

International Nuclear Information System (INIS)

Francois, J.L.; Nunez C, A.

2003-01-01

In this work the design of the reactor nucleus of boiling water using fuel of thorium-uranium is presented. Starting from an integral concept based in a type cover-seed assemble is carried out the design of an equilibrium reload for the nucleus of a reactor like that of the Laguna Verde Central and its are analyzed some of the main design variables like the cycle length, the reload fraction, the burnt fuel, the vacuum distribution, the generation of lineal heat, the margin of shutdown, as well as a first estimation of the fuel cost. The results show that it is feasible to obtain an equilibrium reload, comparable to those that are carried out in the Laguna Verde reactors, with a good behavior of those analyzed variables. The cost of the equilibrium reload designed with the thorium-uranium fuel is approximately 2% high that the uranium reload producing the same energy. It is concluded that it is convenient to include burnable poisons, type gadolinium, in the fuel with the end of improving the reload design, the fuel costs and the margin of shutdown. (Author)

Economics analysis of fuel cycle cost of fusion–fission hybrid reactors based on different fuel cycle strategies

Energy Technology Data Exchange (ETDEWEB)

Zu, Tiejun, E-mail: tiejun@mail.xjtu.edu.cn; Wu, Hongchun; Zheng, Youqi; Cao, Liangzhi

2015-01-15

Highlights: • Economics analysis of fuel cycle cost of FFHRs is carried out. • The mass flows of different fuel cycle strategies are established based on the equilibrium fuel cycle model. • The levelized fuel cycle costs of different fuel cycle strategies are calculated, and compared with current once-through fuel cycle. - Abstract: The economics analysis of fuel cycle cost of fusion–fission hybrid reactors has been performed to compare four fuel cycle strategies: light water cooled blanket burning natural uranium (Strategy A) or spent nuclear fuel (Strategy B), sodium cooled blanket burning transuranics (Strategy C) or minor actinides (Strategy D). The levelized fuel cycle costs (LFCC) which does not include the capital cost, operation and maintenance cost have been calculated based on the equilibrium mass flows. The current once-through (OT) cycle strategy has also been analyzed to serve as the reference fuel cycle for comparisons. It is found that Strategy A and Strategy B have lower LFCCs than OT cycle; although the LFCC of Strategy C is higher than that of OT cycle when the uranium price is at its nominal value, it would become comparable to that of OT cycle when the uranium price reaches its historical peak value level; Strategy D shows the highest LFCC, because it needs to reprocess huge mass of spent nuclear fuel; LFCC is sensitive to the discharge burnup of the nuclear fuel.

Increased fuel burn-up and fuel cycle equilibrium

International Nuclear Information System (INIS)

Debes, M.

2001-01-01

Improvement of nuclear competitiveness will rely mainly on increased fuel performance, with higher burn-up, and reactors sustained life. Regarding spent fuel management, the EDF current policy relies on UO 2 fuel reprocessing (around 850 MTHM/year at La Hague) and MOX recycling to ensure plutonium flux adequacy (around 100 MTHM/year, with an electricity production equivalent to 30 TWh). This policy enables to reuse fuel material, while maintaining global kWh economy with existing facilities. It goes along with current perspective to increase fuel burn-up up to 57 GWday/t mean in 2010. The following presentation describes the consequences of higher fuel burn-up on fuel cycle and waste management and implementation of a long term and global equilibrium for decades in spent fuel management resulting from this strategy. (author)

Efficient cycles for carbon capture CLC power plants based on thermally balanced redox reactors

KAUST Repository

Iloeje, Chukwunwike

2015-10-01

© 2015 Elsevier Ltd. The rotary reactor differs from most alternative chemical looping combustion (CLC) reactor designs because it maintains near-thermal equilibrium between the two stages of the redox process by thermally coupling channels undergoing oxidation and reduction. An earlier study showed that this thermal coupling between the oxidation and reduction reactors increases the efficiency by up to 2% points when implemented in a regenerative Brayton cycle. The present study extends this analysis to alternative CLC cycles with the objective of identifying optimal configurations and design tradeoffs. Results show that the increased efficiency from reactor thermal coupling applies only to cycles that are capable of exploiting the increased availability in the reduction reactor exhaust. Thus, in addition to the regenerative cycle, the combined CLC cycle and the combined-regenerative CLC cycle are suitable for integration with the rotary reactor. Parametric studies are used to compare the sensitivity of the different cycle efficiencies to parameters like pressure ratio, turbine inlet temperature, carrier-gas fraction and purge steam generation. One of the key conclusions from this analysis is that while the optimal efficiency for regenerative CLC cycle was the highest of the three (56% at 3. bars, 1200. °C), the combined-regenerative cycle offers a trade-off that combines a reasonably high efficiency (about 54% at 12. bars, 1200. °C) with much lower gas volumetric flow rate and consequently, smaller reactor size. Unlike the other two cycles, the optimal compressor pressure ratio for the regenerative cycle is weakly dependent on the design turbine inlet temperature. For the regenerative and combined regenerative cycles, steam production in the regenerator below 2× fuel flow rate improves exhaust recovery and consequently, the overall system efficiency. Also, given that the fuel side regenerator flow is unbalanced, it is more efficient to generate steam from the

Economic Analysis of Symbiotic Light Water Reactor/Fast Burner Reactor Fuel Cycles Proposed as Part of the U.S. Advanced Fuel Cycle Initiative (AFCI)

International Nuclear Information System (INIS)

Williams, Kent Alan; Shropshire, David E.

2009-01-01

A spreadsheet-based 'static equilibrium' economic analysis was performed for three nuclear fuel cycle scenarios, each designed for 100 GWe-years of electrical generation annually: (1) a 'once-through' fuel cycle based on 100% LWRs fueled by standard UO2 fuel assemblies with all used fuel destined for geologic repository emplacement, (2) a 'single-tier recycle' scenario involving multiple fast burner reactors (37% of generation) accepting actinides (Pu,Np,Am,Cm) from the reprocessing of used fuel from the uranium-fueled LWR fleet (63% of generation), and (3) a 'two-tier' 'thermal+fast' recycle scenario where co-extracted U,Pu from the reprocessing of used fuel from the uranium-fueled part of the LWR fleet (66% of generation) is recycled once as full-core LWR MOX fuel (8% of generation), with the LWR MOX used fuel being reprocessed and all actinide products from both UO2 and MOX used fuel reprocessing being introduced into the closed fast burner reactor (26% of generation) fuel cycle. The latter two 'closed' fuel cycles, which involve symbiotic use of both thermal and fast reactors, have the advantages of lower natural uranium requirements per kilowatt-hour generated and less geologic repository space per kilowatt-hour as compared to the 'once-through' cycle. The overall fuel cycle cost in terms of $ per megawatt-hr of generation, however, for the closed cycles is 15% (single tier) to 29% (two-tier) higher than for the once-through cycle, based on 'expected values' from an uncertainty analysis using triangular distributions for the unit costs for each required step of the fuel cycle. (The fuel cycle cost does not include the levelized reactor life cycle costs.) Since fuel cycle costs are a relatively small percentage (10 to 20%) of the overall busbar cost (LUEC or 'levelized unit electricity cost') of nuclear power generation, this fuel cycle cost increase should not have a highly deleterious effect on the competitiveness of nuclear power. If the reactor life cycle

Decay profiles of β and γ for a radionuclide inventory in equilibrium cycle of a BWR type reactor

International Nuclear Information System (INIS)

Salaices, M.; Sandoval, S.; Ovando, R.

2007-01-01

Presently work the β and γ radiation decay profiles for a radionuclides inventory in equilibrium cycle of a BWR type reactor is presented. The profiles are presented in terms of decay in the activity of the total inventory as well as of the chemical groups that conform the inventory. In the obtaining of the radionuclides inventory in equilibrium cycle the ORIGEN2 code, version 1 was used, which simulates fuel burnup cycles and it calculates the evolution of the isotopic composition as a result of the burnt one, irradiation and decay of the nuclear fuel. It can be observed starting from the results that the decrease in the activity for the initial inventory and the different chemical groups that conform it is approximately proportional to the base 10 logarithm of the time for the first 24 hours of having concluded the burnt one. It can also be observed that the chemical groups that contribute in more proportion to the total activity of the inventory are the lanthanides-actinides and the transition metals, with 39% and 28%, respectively. The groups of alkaline earth metals, halogens, metalloids, noble gases and alkaline metals, contribute with percentages that go from the 8 to 5%. The groups that less they contribute to the total activity of the inventory they are the non metals and semi-metals with smaller proportions that 1%. The chemical groups that more contribute to the energy of β and γ radiation its are the transition metals and the lanthanides-actinides with a change in the order of importance at the end of the 24 hours period. The case of the halogens is of relevance for the case of the γ radiation energy due that occupying the very near third site to the dimensions of the two previous groups. Additionally, the decay in the activity for the total inventory and the groups that conform it can be simulated by means of order 6 polynomials or smaller than describe its behavior appropriately. The results presented in this work, coupled to a distribution model

Application of load follow operation to equilibrium cycle of OPR1000

Energy Technology Data Exchange (ETDEWEB)

Lee, Kyeongju; Choe, Jiwon; Lee, Deokjung [Interdisciplinary School of Green Energy, Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

2012-10-15

All nuclear power plants in Korea are operated at a base load, that is 100% rated power, and do not rely largely on power tracking control except for startup, shutdown, and some minor problem occurrences. However, if the electricity from nuclear power plants exceeds 50% of total electricity generation according to national energy plan, load follow operation is necessary to efficiently use the electrical energy. But it is very difficult to control the axial power distribution and reactor core reactivity at the same time as needed because of variations in nuclear system parameters. In 1990s, an advanced reactor control algorithm, Mode-K, was developed which uses regulation banks, boron control, and a heavy-worth bank (H-bank). The regulation banks and boron control are used for core reactivity control and the H-bank is used for the control of axial power shape. In this study, reactor core simulations with HELIOS/MASTER code system using Mode-K strategy are applied to the daily load follow operation in equilibrium cycle of OPR1000.

Modeling minor actinide multiple recycling in a lead-cooled fast reactor to demonstrate a fuel cycle without long-lived nuclear waste

Directory of Open Access Journals (Sweden)

Stanisz Przemysław

2015-09-01

Full Text Available The concept of closed nuclear fuel cycle seems to be the most promising options for the efficient usage of the nuclear energy resources. However, it can be implemented only in fast breeder reactors of the IVth generation, which are characterized by the fast neutron spectrum. The lead-cooled fast reactor (LFR was defined and studied on the level of technical design in order to demonstrate its performance and reliability within the European collaboration on ELSY (European Lead-cooled System and LEADER (Lead-cooled European Advanced Demonstration Reactor projects. It has been demonstrated that LFR meets the requirements of the closed nuclear fuel cycle, where plutonium and minor actinides (MA are recycled for reuse, thereby producing no MA waste. In this study, the most promising option was realized when entire Pu + MA material is fully recycled to produce a new batch of fuel without partitioning. This is the concept of a fuel cycle which asymptotically tends to the adiabatic equilibrium, where the concentrations of plutonium and MA at the beginning of the cycle are restored in the subsequent cycle in the combined process of fuel transmutation and cooling, removal of fission products (FPs, and admixture of depleted uranium. In this way, generation of nuclear waste containing radioactive plutonium and MA can be eliminated. The paper shows methodology applied to the LFR equilibrium fuel cycle assessment, which was developed for the Monte Carlo continuous energy burnup (MCB code, equipped with enhanced modules for material processing and fuel handling. The numerical analysis of the reactor core concerns multiple recycling and recovery of long-lived nuclides and their influence on safety parameters. The paper also presents a general concept of the novel IVth generation breeder reactor with equilibrium fuel and its future role in the management of MA.

Equilibrium core layout for the 1000 MW direct cycle HTR (HHT) with hexagonal monolith moulded fuel blocks

Energy Technology Data Exchange (ETDEWEB)

Dworak, A

1973-03-15

The aim of this survey is to calculate an equilibrium Thorium fuel cycle for a 1000 MW HHT-core in off-load refuelling with hexagonal monolith moulded fuel blocks. It was tried to achieve an axial power distribution similar to the advanced pebble-bed reactors (OTTO) by introducing three axial core zones with different heavy metal content and initial enrichment.

Simultaneous loading patterns optimization for two successive cycles of pressurized water reactors

International Nuclear Information System (INIS)

Yamamoto, Akio; Sugimura, Erina; Kitamura, Yasunori; Yamane, Yoshihiro

2004-01-01

In this paper, simultaneous optimization is carried out for successive two cycles of pressurized water reactors. At first, a simplified problem of the simultaneous optimization was studied by assuming the batch-wise power sharing as independent variable, i.e., batch-wise power sharing was optimized without considering corresponding loading patterns. The optimization of the batch-wise power sharing was carried out for the conventional single cycle, the equilibrium cycle and the two successive (tandem) cycles. The analysis indicated that the tandem cycle optimization well reproduce that of the equilibrium cycle optimization, which is considered as a typical case of the true multicycle optimization. Next, simultaneous optimization of loading patterns for tandem cycles is carried out using the simulated annealing method. Since the design space of the tandem cycles optimization is much larger than that of the conventional single cycle optimization, the optimization condition (i.e., number of calculated patterns) are established through sensitivity study. The optimization results are compared with those obtained by the successive single cycle optimizations and it is clarified that the successive single cycle optimization well reproduces the optimization results obtained by the simultaneous optimization if objective functions are appropriately chosen. The above result will be encouraging for the current in-core optimization method since single cycle optimization is utilized due to limitation of computation time. (author)

Void reactivity decomposition for the Sodium-cooled Fast Reactor in equilibrium fuel cycle

Energy Technology Data Exchange (ETDEWEB)

Sun Kaichao, E-mail: kaichao.sun@psi.ch [Paul Scherrer Institut (PSI), 5232 Villigen PSI (Switzerland); Ecole Polytechnique Federale de Lausanne (EPFL), 1015 Lausanne (Switzerland); Krepel, Jiri; Mikityuk, Konstantin; Pelloni, Sandro [Paul Scherrer Institut (PSI), 5232 Villigen PSI (Switzerland); Chawla, Rakesh [Paul Scherrer Institut (PSI), 5232 Villigen PSI (Switzerland); Ecole Polytechnique Federale de Lausanne (EPFL), 1015 Lausanne (Switzerland)

2011-07-15

Highlights: > We analyze the void reactivity effect for three ESFR core fuel cycle states. > The void reactivity effect is decomposed by neutron balance method. > Novelly, the normalization to the integral flux in the active core is applied. > The decomposition is compared with the perturbation theory based results. > The mechanism and the differences of the void reactivity effect are explained. - Abstract: The Sodium-cooled Fast Reactor (SFR) is one of the most promising Generation IV systems with many advantages, but has one dominating neutronic drawback - a positive sodium void reactivity. The aim of this study is to develop and apply a methodology, which should help better understand the causes and consequences of the sodium void effect. It focuses not only on the beginning-of-life (BOL) state of the core, but also on the beginning of open and closed equilibrium (BOC and BEC, respectively) fuel cycle conditions. The deeper understanding of the principal phenomena involved may subsequently lead to appropriate optimization studies. Various voiding scenarios, corresponding to different spatial zones, e.g. node or assembly, have been analyzed, and the most conservative case - the voiding of both inner and outer fuel zones - has been selected as the reference scenario. On the basis of the neutron balance method, the corresponding SFR void reactivity has been decomposed reaction-, isotope-, and energy-group-wise. Complementary results, based on generalized perturbation theory and sensitivity analysis, are also presented. The numerical analysis for both neutron balance and perturbation theory methods has been carried out using appropriate modules of the ERANOS code system. A strong correlation between the flux worth, i.e. the product of flux and adjoint flux, and the void reactivity importance distributions has been found for the node- and assembly-wise voiding scenarios. The neutron balance based decomposition has shown that the void effect is caused mainly by the

Void reactivity decomposition for the Sodium-cooled Fast Reactor in equilibrium fuel cycle

International Nuclear Information System (INIS)

Sun Kaichao; Krepel, Jiri; Mikityuk, Konstantin; Pelloni, Sandro; Chawla, Rakesh

2011-01-01

Highlights: → We analyze the void reactivity effect for three ESFR core fuel cycle states. → The void reactivity effect is decomposed by neutron balance method. → Novelly, the normalization to the integral flux in the active core is applied. → The decomposition is compared with the perturbation theory based results. → The mechanism and the differences of the void reactivity effect are explained. - Abstract: The Sodium-cooled Fast Reactor (SFR) is one of the most promising Generation IV systems with many advantages, but has one dominating neutronic drawback - a positive sodium void reactivity. The aim of this study is to develop and apply a methodology, which should help better understand the causes and consequences of the sodium void effect. It focuses not only on the beginning-of-life (BOL) state of the core, but also on the beginning of open and closed equilibrium (BOC and BEC, respectively) fuel cycle conditions. The deeper understanding of the principal phenomena involved may subsequently lead to appropriate optimization studies. Various voiding scenarios, corresponding to different spatial zones, e.g. node or assembly, have been analyzed, and the most conservative case - the voiding of both inner and outer fuel zones - has been selected as the reference scenario. On the basis of the neutron balance method, the corresponding SFR void reactivity has been decomposed reaction-, isotope-, and energy-group-wise. Complementary results, based on generalized perturbation theory and sensitivity analysis, are also presented. The numerical analysis for both neutron balance and perturbation theory methods has been carried out using appropriate modules of the ERANOS code system. A strong correlation between the flux worth, i.e. the product of flux and adjoint flux, and the void reactivity importance distributions has been found for the node- and assembly-wise voiding scenarios. The neutron balance based decomposition has shown that the void effect is caused mainly

Nuclear fuel cycle cost estimation and sensitivity analysis of unit costs on the basis of an equilibrium model

International Nuclear Information System (INIS)

Kim, S. K.; Ko, W. I.; You, S. R.; Gao, R. X.

2015-01-01

This paper examines the difference in the value of the nuclear fuel cycle cost calculated by the deterministic and probabilistic methods on the basis of an equilibrium model. Calculating using the deterministic method, the direct disposal cost and Pyro-SFR (sodium-cooled fast reactor) nuclear fuel cycle cost, including the reactor cost, were found to be 66.41 mills/kWh and 77.82 mills/kWh, respectively (1 mill = one thousand of a dollar, i.e., 10-3 $). This is because the cost of SFR is considerably expensive. Calculating again using the probabilistic method, however, the direct disposal cost and Pyro-SFR nuclear fuel cycle cost, excluding the reactor cost, were found be 7.47 mills/kWh and 6.40 mills/kWh, respectively, on the basis of the most likely value. This is because the nuclear fuel cycle cost is significantly affected by the standard deviation and the mean of the unit cost that includes uncertainty. Thus, it is judged that not only the deterministic method, but also the probabilistic method, would also be necessary to evaluate the nuclear fuel cycle cost. By analyzing the sensitivity of the unit cost in each phase of the nuclear fuel cycle, it was found that the uranium unit price is the most influential factor in determining nuclear fuel cycle costs.

Core characteristics of fast reactor cycle with simple dry pyrochemical processing

International Nuclear Information System (INIS)

Ikegami, Tetsuo

2008-01-01

Fast reactor core concept and core nuclear characteristics are studied for the application of the simple dry pyrochemical processing for fast reactor mixed oxide spent fuels, that is, the Compound Process Fuel Cycle, large FR core with of loaded fuels are recycled by the simple dry pyrochemical processing. Results of the core nuclear analyses show that it is possible to recycle FR spent fuel once and to have 1.01 of breeding ratio without radial blanket region. The comparison is made among three kinds of recycle fuels, LWR UO 2 spent fuel, LWR MOX spent fuel, and FR spent fuel. The recycle fuels reach an equilibrium state after recycles regardless of their starting heavy metal compositions, and the recycled FR fuel has the lowest radio-activity and the same level of heat generation among the recycle fuels. Therefore, the compound process fuel cycle has flexibility to recycle both LWR spent fuel and FR spent fuel. The concept has a possibility of enhancement of nuclear non-proliferation and process simplification of fuel cycle. (author)

Economic evaluation of fast reactor fuel cycling

International Nuclear Information System (INIS)

Hu Ping; Zhao Fuyu; Yan Zhou; Li Chong

2012-01-01

Economic calculation and analysis of two kinds of nuclear fuel cycle are conducted by check off method, based on the nuclear fuel cycling process and model for fast reactor power plant, and comparison is carried out for the economy of fast reactor fuel cycle and PWR once-through fuel cycle. Calculated based on the current price level, the economy of PWR one-through fuel cycle is better than that of the fast reactor fuel cycle. However, in the long term considering the rising of the natural uranium's price and the development of the post treatment technology for nuclear fuels, the cost of the fast reactor fuel cycle is expected to match or lower than that of the PWR once-through fuel cycle. (authors)

All heavy metals closed-cycle analysis on water-cooled reactors of uranium and thorium fuel cycle systems

International Nuclear Information System (INIS)

Permana, Sidik; Sekimoto, Hiroshi; Waris, Abdul; Takaki, Naoyuki

2009-01-01

Uranium and Thorium fuels as the basis fuel of nuclear energy utilization has been used for several reactor types which produce trans-uranium or trans-thorium as 'by product' nuclear reaction with higher mass number and the remaining uranium and thorium fuels. The utilization of recycled spent fuel as world wide concerns are spent fuel of uranium and plutonium and in some cases using recycled minor actinide (MA). Those fuel schemes are used for improving an optimum nuclear fuel utilization as well to reduce the radioactive waste from spent fuels. A closed-cycle analysis of all heavy metals on water-cooled cases for both uranium and thorium fuel cycles has been investigated to evaluate the criticality condition, breeding performances, uranium or thorium utilization capability and void reactivity condition. Water-cooled reactor is used for the basic design study including light water and heavy water-cooled as an established technology as well as commercialized nuclear technologies. A developed coupling code of equilibrium fuel cycle burnup code and cell calculation of SRAC code are used for optimization analysis with JENDL 3.3 as nuclear data library. An equilibrium burnup calculation is adopted for estimating an equilibrium state condition of nuclide composition and cell calculation is performed for calculating microscopic neutron cross-sections and fluxes in relation to the effect of different fuel compositions, different fuel pin types and moderation ratios. The sensitivity analysis such as criticality, breeding performance, and void reactivity are strongly depends on moderation ratio and each fuel case has its trend as a function of moderation ratio. Heavy water coolant shows better breeding performance compared with light water coolant, however, it obtains less negative or more positive void reactivity. Equilibrium nuclide compositions are also evaluated to show the production of main nuclides and also to analyze the isotopic composition pattern especially

A procedure for searching the equilibrium core of a research reactor

International Nuclear Information System (INIS)

Bakri Arbie; Liem Peng Hong; Prayoto

1996-01-01

A procedure for searching the equilibrium core of a research reactor has been proposed. The searching procedure is based on the relaxation method and has been implemented in Batan-EQUIL-2D in-core fuel management code. The few-group neutron diffusion theory in 2-D, X-Y, and R-Z reactor geometries are adopted as the framework of the code. The successful applicability of the procedure for obtaining the new RSG-GAS (MPR-30) silicide equilibrium core was shown. (author)

Comparative assessment of nuclear fuel cycles. Light-water reactor once-through, classical fast breeder reactor, and symbiotic fast breeder reactor cycles

International Nuclear Information System (INIS)

Hardie, R.W.; Barrett, R.J.; Freiwald, J.G.

1980-06-01

The object of the Alternative Nuclear Fuel Cycle Study is to perform comparative assessments of nuclear power systems. There are two important features of this study. First, this evaluation attempts to encompass the complete, integrated fuel cycle from mining of uranium ore to disposal of waste rather than isolated components. Second, it compares several aspects of each cycle - energy use, economics, technological status, proliferation, public safety, and commercial potential - instead of concentrating on one or two assessment areas. This report presents assessment results for three fuel cycles. These are the light-water reactor once-through cycle, the fast breeder reactor on the classical plutonium cycle, and the fast breeder reactor on a symbiotic cycle using plutonium and 233 U as fissile fuels. The report also contains a description of the methodology used in this assessment. Subsequent reports will present results for additional fuel cycles

Gas reactor international cooperative program interim report. Pebble bed reactor fuel cycle evaluation

International Nuclear Information System (INIS)

1978-09-01

Nuclear fuel cycles were evaluated for the Pebble Bed Gas Cooled Reactor under development in the Federal Republic of Germany. The basic fuel cycle specified for the HTR-K and PNP is well qualified and will meet the requirements of these reactors. Twenty alternate fuel cycles are described, including high-conversion cycles, net-breeding cycles, and proliferation-resistant cycles. High-conversion cycles, which have a high probability of being successfully developed, promise a significant improvement in resource utilization. Proliferation-resistant cycles, also with a high probability of successful development, compare very favorably with those for other types of reactors. Most of the advanced cycles could be adapted to first-generation pebble bed reactors with no significant modifications

The uranium-plutonium breeder reactor fuel cycle

International Nuclear Information System (INIS)

Salmon, A.; Allardice, R.H.

1979-01-01

All power-producing systems have an associated fuel cycle covering the history of the fuel from its source to its eventual sink. Most, if not all, of the processes of extraction, preparation, generation, reprocessing, waste treatment and transportation are involved. With thermal nuclear reactors more than one fuel cycle is possible, however it is probable that the uranium-plutonium fuel cycle will become predominant; in this cycle the fuel is mined, usually enriched, fabricated, used and then reprocessed. The useful components of the fuel, the uranium and the plutonium, are then available for further use, the waste products are treated and disposed of safely. This particular thermal reactor fuel cycle is essential if the fast breeder reactor (FBR) using plutonium as its major fuel is to be used in a power-producing system, because it provides the necessary initial plutonium to get the system started. In this paper the authors only consider the FBR using plutonium as its major fuel, at present it is the type envisaged in all, current national plans for FBR power systems. The corresponding fuel cycle, the uranium-plutonium breeder reactor fuel cycle, is basically the same as the thermal reactor fuel cycle - the fuel is used and then reprocessed to separate the useful components from the waste products, the useful uranium and plutonium are used again and the waste disposed of safely. However the details of the cycle are significantly different from those of the thermal reactor cycle. (Auth.)

A new estimation method for nuclide number densities in equilibrium cycle

International Nuclear Information System (INIS)

Seino, Takeshi; Sekimoto, Hiroshi; Ando, Yoshihira.

1997-01-01

A new method is proposed for estimating nuclide number densities of LWR equilibrium cycle by multi-recycling calculation. Conventionally, it is necessary to spend a large computation time for attaining the ultimate equilibrium state. Hence, the cycle in nearly constant fuel composition has been considered as an equilibrium state which can be achieved by a few of recycling calculations on a simulated cycle operation under a specific fuel core design. The present method uses steady state fuel nuclide number densities as the initial guess for multi-recycling burnup calculation obtained by a continuously fuel supplied core model. The number densities are modified to be the initial number densities for nuclides of a batch supplied fuel. It was found that the calculated number densities could attain to more precise equilibrium state than that of a conventional multi-recycling calculation with a small number of recyclings. In particular, the present method could give the ultimate equilibrium number densities of the nuclides with the higher mass number than 245 Cm and 244 Pu which were not able to attain to the ultimate equilibrium state within a reasonable number of iterations using a conventional method. (author)

Cost aspects of the research reactor fuel cycle

International Nuclear Information System (INIS)

2010-01-01

Research reactors have made valuable contributions to the development of nuclear power, basic science, materials development, radioisotope production for medicine and industry, and education and training. In doing so, they have provided an invaluable service to humanity. Research reactors are expected to make important contributions in the coming decades to further development of the peaceful uses of nuclear technology, in particular for advanced nuclear fission reactors and fuel cycles, fusion, high energy physics, basic research, materials science, nuclear medicine, and biological sciences. However, in the context of decreased public sector support, research reactors are increasingly faced with financial constraints. It is therefore of great importance that their operations are based on a sound understanding of the costs of the complete research reactor fuel cycle, and that they are managed according to sound financial and economic principles. This publication is targeted at individuals and organizations involved with research reactor operations, with the aim of providing both information and an analytical framework for assessing and determining the cost structure of fuel cycle related activities. Efficient management of fuel cycle expenditures is an important component in developing strategies for sustainable future operation of a research reactor. The elements of the fuel cycle are presented with a description of how they can affect the cost efficient operation of a research reactor. A systematic review of fuel cycle choices is particularly important when a new reactor is being planned or when an existing reactor is facing major changes in its fuel cycle structure, for example because of conversion of the core from high enriched uranium (HEU) to low enriched uranium (LEU) fuel, or the changes in spent fuel management provision. Review and optimization of fuel cycle issues is also recommended for existing research reactors, even in cases where research reactor

Study on characteristics for different moderation ratios of heavy water coolant with different reactor types in equilibrium states

International Nuclear Information System (INIS)

Permana, Sidik; Takaki, Naoyuki; Sekimoto, Hiroshi

2005-01-01

Several characteristics for different moderation ratios of heavy water coolant with different reactor types in equilibrium states have been investigated. Performances of PWR and CANDU reactors are compared. A calculation method for determining the required uranium enrichment for criticality of the systems has been developed by coupling the equilibrium fuel cycle burn-up calculation and cell calculation of PIJ module of SRAC2000 code. In the present study, we have compared the characteristics for different moderator to fuel ratio (MFR, 0.1 to 30), different burn-up for CANDU type and four fuels cycle schemes. Nuclide density of 235 U at MFR=1.9 decreases with increasing number of confined HM, while 235 U at higher MFR has the opposite trend. However, the nuclide density of fissile material at higher MFR is lower except 238 U. CANDU type requires lower uranium enrichment and obtains higher conversion ratio than PWR type. Lowest burn-up requires the lowest uranium enrichment and obtains the highest conversion ratio. The breeding condition can be obtained for plutonium recycle cases at MFR=2.1 of Case 4 and MFR=1.4 of Case 3. The natural uranium can be achieved at MFR=14 of plutonium recycle cases, and it can be used easier by increasing number of confined HM. (author)

Neutronic characteristics of linear-assembly breed-and-burn reactors

International Nuclear Information System (INIS)

Petroski, Robert; Forget, Benoit; Forsberg, Charles

2012-01-01

Highlights: ► Simple models used to characterize general behavior of linear-assembly B and B reactors. ► Diffusion theory model developed to explain axial distributions, height vs. reactivity. ► Neutron excess concept reformulated to include linear-assembly B and B reactors. ► Designed model of B and B reactor started using melt-refined B and B reactor used fuel. ► Computed doubling time of fuel cycle requiring no chemical separations. - Abstract: Linear-assembly breed-and-burn (B and B) reactors are B and B reactors that use axially connected assemblies similar to conventional LWR or fast reactor fuel assemblies. Methods for analyzing linear-assembly B and B reactors and their fuel cycles are developed and applied. General neutronic characteristics of linear-assembly B and B reactors are analyzed, including the effects that burnup, shuffling sequence, and radial and axial size have on equilibrium-cycle k-effective. The mechanisms that give rise to a highly peaked axial burnup distribution are explained, and a method for predicting peak burnup vs. k-effective based on infinite-medium depletion calculations is developed. Next, the neutron excess concept from previous studies of B and B reactors is extended to apply to linear-assembly B and B reactors, which allows the amount of starter fuel needed to establish a given equilibrium cycle to be calculated. Several example applications of the neutron excess formulation are given. First, an example model of a linear-assembly B and B reactor is analyzed to find the neutron excess cost of an equilibrium cycle. Second, simple one-dimensional models are used to predict the neutron excess value obtainable from different starter fuel configurations. Finally, these ideas are applied to design a fuel cycle consisting of linear-assembly B and B reactors and fuel recycling via a melt refining process. The neutron excess concept is used to design an appropriate starter fuel configuration made from melt refined fuel, which

Thorium-Based Fuel Cycles in the Modular High Temperature Reactor

Institute of Scientific and Technical Information of China (English)

CHANG Hong; YANG Yongwei; JING Xingqing; XU Yunlin

2006-01-01

Large stockpiles of civil-grade as well as weapons-grade plutonium have been accumulated in the world from nuclear power or other programs of different countries. One alternative for the management of the plutonium is to incinerate it in the high temperature reactor (HTR). The thorium-based fuel cycle was studied in the modular HTR to reduce weapons-grade plutonium stockpiles, while producing no additional plutonium or other transuranic elements. Three thorium-uranium fuel cycles were also investigated. The thorium absorption cross sections of the resolved and unresolved resonances were generated using the ZUT-DGL code based on existing resonance data. The equilibrium core of the modular HTR was calculated and analyzed by means of the code VSOP'94. The results show that the modular HTR can incinerate most of the initially loaded plutonium amounting to about 95.3% net 239Pu for weapons-grade plutonium and can effectively utilize the uranium and thorium in the thorium-uranium fuel cycles.

Convective equilibrium and mixing-length theory for stellarator reactors

International Nuclear Information System (INIS)

Ho, D.D.M.; Kulsrud, R.M.

1985-09-01

In high β stellarator and tokamak reactors, the plasma pressure gradient in some regions of the plasma may exceed the critical pressure gradient set by ballooning instabilities. In these regions, convective cells break out to enhance the transport. As a result, the pressure gradient can rise only slightly above the critical gradient and the plasma is in another state of equilibrium - ''convective equilibrium'' - in these regions. Although the convective transport cannot be calculated precisely, it is shown that the density and temperature profiles in the convective region can still be estimated. A simple mixing-length theory, similar to that used for convection in stellar interiors, is introduced in this paper to provide a qualitative description of the convective cells and to show that the convective transport is highly efficient. A numerical example for obtaining the density and temperature profiles in a stellarator reactor is given

Study on thermodynamic cycle of high temperature gas-cooled reactor

International Nuclear Information System (INIS)

Qu Xinhe; Yang Xiaoyong; Wang Jie

2017-01-01

The development trend of the (very) High temperature gas-cooled reactor is to gradually increase the reactor outlet temperature. The different power conversion units are required at the different reactor outlet temperature. In this paper, for the helium turbine direct cycle and the combined cycle of the power conversion unit of the High temperature gas-cooled reactor, the mathematic models are established, and three cycle plans are designed. The helium turbine direct cycle is a Brayton cycle with recuperator, precooler and intercooler. In the combined cycle plan 1, the topping cycle is a simple Brayton cycle without recuperator, precooler and intercooler, and the bottoming cycle is based on the steam parameters (540deg, 6 MPa) recommended by Siemens. In the combined cycle plan 2, the topping cycle also is a simple Brayton cycle, and the bottoming cycle which is a Rankine cycle with reheating cycle is based on the steam parameters of conventional subcritical thermal power generation (540degC, 18 MPa). The optimization results showed that the cycle efficiency of the combined cycle plan 2 is the highest, the second is the helium turbine direct cycle, and the combined cycle plan 2 is the lowest. When the reactor outlet temperature is 900degC and the pressure ratio is 2.02, the cycle efficiency of the combined cycle plan 2 can reach 49.7%. The helium turbine direct cycle has a reactor inlet temperature above 500degC due to the regenerating cycle, so it requires a cooling circuit for the internal wall of the reactor pressure vessel. When the reactor outlet temperature increases, the increase of the pressure ratio required by the helium turbine direct cycle increases may bring some difficulties to the design and manufacture of the magnetic bearings. For the combined cycle, the reactor inlet temperature can be controlled below than 370degC, so the reactor pressure vessel can use SA533 steel without cooling the internal wall of the reactor pressure vessel. The pressure

Candu reactors with thorium fuel cycles

International Nuclear Information System (INIS)

Hopwood, J.M.; Fehrenbach, P.; Duffey, R.; Kuran, S.; Ivanco, M.; Dyck, G.R.; Chan, P.S.W.; Tyagi, A.K.; Mancuso, C.

2006-01-01

Over the last decade and a half AECL has established a strong record of delivering CANDU 6 nuclear power plants on time and at budget. Inherently flexible features of the CANDU type reactors, such as on-power fuelling, high neutron economy, fuel channel based heat transport system, simple fuel bundle configuration, two independent shut down systems, a cool moderator and a defence-in-depth based safety philosophy provides an evolutionary path to further improvements in design. The immediate milestone on this path is the Advanced CANDU ReactorTM** (ACRTM**), in the form of the ACR-1000TM**. This effort is being followed by the Super Critical Water Reactor (SCWR) design that will allow water-cooled reactors to attain high efficiencies by increasing the coolant temperature above 550 0 C. Adaptability of the CANDU design to different fuel cycles is another technology advantage that offers an additional avenue for design evolution. Thorium is one of the potential fuels for future reactors due to relative abundance, neutronics advantage as a fertile material in thermal reactors and proliferation resistance. The Thorium fuel cycle is also of interest to China, India, and Turkey due to local abundance that can ensure sustainable energy independence over the long term. AECL has performed an assessment of both CANDU 6 and ACR-1000 designs to identify systems, components, safety features and operational processes that may need to be modified to replace the NU or SEU fuel cycles with one based on Thorium. The paper reviews some of these requirements and the associated practical design solutions. These modifications can either be incorporated into the design prior to construction or, for currently operational reactors, during a refurbishment outage. In parallel with reactor modifications, various Thorium fuel cycles, either based on mixed bundles (homogeneous) or mixed channels (heterogeneous) have been assessed for technical and economic viability. Potential applications of a

Development of parallellized higher-order generalized depletion perturbation theory for application in equilibrium cycle optimization

Energy Technology Data Exchange (ETDEWEB)

Geemert, R. van E-mail: rene.vangeemert@psi.ch; Hoogenboom, J.E. E-mail: j.e.hoogenboom@iri.tudelft.nl

2001-09-01

As nuclear fuel economy is basically a multi-cycle issue, a fair way of evaluating reload patterns is to consider their performance in the case of an equilibrium cycle. The equilibrium cycle associated with a reload pattern is defined as the limit fuel cycle that eventually emerges after multiple successive periodic refueling, each time implementing the same reload scheme. Since the equilibrium cycle is the solution of a reload operation invariance equation, it can in principle be found with sufficient accuracy only by applying an iterative procedure, simulating the emergence of the limit cycle. For a design purpose such as the optimization of reload patterns, in which many different equilibrium cycle perturbations (resulting from many different limited changes in the reload operator) must be evaluated, this requires far too much computational effort. However, for very fast calculation of these many different equilibrium cycle perturbations it is also possible to set up a generalized variational approach. This approach results in an iterative scheme that yields the exact perturbation in the equilibrium cycle solution as well, in an accelerated way. Furthermore, both the solution of the adjoint equations occurring in the perturbation theory formalism and the implementation of the optimization algorithm have been parallellized and executed on a massively parallel machine. The combination of parallellism and generalized perturbation theory offers the opportunity to perform very exhaustive, fast and accurate sampling of the solution space for the equilibrium cycle reload pattern optimization problem.

Pulse-density modulation control of chemical oscillation far from equilibrium in a droplet open-reactor system.

Science.gov (United States)

Sugiura, Haruka; Ito, Manami; Okuaki, Tomoya; Mori, Yoshihito; Kitahata, Hiroyuki; Takinoue, Masahiro

2016-01-20

The design, construction and control of artificial self-organized systems modelled on dynamical behaviours of living systems are important issues in biologically inspired engineering. Such systems are usually based on complex reaction dynamics far from equilibrium; therefore, the control of non-equilibrium conditions is required. Here we report a droplet open-reactor system, based on droplet fusion and fission, that achieves dynamical control over chemical fluxes into/out of the reactor for chemical reactions far from equilibrium. We mathematically reveal that the control mechanism is formulated as pulse-density modulation control of the fusion-fission timing. We produce the droplet open-reactor system using microfluidic technologies and then perform external control and autonomous feedback control over autocatalytic chemical oscillation reactions far from equilibrium. We believe that this system will be valuable for the dynamical control over self-organized phenomena far from equilibrium in chemical and biomedical studies.

Analysis of thorium and uranium fuel cycles in an iso-breeder lead fast reactor using extended-EQL3D procedure

International Nuclear Information System (INIS)

Fiorina, Carlo; Krepel, Jiri; Cammi, Antonio; Franceschini, Fausto; Mikityuk, Konstantin; Ricotti, Marco Enrico

2013-01-01

Highlights: ► Extension of EQL3D procedure to calculate radio-toxicity and decay heat. ► Characterization of uranium- and thorium-fueled LFR from BOL to equilibrium. ► Safety improvements for a LFR in a closed thorium cycle. ► Advantages of thorium-fueled LFR in terms of decay heat and radio-toxicity generation. ► Safety, decay heat and radio-toxicity concerns for a Th–Pu beginning-of-life core. - Abstract: Use of thorium in fast reactors has typically been considered as a secondary option, mainly thanks to a possible self-sustaining thorium cycle already in thermal reactors and due to the limited breeding capabilities compared to U–Pu in the fast neutron energy range. In recent years nuclear waste management has become more important, and the thorium option has been reconsidered for the claimed potential to burn transuranic waste and the lower build-up of hazardous isotopes in a closed cycle. To ascertain these claims and their limitations, the fuel cycle isotopic inventory, and associated waste radio-toxicity and decay heat, should be quantified and compared to the case of the uranium cycle using realistic core configurations, with complete recycle of all the actinides. Since the transition from uranium to thorium fuel cycles will likely involve a transuranic burning phase, this transition and the challenges that the evolving fuel actinide composition presents, for instance on reactor feedback parameters, should also be analyzed. In the present paper, these issues are investigated based on core physics analysis of the Lead-cooled Fast Reactor ELSY, performed with the fast reactor ERANOS code and the EQL3D procedure allowing full-core characterization of the equilibrium cycle and the transition cycles. In order to compute radio-toxicity and decay heat, EQL3D has been extended by developing a new module, which has been assessed against ORIGEN-S and is presented here. The capability of the EQL3D procedure to treat full-core 3D geometries allowed to

Aspects of the fast reactors fuel cycle

International Nuclear Information System (INIS)

Zouain, D.M.

1982-06-01

The fuel cycle for fast reactors, is analysed, regarding the technical aspects of the developing of the reprocessing stages and the fuel fabrication. The environmental impact of LMFBRs and the waste management of this cycle are studied. The economic aspects of the fuel cycle, are studied too. Some coments about the Brazilian fast reactors programs are done. (E.G.) [pt

Design of an equilibrium nucleus of a BWR type reactor based in a Thorium-Uranium fuel; Diseno de un nucleo de equilibrio de un reactor tipo BWR basado en un combustible de Torio-Uranio

Energy Technology Data Exchange (ETDEWEB)

Francois, J.L.; Nunez C, A. [Laboratorio de Analisis en Ingenieria de Reactores Nucleares, Facultad de Ingenieria-UNAM, Paseo Cuauhnahuac 8532, Jiutepec, Morelos (Mexico)

2003-07-01

In this work the design of the reactor nucleus of boiling water using fuel of thorium-uranium is presented. Starting from an integral concept based in a type cover-seed assemble is carried out the design of an equilibrium reload for the nucleus of a reactor like that of the Laguna Verde Central and its are analyzed some of the main design variables like the cycle length, the reload fraction, the burnt fuel, the vacuum distribution, the generation of lineal heat, the margin of shutdown, as well as a first estimation of the fuel cost. The results show that it is feasible to obtain an equilibrium reload, comparable to those that are carried out in the Laguna Verde reactors, with a good behavior of those analyzed variables. The cost of the equilibrium reload designed with the thorium-uranium fuel is approximately 2% high that the uranium reload producing the same energy. It is concluded that it is convenient to include burnable poisons, type gadolinium, in the fuel with the end of improving the reload design, the fuel costs and the margin of shutdown. (Author)

Non-equilibrium plasma reactor for natrual gas processing

International Nuclear Information System (INIS)

Shair, F.H.; Ravimohan, A.L.

1974-01-01

A non-equilibrium plasma reactor for natural gas processing into ethane and ethylene comprising means of producing a non-equilibrium chemical plasma wherein selective conversion of the methane in natural gas to desired products of ethane and ethylene at a pre-determined ethane/ethylene ratio in the chemical process may be intimately controlled and optimized at a high electrical power efficiency rate by mixing with a recycling gas inert to the chemical process such as argon, helium, or hydrogen, reducing the residence time of the methane in the chemical plasma, selecting the gas pressure in the chemical plasma from a wide range of pressures, and utilizing pulsed electrical discharge producing the chemical plasma. (author)

Sodium fast reactors with closed fuel cycle

CERN Document Server

Raj, Baldev; Vasudeva Rao, PR 0

2015-01-01

Sodium Fast Reactors with Closed Fuel Cycle delivers a detailed discussion of an important technology that is being harnessed for commercial energy production in many parts of the world. Presenting the state of the art of sodium-cooled fast reactors with closed fuel cycles, this book:Offers in-depth coverage of reactor physics, materials, design, safety analysis, validations, engineering, construction, and commissioning aspectsFeatures a special chapter on allied sciences to highlight advanced reactor core materials, specialized manufacturing technologies, chemical sensors, in-service inspecti

Impact of different moderator ratios with light and heavy water cooled reactors in equilibrium states

International Nuclear Information System (INIS)

Permana, Sidik; Takaki, Naoyuki; Sekimoto, Hiroshi

2006-01-01

As an issue of sustainable development in the world, energy sustainability using nuclear energy may be possible using several different ways such as increasing breeding capability of the reactors and optimizing the fuel utilization using spent fuel after reprocessing as well as exploring additional nuclear resources from sea water. In this present study the characteristics of light and heavy water cooled reactors for different moderator ratios in equilibrium states have been investigated. The moderator to fuel ratio (MFR) is varied from 0.1 to 4.0. Four fuel cycle schemes are evaluated in order to investigate the effect of heavy metal (HM) recycling. A calculation method for determining the required uranium enrichment for criticality of the systems has been developed by coupling the equilibrium fuel cycle burn-up calculation and cell calculation of SRAC 2000 code using nuclear data library from the JENDL 3.2. The results show a thermal spectrum peak appears for light water coolant and no thermal peak for heavy water coolant along the MFR (0.1 ≤ MFR ≤ 4.0). The plutonium quality can be reduced effectively by increasing the MFR and number of recycled HM. Considering the effect of increasing number of recycled HM; it is also effective to reduce the uranium utilization and to increase the conversion ratio. trans-Plutonium production such as americium (Am) and curium (Cm) productions are smaller for heavy water coolant than light water coolant. The light water coolant shows the feasibility of breeding when HM is recycled with reducing the MFR. Wider feasible area of breeding has been obtained when light water coolant is replaced by heavy water coolant

Burnup effect on nuclear fuel cycle cost using an equilibrium model

International Nuclear Information System (INIS)

Youn, S. R.; Kim, S. K.; Ko, W. I.

2014-01-01

The degree of fuel burnup is an important technical parameter to the nuclear fuel cycle, being sensitive and progressive to reduce the total volume of process flow materials and eventually cut the nuclear fuel cycle costs. This paper performed the sensitivity analysis of the total nuclear fuel cycle costs to changes in the technical parameter by varying the degree of burnups in each of the three nuclear fuel cycles using an equilibrium model. Important as burnup does, burnup effect was used among the cost drivers of fuel cycle, as the technical parameter. The fuel cycle options analyzed in this paper are three different fuel cycle options as follows: PWR-Once Through Cycle(PWR-OT), PWR-MOX Recycle, Pyro-SFR Recycle. These fuel cycles are most likely to be adopted in the foreseeable future. As a result of the sensitivity analysis on burnup effect of each three different nuclear fuel cycle costs, PWR-MOX turned out to be the most influenced by burnup changes. Next to PWR-MOX cycle, in the order of Pyro-SFR and PWR-OT cycle turned out to be influenced by the degree of burnup. In conclusion, the degree of burnup in the three nuclear fuel cycles can act as the controlling driver of nuclear fuel cycle costs due to a reduction in the volume of spent fuel leading better availability and capacity factors. However, the equilibrium model used in this paper has a limit that time-dependent material flow and cost calculation is impossible. Hence, comparative analysis of the results calculated by dynamic model hereafter and the calculation results using an equilibrium model should be proceed. Moving forward to the foreseeable future with increasing burnups, further studies regarding alternative material of high corrosion resistance fuel cladding for the overall

Power generation costs for alternate reactor fuel cycles

International Nuclear Information System (INIS)

Smolen, G.R.; Delene, J.G.

1980-09-01

The total electric generating costs at the power plant busbar are estimated for various nuclear reactor fuel cycles which may be considered for power generation in the future. The reactor systems include pressurized water reactors (PWR), heavy-water reactors (HWR), high-temperature gas cooled reactors (HTGR), liquid-metal fast breeder reactors (LMFBR), light-water pre-breeder and breeder reactors (LWPR, LWBR), and a fast mixed spectrum reactor (FMSR). Fuel cycles include once-through, uranium-only recycle, and full recycle of the uranium and plutonium in the spent fuel assemblies. The U 3 O 8 price for economic transition from once-through LWR fuel cycles to both PWR recycle and LMFBR systems is estimated. Electric power generation costs were determined both for a reference set of unit cost parameters and for a range of uncertainty in these parameters. In addition, cost sensitivity parameters are provided so that independent estimations can be made for alternate cost assumptions

Assessment of the thorium fuel cycle in power reactors

International Nuclear Information System (INIS)

Kasten, P.R.; Homan, F.J.; Allen, E.J.

1977-01-01

A study was conducted at Oak Ridge National Laboratory to evaluate the role of thorium fuel cycles in power reactors. Three thermal reactor systems were considered: Light Water Reactors (LWRs); High-Temperature Gas-Cooled Reactors (HTGRs); and Heavy Water Reactors (HWRs) of the Canadian Deuterium Uranium Reactor (CANDU) type; most of the effort was on these systems. A summary comparing thorium and uranium fuel cycles in Fast Breeder Reactors (FBRs) was also compiled

Nonproliferation and safeguard considerations: Pebble Bed reactor fuel cycle evaluation

International Nuclear Information System (INIS)

1978-09-01

Nuclear fuel cycles were evaluated for the Pebble Bed Gas Cooled Reactor under development in the Federal Republic of Germany. The basic fuel cycle specified for the HTR-K and PNP is well qualified and will meet the requirements of these reactors. Twenty alternate fuel cycles are described, including high-conversion cycles, net-breeding cycles, and proliferation-resistant cycles. High-conversion cycles, which have a high probability of being successfully developed, promise a significant improvement in resource utilization. Proliferation-resistant cycles, also with a high probability of successful development, conpare very favorably with those for other types of reactors. Most of the advanced cycles could be adapted to first-generation pebble bed reactors with no significant modifications

In core-fuel management in approach to equilibrium of WWER-440 reactor; Prelazni rezim iskoriscenja goriva u nuklearnom reaktoru tipa VVER-440

Energy Technology Data Exchange (ETDEWEB)

Marinkovic, N [Boris Kidric Institute of nuclear sciences, Vinca, Belgrade (Yugoslavia)

1978-07-01

For the need of in core fuel management and prediction of fuel cycle costs as well as operating of a nuclear power plant behaviour of main physical parameters and refueling scheme during approach to equilibrium operation are indispensable. An estimation of a refueling scheme during forst six years of exploitation for a commercially proven PWR reactor of WWER-440 type is shown in this paper. (author)

Homogeneous Thorium Fuel Cycles in Candu Reactors

Energy Technology Data Exchange (ETDEWEB)

Hyland, B.; Dyck, G.R.; Edwards, G.W.R.; Magill, M. [Chalk River Laboratories, Atomic Energy of Canada Limited (Canada)

2009-06-15

The CANDU{sup R} reactor has an unsurpassed degree of fuel-cycle flexibility, as a consequence of its fuel-channel design, excellent neutron economy, on-power refueling, and simple fuel bundle [1]. These features facilitate the introduction and full exploitation of thorium fuel cycles in Candu reactors in an evolutionary fashion. Because thorium itself does not contain a fissile isotope, neutrons must be provided by adding a fissile material, either within or outside of the thorium-based fuel. Those same Candu features that provide fuel-cycle flexibility also make possible many thorium fuel-cycle options. Various thorium fuel cycles can be categorized by the type and geometry of the added fissile material. The simplest of these fuel cycles are based on homogeneous thorium fuel designs, where the fissile material is mixed uniformly with the fertile thorium. These fuel cycles can be competitive in resource utilization with the best uranium-based fuel cycles, while building up a 'mine' of U-233 in the spent fuel, for possible recycle in thermal reactors. When U-233 is recycled from the spent fuel, thorium-based fuel cycles in Candu reactors can provide substantial improvements in the efficiency of energy production from existing fissile resources. The fissile component driving the initial fuel could be enriched uranium, plutonium, or uranium-233. Many different thorium fuel cycle options have been studied at AECL [2,3]. This paper presents the results of recent homogeneous thorium fuel cycle calculations using plutonium and enriched uranium as driver fuels, with and without U-233 recycle. High and low burnup cases have been investigated for both the once-through and U-233 recycle cases. CANDU{sup R} is a registered trademark of Atomic Energy of Canada Limited (AECL). 1. Boczar, P.G. 'Candu Fuel-Cycle Vision', Presented at IAEA Technical Committee Meeting on 'Fuel Cycle Options for LWRs and HWRs', 1998 April 28 - May 01, also Atomic Energy

Research reactors fuel cycle problems and dilemma

International Nuclear Information System (INIS)

Romano, R.

2004-01-01

During last 10 years, some problems appeared in different steps of research reactors fuel cycle. Actually the majority of these reactors have been built in the 60s and these reactors were operated during all this long period in a cycle with steps which were dedicated to this activity. Progressively and for reasons often economical, certain steps of the cycle became more and more difficult to manage due to closing of some specialised workshops in the activities of scraps recycling, irradiated fuel reprocessing, even fuel fabrication. Other steps of the cycle meet or will meet difficulties, in particular supplying of fissile raw material LEU or HEU because this material was mostly produced in enrichment units existing mainly for military reason. Rarefaction of fissile material lead to use more and more enriched uraniums said 'of technical quality', that is to say which come from mixing of varied qualities of enriched material, containing products resulting from reprocessing. Actually, problems of end of fuel cycle are increased, either consisting of intermediary storage on the site of reactor or on specialised sites, or consisting of reprocessing. This brief summary shows most difficulties which are met today by a major part of industrials of the fuel cycle in the exercise of their activities

Proposed fuel cycle for the Integral Fast Reactor

International Nuclear Information System (INIS)

Burris, L.; Walters, L.C.

1985-01-01

One of the key features of ANL's Integral Fast Reactor (IFR) concept is a close-coupled fuel cycle. The proposed fuel cycle is similar to that demonstrated over the first five to six years of operation of EBR-II, when a fuel cycle facility adjacent to EBR-II was operated to reprocess and refabricate rapidly fuel discharged from the EBR-II. Locating the IFR and its fuel cycle facility on the same site makes the IFR a self-contained system. Because the reactor fuel and the uranium blanket are metals, pyrometallurgical processes (shortned to ''pyroprocesses'') have been chosen. The objectives of the IFR processes for the reactor fuel and blanket materials are to (1) recover fissionable materials in high yield; (2) remove fission products adequately from the reactor fuel, e.g., a decontamination factor of 10 to 100; and (3) upgrade the concentration of plutonium in uranium sufficiently to replenish the fissile-material content of the reactor fuel. After the fuel has been reconstituted, new fuel elements will be fabricated for recycle to the reactor

Future fuel cycle and reactor strategies

International Nuclear Information System (INIS)

Meneley, D.A.

1999-01-01

Within the framework of the 1997 IAEA Symposium 'Future Fuel Cycle and Reactor Strategies Adjusting to New Realities', Working Group No.3 produced a Key Issues paper addressing the title of the symposium. The scope of the Key Issues paper included those factors that are expected to remain or become important in the time period from 2015 to 2050, considering all facets of nuclear energy utilization from ore extraction to final disposal of waste products. The paper addressed the factors influencing the choice of reactor and fuel cycle. It then addressed the quantitatively largest category of reactor types expected to be important during the period; that is, thermal reactors burning uranium and plutonium fuel. The fast reactor then was discussed both as a stand-alone technology and as might be used in combination with thermal reactors. Thorium fuel use was discussed briefly. The present paper includes of a digest of the Key Issues Paper. Some comparisons arc made between the directions suggested in that paper and those indicated by the Abstracts of this Technical Committee Meeting- Recommendations are made for work which might be undertaken in the short and medium time frames, to ensure that fuel cycle technologies and processes established by the year 2050 will support the continuation of nuclear energy applications in the long term. (author)

Burnup performance of OTTO cycle pebble bed reactors with ROX fuel

International Nuclear Information System (INIS)

Ho, Hai Quan; Obara, Toru

2015-01-01

Highlights: • A 300 MW t Small Pebble Bed Reactor with Rock-like oxide fuel is proposed. • Using ROX fuel can achieve high discharged burnup of spent fuel. • High geological stability can be expected in direct disposal of the spent ROX fuel. • The Pebble Bed Reactor with ROX fuel can be critical at steady state operation. • All the reactor designs have a negative temperature coefficient. - Abstract: A pebble bed high-temperature gas-cooled reactor (PBR) with rock-like oxide (ROX) fuel was designed to achieve high discharged burnup and improve the integrity of the spent fuel in geological disposal. The MCPBR code with a JENDL-4.0 library, which developed the analysis of the Once-Through-Then-Out (OTTO) cycle in PBR, was used to perform the criticality and burnup analysis. Burnup calculations for eight cases were carried out for both ROX fuel and a UO 2 fuel reactor with different heavy-metal loading conditions. The effective multiplication factor of all cases approximately equalled unity in the equilibrium condition. The ROX fuel reactor showed lower FIFA than the UO 2 fuel reactor at the same heavy-metal loading, about 5–15%. However, the power peaking factor and maximum power per fuel ball in the ROX fuel core were lower than that of UO 2 fuel core. This effect makes it possible to compensate for the lower-FIFA disadvantage in a ROX fuel core. All reactor designs had a negative temperature coefficient that is needed for the passive safety features of a pebble bed reactor

Supercritical Water Reactor Cycle for Medium Power Applications

International Nuclear Information System (INIS)

BD Middleton; J Buongiorno

2007-01-01

Scoping studies for a power conversion system based on a direct-cycle supercritical water reactor have been conducted. The electric power range of interest is 5-30 MWe with a design point of 20 MWe. The overall design objective is to develop a system that has minimized physical size and performs satisfactorily over a broad range of operating conditions. The design constraints are as follows: Net cycle thermal efficiency (ge)20%; Steam turbine outlet quality (ge)90%; and Pumping power (le)2500 kW (at nominal conditions). Three basic cycle configurations were analyzed. Listed in order of increased plant complexity, they are: (1) Simple supercritical Rankine cycle; (2) All-supercritical Brayton cycle; and (3) Supercritical Rankine cycle with feedwater preheating. The sensitivity of these three configurations to various parameters, such as reactor exit temperature, reactor pressure, condenser pressure, etc., was assessed. The Thermoflex software package was used for this task. The results are as follows: (a) The simple supercritical Rankine cycle offers the greatest hardware simplification, but its high reactor temperature rise and reactor outlet temperature may pose serious problems from the viewpoint of thermal stresses, stability and materials in the core. (b) The all-supercritical Brayton cycle is not a contender, due to its poor thermal efficiency. (c) The supercritical Rankine cycle with feedwater preheating affords acceptable thermal efficiency with lower reactor temperature rise and outlet temperature. (d) The use of a moisture separator improves the performance of the supercritical Rankine cycle with feedwater preheating and allows for a further reduction of the reactor outlet temperature, thus it was selected for the next step. Preliminary engineering design of the supercritical Rankine cycle with feedwater preheating and moisture separation was performed. All major components including the turbine, feedwater heater, feedwater pump, condenser, condenser pump

Potentialities of the molten salt reactor concept for a sustainable nuclear power production based on thorium cycle in epithermal spectrum

International Nuclear Information System (INIS)

Nuttin, Alexis

2002-01-01

In the case of a significant nuclear contribution to world energy needs, the problem of present nuclear waste management pose the sustainability of the PWR fuel cycle back into question. Studies on storage and incineration of these wastes should therefore go hand in hand with studies on innovative systems dedicated to a durable nuclear energy production, as reliable, clean and safe as possible. We are here interested in the concept of molten salt reactor, whose fuel is liquid. This particularity allows an online pyrochemical reprocessing which gives the possibility to overcome some neutronic limits. In the late sixties, the MSBR (Molten Salt Breeder Reactor) project of a graphite-moderated fluoride molten salt reactor proved thus that breeding is attainable with thorium in a thermal spectrum, provided that the online reprocessing is appropriate. By means of simulation tools developed around the Monte Carlo code MCNP, we first re-evaluate the performance of a reference system, which is inspired by the MSBR project. The complete study of the pre-equilibrium transient of this 2,500 MWth reactor, started with 232 Th/ 233 U fuel, allows us to validate our reference choices. The obtained equilibrium shows an important reduction of inventories and induced radio-toxicities in comparison with the other possible fuel cycles. The online reprocessing is efficient enough to make the system breed, with a doubling time of about thirty years at equilibrium. From the reference system, we then test different options in terms of neutron economy, transmutation and control of reactivity. We find that the online reprocessing brings most of its flexibility to this system, which is particularly well adapted to power generation with thorium. The study of transition scenarios to this fuel cycle quantifies the limits of a possible deployment from the present French power stock, and finally shows that a rational management of the available plutonium would be necessary in any case. (author)

Elements of nuclear reactor fueling theory

International Nuclear Information System (INIS)

Egan, M.R.

1984-01-01

Starting with a review of the simple batch size effect, a more general theory of nuclear fueling is derived to describe the behaviour and physical requirements of operating cycle sequences and fueling strategies having practical use in fuel management. The generalized theory, based on linear reactivity modeling, is analytical and represents the effects of multiple-stream, multiple-depletion-batch fueling configurations in systems employing arbitrary, non-integer batch size strategies, and containing fuel with variable energy generation rates. Reactor operating cycles and cycle sequences are represented with realistic structure that includes the effects of variable cycle energy production, cycle lengths, end-of-cycle operating extensions and manoeuvering allowances. Results of the analytical theory are first applied to the special case of degenerate equilibrium cycle sequences, yielding several fundamental principles related to the selection of refueling strategy. Numerical evaluations of degenerate equilibrium cycle sequences are then performed for a typical PWR core, and accompanying fuel cycle costs are calculated. The impact of design and operational limits as constraints on the performance mappings for this reactor are also studied with respect to achieving improved cost performance from the once-through fuel cycle. The dynamics of transition cycle sequences are then examined using the generalized theory. Proof of the existence of non-degenerate equilibrium cycle sequences is presented when the mechanics of the fixed reload batch size strategy are developed analytically for transition sequences. Finally, an analysis of the fixed reload enrichment strategy demonstrates the potential for convergence of the transition sequence to a fully degenerate equilibrium sequence. (author)

Fast Reactor Fuel Cycle Cost Estimates for Advanced Fuel Cycle Studies

International Nuclear Information System (INIS)

Harrison, Thomas

2013-01-01

Presentation Outline: • Why Do I Need a Cost Basis?; • History of the Advanced Fuel Cycle Cost Basis; • Description of the Cost Basis; • Current Work; • Fast Reactor Fuel Cycle Applications; • Sample Fuel Cycle Cost Estimate Analysis; • Future Work

Waste arisings from a high-temperature reactor with a uranium-thorium fuel cycle

International Nuclear Information System (INIS)

1979-09-01

This paper presents an equilibrium-recycle condition flow sheet for a high-temperature gas-cooled reactor (HTR) fuel cycle which uses thorium and high-enriched uranium (93% U-235) as makeup fuel. INFCE Working Group 7 defined percentage losses to various waste streams are used to adjust the heavy-element mass flows per gigawatt-year of electricity generated. Thorium and bred U-233 are recycled following Thorex reprocessing. Fissile U-235 is recycled one time following Purex reprocessing and then is discarded to waste. Plutonium and other transuranics are discarded to waste. Included are estimates of volume, radioactivity, and heavy-element content of wastes arising from HTR fuel element fabrication; HTR operation, maintenance, and decommissioning; and reprocessing spent fuel where the waste is unique to the HTR fuel cycle

Fuel cycle problems in fusion reactors

International Nuclear Information System (INIS)

Hickman, R.G.

1976-01-01

Fuel cycle problems of fusion reactors evolve around the breeding, recovery, containment, and recycling of tritium. These processes are described, and their implications and alternatives are discussed. Technically, fuel cycle problems are solvable; economically, their feasibility is not yet known

Pulse-density modulation control of chemical oscillation far from equilibrium in a droplet open-reactor system

OpenAIRE

Sugiura, Haruka; Ito, Manami; Okuaki, Tomoya; Mori, Yoshihito; Kitahata, Hiroyuki; Takinoue, Masahiro

2016-01-01

The design, construction and control of artificial self-organized systems modelled on dynamical behaviours of living systems are important issues in biologically inspired engineering. Such systems are usually based on complex reaction dynamics far from equilibrium; therefore, the control of non-equilibrium conditions is required. Here we report a droplet open-reactor system, based on droplet fusion and fission, that achieves dynamical control over chemical fluxes into/out of the reactor for c...

Strategies of operation cycles in BWR type reactors

International Nuclear Information System (INIS)

Molina, D.; Sendino, F.

1996-01-01

The article analyzes the operation cycles in BWR type reactors. The cycle size of operation is the consequence on the optimization process of the costs with the technical characteristics of nuclear fuel and the characteristics of demand and production. The authors analyze the cases of Garona NP and Cofrentes NP, both with BWR reactors. (Author)

A study on improving the performance of a research reactor's equilibrium core

Directory of Open Access Journals (Sweden)

Muhammad Atta

2013-01-01

Full Text Available Utilizing low enriched uranium silicide fuel (U3Si2-Al of existing uranium density (3.285 g/cm3, different core configurations have been studied in search of an equilibrium core with an improved performance for the Pakistan Research Reactor-1. Furthermore, we have extended our analysis to the performance of higher density silicide fuels with a uranium density of 4.0 and 4.8 U g/cm3. The criterion used in selecting the best performing core was that of “unit flux time cycle length per 235U mass per cycle”. In order to analyze core performance by improving neutron moderation, utilizing higher-density fuel, the effect of the coolant channel width was also studied by reducing the number of plates in the standard/control fuel element. Calculations employing computer codes WIMSD/4 and CITATION were performed. A ten energy group structure for fission neutrons was used for the generation of microscopic cross-sections through WIMSD/4. To search the equilibrium core, two-dimensional core modelling was performed in CITATION. Performance indicators have shown that the higher-density uranium silicide-fuelled core (U density 4.8 g/cm3 without any changes in standard/control fuel elements, comprising of 15 standard and 4 control fuel elements, is the best performing of all analyzed cores.

Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems

Energy Technology Data Exchange (ETDEWEB)

D. E. Shropshire

2009-01-01

The Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems, prepared to support the U.S. Advanced Fuel Cycle Initiative (AFCI) systems analysis, provides a technology-oriented baseline system cost comparison between the open fuel cycle and closed fuel cycle systems. The intent is to understand their overall cost trends, cost sensitivities, and trade-offs. This analysis also improves the AFCI Program’s understanding of the cost drivers that will determine nuclear power’s cost competitiveness vis-a-vis other baseload generation systems. The common reactor-related costs consist of capital, operating, and decontamination and decommissioning costs. Fuel cycle costs include front-end (pre-irradiation) and back-end (post-iradiation) costs, as well as costs specifically associated with fuel recycling. This analysis reveals that there are large cost uncertainties associated with all the fuel cycle strategies, and that overall systems (reactor plus fuel cycle) using a closed fuel cycle are about 10% more expensive in terms of electricity generation cost than open cycle systems. The study concludes that further U.S. and joint international-based design studies are needed to reduce the cost uncertainties with respect to fast reactor, fuel separation and fabrication, and waste disposition. The results of this work can help provide insight to the cost-related factors and conditions needed to keep nuclear energy (including closed fuel cycles) economically competitive in the U.S. and worldwide. These results may be updated over time based on new cost information, revised assumptions, and feedback received from additional reviews.

Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems

International Nuclear Information System (INIS)

Shropshire, D.E.

2009-01-01

The Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems, prepared to support the U.S. Advanced Fuel Cycle Initiative (AFCI) systems analysis, provides a technology-oriented baseline system cost comparison between the open fuel cycle and closed fuel cycle systems. The intent is to understand their overall cost trends, cost sensitivities, and trade-offs. This analysis also improves the AFCI Program's understanding of the cost drivers that will determine nuclear power's cost competitiveness vis-a-vis other baseload generation systems. The common reactor-related costs consist of capital, operating, and decontamination and decommissioning costs. Fuel cycle costs include front-end (pre-irradiation) and back-end (post-irradiation) costs, as well as costs specifically associated with fuel recycling. This analysis reveals that there are large cost uncertainties associated with all the fuel cycle strategies, and that overall systems (reactor plus fuel cycle) using a closed fuel cycle are about 10% more expensive in terms of electricity generation cost than open cycle systems. The study concludes that further U.S. and joint international-based design studies are needed to reduce the cost uncertainties with respect to fast reactor, fuel separation and fabrication, and waste disposition. The results of this work can help provide insight to the cost-related factors and conditions needed to keep nuclear energy (including closed fuel cycles) economically competitive in the U.S. and worldwide. These results may be updated over time based on new cost information, revised assumptions, and feedback received from additional reviews.

Fast reactors fuel Cycle: State in Europe

International Nuclear Information System (INIS)

1991-01-01

In this SFEN day we treat all aspects (economics-reactor cores, reprocessing, experience return) of the LMFBR fuel cycle in Europe and we discuss about the development of this type of reactor (EFR project) [fr

Future fuel cycle development for CANDU reactors

International Nuclear Information System (INIS)

Hatcher, S.R.; McDonnell, F.N.; Griffiths, J.; Boczar, P.G.

1987-01-01

The CANDU reactor has proven to be safe and economical and has demonstrated outstanding performance with natural uranium fuel. The use of on-power fuelling, coupled with excellent neutron economy, leads to a very flexible reactor system with can utilize a wide variety of fuels. The spectrum of fuel cycles ranges from natural uranium, through slightly enriched uranium, to plutonium and ultimately thorium fuels which offer many of the advantages of the fast breeder reactor system. CANDU can also burn the recycled uranium and/or the plutonium from fuel discharged from light water reactors. This synergistic relationship could obviate the need to re-enrich the reprocessed uranium and allow a simpler reprocessing scheme. Fule management strategies that will permit future fuel cycles to be used in existing CANDU reactors have been identified. Evolutionary design changes will lead to an even greater flexibility, which will guarantee the continued success of the CANDU system. (author)

Verify Super Double-Heterogeneous Spherical Lattice Model for Equilibrium Fuel Cycle Analysis AND HTR Spherical Super Lattice Model for Equilibrium Fuel Cycle Analysis

International Nuclear Information System (INIS)

Gray S. Chang

2005-01-01

The currently being developed advanced High Temperature gas-cooled Reactors (HTR) is able to achieve a simplification of safety through reliance on innovative features and passive systems. One of the innovative features in these HTRs is reliance on ceramic-coated fuel particles to retain the fission products even under extreme accident conditions. Traditionally, the effect of the random fuel kernel distribution in the fuel pebble/block is addressed through the use of the Dancoff correction factor in the resonance treatment. However, the Dancoff correction factor is a function of burnup and fuel kernel packing factor, which requires that the Dancoff correction factor be updated during Equilibrium Fuel Cycle (EqFC) analysis. An advanced KbK-sph model and whole pebble super lattice model (PSLM), which can address and update the burnup dependent Dancoff effect during the EqFC analysis. The pebble homogeneous lattice model (HLM) is verified by the burnup characteristics with the double-heterogeneous KbK-sph lattice model results. This study summarizes and compares the KbK-sph lattice model and HLM burnup analyzed results. Finally, we discuss the Monte-Carlo coupling with a fuel depletion and buildup code--ORIGEN-2 as a fuel burnup analysis tool and its PSLM calculated results for the HTR EqFC burnup analysis

Reference equilibrium core with central flux irradiation facility for Pakistan research reactor-1

International Nuclear Information System (INIS)

Israr, M.; Shami, Qamar-ud-din; Pervez, S.

1997-11-01

In order to assess various core parameters a reference equilibrium core with Low Enriched Uranium (LEU) fuel for Pakistan Research Reactor (PARR-1) was assembled. Due to increased volume of reference core, the average neutron flux reduced as compared to the first higher power operation. To get a higher neutron flux an irradiation facility was created in centre of the reference equilibrium core where the advantage of the neutron flux peaking was taken. Various low power experiments were performed in order to evaluate control rods worth and neutron flux mapping inside the core. The neutron flux inside the central irradiation facility almost doubled. With this arrangement reactor operation time was cut down from 72 hours to 48 hours for the production of the required specific radioactivity. (author)

Advanced fuel cycles for WWER-1000 reactors

International Nuclear Information System (INIS)

Semchenkov, Y. M.; Pavlovichev, A. M.; Pavlov, V. I.; Spirkin, E. I.; Styrin, Y. A.; Kosourov, E. K.

2007-01-01

Main stages of Russian uranium fuel development regarding improvement of safety and economics of fuel load operation are presented. Intervals of possible changes in fuel cycle duration have been demonstrated for the use of current and perspective fuel. Examples of equilibrium fuel load patterns have been demonstrated and main core neutronics parameters have been presented. Problems on the use of axial blankets with reduced enrichment in WWER-1000 fuel assemblies are considered. Some results are presented regarding core neutronic characteristics of WWER-1000 at the use of regenerated uranium and uranium-plutonium fuel. Examples of equilibrium fuel cycles for the core partially loaded with MOX fuel from weapon-grade plutonium are also considered (Authors)

The integral fast reactor fuel cycle

International Nuclear Information System (INIS)

Chang, Y.I.

1990-01-01

The liquid-metal reactor (LMR) has the potential to extend the uranium resource by a factor of 50 to 100 over current commercial light water reactors (LWRs). In the integral fast reactor (IFR) development program, the entire reactor system - reactor, fuel cycle, and waste process - is being developed and optimized at the same time as a single integral entity. A key feature of the IFR concept is the metallic fuel. The lead irradiation tests on the new U-Pu-Zr metallic fuel in the Experimental Breeder Reactor II have surpassed 185000 MWd/t burnup, and its high burnup capability has now been fully demonstrated. The metallic fuel also allows a radically improved fuel cycle technology. Pyroprocessing, which utilizes high temperatures and molten salt and molten metal solvents, can be advantageously utilized for processing metal fuels because the product is metal suitable for fabrication into new fuel elements. Direct production of a metal product avoids expensive and cumbersome chemical conversion steps that would result from use of the conventional Purex solvent extraction process. The key step in the IFR process is electrorefining, which provides for recovery of the valuable fuel constituents, uranium and plutonium, and for removal of fission products. A notable feature of the IFR process is that the actinide elements accompany plutonium through the process. This results in a major advantage in the high-level waste management

A model for a countercurrent gas—solid—solid trickle flow reactor for equilibrium reactions. The methanol synthesis

NARCIS (Netherlands)

Westerterp, K.R.; Kuczynski, M.

1987-01-01

The theoretical background for a novel, countercurrent gas—solid—solid trickle flow reactor for equilibrium gas reactions is presented. A one-dimensional, steady-state reactor model is developed. The influence of the various process parameters on the reactor performance is discussed. The physical

LEU fuel cycle analyses for the Belgian BR2 Research Reactor

International Nuclear Information System (INIS)

Deen, J.R.; Snelgrove, J.L.

1988-01-01

Equilibrium fuel cycle characteristics were calculated for reference HEU and two proposed LEU fuel cycles using an 11-group diffusion-theory neutron flux solution in hexagonal-Z geometry. The diffusion theory model was benchmarked with a detailed Monte Carlo core model. The two proposed LEU fuel designs increased the 235 U loading 20% and the fuel meat volume 51%. The first LEU design used 10 B as a burnable absorber. Either proposed LEU fuel element would provide equilibrium fuel cycle characteristics similar to those of the HEU fuel cycle. Irradiation rates of Co control followers and Ir disks in the center of the core were reduced 6 ± 1% in the LEU equilibrium core compared to reference HEU core. 11 refs., 4 figs., 5 tabs

Thermodynamic cycle calculations for a pumped gaseous core fission reactor

International Nuclear Information System (INIS)

Kuijper, J.C.; Van Dam, H.

1991-01-01

Finite and 'infinitesimal' thermodynamic cycle calculations have been performed for a 'solid piston' model of a pumped Gaseous Core Fission Reactor with dissociating reactor gas, consisting of Uranium, Carbon and Fluorine ('UCF'). In the finite cycle calculations the influence has been investigated of several parameters on the thermodynamics of the system, especially on the attainable direct (nuclear to electrical) energy conversion efficiency. In order to facilitate the investigation of the influence of dissociation, a model gas, 'Modelium', was developed, which approximates, in a simplified, analytical way, the dissociation behaviour of the 'real' reactor gas. Comparison of the finite cycle calculation results with those of a so-called infinitesimal Otto cycle calculation leads to the conclusion that the conversion efficiency of a finite cycle can be predicted, without actually performing the finite cycle calculation, with reasonable accuracy, from the so-called 'infinitesimal efficiency factor', which is determined only by the thermodynamic properties of the reactor gas used. (author)

Economic Analysis on Direct Use of Spent Pressurized Water Reactor Fuel in CANDU Reactors - IV: DUPIC Fuel Cycle Cost

International Nuclear Information System (INIS)

Ko, Won Il; Choi, Hangbok; Yang, Myung Seung

2001-01-01

This study examines the economics of the DUPIC fuel cycle using unit costs of fuel cycle components estimated based on conceptual designs. The fuel cycle cost (FCC) was calculated by a deterministic method in which reference values of fuel cycle components are used. The FCC was then analyzed by a Monte Carlo simulation to get the uncertainty of the FCC associated with the unit costs of the fuel cycle components. From the deterministic analysis on the equilibrium fuel cycle model, the DUPIC FCC was estimated to be 6.21 to 6.34 mills/kW.h for DUPIC fuel options, which is a little smaller than that of the once-through FCC by 0.07 to 0.27 mills/kW.h. Considering the uncertainty (0.40 to 0.44 mills/kW.h) of the FCC estimated by the Monte Carlo simulation method, the cost difference between the DUPIC and once-through fuel cycle is negligible. On the other hand, the material balance calculation has shown that the DUPIC fuel cycle can save natural uranium resources by ∼20% and reduce the spent fuel arising by ∼65% compared with the once-through fuel cycle. In conclusion, the DUPIC fuel cycle is comparable with the once-through fuel cycle from the viewpoint of FCC. In the future, it should be important to consider factors such as the environmental benefit owing to natural uranium savings, the capability of reusing spent pressurized water reactor fuel, and the safeguardability of the fuel cycle when deciding on an advanced nuclear fuel cycle option

Preparations for the Integral Fast Reactor fuel cycle demonstration

International Nuclear Information System (INIS)

Lineberry, M.J.; Phipps, R.D.

1989-01-01

Modifications to the Hot Fuel Examination Facility-South (HFEF/S) have been in progress since mid-1988 to ready the facility for demonstration of the unique Integral Fast Reactor (IFR) pyroprocess fuel cycle. This paper updates the last report on this subject to the American Nuclear Society and describes the progress made in the modifications to the facility and in fabrication of the new process equipment. The IFR is a breeder reactor, which is central to the capability of any reactor concept to contribute to mitigation of environmental impacts of fossil fuel combustion. As a fast breeder, fuel of course must be recycled in order to have any chance of an economical fuel cycle. The pyroprocess fuel cycle, relying on a metal alloy reactor fuel rather than oxide, has the potential to be economical even at small-scale deployment. Establishing this quantitatively is one important goal of the IFR fuel cycle demonstration

Fuel cycle cost analysis on molten-salt reactors

International Nuclear Information System (INIS)

Shimazu, Yoichiro

1976-01-01

An evaluation is made of the fuel cycle costs for molten-salt reactors (MSR's), developed at Oak Ridge National Laboratory. Eight combinations of conditions affecting fuel cycle costs are compared, covering 233 U-Th, 235 U-Th and 239 Pu-Th fuels, with and without on-site continuous fuel reprocessing. The resulting fuel cycle costs range from 0.61 to 1.18 mill/kWh. A discussion is also given on the practicability of these fuel cycles. The calculations indicate that somewhat lower fuel cycle costs can be expected from reactor operation in converter mode on 235 U make-up with fuel reprocessed in batches every 10 years to avoid fission product precipitation, than from operation as 233 U-Th breeder with continuous reprocessing. (auth.)

A Novel Fuel/Reactor Cycle to Implement the 300 Years Nuclear Waste Policy Approach - 12377

Energy Technology Data Exchange (ETDEWEB)

Carelli, M.D.; Franceschini, F.; Lahoda, E.J. [Westinghouse Electric Company LLC., Cranberry Township, PA (United States); Petrovic, B. [Georgia Institute of Technology, Atlanta, GA (United States)

2012-07-01

A thorium-based fuel cycle system can effectively burn the currently accumulated commercial used nuclear fuel and move to a sustainable equilibrium where the actinide levels in the high level waste are low enough to yield a radiotoxicity after 300 years lower than that of the equivalent uranium ore. The second step of the Westinghouse approach to solving the waste 'problem' has been completed. The thorium fuel cycle has indeed the potential of burning the legacy TRU and achieve the waste objective proposed. Initial evaluations have been started for the third step, development and selection of appropriate reactors. Indications are that the probability of show-stoppers is rather remote. It is, therefore, believed that development of the thorium cycle and associated technologies will provide a permanent solution to the waste management. Westinghouse is open to the widest collaboration to make this a reality. (authors)

Safeguards operations in the integral fast reactor fuel cycle

International Nuclear Information System (INIS)

Goff, K.M.; Benedict, R.W.; Brumbach, S.B.; Dickerman, C.E.; Tompot, R.W.

1994-01-01

Argonne National Laboratory is currently demonstrating the fuel cycle for the Integral Fast Reactor (IFR), an advanced reactor concept that takes advantage of the properties of metallic fuel and liquid metal cooling to offer significant improvements in reactor safety, operation, fuel-cycle economics, environmental protection, and safeguards. The IFR fuel cycle employs a pyrometallurgical process using molten salts and liquid metals to recover actinides from spent fuel. The safeguards aspects of the fuel cycle demonstration must be approved by the United States Department of Energy, but a further goal of the program is to develop a safeguards system that could gain acceptance from the Nuclear Regulatory Commission and International Atomic Energy Agency. This fuel cycle is described with emphasis on aspects that differ from aqueous reprocessing and on its improved safeguardability due to decreased attractiveness and diversion potential of all process streams, including the fuel product

A BWR 24-month cycle analysis using multicycle techniques

International Nuclear Information System (INIS)

Hartley, K.D.

1993-01-01

Boiling water reactor (BWR) fuel cycle design analyses have become increasingly challenging in the past several years. As utilities continue to seek improved capacity factors, reduced power generation costs, and reduced outage costs, longer cycle lengths and fuel design optimization become important considerations. Accurate multicycle analysis techniques are necessary to determine the viability of fuel designs and cycle operating strategies to meet reactor operating requirements, e.g., meet thermal and reactivity margin constraints, while minimizing overall fuel cycle costs. Siemens Power Corporation (SPC), Nuclear Division, has successfully employed multi-cycle analysis techniques with realistic rodded cycle depletions to demonstrate equilibrium fuel cycle performance in 24-month cycles. Analyses have been performed by a BWR/5 reactor, at both rated and uprated power conditions

Economic analysis of fast reactor fuel cycle with different modes

International Nuclear Information System (INIS)

Ding Xiaoming

2014-01-01

Because of limitations on the access to technical and economic data and the lack of effective verification, the lack of in-depth study on the economy of fast reactor fuel cycle in China. This paper introduces the analysis and calculation results of the levelized cost of electricity (LCOE) under three different fuel cycle modes including fast reactor fuel cycle carried out by Massachusetts Institute of Technology (MIT). The author used the evaluation method and hypothesis parameters provided by the MIT to carry out the sensitivity analysis for the impact of the overnight cost, the discount rate and changes of uranium price on the LCOE under three fuel cycle modes. Finally, some suggestions are proposed on the study of economy in China's fast reactor fuel cycle. (authors)

Parametric Investigation of Brayton Cycle for High Temperature Gas-Cooled Reactor

International Nuclear Information System (INIS)

Chang Oh

2004-01-01

The Idaho National Engineering and Environmental Laboratory (INEEL) is investigating a Brayton cycle efficiency improvement on a high temperature gas-cooled reactor (HTGR) as part of Generation-IV nuclear engineering research initiative. In this project, we are investigating helium Brayton cycles for the secondary side of an indirect energy conversion system. Ultimately we will investigate the improvement of the Brayton cycle using other fluids, such as supercritical carbon dioxide. Prior to the cycle improvement study, we established a number of baseline cases for the helium indirect Brayton cycle. These cases look at both single-shaft and multiple-shaft turbomachinery. The baseline cases are based on a 250 MW thermal pebble bed HTGR. The results from this study are applicable to other reactor concepts such as a very high temperature gas-cooled reactor (VHTR), fast gas-cooled reactor (FGR), supercritical water reactor (SWR), and others. In this study, we are using the HYSYS computer code for optimization of the helium Brayton cycle. Besides the HYSYS process optimization, we performed parametric study to see the effect of important parameters on the cycle efficiency. For these parametric calculations, we use a cycle efficiency model that was developed based on the Visual Basic computer language. As a part of this study we are currently investigated single-shaft vs. multiple shaft arrangement for cycle efficiency and comparison, which will be published in the next paper. The ultimate goal of this study is to use supercritical carbon dioxide for the HTGR power conversion loop in order to improve the cycle efficiency to values great than that of the helium Brayton cycle. This paper includes preliminary calculations of the steady state overall Brayton cycle efficiency based on the pebble bed reactor reference design (helium used as the working fluid) and compares those results with an initial calculation of a CO2 Brayton cycle

Optimum fuel use in PWR reactors

International Nuclear Information System (INIS)

Neubauer, W.

1979-07-01

An optimization program was developed to calculate minimum-cost refuelling schedules for PWR reactors. Optimization was made over several cycles, without any constraints (equilibrium cycle). In developing the optimization program, special consideration was given to an individual treatment of every fuel element and to a sufficiently accurate calculation of all the data required for safe reactor operation. The results of the optimization program were compared with experimental values obtained at Obrigheim nuclear power plant. (orig.) [de

Integrated fuel-cycle models for fast breeder reactors

International Nuclear Information System (INIS)

Ott, K.O.; Maudlin, P.J.

1981-01-01

Breeder-reactor fuel-cycle analysis can be divided into four different areas or categories. The first category concerns questions about the spatial variation of the fuel composition for single loading intervals. Questions of the variations in the fuel composition over several cycles represent a second category. Third, there is a need for a determination of the breeding capability of the reactor. The fourth category concerns the investigation of breeding and long-term fuel logistics. Two fuel-cycle models used to answer questions in the third and fourth area are presented. The space- and time-dependent actinide balance, coupled with criticality and fuel-management constraints, is the basis for both the Discontinuous Integrated Fuel-Cycle Model and the Continuous Integrated Fuel-Cycle Model. The results of the continuous model are compared with results obtained from detailed two-dimensional space and multigroup depletion calculations. The continuous model yields nearly the same results as the detailed calculation, and this is with a comparatively insignificant fraction of the computational effort needed for the detailed calculation. Thus, the integrated model presented is an accurate tool for answering questions concerning reactor breeding capability and long-term fuel logistics. (author)

Waste disposal from the light water reactor fuel cycle

International Nuclear Information System (INIS)

Costello, J.M.; Hardy, C.J.

1981-05-01

Alternative nuclear fuel cycles for support of light water reactors are described and wastes containing naturally occurring or artificially produced radioactivity reviewed. General principles and objectives in radioactive waste management are outlined, and methods for their practical application to fuel cycle wastes discussed. The paper concentrates upon management of wastes from upgrading processes of uranium hexafluoride manufacture and uranium enrichment, and, to a lesser extent, nuclear power reactor wastes. Some estimates of radiological dose commitments and health effects from nuclear power and fuel cycle wastes have been made for US conditions. These indicate that the major part of the radiological dose arises from uranium mining and milling, operation of nuclear reactors, and spent fuel reprocessing. However, the total dose from the fuel cycle is estimated to be only a small fraction of that from natural background radiation

Core performance of equilibrium fast reactors for different coolant materials and fuel types

International Nuclear Information System (INIS)

Mizutani, Akihiko; Sekimoto, Hiroshi

1998-01-01

Parametric studies with several coolant and fuel materials in the equilibrium state are performed for fast reactors in which natural uranium is fed and all of the actinides are confined. Sodium, sodium-potassium, lead, lead-bismuth and helium coolant materials, and oxide, nitride and metal fuels are employed to compare the neutronic characteristics in the equilibrium state. As to the criticality performance, sodium-potassium shows the best performance among the liquid metal coolants and the metallic fuel indicates the best performance

The plutonium fuel cycles

International Nuclear Information System (INIS)

Pigford, T.H.; Ang, K.P.

1975-01-01

The quantities of plutonium and other fuel actinides have been calculated for equilibrium fuel cycles for 1000-MW water reactors fueled with slightly enriched uranium, water reactors fueled with plutonium and natural uranium, fast-breder reactors, gas-cooled reactors fueled with thorium and highly enriched uranium, and gas-cooled reactors fueled with thorium, plutonium and recycled uranium. The radioactivity quantities of plutonium, americium and curium processed yearly in these fuel cycles are greatest for the water reactors fueled with natural uranium and recycled plutonium. The total amount of actinides processed is calculated for the predicted future growth of the U.S. nuclear power industry. For the same total installed nuclear power capacity, the introduction of the plutonium breeder has little effect upon the total amount of plutonium in this century. The estimated amount of plutonium in the low-level process wastes in the plutonium fuel cycles is comparable to the amount of plutonium in the high-level fission product wastes. The amount of plutonium processed in the nuclear fuel cycles can be considerably reduced by using gas-cooled reactors to consume plutonium produced in uranium-fueled water reactors. These, and other reactors dedicated for plutonium utilization, could be co-located with facilities for fuel reprocessing ad fuel fabrication to eliminate the off-site transport of separated plutonium. (author)

Enriched uranium cycles in pressurized heavy water reactors

International Nuclear Information System (INIS)

Mazzola, A.

1994-01-01

A study was made on the substitution of natural uranium with enriched and on plutonium recycle in unmodified PHWRs (pressure vessel reactor). Results clearly show the usefulness of enriched fuel utilisation for both uranium ore consumption (savings of 30% around 1.3% enrichment) and decreasing fuel cycle coasts. This is also due to a better plutonium exploitation during the cycle. On the other hand plutonium recycle in these reactors via MOX-type fuel appears economically unfavourable under any condition

The relationship between natural uranium and advanced fuel cycles in CANDU reactors

International Nuclear Information System (INIS)

Lane, A.D.; McDonnell, F.N.; Griffiths, J.

1988-11-01

CANDU is the most uranium-economic type of thermal power reactor, and is the only type used in Canada. CANDU reactors consume approximately 15% of Canadian uranium production and support a fuel service industry valued at ∼$250 M/a. In addition to their once-through, natural-uranium fuel cycle, CANDU reactors are capable of operating with slightly-enriched uranium (SEU), uranium-plutonium and thorium cycles, more efficiently than other reactors. Only SEU is economically attractive in Canada now, but the other cycles are of interest to countries without indigenous fuel resources. A program is underway to establish the fuel technologies necessary for the use of SEU and the other fuel cycles in CANDU reactors. 22 refs

Safety considerations in the fast reactor fuel cycle

International Nuclear Information System (INIS)

Baker, A.R.; Burton, W.R.; Taylor, H.A.

1977-01-01

The fuel cycle safety problems for fast reactors, as compared with thermal reactors, are enhanced by the higher fissile content and heat rating of the fuel. Additionally recycling leads to the build up of substantial isotopes which contribute to the alpha and neutron hazards. The plutonium arisings in a nuclear power reactor programme extending into the next century are discussed. A requirement is to be able to return the product plutonium to a reactor about 9 months after the end of irradiation and it is anticipated that progress will be made slowly towards this fuel cycle, having regard to the necessity for maintaining safe and reliable operations. Consideration of the steps in the fuel cycle has indicated that it will be best to store the irradiated fuel on the reactor sites while I131 decays and decay heat falls before transporting and a suitable transport flask is being developed. Reprocessing development work is aimed at the key area of fuel breakdown, the inter-relation of the fuel characteristics on the dissolution of the plutonium and a solvent extract cycle leading to a product suitable for a co-located fabrication plant. Because of the high activity of recycled fuel it is considered that fabrication must move to a fully remote operation as is already the case for reprocessing, and a gel precipitation process producing a vibro compacted fuel is under development for this purpose. The waste streams from the processing plants must be minimised, processed for recovery of plutonium where applicable and then conditioned so that the final products released from the processing cycle are acceptable for ultimate disposal. The safety aspects reviewed cover protection of operators, containment of radioactive materials, criticality and regulation of discharges to the environment

V.S.O.P. (99/09) computer code system for reactor physics and fuel cycle simulation. Version 2009

Energy Technology Data Exchange (ETDEWEB)

Ruetten, H.J.; Haas, K.A.; Brockmann, H.; Ohlig, U.; Pohl, C.; Scherer, W.

2010-07-15

V.S.O.P. (99/ 09) represents the further development of V.S.O.P. (99/ 05). Compared to its precursor, the code system has been improved again in many details. The main motivation for this new code version was to update the basic nuclear libraries used by the code system. Thus, all cross section libraries involved in the code have now been based on ENDF/B-VII. V.S.O.P. is a computer code system for the comprehensive numerical simulation of the physics of thermal reactors. It implies the setup of the reactor and of the fuel element, processing of cross sections, neutron spectrum evaluation, neutron diffusion calculation in two or three dimensions, fuel burnup, fuel shuffling, reactor control, thermal hydraulics and fuel cycle costs. The thermal hydraulics part (steady state and time-dependent) is restricted to gas-cooled reactors and to two spatial dimensions. The code can simulate the reactor operation from the initial core towards the equilibrium core. This latest code version was developed and tested under the WINDOWS-XP - operating system. (orig.)

V.S.O.P. (99/09) computer code system for reactor physics and fuel cycle simulation. Version 2009

International Nuclear Information System (INIS)

Ruetten, H.J.; Haas, K.A.; Brockmann, H.; Ohlig, U.; Pohl, C.; Scherer, W.

2010-07-01

V.S.O.P. (99/ 09) represents the further development of V.S.O.P. (99/ 05). Compared to its precursor, the code system has been improved again in many details. The main motivation for this new code version was to update the basic nuclear libraries used by the code system. Thus, all cross section libraries involved in the code have now been based on ENDF/B-VII. V.S.O.P. is a computer code system for the comprehensive numerical simulation of the physics of thermal reactors. It implies the setup of the reactor and of the fuel element, processing of cross sections, neutron spectrum evaluation, neutron diffusion calculation in two or three dimensions, fuel burnup, fuel shuffling, reactor control, thermal hydraulics and fuel cycle costs. The thermal hydraulics part (steady state and time-dependent) is restricted to gas-cooled reactors and to two spatial dimensions. The code can simulate the reactor operation from the initial core towards the equilibrium core. This latest code version was developed and tested under the WINDOWS-XP - operating system. (orig.)

Logistics of the research reactor fuel cycle: AREVA solutions

International Nuclear Information System (INIS)

Ohayon, David; Halle, Laurent; Naigeon, Philippe; Falgoux, Jean-Louis; Franck Obadia, Franck; Auziere, Philippe

2005-01-01

The AREVA Group Companies offer comprehensive solutions for the entire fuel cycle of Research Reactors comply with IAEA standards. CERCA and Cogema Logistics have developed a full partnership in the front end cycle. In the field of uranium CERCA and Cogema Logistics have the long term experience of the shipment from Russia, USA to the CERCA plant.. Since 1960, CERCA has manufactured over 300,000 fuel plates and 15,000 fuel elements of more than 70 designs. These fuel elements have been delivered to 40 research reactors in 20 countries. For the Back-End stage, Cogema and Cogema Logistics propose customised solutions and services for international shipments. Cogema Logistics has developed a new generation of packaging to meet the various needs and requirements of the Laboratories and Research Reactors all over the world, and complex regulatory framework. Comprehensive assistance dedicated, services, technical studies, packaging and transport systems are provided by AREVA for every step of research reactor fuel cycle. (author)

Study on the fuel cycle cost of gas turbine high temperature reactor (GTHTR300). Contract research

Energy Technology Data Exchange (ETDEWEB)

Takei, Masanobu; Katanishi, Shoji; Nakata, Tetsuo; Kunitomi, Kazuhiko [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Oda, Takefumi; Izumiya, Toru [Nuclear Fuel Industries, Ltd., Tokyo (Japan)

2002-11-01

In the basic design of gas turbine high temperature reactor (GTHTR300), reduction of the fuel cycle cost has a large benefit of improving overall plant economy. Then, fuel cycle cost was evaluated for GTHTR300. First, of fuel fabrication for high-temperature gas cooled reactor, since there was no actual experience with a commercial scale, a preliminary design for a fuel fabrication plant with annual processing of 7.7 ton-U sufficient four GTHTR300 was performed, and fuel fabrication cost was evaluated. Second, fuel cycle cost was evaluated based on the equilibrium cycle of GTHTR300. The factors which were considered in this cost evaluation include uranium price, conversion, enrichment, fabrication, storage of spent fuel, reprocessing, and waste disposal. The fuel cycle cost of GTHTR300 was estimated at about 1.07 yen/kWh. If the back-end cost of reprocessing and waste disposal is included and assumed to be nearly equivalent to LWR, the fuel cycle cost of GTHTR300 was estimated to be about 1.31 yen/kWh. Furthermore, the effects on fuel fabrication cost by such of fuel specification parameters as enrichment, the number of fuel types, and the layer thickness were considered. Even if the enrichment varies from 10 to 20%, the number of fuel types change from 1 to 4, the 1st layer thickness of fuel changes by 30 {mu}m, or the 2nd layer to the 4th layer thickness of fuel changes by 10 {mu}m, the impact on fuel fabrication cost was evaluated to be negligible. (author)

Dynamic analysis of Korean nuclear fuel cycle with fast reactor systems

International Nuclear Information System (INIS)

Jeong, Chang Joon

2004-12-01

The Korean nuclear fuel cycle scenario was analyzed by the dynamic analysis method, including Pressurized Water Reactor (PWR), Canadian Deuterium Uranium (CANDU) and fast reactor systems. For the once-through fuel cycle model, the existing nuclear power plant construction plan was considered up to 2016, while the nuclear demand growth rate from the year 2016 was assumed to be 1%. After setting up the once-through fuel cycle model, the Korea Advanced Liquid Metal Reactor (KALIMER) scenario was modeled to investigate the fuel cycle parameters. For the analysis of the fast reactor fuel cycle, both KAILMER-150 and KALIMER-600 reactors were considered. In this analysis, the spent fuel inventory as well as the amount of plutonium, Minor Actinides (MA) and Fission Products (FP) of the recycling fuel cycle was estimated and compared to that of the once-through fuel cycle. Results of the once-through fuel cycle calculation showed that the demand grows up to 64 GWe and total amount of spent fuel would be ∼102 kt in 2100. If the KALIMER scenario is implemented, the total spent fuel inventory can be reduced by ∼80%. However it was found that the KALIMER scenario does not contribute to reduce the amount of MA and FP, which is important when designing a repository. For the further destruction of MA, an actinide burner can be considered in the future nuclear fuel cycle

Once-through cycle, supercritical-pressure light water cooled reactor concept

Energy Technology Data Exchange (ETDEWEB)

Oka, Y.; Koshizuka, S. [Tokyo Univ., Tokai, Ibaraki (Japan). Nuclear Engineering Research Lab

2001-07-01

Concept of once-through cycle, supercritical-pressure light water cooled reactors was developed. The research covered major aspects of conceptual design such as cores of thermal and fast reactors, plant system and heat balance, safety system and criteria, accident and transient analysis, LOCA, PSA, plant control and start-up. The advantages of the reactor lie in the compactness of the plant from high specific enthalpy of supercritical water, the simplicity of the once-through cycle and the experiences of major component technologies which are based on supercritical fossil-fired power plants and LWRs. The operating temperatures of the major components are within the experience in spite of high coolant outlet temperature. The once-through cycle is compatible with the tight fuel lattice fast reactor because of high head pumps and small coolant flow rate. (author)

Once-through cycle, supercritical-pressure light water cooled reactor concept

International Nuclear Information System (INIS)

Oka, Y.; Koshizuka, S.

2001-01-01

Concept of once-through cycle, supercritical-pressure light water cooled reactors was developed. The research covered major aspects of conceptual design such as cores of thermal and fast reactors, plant system and heat balance, safety system and criteria, accident and transient analysis, LOCA, PSA, plant control and start-up. The advantages of the reactor lie in the compactness of the plant from high specific enthalpy of supercritical water, the simplicity of the once-through cycle and the experiences of major component technologies which are based on supercritical fossil-fired power plants and LWRs. The operating temperatures of the major components are within the experience in spite of high coolant outlet temperature. The once-through cycle is compatible with the tight fuel lattice fast reactor because of high head pumps and small coolant flow rate. (author)

Once-through uranium thorium fuel cycle in CANDU reactors

International Nuclear Information System (INIS)

Ozdemir, S.; Cubukcu, E.

2000-01-01

In this study, the performance of the once-through uranium-thorium fuel cycle in CANDU reactors is investigated. (Th-U)O 2 is used as fuel in all fuel rod clusters where Th and U are mixed homogeneously. CANDU reactors have the advantage of being capable of employing various fuel cycle options because of its good neutron economy, continuous on line refueling ability and axial fuel replacement possibility. For lattice cell calculations transport code WIMS is used. WIMS cross-section library is modified to achieve precise lattice cell calculations. For various enrichments and Th-U mixtures, criticality, heavy element composition changes, diffusion coefficients and cross-sections are calculate. Reactor core is modeled by using the diffusion code CITATION. We conclude that an overall saving of 22% in natural uranium demand can be achieved with the use of Th cycle. However, slightly enriched U cycle still consumes less natural Uranium and is a lot less complicated. (author)

Effect of DUPIC cycle on CANDU reactor safety parameters

Energy Technology Data Exchange (ETDEWEB)

Mohamed, Nader M. A. [Atomic Energy Authority, ETRR-2, Cairo (Egypt); Badawi, Alya [Dept. of Nuclear and Radiation Engineering, Alexandria University, Alexandria (Egypt)

2016-10-15

Although, the direct use of spent pressurized water reactor (PWR) fuel in CANda Deuterium Uranium (CANDU) reactors (DUPIC) cycle is still under investigation, DUPIC cycle is a promising method for uranium utilization improvement, for reduction of high level nuclear waste, and for high degree of proliferation resistance. This paper focuses on the effect of DUPIC cycle on CANDU reactor safety parameters. MCNP6 was used for lattice cell simulation of a typical 3,411 MWth PWR fueled by UO{sub 2} enriched to 4.5w/o U-235 to calculate the spent fuel inventories after a burnup of 51.7 MWd/kgU. The code was also used to simulate the lattice cell of CANDU-6 reactor fueled with spent fuel after its fabrication into the standard 37-element fuel bundle. It is assumed a 5-year cooling time between the spent fuel discharges from the PWR to the loading into the CANDU-6. The simulation was carried out to calculate the burnup and the effect of DUPIC fuel on: (1) the power distribution amongst the fuel elements of the bundle; (2) the coolant void reactivity; and (3) the reactor point-kinetics parameters.

Pebble Bed Reactor: core physics and fuel cycle analysis

Energy Technology Data Exchange (ETDEWEB)

Vondy, D.R.; Worley, B.A.

1979-10-01

The Pebble Bed Reactor is a gas-cooled, graphite-moderated high-temperature reactor that is continuously fueled with small spherical fuel elements. The projected performance was studied over a broad range of reactor applicability. Calculations were done for a burner on a throwaway cycle, a converter with recycle, a prebreeder and breeder. The thorium fuel cycle was considered using low, medium (denatured), and highly enriched uranium. The base calculations were carried out for electrical energy generation in a 1200 MW/sub e/ plant. A steady-state, continuous-fueling model was developed and one- and two-dimensional calculations were used to characterize performance. Treating a single point in time effects considerable savings in computer time as opposed to following a long reactor history, permitting evaluation of reactor performance over a broad range of design parameters and operating modes.

Basic researches on thermo-hydraulic non-equilibrium phenomena related to nuclear reactor safety

International Nuclear Information System (INIS)

Sakurai, Akira; Kataoka, Isao; Aritomi, Masanori.

1989-01-01

A review was made of recent developments of fundamental researches on thermo-hydraulic non-equilibrium phenomena related to light water reactor safety, in relation to problems to be solved for the improvement of safety analysis codes. As for the problems related to flow con ditions, fundamental researches on basic conservation equations and constitutive equations for transient two-phase flow were reviewed. Regarding to the problems related to thermal non-equilibrium phenomena, fundamental researches on film boiling in pool and forced convection, transient boiling heat transfer and flow behavior caused by pressure transients were reviewed. (author)

A prospect of fast reactor and related fuel cycle in Japan

Energy Technology Data Exchange (ETDEWEB)

Nagata, Takashi [Japan Atomic Energy Agency, Ibaraki (Japan)

2009-04-15

JAEA has launched a new project 'Fast Reactor Cycle Technology Development'(FaCT) in cooperation with electric utilities. In this FaCT project, a combination of 'the Japanese sodium cooled loop type fast reactor with oxide fuel, the advanced aqueous reprocessing, and the simplified palletizing fuel fabrication systems' is adopted, where many innovative technologies with technical challenging issues are actively used in order to provide significant improvements in economic competitiveness, and enhancement of safety and reliability, sustainability, and nonproliferation. Fast reactor cycle technology will provide harmonic solutions for global issues of energy resources and environments, and is expected to contribute to sustainable development of the future society. Therefore, it was selected as one of key technologies of national importance in the third term (JPY2006-2010) 'Science and Technology Basic Plan' in March 2006 in Japan. The 'Nuclear Energy National Plan' in August 2006 states start up of a demonstration FR by around 2025 and deployment of a commercial FR before 2050, and start operating fuel cycle facilities when these reactors achieve consistency. Accordingly, we will decide about the adoption of innovative technologies by judging their applicability by 2010, and present the conceptual designs of commercial and demonstration FR cycle facilities by 2015 with the R and D plans to realize. In developing the FR cycle, 5 Party council, which consists of MEXt, MITI, electricity utilities, manufacturers, and JAEA, was established in July 2006 for moving forward on the commercialization smoothly. In this framework, users' requirements for the future R and D, a scenario of transition from light water reactor cycle to sodium cooled FR cycle, international collaboration, development schedule, demonstration steps, and so on are discussed. In this presentation, a prospect concerning the system design features of JSFR and a

A prospect of fast reactor and related fuel cycle in Japan

International Nuclear Information System (INIS)

Nagata, Takashi

2009-01-01

JAEA has launched a new project 'Fast Reactor Cycle Technology Development'(FaCT) in cooperation with electric utilities. In this FaCT project, a combination of 'the Japanese sodium cooled loop type fast reactor with oxide fuel, the advanced aqueous reprocessing, and the simplified palletizing fuel fabrication systems' is adopted, where many innovative technologies with technical challenging issues are actively used in order to provide significant improvements in economic competitiveness, and enhancement of safety and reliability, sustainability, and nonproliferation. Fast reactor cycle technology will provide harmonic solutions for global issues of energy resources and environments, and is expected to contribute to sustainable development of the future society. Therefore, it was selected as one of key technologies of national importance in the third term (JPY2006-2010) 'Science and Technology Basic Plan' in March 2006 in Japan. The 'Nuclear Energy National Plan' in August 2006 states start up of a demonstration FR by around 2025 and deployment of a commercial FR before 2050, and start operating fuel cycle facilities when these reactors achieve consistency. Accordingly, we will decide about the adoption of innovative technologies by judging their applicability by 2010, and present the conceptual designs of commercial and demonstration FR cycle facilities by 2015 with the R and D plans to realize. In developing the FR cycle, 5 Party council, which consists of MEXt, MITI, electricity utilities, manufacturers, and JAEA, was established in July 2006 for moving forward on the commercialization smoothly. In this framework, users' requirements for the future R and D, a scenario of transition from light water reactor cycle to sodium cooled FR cycle, international collaboration, development schedule, demonstration steps, and so on are discussed. In this presentation, a prospect concerning the system design features of JSFR and a summary of the above R and D progresses for

Preliminary Neutronic Design of High Burnup OTTO Cycle Pebble Bed Reactor

OpenAIRE

Setiadipura, T; Irwanto, D; Zuhair, Zuhair

2015-01-01

The pebble bed type High Temperature Gas-cooled Reactor (HTGR) is among the interesting nuclear reactor designs in terms of safety and flexibility for co-generation applications. In addition, the strong inherent safety characteristics of the pebble bed reactor (PBR) which is based on natural mechanisms improve the simplicity of the PBR design, in particular for the Once-Through-Then-Out (OTTO) cycle PBR design. One of the important challenges of the OTTO cycle PBR design, and nuclear reactor ...

Nuclear reactor fuel cycle technology with pyroelectrochemical processes

International Nuclear Information System (INIS)

Skiba, O.V.; Maershin, A.A.; Bychkov, A.V.; Zhdanov, A.N.; Kislyj, V.A.; Vavilov, S.K.; Babikov, L.G.

1999-01-01

A group of dry technologies and processes of vibro-packing granulated fuel in combination with unique properties of vibro-packed FEs make it possible to implement a new comprehensive approach to the fuel cycle with plutonium fuel. Testing of a big number of FEs with vibro-packed U-Pu oxide fuel in the BOR-60 reactor, successful testing of experimental FSAs in the BN-600 rector, reliable operation of the experimental and research complex facilities allow to make the conclusion about a real possibility to develop a safe, economically beneficial U-Pu fuel cycle based on the technologies enumerated above and to use both reactor-grade and weapon-grade plutonium in nuclear reactors with a reliable control and accounting system [ru

A Fast Numerical Method for the Calculation of the Equilibrium Isotopic Composition of a Transmutation System in an Advanced Fuel Cycle

Directory of Open Access Journals (Sweden)

F. Álvarez-Velarde

2012-01-01

Full Text Available A fast numerical method for the calculation in a zero-dimensional approach of the equilibrium isotopic composition of an iteratively used transmutation system in an advanced fuel cycle, based on the Banach fixed point theorem, is described in this paper. The method divides the fuel cycle in successive stages: fuel fabrication, storage, irradiation inside the transmutation system, cooling, reprocessing, and incorporation of the external material into the new fresh fuel. The change of the fuel isotopic composition, represented by an isotope vector, is described in a matrix formulation. The resulting matrix equations are solved using direct methods with arbitrary precision arithmetic. The method has been successfully applied to a double-strata fuel cycle with light water reactors and accelerator-driven subcritical systems. After comparison to the results of the EVOLCODE 2.0 burn-up code, the observed differences are about a few percents in the mass estimations of the main actinides.

Fuel-management simulations for once-through thorium fuel cycle in CANDU reactors

International Nuclear Information System (INIS)

Chan, P.S.W.; Boczar, P.G.; Ellis, R.J.; Ardeshiri, F.

1999-01-01

High neutron economy, on-power refuelling and a simple fuel bundle design result in unsurpassed fuel cycle flexibility for CANDU reactors. These features facilitate the introduction and exploitation of thorium fuel cycles in existing CANDU reactors in an evolutionary fashion. Detailed full-core fuel-management simulations concluded that a once-through thorium fuel cycle can be successfully implemented in an existing CANDU reactor without requiring major modifications. (author)

An overview of the potential of the CANDU reactor as a thermal breeder

International Nuclear Information System (INIS)

Slater, J.B.

1977-02-01

This paper is concerned with the use of thorium as a fuel in the existing CANDU concept. The neutron balance of the reactor core is analyzed and an assessment is made of the potential for development of a thermal 'breeder' reactor system. It is concluded that while the SSET cycle (i.e. self-sufficient equilibrium thorium cycle) appears feasible, there is little potential for developing a significant 'breeding' fuel cycle if current reactor operating capability and capital costs are to be maintained. (author)

Equilibrium and pre-equilibrium emissions in proton-induced ...

Indian Academy of Sciences (India)

necessary for the domain of fission-reactor technology for the calculation of nuclear transmutation ... tions occur in three stages: INC, pre-equilibrium and equilibrium (or compound. 344. Pramana ... In the evaporation phase of the reaction, the.

Features of supercritical carbon dioxide Brayton cycle coupled with reactor

International Nuclear Information System (INIS)

Duan Chengjie; Wang Jie; Yang Xiaoyong

2010-01-01

In order to obtain acceptable cycle efficiency, current helium gas turbine power cycle technology needs high cycle temperature which means that the cycle needs high core-out temperature. The technology has high requirements on reactor structure and fuel elements materials, and also on turbine manufacture. While utilizing CO 2 as cycle working fluid, it can guarantee to lower the cycle temperature and turbo machine Janume but achieve the same cycle efficiency, so as to enhance the safety and economy of reactor. According to the laws of thermodynamics, a calculation model of supercritical CO 2 power cycle was established to analyze the feature, and the decisive parameters of the cycle and also investigate the effect of each parameter on the cycle efficiency in detail were obtained. The results show that supercritical CO 2 power cycle can achieve quite satisfied efficiency at a lower cycle highest temperature than helium cycle, and CO 2 is a promising working fluid. (authors)

SPES, Fuel Cycle Optimization for LWR

International Nuclear Information System (INIS)

1973-01-01

1 - Nature of physical problem solved: Determination of optimal fuel cycle at equilibrium for a light water reactor taking into account batch size, fuel enrichment, de-rating, shutdown time, cost of replacement energy. 2 - Method of solution: Iterative method

Evaluation of remote maintenance schemes by plasma equilibrium analysis in Tokamak DEMO reactor

International Nuclear Information System (INIS)

Utoh, Hiroyasu; Tobita, Kenji; Asakura, Nobuyuki; Sakamoto, Yoshiteru

2014-01-01

Highlights: • The remote maintenance schemes in DEMO reactor were evaluated by the plasma equilibrium analysis. • Horizontal sector transport maintenance scheme requires the largest total PF coil current. • The difference of total PF coil current for MHD equilibrium in between the large segmented divertor maintenance and the segmentalized divertor maintenance was about 10%. - Abstract: The remote maintenance schemes in a DEMO reactor are categorized by insertion direction, blanket segmentation, and divertor maintenance scheme, and are quantitatively evaluated by analysing the plasma equilibrium. The positions of the poloidal field (PF) coil are limited by the size of the toroidal field (TF) coil and the maintenance port layout of each remote maintenance scheme. Because the PF coils are located near the larger TF coil and far from the plasma surface, the horizontal sector transport maintenance scheme requires the largest part of total PF coil current, 25% larger than that required for separated sector transport using vertical maintenance ports with segmented divertor maintenance (SDM). In the unsegmented divertor maintenance (UDM) scheme, the total magnetic stored energy in the PF coils at plasma equilibrium is about 30% larger than that stored in the SDM scheme, but the time required for removal and installation of all the divertor cassettes in the UDM scheme is roughly a third of that required in the SDM scheme because the number of divertor cassettes in the UDM scheme is a third of that in the SDM scheme. From the viewpoint of simple maintenance operations, the merit of the UDM scheme has more merit than the SDM scheme

IAEA Activities in the Area of Fast Reactors and Related Fuels and Fuel Cycles

International Nuclear Information System (INIS)

Monti, S.; Basak, U.; Dyck, G.; Inozemtsev, V.; Toti, A.; Zeman, A.

2013-01-01

Summary: • The IAEA role to support fast reactors and associated fuel cycle development programmes; • Main IAEA activities on fast reactors and related fuel and fuel cycle technology; • Main IAEA deliverables on fast reactors and related fuel and fuel cycle technology

Energy storage for tokamak reactor cycles

International Nuclear Information System (INIS)

Buchanan, C.H.

1979-01-01

The inherent characteristic of a tokamak reactor requiring periodic plasma quench and reignition introduces the problem of energy storage to permit continuous electrical output to the power grid. The cycle under consideration in this paper is a 1000 second burn followed by a 100 second reignition phase. The physical size of a typical toroidal plasma reaction chamber for a tokamak reactor has been described earlier. The thermal energy storage requirements described in this reference will serve as a basis for much of the ensuing discussion

Transuranic material recovery in the Integral Fast Reactor fuel cycle demonstration

International Nuclear Information System (INIS)

Benedict, R.W.; Goff, K.M.

1993-01-01

The Integral Fast Reactor is an innovative liquid metal reactor concept that is being developed by Argonne National Laboratory. It takes advantage of the properties of metallic fuel and liquid metal cooling to offer significant improvements in reactor safety, operation, fuel cycle economics, environmental protection, and safeguards. The plans for demonstrating the IFR fuel cycle, including its waste processing options, by processing irradiated fuel from the Experimental Breeder Reactor-II fuel in its associated Fuel Cycle Facility have been developed for the first refining series. This series has been designed to provide the data needed for the further development of the IFR program. An important piece of the data needed is the recovery of TRU material during the reprocessing and waste operations

Benefits of cycle stretchout in pressurized water reactor extended-burnup fuel cycles

International Nuclear Information System (INIS)

Matzie, R.A.; Leung, D.C.; Liu, Y.; Beekmann, R.W.

1981-01-01

Nuclear reactors are inherently capable of operating for a substantial period beyond their nominal end of cycle (EOC) as a result of negative moderator and fuel temperature coefficients and the decrease in xenon poisoning with lower core power levels. This inherent capability can be used to advantage to reduce annual uranium makeup requirements and cycle energy costs by the use of planned EOC stretchout. This paper discusses the fuel utilization efficiency and economics of both the five-batch, extended-burnup cycle and the three-batch, standard-burnup cycle, which can be improved by employing planned EOC (end of cycle) stretchout. 11 refs

Preliminary Design of S-CO2 Brayton Cycle for KAIST Micro Modular Reactor

International Nuclear Information System (INIS)

Kim, Seong Gu; Kim, Min Gil; Bae, Seong Jun; Lee, Jeong Ik

2013-01-01

This paper suggests a complete modular reactor with an innovative concept of reactor cooling by using a supercritical carbon dioxide directly. Authors propose the supercritical CO 2 Brayton cycle (S-CO 2 cycle) as a power conversion system to achieve small volume of power conversion unit (PCU) and to contain the core and PCU in one vessel for the full modularization. This study suggests a conceptual design of small modular reactor including PCU which is named as KAIST Micro Modular Reactor (MMR). As a part of ongoing research of conceptual design of KAIST MMR, preliminary design of power generation cycle was performed in this study. Since the targets of MMR are full modularization of a reactor system with S-CO 2 coolant, authors selected a simple recuperated S-CO 2 Brayton cycle as a power conversion system for KAIST MMR. The size of components of the S-CO 2 cycle is much smaller than existing helium Brayton cycle and steam Rankine cycle, and whole power conversion system can be contained with core and safety system in one containment vessel. From the investigation of the power conversion cycle, recompressing recuperated cycle showed higher efficiency than the simple recuperated cycle. However the volume of heat exchanger for recompressing cycle is too large so more space will be occupied by heat exchanger in the recompressing cycle than the simple recuperated cycle. Thus, authors consider that the simple recuperated cycle is more suitable for MMR. More research for the KAIST MMR will be followed in the future and detailed information of reactor core and safety system will be developed down the road. More refined cycle layout and design of turbomachinery and heat exchanger will be performed in the future study

Application of the complex equilibrium code QUIL to cesium-impurity equilibria in the primary coolant of high-temperature gas-cooled reactors

International Nuclear Information System (INIS)

Feber, R.D.; Lunsford, J.L.; Stark, W.A. Jr.

1976-05-01

An equilibrium analysis has been made of the fission-product cesium in the primary coolant loop of the high-temperature gas-cooled reactor (HTGR). The species distributions that result at equilibrium have been calculated for various conditions of reactor operation. The cesium species considered were the monomer, dimer, oxides, hydroxides, and the hydride. The effect of cesium sorption isotherms on graphite also was included in the analysis. During normal reactor operations, the abundant species of cesium were calculated to be elemental cesium, Cs, and the monomeric hydroxide, CsOH. Under most conditions of steam ingress, the abundant species was calculated to be CsOH. Cesium adsorbed onto graphite was stable under all steam-ingress conditions considered. Thermal transients above 1500 0 K were required for equilibrium transport of cesium from the core to the coolant. The analysis was carried out using the complex equilibrium code QUIL, designed and written with special emphasis on features that make it applicable to the fission-product problem

Supercritical-pressure, once-through cycle light water cooled reactor concept

International Nuclear Information System (INIS)

Oka, Yoshiaki; Koshizuka, Seiichi

2001-01-01

The purpose of the study is to develop new reactor concepts for the innovation of light water reactors (LWR) and fast reactors. Concept of the once-through coolant cycle, supercritical-pressure light water cooled reactor was developed. Major aspects of reactor design and safety were analysed by the computer codes which were developed by ourselves. It includes core design of thermal and fast reactors, plant system, safety criteria, accident and transient analysis, LOCA, PSA, plant control, start up and stability. High enthalpy rise as supercritical boiler was achieved by evaluating the cladding temperature directly during transients. Fundamental safety principle of the reactor is monitoring coolant flow rate instead of water level of LWR. The reactor system is compact and simple because of high specific enthalpy of supercritical water and the once-through cycle. The major components are similar to those of LWR and supercritical thermal plant. Their temperature are within the experiences in spite of the high outlet coolant temperature. The reactor is compatible with tight fuel lattice fast reactor because of the high head pumps and low coolant flow rate. The power rating of the fast reactor is higher than the that of thermal reactor because of the high power density. (author)

A study on different thermodynamic cycle schemes coupled with a high temperature gas-cooled reactor

International Nuclear Information System (INIS)

Qu, Xinhe; Yang, Xiaoyong; Wang, Jie

2017-01-01

Highlights: • The features of three different power generation schemes, including closed Brayton cycle, non-reheating combined cycle and reheating combined cycle, coupled with high temperature gas-cooled reactor (HTGR) were investigated and compared. • The effects and mechanism of reactor core outlet temperature, compression ratio and other key parameters over cycle characteristics were analyzed by the thermodynamic models.. • It is found that reheated combined cycle has the highest efficiency. Reactor outlet temperature and main steam parameters are key factors to improve the cycle’s performance. - Abstract: With gradual increase in reactor outlet temperature, the efficient power conversion technology has become one of developing trends of (very) high temperature gas-cooled reactors (HTGRs). In this paper, different cycle power generation schemes for HTGRs were systematically studied. Physical and mathematical models were established for these three cycle schemes: closed Brayton cycle, simple combined cycle, and reheated combined cycle. The effects and mechanism of key parameters such as reactor core outlet temperature, reactor core inlet temperature and compression ratio on the features of these cycles were analyzed. Then, optimization results were given with engineering restrictive conditions, including pinch point temperature differences. Results revealed that within the temperature range of HTGRs (700–900 °C), the reheated combined cycle had the highest efficiency, while the simple combined cycle had the lowest efficiency (900 °C). The efficiencies of the closed Brayton cycle, simple combined cycle and reheated combined cycle are 49.5%, 46.6% and 50.1%, respectively. These results provide insights on the different schemes of these cycles, and reveal the effects of key parameters on performance of these cycles. It could be helpful to understand and develop a combined cycle coupled with a high temperature reactor in the future.

Multi-faceted evaluation for nuclear fuel cycles with transmutation

International Nuclear Information System (INIS)

Nishihara, Kenji

2015-03-01

Environment impact, economy and proliferation resistance were estimated for nuclear fuel cycles involving transmutation by fast reactor and accelerator-driven system in equilibrium state. As a result, the transmutation scenario using only fast reactor was superior to the scenarios combined with accelerator-driven system in all estimation, but the differences were insignificant. (author)

On the vapor-liquid equilibrium in hydroprocessing reactors

Energy Technology Data Exchange (ETDEWEB)

Chen, J.; Munteanu, M.; Farooqi, H. [National Centre for Upgrading Technology, Devon, AB (Canada)

2009-07-01

When petroleum distillates undergo hydrotreating and hydrocracking, the feedstock and hydrogen pass through trickle-bed catalytic reactors at high temperatures and pressures with large hydrogen flow. As such, the oil is partially vaporized and the hydrogen is partially dissolved in liquid to form a vapor-liquid equilibrium (VLE) system with both vapor and liquid phases containing oil and hydrogen. This may result in considerable changes in flow rates, physical properties and chemical compositions of both phases. Flow dynamics, mass transfer, heat transfer and reaction kinetics may also be modified. Experimental observations of VLE behaviours in distillates with different feedstocks under a range of operating conditions were presented. In addition, VLE was predicted along with its effects on distillates in pilot and commercial scale plants. tabs., figs.

Investigation of boiling water reactor stability and limit-cycle amplitude

International Nuclear Information System (INIS)

Damiano, B.; March-Leuba, J.A.; Euler, J.A.

1991-01-01

Galerkin's method has been applied to a boiling water reactor (BWR) dynamics model consisting of the point kinetics equations, which describe the neutronics, and a feedback transfer function, which describes the thermal hydraulics. The result is a low-order approximate solution describing BWR behavior during small-amplitude limit-cycle oscillations. The approximate solution has been used to obtain a stability condition, show that the average reactor power must increase during limit-cycle oscillations, and qualitatively determine how changes in transfer function values affect the limit-cycle amplitude. 6 refs., 2 figs., 2 tabs

Operation of CANDU power reactor in thorium self-sufficient fuel cycle

Indian Academy of Sciences (India)

This paper presents the results of calculations for CANDU reactor operation in thorium fuel cycle. Calculations are performed to estimate the feasibility of operation of heavy-water thermal neutron power reactor in self-sufficient thorium cycle. Parameters of active core and scheme of fuel reloading were considered to be the ...

Once-through thorium cycles in Candu reactors

International Nuclear Information System (INIS)

Milgram, M.S.

1982-01-01

In once-through thorium cycles pure thorium fuel bundles can be irradiated conjointly with uranium fuel bundles in a CANDU reactor with parameters judiciously chosen such that the overall fuel cycle cost is competitive with other possibilities - notably low-enriched uranium. Uranium 233 can be created and stockpiled for possible future use with no imperative that it be used unless future conditions warrant, and a stockpile can be begun independently of the state of reprocessing technology. The existence and general properties of these cycles are discussed

Future reactors and their fuel cycle

International Nuclear Information System (INIS)

Rastoin, J.

1990-01-01

Known world reserves of oil and natural gas may only last another 50 years and therefore nuclear energy will become more important in the future. Industrialised countries should also be encouraged to conserve their oil reserves to make better use of them and share them with less developed countries. France already produces 30% or more of its primary energy from uranium in the form of nuclear generated electricity. France has therefore accumulated considerable expertise in all aspects of the nuclear fuel cycle. Each stage of the fuel cycle, extraction, enrichment, fuel fabrication, fissile material utilisation, reprocessing and waste storage is discussed. The utilisation of fissile material is the most important stage and this is considered in more detail under headings: increase in burn-up, spectral shift, plutonium utilisation including recycling in pressurized water reactors and fast reactors and utilisation of reprocessed uranium. It is concluded that nuclear power for electricity production will be widely used throughout the world in the future. (UK)

Power conversion systems based on Brayton cycles for fusion reactors

International Nuclear Information System (INIS)

Linares, J.I.; Herranz, L.E.; Moratilla, B.Y.; Serrano, I.P.

2011-01-01

This paper investigates Brayton power cycles for fusion reactors. Two working fluids have been explored: helium in classical configurations and CO 2 in recompression layouts (Feher cycle). Typical recuperator arrangements in both cycles have been strongly constrained by low temperature of some of the energy thermal sources from the reactor. This limitation has been overcome in two ways: with a combined architecture and with dual cycles. Combined architecture couples the Brayton cycle with a Rankine one capable of taking advantage of the thermal energy content of the working fluid after exiting the turbine stage (iso-butane and steam fitted best the conditions of the He and CO 2 cycles, respectively). Dual cycles set a specific Rankine cycle to exploit the lowest quality thermal energy source, allowing usual recuperator arrangements in the Brayton cycle. The results of the analyses indicate that dual cycles could reach thermal efficiencies around 42.8% when using helium, whereas thermal performance might be even better (46.7%), if a combined CO 2 -H 2 O cycle was set.

Nuclear reactors and fuel cycle

International Nuclear Information System (INIS)

2014-01-01

to contribute in improving the quality of life of the Brazilian people. The nuclear fuel cycle is a series of steps involved in the production and use of fuel for nuclear reactors. The Laboratories of Chemistry and Environmental Diagnosis Center, CQMA, support the demand of Nuclear Fuel Cycle Program providing chemical characterization of uranium compounds and other related materials. In this period the Research Reactor Center (CRPq) concentrated efforts on improving equipment and systems to enable the IEA-R1 research reactor to operate at higher power, increasing the capacity of radioisotopes production, samples irradiation, tests and experiments. (author)

Nuclear reactors and fuel cycle

Energy Technology Data Exchange (ETDEWEB)

NONE

2014-07-01

fulfill its mission that is to contribute in improving the quality of life of the Brazilian people. The nuclear fuel cycle is a series of steps involved in the production and use of fuel for nuclear reactors. The Laboratories of Chemistry and Environmental Diagnosis Center, CQMA, support the demand of Nuclear Fuel Cycle Program providing chemical characterization of uranium compounds and other related materials. In this period the Research Reactor Center (CRPq) concentrated efforts on improving equipment and systems to enable the IEA-R1 research reactor to operate at higher power, increasing the capacity of radioisotopes production, samples irradiation, tests and experiments. (author)

Long-term scenarios of power reactors and fuel cycle development and the role of reduced moderation water reactors

International Nuclear Information System (INIS)

Sato, Osamu; Tatematsu, Kenji; Tanaka, Yoji

2000-01-01

Reduced moderation spectrum reactor is one of water cooled type reactors in future, which is based on the advanced technology of conventional nuclear power plants. The reduced moderation water reactor (RMWR) has various advantages, such as effective utilization of uranium resources, high conversion ratio, high burn-up, long-term cycle operation, and multiple recycle of plutonium. The RMWR is expected to be a substitute of fast breeder reactor (FBR) of which the development encounters with some technical and financial difficulties, and discontinues in many countries. The role of the RMWR on long-term scenarios of power reactor and fuel cycle development in Japan is investigated from the point of view of uranium resource needed. The consumption of natural uranium needed up to the year 2200 is calculated on various assumptions for the following three cases: (1) no breeder reactor; plutonium-thermal cycle in conventional light water reactor, (2) introduction of the FBR, and (3) introduction of the RMWR. The amounts of natural uranium consumption depends largely on the conversion ratio and plutonium quantity needed of a reactor type. The RMWR has a possibility as a substitute technology of the FBR with the improvement of conversion ratio and high burn-up. (Suetake, M.)

Long-term scenarios of power reactors and fuel cycle development and the role of reduced moderation water reactors

Energy Technology Data Exchange (ETDEWEB)

Sato, Osamu; Tatematsu, Kenji; Tanaka, Yoji [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2000-06-01

Reduced moderation spectrum reactor is one of water cooled type reactors in future, which is based on the advanced technology of conventional nuclear power plants. The reduced moderation water reactor (RMWR) has various advantages, such as effective utilization of uranium resources, high conversion ratio, high burn-up, long-term cycle operation, and multiple recycle of plutonium. The RMWR is expected to be a substitute of fast breeder reactor (FBR) of which the development encounters with some technical and financial difficulties, and discontinues in many countries. The role of the RMWR on long-term scenarios of power reactor and fuel cycle development in Japan is investigated from the point of view of uranium resource needed. The consumption of natural uranium needed up to the year 2200 is calculated on various assumptions for the following three cases: (1) no breeder reactor; plutonium-thermal cycle in conventional light water reactor, (2) introduction of the FBR, and (3) introduction of the RMWR. The amounts of natural uranium consumption depends largely on the conversion ratio and plutonium quantity needed of a reactor type. The RMWR has a possibility as a substitute technology of the FBR with the improvement of conversion ratio and high burn-up. (Suetake, M.)

Characteristics of fast reactor core designs and closed fuel cycle

International Nuclear Information System (INIS)

Poplavsky, V.M.; Eliseev, V.A.; Matveev, V.I.; Khomyakov, Y.S.; Tsyboulya, A.M.; Tsykunov, A.G.; Chebeskov, A.N.

2007-01-01

On the basis of the results of recent studies, preliminary basic requirements related to characteristics of fast reactor core and nuclear fuel cycle were elaborated. Decreasing reactivity margin due to approaching breeding ratio to 1, requirements to support non-proliferation of nuclear weapons, and requirements to decrease amount of radioactive waste are under consideration. Several designs of the BN-800 reactor core have been studied. In the case of MOX fuel it is possible to reach a breeding ratio about 1 due to the use of larger size of fuel elements with higher fuel density. Keeping low axial fertile blanket that would be reprocessed altogether with the core, it is possible to set up closed fuel cycle with the use of own produced plutonium only. Conceptual core designs of advanced commercial reactor BN-1800 with MOX and nitride fuel are also under consideration. It has been shown that it is expedient to use single enrichment fuel core design in this reactor in order to reach sufficient flattening and stability of power rating in the core. The main feature of fast reactor fuel cycle is a possibility to utilize plutonium and minor actinides which are the main contributors to the long-living radiotoxicity in irradiated nuclear fuel. The results of comparative analytical studies on the risk of plutonium proliferation in case of open and closed fuel cycle of nuclear power are also presented in the paper. (authors)

Conversion ratio in epithermal PWR, in thorium and uranium cycle

International Nuclear Information System (INIS)

Barroso, D.E.G.

1982-01-01

Results obtained for the conversion ratio in PWR reactors with close lattices, operating in thorium and uranium cycles, are presented. The study of those reactors is done in an unitary fuel cell of the lattices with several ratios V sub(M)/V sub(F), considering only the equilibrium cycles and adopting a non-spatial depletion calculation model, aiming to simulate mass flux of reactor heavy elements in the reactor. The neutronic analysis and the cross sections generation are done with Hammer computer code, with one critical apreciation about the application of this code in epithermal systems and with modifications introduced in the library of basic data. (E.G.) [pt

Current status and perspective of advanced loop type fast reactor in fast reactor cycle technology development project

International Nuclear Information System (INIS)

Niwa, Hajime; Aoto, Kazumi; Morishita, Masaki

2007-01-01

After selecting the combination of the sodium-cooled fast reactor (SFR) with oxide fuel, the advanced aqueous reprocessing and the simplified pelletizing fuel fabrication as the most promising concept of FR cycle system, 'Feasibility Study on Commercialized Fast Reactor Cycle Systems' was finalized in 2006. Instead, a new project, Fast Reactor Cycle Technology Development Project (FaCT Project) was launched in Japan focusing on development of the selected concepts. This paper describes the current status and perspective of the advanced loop type SFR system in the FaCT Project, especially on the design requirements, current design as well as the related innovative technologies together with the development road-map. Some considerations on advantages of the advanced loop type design are also described. (authors)

Advanced CANDU reactors fuel analysis through optimal fuel management at approach to refuelling equilibrium

International Nuclear Information System (INIS)

Tingle, C.P.; Bonin, H.W.

1999-01-01

The analysis of alternate CANDU fuels along with natural uranium-based fuel was carried out from the view point of optimal in-core fuel management at approach to refuelling equilibrium. The alternate fuels considered in the present work include thorium containing oxide mixtures (MOX), plutonium-based MOX, and Pressurised Water Reactor (PWR) spent fuel recycled in CANDU reactors (Direct Use of spent PWR fuel in CANDU (DUPIC)); these are compared with the usual natural UO 2 fuel. The focus of the study is on the 'Approach to Refuelling Equilibrium' period which immediately follows the initial commissioning of the reactor. The in-core fuel management problem for this period is treated as an optimization problem in which the objective function is the refuelling frequency to be minimized by adjusting the following decision variables: the channel to be refuelled next, the time of the refuelling and the number of fresh fuel bundles to be inserted in the channel. Several constraints are also included in the optimisation problem which is solved using Perturbation Theory. Both the present 37-rod CANDU fuel bundle and the proposed CANFLEX bundle designs are part of this study. The results include the time to reach refuelling equilibrium from initial start-up of the reactor, the average discharge burnup, the average refuelling frequency and the average channel and bundle powers relative to natural UO 2 . The model was initially tested and the average discharge burnup for natural UO 2 came within 2% of the industry accepted 199 MWh/kgHE. For this type of fuel, the optimization exercise predicted the savings of 43 bundles per full power year. In addition to producing average discharge burnups and other parameters for the advanced fuels investigated, the optimisation model also evidenced some problem areas like high power densities for fuels such as the DUPIC. Perturbation Theory has proven itself to be an accurate and valuable optimization tool in predicting the time between

Improving fuel cycle design and safety characteristics of a gas cooled fast reactor

NARCIS (Netherlands)

van Rooijen, W.F.G.

2006-01-01

This research concerns the fuel cycle and safety aspects of a Gas Cooled Fast Reactor, one of the so-called "Generation IV" nuclear reactor designs. The Generation IV Gas Cooled Fast Reactor uses helium as coolant at high temperature. The goal of the GCFR is to obtain a "closed nuclear fuel cycle",

Simulated first operating campaign for the Integral Fast Reactor fuel cycle demonstration

International Nuclear Information System (INIS)

Goff, K.M.; Mariani, R.D.; Benedict, R.W.; Park, K.H.; Ackerman, J.P.

1993-01-01

This report discusses the Integral Fast Reactor (IFR) which is an innovative liquid-metal-cooled reactor concept that is being developed by Argonne National Laboratory. It takes advantage of the properties of metallic fuel and liquid-metal cooling to offer significant improvements in reactor safety, operation, fuel cycle-economics, environmental protection, and safeguards. Over the next few years, the IFR fuel cycle will be demonstrated at Argonne-West in Idaho. Spent fuel from the Experimental Breeder Reactor II (EBR-II) win be processed in its associated Fuel Cycle Facility (FCF) using a pyrochemical method that employs molten salts and liquid metals in an electrorefining operation. As part of the preparation for the fuel cycle demonstration, a computer code, PYRO, was developed at Argonne to model the electrorefining operation using thermodynamic and empirical data. This code has been used extensively to evaluate various operating strategies for the fuel cycle demonstration. The modeled results from the first operating campaign are presented. This campaign is capable of processing more than enough material to refuel completely the EBR-II core

An enhanced search algorithm for the charged fuel enrichment in equilibrium cycle analysis of REBUS-3

International Nuclear Information System (INIS)

Park, Tongkyu; Yang, Won Sik; Kim, Sang-Ji

2017-01-01

Highlights: • An enhanced search algorithm for charged fuel enrichment was developed for equilibrium cycle analysis with REBUS-3. • The new search algorithm is not sensitive to the user-specified initial guesses. • The new algorithm reduces the computational time by a factor of 2–3. - Abstract: This paper presents an enhanced search algorithm for the charged fuel enrichment in equilibrium cycle analysis of REBUS-3. The current enrichment search algorithm of REBUS-3 takes a large number of iterations to yield a converged solution or even terminates without a converged solution when the user-specified initial guesses are far from the solution. To resolve the convergence problem and to reduce the computational time, an enhanced search algorithm was developed. The enhanced algorithm is based on the idea of minimizing the number of enrichment estimates by allowing drastic enrichment changes and by optimizing the current search algorithm of REBUS-3. Three equilibrium cycle problems with recycling, without recycling and of high discharge burnup were defined and a series of sensitivity analyses were performed with a wide range of user-specified initial guesses. Test results showed that the enhanced search algorithm is able to produce a converged solution regardless of the initial guesses. In addition, it was able to reduce the number of flux calculations by a factor of 2.9, 1.8, and 1.7 for equilibrium cycle problems with recycling, without recycling, and of high discharge burnup, respectively, compared to the current search algorithm.

A fuel management study and cycle nuclear design for PW reactors

International Nuclear Information System (INIS)

Minguez, E.; Ahnert, C.; Aragones, J. M.; Corella, M. R.

1975-01-01

A reference reactor was chosen to do a general study involving Fuel Management Evaluations of several cycles, and Design Calculations of cycles already performed, according to a calculation scheme set up in the Reactor Technology Division of the J.E.N., using some computer codes acquired to foreign sources and other ones developed in the J.E.N. (Author) 5 refs

A fuel management study and cycle nuclear design for PW reactors

Energy Technology Data Exchange (ETDEWEB)

Minguez, E; Ahnert, C; Aragones, J M; Corella, M R

1975-07-01

A reference reactor was chosen to do a general study involving Fuel Management Evaluations of several cycles, and Design Calculations of cycles already performed, according to a calculation scheme set up in the Reactor Technology Division of the J.E.N., using some computer codes acquired to foreign sources and other ones developed in the J.E.N. (Author) 5 refs.

Fuel cycle options for light water reactors and heavy water reactors. Proceedings of a technical committee meeting

International Nuclear Information System (INIS)

1999-11-01

In the second half of the 20th century nuclear power has evolved from the research and development environment to an industry that supplies 16% of the world's electricity. By the end of 1997, over 8500 reactor-years of operating experience had been accumulated. Global environmental change, and the continuing increase in global energy supply required to provide increasing populations with an improving standard of living, make the contribution from nuclear energy even more important for the next century. For nuclear power to achieve its full potential and make its needed contribution, it must be safe, economical, reliable and sustainable. All of these factors can be enhanced by judicious choice and development of advanced fuel cycle options. The Technical Committee Meeting (TCM) on Fuel Cycle Options for Light Water Reactors and Heavy Water Reactors was hosted by Atomic Energy of Canada Limited (AECL) on behalf of the Canadian Government and was jointly conducted within the frame of activities of the IAEA International Working Group on Advanced Technologies for Light Water Reactors (IWG-LWR) and the IAEA International Working Group on Advanced Technologies for Heavy Water Reactors (IWG-HWR). The TCM provided the opportunity to have in-depth discussions on important technical topics which were highlighted in the International Symposium on Nuclear Fuel Cycle and Reactor Strategies: Adjusting to New Realities, held in Vienna, 3-6 June 1997. The main results and conclusions of the TCM were presented as input for discussion at the first meeting of the IAEA newly formed International Working Group on Fuel Cycle Options

Trends and Developments for Fast Neutron Reactors and Related Fuel Cycles

International Nuclear Information System (INIS)

Carré, Frank

2013-01-01

• FR13 – A unique and dedicated framework to share updates on national programs of Fast Reactor developments, projects of new builds and plans for the future: - Near term projects of sodium and lead-alloy Fast Reactors; - Gen-IV visions of sodium-cooled and alternative types of Fast Neutron Reactors (GFR, LFR…). • FR13 – A special emphasis put on Fast Reactor Safety, Sustainability of nuclear fuel cycle and Young Generation perspective. • FR13 – A catalyst for further collaborations and alliances: - To share visions of goals and advisable options for future Fast Reactors and Nuclear Fuel Cycle; - To share cost of R&D and large demonstrations (safety, security, recycling); - To progress towards harmonized international standards; - To integrate national projects into a consistent international roadmap

A preliminary study of a D-T tokamak fusion reactor with advanced blanket using the compact fusion advanced Brayton (CFAB) cycle

International Nuclear Information System (INIS)

Yoshikawa, K.; Ishikawa, M.; Umoto, J.; Fukuyama, A.; Mitarai, O.; Okamoto, M.; Sekimoto, H.; Nagatsu, M.

1995-01-01

Preliminary key issues for a synchrotron radiation-enhanced compact fusion advanced Brayton (CFAB) cycle fusion reactor similar to the CFAR (compact fusion advanced Rankine) cycle reactor are presented. These include plasma operation windows as a function of the first wall reflectivity and related issues, to estimate an allowance for deterioration of the first wall reflectivity due to dpa effects. It was found theoretically that first wall reflectivities down to 0.8 are still adequate for operation at an energy confinement scaling of 3 times Kaye-Goldston. Measurements of the graphite first wall reflectivities at Nagoya University indicate excellent reflectivities in excess of 90% for CC-312, PCC-2S, and PD-330S in the submillimeter regime, even at high temperatures in excess of 1000K. Some engineering issues inherent to the CFAB cycle are also discussed briefly in comparison with the CFAR cycle which uses hazardous limited-resource materials but is capable of using mercury as coolant for high heat removal. The CFAB cycle using helium coolant is found to achieve higher net plant conversion efficiencies in excess 60% using a non-equilibrium magnetohydrodynamic disk generator in the moderate pressure range, even at the cost of a relatively large pumping power, and at the penalty of high temperature materials, although excellent heat removal characteristics in the moderate pressure range need to be guaranteed in the future. (orig.)

Improvement of system code importing evaluation of Life Cycle Analysis of tokamak fusion power reactors

International Nuclear Information System (INIS)

Kobori, Hikaru; Kasada, Ryuta; Hiwatari, Ryoji; Konishi, Satoshi

2016-01-01

Highlights: • We incorporated the Life Cycle Analysis (LCA) of tokamak type DEMO reactor and following commercial reactors as an extension of a system code. • We calculated CO_2 emissions from reactor construction, operation and decommissioning that is considered as a major environmental cost. • We found that the objective of conceptual design of the tokamak fusion power reactor is moved by changing evaluation index. • The tokamak fusion reactor can reduce CO_2 emissions in the life cycle effectively by reduction of the amount involved in the replacement of internal components. • The tokamak fusion reactor achieves under 0.174$/kWh electricity cost, the tokamak fusion reactor is contestable with 1500 degrees-class LNG-fired combined cycle power plant. - Abstract: This study incorporate the Life Cycle Analysis (LCA) of tokamak type DEMO reactor and following commercial reactors as an extension of a system code to calculate CO_2 emissions from reactor construction, operation and decommissioning that is considered as a major environmental cost. Competitiveness of tokamak fusion power reactors is expected to be evaluated by the cost and environmental impact represented by the CO_2 emissions, compared with present and future power generating systems such as fossil, nuclear and renewables. Result indicated that (1) The objective of conceptual design of the tokamak fusion power reactor is moved by changing evaluation index. (2) The tokamak fusion reactor can reduce CO_2 emissions in the life cycle effectively by reduction of the amount involved in the replacement of internal components. (3) The tokamak fusion reactor achieves under 0.174$/kWh electricity cost, the tokamak fusion reactor is contestable with 1500 degrees-class LNG-fired combined cycle power plant.

Improvement of system code importing evaluation of Life Cycle Analysis of tokamak fusion power reactors

Energy Technology Data Exchange (ETDEWEB)

Kobori, Hikaru [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Kasada, Ryuta, E-mail: r-kasada@iae.kyoto-u.ac.jp [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Hiwatari, Ryoji [Central Research Institute of Electric Power Industry, Tokyo (Japan); Konishi, Satoshi [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan)

2016-11-01

Highlights: • We incorporated the Life Cycle Analysis (LCA) of tokamak type DEMO reactor and following commercial reactors as an extension of a system code. • We calculated CO{sub 2} emissions from reactor construction, operation and decommissioning that is considered as a major environmental cost. • We found that the objective of conceptual design of the tokamak fusion power reactor is moved by changing evaluation index. • The tokamak fusion reactor can reduce CO{sub 2} emissions in the life cycle effectively by reduction of the amount involved in the replacement of internal components. • The tokamak fusion reactor achieves under 0.174$/kWh electricity cost, the tokamak fusion reactor is contestable with 1500 degrees-class LNG-fired combined cycle power plant. - Abstract: This study incorporate the Life Cycle Analysis (LCA) of tokamak type DEMO reactor and following commercial reactors as an extension of a system code to calculate CO{sub 2} emissions from reactor construction, operation and decommissioning that is considered as a major environmental cost. Competitiveness of tokamak fusion power reactors is expected to be evaluated by the cost and environmental impact represented by the CO{sub 2} emissions, compared with present and future power generating systems such as fossil, nuclear and renewables. Result indicated that (1) The objective of conceptual design of the tokamak fusion power reactor is moved by changing evaluation index. (2) The tokamak fusion reactor can reduce CO{sub 2} emissions in the life cycle effectively by reduction of the amount involved in the replacement of internal components. (3) The tokamak fusion reactor achieves under 0.174$/kWh electricity cost, the tokamak fusion reactor is contestable with 1500 degrees-class LNG-fired combined cycle power plant.

Plutonium fuel cycles in the spectral shift controlled reactor

International Nuclear Information System (INIS)

Sider, F.M.; Matzie, R.A.

1980-01-01

The spectral shift controlled reactor (SSCR) controls excess core reactivity during an operating cycle through the use of variable heavy water concentrations in the moderator. With heavy water in the coolant, the neutron spectrum is shifted to higher energy levels, thus increasing fertile conversion. In addition, since heavy water obviates the need for soluble boron, neutron losses to control poison are eliminated. As a result, better resource utilization is obtained in the SSCR employing plutonium fuel cycles compared to similarly fueled pressurized water reactors (PWRs). The SSCR, however, is not competitive with the PWR due to higher capital costs, operation and maintenance costs, and the heavy water costs, which outweigh the fuel cycle cost savings. The SSCR may become an attractive alternative to the PWR if uranium prices increase substantially

Evaluation of plutonium, uranium, and thorium use in power reactor fuel cycles

International Nuclear Information System (INIS)

Kasten, P.R.; Homan, F.J.

1977-01-01

The increased cost of uranium and separative work has increased the attractiveness of plutonium use in both uranium and thorium fuel cycles in thermal reactors. A technology, fuel utilization, and economic evaluation is given for uranium and thorium fuel cycles in various reactor types, along with the use of plutonium and 238 U. Reactors considered are LWRs, HWRs, LWBRs, HTGRs, and FBRs. Key technology factors are fuel irradiation performance and associated physical property values. Key economic factors are unit costs for fuel fabrication and reprocessing, and for refabrication of recycle fuels; consistent cost estimates are utilized. In thermal reactors, the irradiation performance of ceramic fuels appears to be satisfactory. At present costs for uranium ore and separative work, recycle of plutonium with thorium rather than uranium is preferable from fuel utilization and economic viewpoints. Further, the unit recovery cost of plutonium is lower from LWR fuels than from natural-uranium HWR fuels; use of LWR product permits plutonium/thorium fueling to compete with uranium cycles. Converting uranium cycles to thorium cycles increases the energy which can be extracted from a given uranium resource. Thus, additional fuel utilization improvement can be obtained by fueling all thermal reactors with thorium, but this requires use of highly enriched uranium; use of 235 U with thorium is most economic in HTGRs followed by HWRs and then LWRs. Marked improvement in long-term fuel utilization can be obtained through high thorium loadings and short fuel cycle irradiations as in the LWBR, but this imposes significant economic penalties. Similar operating modes are possible in HWRs and HTGRs. In fast reactors, use of the plutonium-uranium cycle gives advantageous fuel resource utilization in both LMFBRs and GCFRs; use of the thorium cycle provides more negative core reactivity coefficients and more flexibility relative to use of recycle fuels containing uranium of less than 20

Power cycling experiments in INR-TRIGA-SSR Reactor

International Nuclear Information System (INIS)

Dumitru, M.

2008-01-01

The in-reactor experimental program started this summer with some power cycling experiments to provide date on fuel behaviour under abnormal reactor operating conditions. The paper describes the irradiation device, its operational features and an original 'under-flux' movement system. Also, there are presented main data of irradiation device (pressure, flow, temperature, construction), in-pile section, location, sample, instrumentation, experimental sequences and operating data of Interest for the experimenters. (author)

Transitioning nuclear fuel cycles with uncertain fast reactor costs

Energy Technology Data Exchange (ETDEWEB)

Phathanapirom, U.B., E-mail: bphathanapirom@utexas.edu; Schneider, E.A.

2016-06-15

This paper applies a novel decision making methodology to a case study involving choices leading to the transition from the current once-through light water reactor fuel cycle to one relying on continuous recycle of plutonium and minor actinides in fast reactors in the face of uncertain fast reactor capital costs. Unique to this work is a multi-stage treatment of a range of plausible trajectories for the evolution of fast reactor capital costs over time, characterized by first-of-a-kind penalties as well as time- and unit-based learning. The methodology explicitly incorporates uncertainties in key parameters into the decision-making process by constructing a stochastic model and embedding uncertainties as bifurcations in the decision tree. “Hedging” strategies are found by applying a choice criterion to select courses of action which mitigate “regrets”. These regrets are calculated by evaluating the performance of all possible transition strategies for every feasible outcome of the uncertain parameter. The hedging strategies are those that preserve the most flexibility for adjusting the fuel cycle strategy in response to new information as uncertainties are resolved.

Transitioning nuclear fuel cycles with uncertain fast reactor costs

International Nuclear Information System (INIS)

Phathanapirom, U.B.; Schneider, E.A.

2016-01-01

This paper applies a novel decision making methodology to a case study involving choices leading to the transition from the current once-through light water reactor fuel cycle to one relying on continuous recycle of plutonium and minor actinides in fast reactors in the face of uncertain fast reactor capital costs. Unique to this work is a multi-stage treatment of a range of plausible trajectories for the evolution of fast reactor capital costs over time, characterized by first-of-a-kind penalties as well as time- and unit-based learning. The methodology explicitly incorporates uncertainties in key parameters into the decision-making process by constructing a stochastic model and embedding uncertainties as bifurcations in the decision tree. “Hedging” strategies are found by applying a choice criterion to select courses of action which mitigate “regrets”. These regrets are calculated by evaluating the performance of all possible transition strategies for every feasible outcome of the uncertain parameter. The hedging strategies are those that preserve the most flexibility for adjusting the fuel cycle strategy in response to new information as uncertainties are resolved.

Impact of partitioning and transmutation on high-level waste disposal for the fast breeder reactor fuel cycle

International Nuclear Information System (INIS)

Nishihara, Kenji; Oigawa, Hiroyuki; Nakayama, Shinichi; Ono, Kiyoshi; Shiotani, Hiroki

2010-01-01

The impact of partitioning and/or transmutation (PT) technology on high-level waste management was investigated for the equilibrium state of several potential fast breeder reactor (FBR) fuel cycles. Three different fuel cycle scenarios involving PT technology were analyzed: 1) partitioning process only (separation of some fission products), 2) transmutation process only (separation and transmutation of minor actinides), and 3) both partitioning and transmutation processes. The conventional light water reactor (LWR) fuel cycle without PT technology, on which the current repository design is based, was also included for comparison. We focused on the thermal constraints in a geological repository and determined the necessary predisposal storage quantities and time periods (by defining a storage capacity index) for several predefined emplacement configurations through transient thermal analysis. The relation between this storage capacity index and the required repository emplacement area was obtained. We found that the introduction of the FBR fuel cycle without PT can yield a 35% smaller repository per unit electricity generation than the LWR fuel cycle, although the predisposal storage period is prolonged from 50 years for the LWR fuel cycle to 65 years for the FBR fuel cycle without PT. The introduction of the partitioning-only process does not result in a significant reduction of the repository emplacement area from that for the FBR fuel cycle without PT, but the introduction of the transmutation-only process can reduce the emplacement area by a factor of 5 when the storage period is extended from 65 to 95 years. When a coupled partitioning and transmutation system is introduced, the repository emplacement area can be reduced by up to two orders of magnitude by assuming a predisposal storage of 60 years for glass waste and 295 years for calcined waste containing the Sr and Cs fraction. The storage period of 295 years for the calcined waste does not require a large

Parametric analyses of single-zone thorium-fueled molten salt reactor fuel cycle options

International Nuclear Information System (INIS)

Powers, J.J.; Worrall, A.; Gehin, J.C.; Harrison, T.J.; Sunny, E.E.

2013-01-01

Analyses of fuel cycle options based on thorium-fueled Molten Salt Reactors (MSRs) have been performed in support of fuel cycle screening and evaluation activities for the United States Department of Energy. The MSR options considered are based on thermal spectrum MSRs with 3 different separations levels: full recycling, limited recycling, and 'once-through' operation without active separations. A single-fluid, single-zone 2250 MWth (1000 MWe) MSR concept consisting of a fuel-bearing molten salt with graphite moderator and reflectors was used as the basis for this study. Radiation transport and isotopic depletion calculations were performed using SCALE 6.1 with ENDF/B-VII nuclear data. New methodology developed at Oak Ridge National Laboratory (ORNL) enables MSR analysis using SCALE, modeling material feed and removal by taking user-specified parameters and performing multiple SCALE/TRITON simulations to determine the resulting equilibrium operating conditions. Parametric analyses examined the sensitivity of the performance of a thorium MSR to variations in the separations efficiency for protactinium and fission products. Results indicate that self-sustained operation is possible with full or limited recycling but once-through operation would require an external neutron source. (authors)

Analysis of an accelerator-driven subcritical light water reactor

International Nuclear Information System (INIS)

Kruijf, W.J.M. de; Wakker, P.H.; Wetering, T.F.H. van de; Verkooijen, A.H.M.

1997-01-01

An analysis of the basic characteristics of an accelerator-driven light water reactor has been made. The waste in the nuclear fuel cycle is considerably less than in the light water reactor open fuel cycle. This is mainly caused by the use of equilibrium nuclear fuel in the reactor. The accelerator enables the use of a fuel composition with infinite multiplication factor k ∞ < 1. The main problem of the use of this type of fuel is the strongly peaked flux distribution in the reactor core. A simple analytical model shows that a large core is needed with a high peak power factor in order to generate net electric energy. The fuel in the outer regions of the reactor core is used very poorly. 7 refs., 4 figs., 1 tab

Decay profiles of {beta} and {gamma} for a radionuclide inventory in equilibrium cycle of a BWR type reactor; Perfiles de decaimiento de radiacion {beta} y {gamma} para un inventario de radionuclidos en ciclo de equilibrio de un reactor tipo BWR

Energy Technology Data Exchange (ETDEWEB)

Salaices, M.; Sandoval, S.; Ovando, R. [Instituto de Investigaciones Electricas. Gerencia de Energia Nuclear, Av. Reforma 113 Col. Palmira. 62490 Cuernavaca, Morelos (Mexico)]. e-mail: sal@iie.org.mx

2007-07-01

Presently work the {beta} and {gamma} radiation decay profiles for a radionuclides inventory in equilibrium cycle of a BWR type reactor is presented. The profiles are presented in terms of decay in the activity of the total inventory as well as of the chemical groups that conform the inventory. In the obtaining of the radionuclides inventory in equilibrium cycle the ORIGEN2 code, version 1 was used, which simulates fuel burnup cycles and it calculates the evolution of the isotopic composition as a result of the burnt one, irradiation and decay of the nuclear fuel. It can be observed starting from the results that the decrease in the activity for the initial inventory and the different chemical groups that conform it is approximately proportional to the base 10 logarithm of the time for the first 24 hours of having concluded the burnt one. It can also be observed that the chemical groups that contribute in more proportion to the total activity of the inventory are the lanthanides-actinides and the transition metals, with 39% and 28%, respectively. The groups of alkaline earth metals, halogens, metalloids, noble gases and alkaline metals, contribute with percentages that go from the 8 to 5%. The groups that less they contribute to the total activity of the inventory they are the non metals and semi-metals with smaller proportions that 1%. The chemical groups that more contribute to the energy of {beta} and {gamma} radiation its are the transition metals and the lanthanides-actinides with a change in the order of importance at the end of the 24 hours period. The case of the halogens is of relevance for the case of the {gamma} radiation energy due that occupying the very near third site to the dimensions of the two previous groups. Additionally, the decay in the activity for the total inventory and the groups that conform it can be simulated by means of order 6 polynomials or smaller than describe its behavior appropriately. The results presented in this work, coupled

The feasibility study on commercialized fast reactor cycle system

International Nuclear Information System (INIS)

Noda, Hiroshi

2002-01-01

The feasibility study on commercialized Fast Reactor cycle system (FS) has been carried out by a joint team with the participation of all parties concerned in Japan since July, 1999. It aims to clarify various perspectives for commercialized fast reactor cycle system and also suggest development strategies to diverse social needs in the 21 st century. The FS consists of several phases. The phase 1 has progressed as planned and the highly feasible candidate concepts with innovative technologies have been screened out among a wide variety of concepts. During the phase 2, approximately five years after the phase 1, the in-depth design studies and engineering scale tests of key technologies are being conducted to verify and validate the feasibility of screened candidate concepts. At the end of the phase 2, a few promising concepts will be selected with their R and D tasks. The paper describes the results of the phase 1, the activities of the phase 2 and the new concept related to the fast reactor fuel cycle focusing on the reduction in environmental burden, which is one of key factors to sustain the nuclear power generation in the 21 st century

Concept of innovative water reactor for flexible fuel cycle (FLWR)

International Nuclear Information System (INIS)

Iwamura, T.; Uchikawa, S.; Okubo, T.; Kugo, T.; Akie, H.; Nakatsuka, T.

2005-01-01

In order to ensure sustainable energy supply in the future based on the matured Light Water Reactor (LWR) and coming LWR-Mixed Oxide (MOX) technologies, a concept of Innovative Water Reactor for Flexible Fuel Cycle (FLWR) has been investigated in Japan Atomic Energy Research Institute (JAERI). The concept consists of two parts in the chronological sequence. The first part realizes a high conversion type core concept, which is basically intended to keep the smooth technical continuity from current LWR and coming LWR-MOX technologies without significant gaps in technical point of view. The second part represents the Reduced-Moderation Water Reactor (RMWR) core concept, which realizes a high conversion ratio over 1.0 being useful for the long-term sustainable energy supply through plutonium multiple recycling based on the well-experienced LWR technologies. The key point is that the two core concepts utilize the compatible and the same size fuel assemblies, and hence, the former concept can proceed to the latter in the same reactor system, based flexibly on the fuel cycle circumstances during the reactor operation period around 60 years. At present, since the fuel cycle for the plutonium multiple recycling with MOX fuel reprocessing has not been realized yet, reprocessed plutonium from the LWR spent fuel is to be utilized in LWR-MOX. After this stage, the first part of FLWR, i.e. the high conversion type, can be introduced as a replacement of LWR or LWR-MOX. Since the plutonium inventory of FLWR is much larger, the number of the reactor with MOX fuel will be significantly reduced compared to the LWR-MOX utilization. The size of the fuel assembly for the first part is the same as in the RMWR concept, i.e. the hexagonal fuel assembly with the inner face-to-face distance of about 200 mm. Fuel rods are arranged in the triangular lattice with a relatively wide gap size around 3 mm between rods, and the effective MOX length is less than 1.5 m without using the blanket. When

Physics calculations for the Clinch River Breeder Reactor

International Nuclear Information System (INIS)

Kalimullah; Kier, P.H.; Hummel, H.H.

1977-06-01

Calculations of distributions of power and sodium void reactivity, unvoided and voided Doppler coefficients and steel and fuel worths have been performed using diffusion theory and first-order perturbation theory for the LWR discharge Pu-fueled CRBR at BOL, the FFTF-grade Pu-fueled CRBR at BOL and for the beginning and end of equilibrium cycle of the LWR-Pu-fueled CRBR. The results of the burnup and breeding ratio calculations performed for obtaining the reactor compositions during the equilibrium cycle are also reported. Effects of sodium and steel contents on the distributions of sodium void reactivity and steel worth have also been studied. Errors and uncertainties in the reactivity coefficients due to cross-sections and the two-dimensional geometric representations of the reactor used in the calculations have also been estimated. Comparisons of the results with those in the CRBR PSAR are also discussed

Heavy water reactors on the once-through uranium cycle

International Nuclear Information System (INIS)

1978-05-01

This paper presents preliminary technical and economic data to INFCE on the once-through uranium fuel cycle for use in early comparisons of alternate nuclear systems. The denatured thorium fuel cycle is discussed in a companion paper. Information for this paper was developed under an ongoing program, and more complete reporting of the evaluation of the heavy water reactor and its fuel cycles is planned toward the end of the year

Neutronic study of heavy nucleus produced in nuclear reactor fuel cycle

International Nuclear Information System (INIS)

Giacometti, A.

1978-01-01

Importance of minor actinides (U, Np, Pu, Am and Cm isotopes) PWR and fast neutron reactors and their associated fuel cycle is examined in this thesis. The amount of actinides formed in the various types of fuels or reactors are given. The different ways of formation and their importance are described. Modifications of the core reactivity due to actinides are shown. After a review of the fuel cycle (enrichment, fabrication, reprocessing, transport) actinide evolution outside the core is described and main problems concerning radioactivity in the different steps of the cycle or long term storage are underlined [fr

Technical Meeting on Fast Reactors and Related Fuel Cycle Facilities with Improved Economic Characteristics. Presentations

International Nuclear Information System (INIS)

2013-01-01

The objectives of the meeting were: • To identify the main issues and technical features that affect capital and energy production costs of fast reactors and related fuel cycle facilities; • To present fast reactor concepts and designs with enhanced economic characteristics, as well as innovative technical solutions (components, subsystems, etc.) that have the potential to reduce the capital costs of fast reactors and related fuel cycle facilities; • To present energy models and advanced tools for the cost assessment of innovative fast reactors and associated nuclear fuel cycles; • To discuss the results of studies and ongoing R&D activities that address cost reduction and the future economic competitiveness of fast reactors; • To identify research and technology development needs in the field, also in view of new IAEA initiatives to help and support Member States in improving the economic competitiveness of fast reactors and associated nuclear fuel cycles

Thorium--uranium cycle ICF hybrid concept

International Nuclear Information System (INIS)

Frank, T.G.

1978-01-01

The results of preliminary studies of a laser-driven fusion-fission hybrid concept utilizing the 232 Th- 233 U breeding cycle are reported. Neutron multiplication in the breeding blanket is provided by a region containing 238 UO 2 and the equilibrium concentration of 239 PuO 2 . Established fission reactor technology is utilized to determine limits on operating conditions for high-temperature fuels and structures. The implications of nonproliferation policies for the operation of fusion-fission hybrid reactors are discussed

Potential application of Rankine and He-Brayton cycles to sodium fast reactors

International Nuclear Information System (INIS)

Perez-Pichel, G.D.; Linares, J.I.; Herranz, L.E.; Moratilla, B.Y.

2011-01-01

Highlights: → This paper has been focused on thermal efficiency of several Rankine and Brayton cycles for SFR. → A sub-critical Rankine configuration could reach a thermal efficiency higher than 43%. → It could be increased to almost 45% using super-critical configurations. → Brayton cycles thermal performance can be enhanced by adding a super-critical organic fluid Rankine cycle. → The moderate coolant temperature at the reactor makes Brayton configurations have poorer. - Abstract: Traditionally all the demos and/or prototypes of the sodium fast reactor (SFR) technology with power output, have used a steam sub-critical Rankine cycle. Sustainability requirement of Gen. IV reactors recommends exploring alternate power cycle configurations capable of reaching high thermal efficiency. By adopting the anticipated working parameters of next SFRs, this paper investigates the potential of some Rankine and He-Brayton layouts to reach thermal efficiencies as high as feasible, so that they could become alternates for SFR reactor balance of plant. The assessment has encompassed from sub-critical to super-critical Rankine cycles and combined cycles based on He-Brayton gas cycles of different complexity coupled to Organic Rankine Cycles. The sub-critical Rankine configuration reached at thermal efficiency higher than 43%, which has been shown to be a superior performance than any of the He-Brayton configurations analyzed. By adopting a super-critical Rankine arrangement, thermal efficiency would increase less than 1.5%. In short, according to the present study a sub-critical layout seems to be the most promising configuration for all those upcoming prototypes to be operated in the short term (10-15 years). The potential of super-critical CO 2 -Brayton cycles should be explored for future SFRs to be deployed in a longer run.

Thermal analysis of heat and power plant with high temperature reactor and intermediate steam cycle

Directory of Open Access Journals (Sweden)

Fic Adam

2015-03-01

Full Text Available Thermal analysis of a heat and power plant with a high temperature gas cooled nuclear reactor is presented. The main aim of the considered system is to supply a technological process with the heat at suitably high temperature level. The considered unit is also used to produce electricity. The high temperature helium cooled nuclear reactor is the primary heat source in the system, which consists of: the reactor cooling cycle, the steam cycle and the gas heat pump cycle. Helium used as a carrier in the first cycle (classic Brayton cycle, which includes the reactor, delivers heat in a steam generator to produce superheated steam with required parameters of the intermediate cycle. The intermediate cycle is provided to transport energy from the reactor installation to the process installation requiring a high temperature heat. The distance between reactor and the process installation is assumed short and negligable, or alternatively equal to 1 km in the analysis. The system is also equipped with a high temperature argon heat pump to obtain the temperature level of a heat carrier required by a high temperature process. Thus, the steam of the intermediate cycle supplies a lower heat exchanger of the heat pump, a process heat exchanger at the medium temperature level and a classical steam turbine system (Rankine cycle. The main purpose of the research was to evaluate the effectiveness of the system considered and to assess whether such a three cycle cogeneration system is reasonable. Multivariant calculations have been carried out employing the developed mathematical model. The results have been presented in a form of the energy efficiency and exergy efficiency of the system as a function of the temperature drop in the high temperature process heat exchanger and the reactor pressure.

Wastes from selected activities in two light-water reactor fuel cycles

International Nuclear Information System (INIS)

Palmer, C.R.; Hill, O.F.

1980-07-01

This report presents projected volumes and radioactivities of wastes from the production of electrical energy using light-water reactors (LWR). The projections are based upon data developed for a recent environmental impact statement in which the transuranic wastes (i.e., those wastes containing certain long-lived alpha emitters at concentrations of at least 370 becquerels, or 10 nCi, per gram of waste) from fuel cycle activities were characterized. In addition, since the WG.7 assumed that all fuel cycle wastes except mill tailings are placed in a mined geologic repository, the nontransuranic wastes from several activities are included in the projections reported. The LWR fuel cycles considered are the LWR, once-through fuel cycle (Strategy 1), in which spent fuel is packaged in metal canisters and then isolated in geologic formations; and the LWR U/Pu recycle fuel cycle (Strategy 2), wherein spent fuel is reprocessed for recovery and recycle of uranium and plutonium in LWRs. The wastes projected for the two LWR fuel cycles are summarized. The reactor operations and decommissioning were found to dominate the rate of waste generation in each cycle. These activities account for at least 85% of the fuel cycle waste volume (not including head-end wastes) when normalized to per unit electrical energy generated. At 10 years out of reactor, however, spent fuel elements in Strategy 1 represent 98% of the fuel cycle activity but only 4% of the volume. Similarly, the packaged high-level waste, fuel hulls and hardware in Strategy 2 concentrate greater than 95% of the activity in 2% of the waste volume

Reactor physics and reactor strategy investigations into the fissionable material economy of the thorium and uranium cycle in fast breeder reactors and high temperature reactors

International Nuclear Information System (INIS)

Schikorr, W.M.

In this work the properties governing the fissionable material economy of the uranium and thorium cycles are investigated for the advanced reactor types currently under development - the fast breeder reactor (FBR) and the high temperature reactor (HTR) - from the point of view of the optimum utilization of the available nuclear fuel reserves and the continuance of supply of these reserves. For this purpose, the two reactor types are first of all considered individually and are subsequently discussed as a complementary overall system

Steam water cycle chemistry of liquid metal cooled innovative nuclear power reactors

International Nuclear Information System (INIS)

Yurmanov, Victor; Lemekhov, Vadim; Smykov, Vladimir

2012-09-01

The Federal Target Program (FTP) of Russian Federation 'Nuclear Energy Technologies of the New Generation for 2010-2015 and for Perspective up to 2020' is aimed at development of advanced nuclear energy technologies on the basis of closed fuel cycle with fast reactors. There are advanced fast reactor technologies of the 4. generation with liquid metal cooled reactors. Development stages of maturity of fast sodium cooled reactor technology in Russia includes experimental reactors BR-5/10 (1958-2002) and BOR-60 (since 1969), nuclear power plants (NPPs) with BN-350 (1972-1999), BN-600 (since 1980), BN-800 (under construction), BN-1200 (under development). Further stage of development of fast sodium cooled reactor technology in Russia is commercialization. Lead-bismuth eutectic fast reactor technology has been proven at industrial scale for nuclear submarines in former Soviet Union. Lead based technology is currently under development and need for experimental justification. Current status and prospects of State Corporation 'Rosatom' participation in GIF activities was clarified at the 31. Meeting of Policy Group of the International Forum 'Generation-IV', Moscow, May 12-13, 2011. In June, 2010, 'Rosatom' joined the Sodium Fast Reactor Arrangement as an authorized representative of the Russian Government. It was also announced the intention of 'Rosatom' to sign the Memorandum on Lead Fast Reactor based on Russia's experience with lead-bismuth and lead cooled fast reactors. In accordance with the above FTP some innovative liquid metal cooled reactors of different design are under development in Russia. Gidropress, well known as WER designer, develops innovative lead-bismuth eutectic cooled reactor SVBR-100. NIKIET develops innovative lead cooled reactor BRESTOD-300. Some other nuclear scientific centres are also involved in this activity, e.g. Research and Development Institute for Power Engineering (RDIPE). Optimum

The IAEA's international project on innovative nuclear reactors and fuel cycles (INPRO)

International Nuclear Information System (INIS)

Kuptiz, Juergen; )

2002-01-01

This paper presents the IAEA International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO). It defines its rationale, key objectives and specifies the organizational structure. The IAEA General Conference (2000) has invited all interested Member states to combine their efforts under the aegis of the Agency in considering the issues of the nuclear fuel cycle, in particular by examining innovative and proliferation-resistant nuclear technology and invited Member states to consider to contribute to a task force on innovative nuclear reactors and fuel cycle

Study of an optimal configuration of a transmutation reactor based on a low-aspect-ratio tokamak

Energy Technology Data Exchange (ETDEWEB)

Hong, Bong Guen, E-mail: bghong@jbnu.ac.kr [Department of Quantum System Engineering, Chonbuk National University, 567 Baekje-daero, Jeonju, Jeonbuk 54896 (Korea, Republic of); Kim, Hoseok [Department of Applied Plasma Engineering, Chonbuk National University, 567 Baekje-daero, Jeonju, Jeonbuk 54896 (Korea, Republic of)

2016-11-15

Highlights: • Optimum configuration of a transmutation reactor based on a low aspect ratio tokamak was found. • Inboard and outboard radial build are determined by plasma physics, engineering and neutronics constraints. • Radial build and equilibrium fuel cycle play a major role in determining the transmutation characteristics. - Abstract: We determine the optimal configuration of a transmutation reactor based on a low-aspect-ratio tokamak. For self-consistent determination of the radial build of the reactor components, we couple a tokamak systems analysis with a radiation transport calculation. The inboard radial build of the reactor components is obtained from plasma physics and engineering constraints, while outboard radial builds are mainly determined by constraints on neutron multiplication, the tritium-breeding ratio, and the power density. We show that the breeding blanket model has an effect on the radial build of a transmutation blanket. A burn cycle has to be determined to keep the fast neutron fluence plasma-facing material below its radiation damage limit. We show that the radial build of the transmutation reactor components and the equilibrium fuel cycle play a major role in determining the transmutation characteristics.

Effect of reduced enrichment on the fuel cycle for research reactors

International Nuclear Information System (INIS)

Travelli, A.

1982-01-01

The new fuels developed by the RERTR Program and by other international programs for application in research reactors with reduced uranium enrichment (<20% EU) are discussed. It is shown that these fuels, combined with proper fuel-element design and fuel-management strategies, can provide at least the same core residence time as high-enrichment fuels in current use, and can frequently significantly extend it. The effect of enrichment reduction on other components of the research reactor fuel cycle, such as uranium and enrichment requirements, fuel fabrication, fuel shipment, and reprocessing are also briefly discussed with their economic implications. From a systematic comparison of HEU and LEU cores for the same reference research reactor, it is concluded that the new fuels have a potential for reducing the research reactor fuel cycle costs while reducing, at the same time, the uranium enrichment of the fuel

A study on the criticality search of transuranium recycling BWR core by adjusting supplied fuel composition in equilibrium state

International Nuclear Information System (INIS)

Seino, Takeshi; Sekimoto, Hiroshi

1997-01-01

There have been some difficulties in carrying out an extensive evaluation of the equilibrium state of Light Water Reactor (LWR) recycling operations keeping their fixed criticality condition using conventional design codes, because of the complexity of their calculational model for practical fuel and core design and because of a large amount of calculation time. This study presents an efficient approach to secure the criticality in an equilibrium cycle by adjusting a supplied fuel composition. The criticality search is performed by the use of fuel importance obtained from the equilibrium adjoint to a continuously fuel supplied core burnup equation. Using this method, some numerical analyses were carried out in order to evaluate the mixed oxide (MOX) fuel composition of equilibrium Boiling Water Reactor (BWR) cores satisfying the criticality requirement. The results showed the comprehensive and quantitative characteristics on the equilibrium cores confining transuranium for different MOX fuel loading fractions and irradiating conditions. (author)

Flexible fuel cycle initiative for the transition period from current reactors to next generation reactors

International Nuclear Information System (INIS)

Yamashita, Junichi; Fukasawa, Tetsuo; Hoshino, Kuniyoshi; Kawamura, Fumio; Shiina, Kouji; Sasahira, Akira

2005-01-01

A sustainable electricity supply by fast breeder reactors (FBRs) is essential to ensure energy security and prevent global warming. Transition from light water reactors (LWRs) to FBRs and establishment of an FBR cycle are indispensable, which requires plutonium (Pu) for the introduction of FBRs. The authors propose advanced system called 'Flexible Fuel Cycle Initiative (FFCI)' which can respond flexibly the future expected technical and social uncertainties, can hold no surplus Pu, and can achieve an economical FBR cycle. In the new concept of FFCI, 2nd LWR reprocessing which would succeed Rokkasho Reprocessing Plant is a simple facility to carry out only uranium (U) removal and residual 'recycle material' is stored or utilized. According to FBRs introduction status, recycle material is immediately treated in an FBR reprocessing to fabricate FBR fuel or temporarily stored for the utilization in FBRs at necessary timing. FFCI has high flexibility by having several options for future uncertainties by the introduction of recycle material as a buffer material between LWR and FBR cycles. (author)

Dynamic Analysis of the Thorium Fuel Cycle in CANDU Reactors

International Nuclear Information System (INIS)

Jeong, Chang Joon; Park, Chang Je

2006-02-01

The thorium fuel recycle scenarios through the Canada deuterium uranium (CANDU) reactor have been analyzed for two types of thorium fuel: homogeneous ThO 2 UO 2 and ThO 2 UO 2 -DUPIC fuels. The recycling is performed through the dry process fuel technology which has a proliferation resistance. For the once-through fuel cycle model, the existing nuclear power plant construction plan was considered up to 2016, while the nuclear demand growth rate from the year 2016 was assumed to be 0%. After setting up the once-through fuel cycle model, the thorium fuel CANDU reactor was modeled to investigate the fuel cycle parameters. In this analysis, the spent fuel inventory as well as the amount of plutonium, minor actinides and fission products of the multiple recycling fuel cycle were estimated and compared to those of the once-through fuel cycle. From the analysis results, it was found that the closed or partially closed thorium fuel cycle can be constructed through the dry process technology. Also, it is known that both the homogeneous and heterogeneous thorium fuel cycles can reduce the SF accumulation and save the natural uranium resource compared with the once-through cycle. From the material balance view point, the heterogeneous thorium fuel cycle seems to be more feasible. It is recommended, however, the economic analysis should be performed in future

Dynamic Analysis of the Thorium Fuel Cycle in CANDU Reactors

Energy Technology Data Exchange (ETDEWEB)

Jeong, Chang Joon; Park, Chang Je

2006-02-15

The thorium fuel recycle scenarios through the Canada deuterium uranium (CANDU) reactor have been analyzed for two types of thorium fuel: homogeneous ThO{sub 2}UO{sub 2} and ThO{sub 2}UO{sub 2}-DUPIC fuels. The recycling is performed through the dry process fuel technology which has a proliferation resistance. For the once-through fuel cycle model, the existing nuclear power plant construction plan was considered up to 2016, while the nuclear demand growth rate from the year 2016 was assumed to be 0%. After setting up the once-through fuel cycle model, the thorium fuel CANDU reactor was modeled to investigate the fuel cycle parameters. In this analysis, the spent fuel inventory as well as the amount of plutonium, minor actinides and fission products of the multiple recycling fuel cycle were estimated and compared to those of the once-through fuel cycle. From the analysis results, it was found that the closed or partially closed thorium fuel cycle can be constructed through the dry process technology. Also, it is known that both the homogeneous and heterogeneous thorium fuel cycles can reduce the SF accumulation and save the natural uranium resource compared with the once-through cycle. From the material balance view point, the heterogeneous thorium fuel cycle seems to be more feasible. It is recommended, however, the economic analysis should be performed in future.

A New Dynamic Model for Nuclear Fuel Cycle System Analysis

International Nuclear Information System (INIS)

Choi, Sungyeol; Ko, Won Il

2014-01-01

The evaluation of mass flow is a complex process where numerous parameters and their complex interaction are involved. Given that many nuclear power countries have light and heavy water reactors and associated fuel cycle technologies, the mass flow analysis has to consider a dynamic transition from the open fuel cycle to other cycles over decades or a century. Although an equilibrium analysis provides insight concerning the end-states of fuel cycle transitions, it cannot answer when we need specific management options, whether the current plan can deliver these options when needed, and how fast the equilibrium can be achieved. As a pilot application, the government brought several experts together to conduct preliminary evaluations for nuclear fuel cycle options in 2010. According to Table 1, they concluded that the closed nuclear fuel cycle has long-term advantages over the open fuel cycle. However, it is still necessary to assess these options in depth and to optimize transition paths of these long-term options with advanced dynamic fuel cycle models. A dynamic simulation model for nuclear fuel cycle systems was developed and its dynamic mass flow analysis capability was validated against the results of existing models. This model can reflects a complex combination of various fuel cycle processes and reactor types, from once-through to multiple recycling, within a single nuclear fuel cycle system. For the open fuel cycle, the results of the developed model are well matched with the results of other models

Three-batch reloading scheme for IRIS reactor extended cycles

International Nuclear Information System (INIS)

Jecmenica, R.; Pevec, D.; Grgic, D.

2004-01-01

To fully exploit the IRIS reactor optimized maintenance, and at the same time improve fuel utilization, a core design enabling a 4-year operating cycle together with a three-batch reloading scheme is desirable. However, this requires not only the increased allowed burnup but also use of fuel with uranium oxide enriched beyond 5%. This paper considers three-batch reloading scheme for a 4-year operating cycle with the assumptions of increased discharge burnup and fuel enrichment beyond 5%. Calculational model of IRIS reactor core has been developed based on FER FA2D code for group constants generation and NRC's PARCS nodal code for global core analysis. Studies have been performed resulting in a preliminary design of a three-batch core configuration for the first cycle. It must be emphasized that this study is outside the current IRIS licensing efforts, which rely on the present fuel technology (enrichment below 5%), but it is of long-term interest for potential future IRIS design upgrades. (author)

Analysis of neutronic parameters of AP1000 core for 18 month and 16/20 month cycle schemes using CASMO4E and SIMULATE-3 codes

International Nuclear Information System (INIS)

Nawaz Amjad; Yoshikawa, Hidekazu; Ming Yang

2015-01-01

AP1000 reactor is designed for 18 month of operating cycle. The core can also be used for 16/20 months of operating cycle. This study is performed to analyze and compare the neutronic parameters of typical AP1000 reactor core for 18 month and 16/20 month alternate cycle lengths. CASMO4E and SIMULATE-3 code package is used for the analysis of initial and equilibrium cores. The key reactor physics safety parameters were analyzed including power peaking factors, core radial and axial power distribution and core reactivity feedback coefficients. Moreover, the analysis of fuel depletion, fission product buildup and burnable poison behaviour with burnup is also analyzed. Full 2-D fuel assembly model in CASMO4E and full 3-D core model in SIMULATE-3 is employed to examine core performance and safety parameters. In order to evaluate the equilibrium core neutronic parameters, the equilibrium core model is attained by performing burnup analysis from initial to equilibrium cycle, where optimized transition core design is obtained so that the power peaking factors remain within designed limits. The MTC for higher concentration of critical boron concentrations is slightly positive at lower moderator temperatures. However, it remains negative at operating temperature ranges. The radial core relative power distribution indicates that low leakage capability of initial and equilibrium cores is reduced at EOC. (author)

Fast molten salt reactor-transmuter for closing nuclear fuel cycle on minor actinides

International Nuclear Information System (INIS)

Dudnikov, A. A.; Alekseev, P. N.; Subbotin, S. A.

2007-01-01

Creation fast critical molten salt reactor for burning-out minor actinides and separate long-living fission products in the closed nuclear fuel cycle is the most perspective and actual direction. The reactor on melts salts - molten salt homogeneous reactor with the circulating fuel, working as burner and transmuter long-living radioactive nuclides in closed nuclear fuel cycle, can serve as an effective ecological cordon from contamination of the nature long-living radiotoxic nuclides. High-flux fast critical molten-salt nuclear reactors in structure of the closed nuclear fuel cycle of the future nuclear power can effectively burning-out / transmute dangerous long-living radioactive nuclides, make radioisotopes, partially utilize plutonium and produce thermal and electric energy. Such reactor allows solving the problems constraining development of large-scale nuclear power, including fueling, minimization of radioactive waste and non-proliferation. Burning minor actinides in molten salt reactor is capable to facilitate work solid fuel power reactors in system NP with the closed nuclear fuel cycle and to reduce transient losses at processing and fabrications fuel pins. At substantiation MSR-transmuter/burner as solvents fuel nuclides for molten-salt reactors various salts were examined, for example: LiF - BeF2; NaF - LiF - BeF2; NaF-LiF ; NaF-ZrF4 ; LiF-NaF -KF; NaCl. RRC 'Kurchatov institute' together with other employees have developed the basic design reactor installations with molten salt reactor - burner long-living nuclides for fluoride fuel composition with the limited solubility minor actinides (MAF3 10 mol %) allows to develop in some times more effective molten salt reactor with fast neutron spectrum - burner/ transmuter of the long-living radioactive waste. In high-flux fast reactors on melts salts within a year it is possible to burn ∼300 kg minor actinides per 1 GW thermal power of reactor. The technical and economic estimation given power

A system dynamics model for tritium cycle of pulsed fusion reactor

International Nuclear Information System (INIS)

Zhu, Zuolong; Nie, Baojie; Chen, Dehong

2017-01-01

As great challenges and uncertainty exist in achieving steady plasma burning, pulsed plasma burning may be a potential scenario for fusion engineering test reactor, even for fusion DEMOnstration reactor. In order to analyze dynamic tritium inventory and tritium self-sufficiency for pulsed fusion systems, a system dynamics model of tritium cycle was developed on the basis of earlier version of Tritium Analysis program for fusion System (TAS). The model was verified with TRIMO, which was developed by KIT in Germany. Tritium self-sufficiency and dynamic tritium inventory assessment were performed for a typical fusion engineering test reactor. The verification results show that the system dynamics model can be used for tritium cycle analysis of pulsed fusion reactor with sufficient reliability. The assessment results of tritium self-sufficiency indicate that the fusion reactor might only need several hundred gram tritium to startup if achieved high efficient tritium handling ability (Referred ITER: 1 h). And the initial tritium startup inventory in pulsed fusion reactor is determined by the combined influence of pulse length, burn availability, and tritium recycle time. Meanwhile, tritium self-sufficiency can be achieved under the defined condition.

A system dynamics model for tritium cycle of pulsed fusion reactor

Energy Technology Data Exchange (ETDEWEB)

Zhu, Zuolong; Nie, Baojie [Key Laboratory of Neutronics and Radiation Safety, Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); University of Science and Technology of China, Hefei, Anhui, 230027 (China); Chen, Dehong, E-mail: dehong.chen@fds.org.cn [Key Laboratory of Neutronics and Radiation Safety, Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China)

2017-05-15

As great challenges and uncertainty exist in achieving steady plasma burning, pulsed plasma burning may be a potential scenario for fusion engineering test reactor, even for fusion DEMOnstration reactor. In order to analyze dynamic tritium inventory and tritium self-sufficiency for pulsed fusion systems, a system dynamics model of tritium cycle was developed on the basis of earlier version of Tritium Analysis program for fusion System (TAS). The model was verified with TRIMO, which was developed by KIT in Germany. Tritium self-sufficiency and dynamic tritium inventory assessment were performed for a typical fusion engineering test reactor. The verification results show that the system dynamics model can be used for tritium cycle analysis of pulsed fusion reactor with sufficient reliability. The assessment results of tritium self-sufficiency indicate that the fusion reactor might only need several hundred gram tritium to startup if achieved high efficient tritium handling ability (Referred ITER: 1 h). And the initial tritium startup inventory in pulsed fusion reactor is determined by the combined influence of pulse length, burn availability, and tritium recycle time. Meanwhile, tritium self-sufficiency can be achieved under the defined condition.

Technical Meeting on Fast Reactors and Related Fuel Cycle Facilities with Improved Economic Characteristics. Working Material

International Nuclear Information System (INIS)

2013-01-01

The objectives of the meeting were: - To identify the main issues and technical features that affect capital and energy production costs of fast reactors and related fuel cycle facilities; - To present fast reactor concepts and designs with enhanced economic characteristics, as well as innovative technical solutions (components, subsystems, etc.) that have the potential to reduce the capital costs of fast reactors and related fuel cycle facilities; - To present energy models and advanced tools for the cost assessment of innovative fast reactors and associated nuclear fuel cycles; - To discuss the results of studies and on-going R&D activities that address cost reduction and the future economic competitiveness of fast reactors; and - To identify research and technology development needs in the field, also in view of new IAEA initiatives to help and support Member States in improving the economic competitiveness of fast reactors and associated nuclear fuel cycles

Fuel Cycle System Analysis Handbook

International Nuclear Information System (INIS)

Piet, Steven J.; Dixon, Brent W.; Gombert, Dirk; Hoffman, Edward A.; Matthern, Gretchen E.; Williams, Kent A.

2009-01-01

This Handbook aims to improve understanding and communication regarding nuclear fuel cycle options. It is intended to assist DOE, Campaign Managers, and other presenters prepare presentations and reports. When looking for information, check here. The Handbook generally includes few details of how calculations were performed, which can be found by consulting references provided to the reader. The Handbook emphasizes results in the form of graphics and diagrams, with only enough text to explain the graphic, to ensure that the messages associated with the graphic is clear, and to explain key assumptions and methods that cause the graphed results. Some of the material is new and is not found in previous reports, for example: (1) Section 3 has system-level mass flow diagrams for 0-tier (once-through), 1-tier (UOX to CR=0.50 fast reactor), and 2-tier (UOX to MOX-Pu to CR=0.50 fast reactor) scenarios - at both static and dynamic equilibrium. (2) To help inform fast reactor transuranic (TRU) conversion ratio and uranium supply behavior, section 5 provides the sustainable fast reactor growth rate as a function of TRU conversion ratio. (3) To help clarify the difference in recycling Pu, NpPu, NpPuAm, and all-TRU, section 5 provides mass fraction, gamma, and neutron emission for those four cases for MOX, heterogeneous LWR IMF (assemblies mixing IMF and UOX pins), and a CR=0.50 fast reactor. There are data for the first 10 LWR recycle passes and equilibrium. (4) Section 6 provides information on the cycle length, planned and unplanned outages, and TRU enrichment as a function of fast reactor TRU conversion ratio, as well as the dilution of TRU feedstock by uranium in making fast reactor fuel. (The recovered uranium is considered to be more pure than recovered TRU.) The latter parameter impacts the required TRU impurity limits specified by the Fuels Campaign. (5) Section 7 provides flows for an 800-tonne UOX separation plant. (6) To complement 'tornado' economic uncertainty

A study on the criticality search of transuranium recycling BWR core by adjusting supplied fuel composition in equilibrium state

International Nuclear Information System (INIS)

Seino, Takeshi; Sekimoto, Hiroshi

1998-01-01

There have been some difficulties in carrying out an extensive evaluation of the equilibrium state of Light Water Reactor (LWR) recycling operations keeping their fixed criticality condition using conventional design codes because of the complexity of their calculation model for practical fuel and core design and because of a large amount of calculation time. This study presents an efficient approach to secure the criticality in an equilibrium cycle by adjusting a supplied fuel composition. The criticality search is performed by the use of fuel importance obtained from the equation adjoint to a continuously fuel supplied core burnup equation. Using this method, some numerical analyses were carried out in order to evaluate the mixed oxide (MOX) fuel composition of equilibrium Boiling Water Reactor (BWR) cores satisfying the criticality requirement. The results showed the comprehensive and quantitative characteristics on the equilibrium cores confining transuraniums for different MOX fuel loading fractions and irradiating conditions

Pebble bed reactor fuel cycle optimization using particle swarm algorithm

Energy Technology Data Exchange (ETDEWEB)

Tavron, Barak, E-mail: btavron@bgu.ac.il [Planning, Development and Technology Division, Israel Electric Corporation Ltd., P.O. Box 10, Haifa 31000 (Israel); Shwageraus, Eugene, E-mail: es607@cam.ac.uk [Department of Engineering, University of Cambridge, Trumpington Street, Cambridge CB2 1PZ (United Kingdom)

2016-10-15

Highlights: • Particle swarm method has been developed for fuel cycle optimization of PBR reactor. • Results show uranium utilization low sensitivity to fuel and core design parameters. • Multi-zone fuel loading pattern leads to a small improvement in uranium utilization. • Thorium mixes with highly enriched uranium yields the best uranium utilization. - Abstract: Pebble bed reactors (PBR) features, such as robust thermo-mechanical fuel design and on-line continuous fueling, facilitate wide range of fuel cycle alternatives. A range off fuel pebble types, containing different amounts of fertile or fissile fuel material, may be loaded into the reactor core. Several fuel loading zones may be used since radial mixing of the pebbles was shown to be limited. This radial separation suggests the possibility to implement the “seed-blanket” concept for the utilization of fertile fuels such as thorium, and for enhancing reactor fuel utilization. In this study, the particle-swarm meta-heuristic evolutionary optimization method (PSO) has been used to find optimal fuel cycle design which yields the highest natural uranium utilization. The PSO method is known for solving efficiently complex problems with non-linear objective function, continuous or discrete parameters and complex constrains. The VSOP system of codes has been used for PBR fuel utilization calculations and MATLAB script has been used to implement the PSO algorithm. Optimization of PBR natural uranium utilization (NUU) has been carried out for 3000 MWth High Temperature Reactor design (HTR) operating on the Once Trough Then Out (OTTO) fuel management scheme, and for 400 MWth Pebble Bed Modular Reactor (PBMR) operating on the multi-pass (MEDUL) fuel management scheme. Results showed only a modest improvement in the NUU (<5%) over reference designs. Investigation of thorium fuel cases showed that the use of HEU in combination with thorium results in the most favorable reactor performance in terms of

Pebble bed reactor fuel cycle optimization using particle swarm algorithm

International Nuclear Information System (INIS)

Tavron, Barak; Shwageraus, Eugene

2016-01-01

Highlights: • Particle swarm method has been developed for fuel cycle optimization of PBR reactor. • Results show uranium utilization low sensitivity to fuel and core design parameters. • Multi-zone fuel loading pattern leads to a small improvement in uranium utilization. • Thorium mixes with highly enriched uranium yields the best uranium utilization. - Abstract: Pebble bed reactors (PBR) features, such as robust thermo-mechanical fuel design and on-line continuous fueling, facilitate wide range of fuel cycle alternatives. A range off fuel pebble types, containing different amounts of fertile or fissile fuel material, may be loaded into the reactor core. Several fuel loading zones may be used since radial mixing of the pebbles was shown to be limited. This radial separation suggests the possibility to implement the “seed-blanket” concept for the utilization of fertile fuels such as thorium, and for enhancing reactor fuel utilization. In this study, the particle-swarm meta-heuristic evolutionary optimization method (PSO) has been used to find optimal fuel cycle design which yields the highest natural uranium utilization. The PSO method is known for solving efficiently complex problems with non-linear objective function, continuous or discrete parameters and complex constrains. The VSOP system of codes has been used for PBR fuel utilization calculations and MATLAB script has been used to implement the PSO algorithm. Optimization of PBR natural uranium utilization (NUU) has been carried out for 3000 MWth High Temperature Reactor design (HTR) operating on the Once Trough Then Out (OTTO) fuel management scheme, and for 400 MWth Pebble Bed Modular Reactor (PBMR) operating on the multi-pass (MEDUL) fuel management scheme. Results showed only a modest improvement in the NUU (<5%) over reference designs. Investigation of thorium fuel cases showed that the use of HEU in combination with thorium results in the most favorable reactor performance in terms of

Assessment of the Dry Processed Oxide Fuel in Liquid Metal Fast Reactors

Energy Technology Data Exchange (ETDEWEB)

Roh, Gyu Hong; Choi, Hang Bok

2005-09-15

The neutronic feasibility of the dry process oxide fuel was assessed for the sodium-cooled and lead-cooled fast reactors (SFR and LFR, respectively), which were recommended as Generation-IV (Gen-IV) reactor systems by the Gen-IV international forum. The reactor analysis was performed for the equilibrium fuel cycle of two core configurations: Hybrid BN-600 benchmark core with an enlarged lattice pitch and a modified BN-600 core. The dry process technology assumed in this study is the molten-salt process, which was developed by Russian scientists for recycling oxide fuels. The core calculation was performed by the REBUS-3 code and the reactor characteristics such as the transuranic (TRU) enrichment, breeding ratio, peak linear power, burnup reactivity swing, etc. were calculated for the equilibrium core under a fixed fuel management scheme. The results showed that a fissile self-sustainable breakeven core was achievable without blanket fuels when the fuel volume fraction was {approx}50% and most of the fission products were removed. If the design criteria used in this study is proved to be acceptable through a detailed physics design and thermal hydraulic analysis in the future, it is practically possible to construct an equilibrium fuel cycle of the SFR and LFR systems based on the oxide fuel by utilizing the dry process technology.

Assessment of the Dry Processed Oxide Fuel in Liquid Metal Fast Reactors

International Nuclear Information System (INIS)

Roh, Gyu Hong; Choi, Hang Bok

2005-09-01

The neutronic feasibility of the dry process oxide fuel was assessed for the sodium-cooled and lead-cooled fast reactors (SFR and LFR, respectively), which were recommended as Generation-IV (Gen-IV) reactor systems by the Gen-IV international forum. The reactor analysis was performed for the equilibrium fuel cycle of two core configurations: Hybrid BN-600 benchmark core with an enlarged lattice pitch and a modified BN-600 core. The dry process technology assumed in this study is the molten-salt process, which was developed by Russian scientists for recycling oxide fuels. The core calculation was performed by the REBUS-3 code and the reactor characteristics such as the transuranic (TRU) enrichment, breeding ratio, peak linear power, burnup reactivity swing, etc. were calculated for the equilibrium core under a fixed fuel management scheme. The results showed that a fissile self-sustainable breakeven core was achievable without blanket fuels when the fuel volume fraction was ∼50% and most of the fission products were removed. If the design criteria used in this study is proved to be acceptable through a detailed physics design and thermal hydraulic analysis in the future, it is practically possible to construct an equilibrium fuel cycle of the SFR and LFR systems based on the oxide fuel by utilizing the dry process technology

Advance reactor and fuel-cycle systems--potentials and limitations for United States utilities

International Nuclear Information System (INIS)

Zebroski, E.L.; Williams, R.F.

1979-01-01

This paper reviews the potential benefits and limitations of advance reactor and fuel-cycle systems for United States utilities. The results of the review of advanced technologies show that for the near and midterm, the only advance reactor and fuel-cycle system with significant potential for United States utilities is the current LWR, and evolutionary, not revolutionary, enhancements. For the long term, the liquid-metal breeder reactor continues to be the most promising advance nuclear option. The major factors leading to this conclusion are summarized

Investigation of thermodynamic cycle for generic 1200 MW{sub el} pressure channel reactor with nuclear steam superheat

Energy Technology Data Exchange (ETDEWEB)

Vincze, A.; Sidawi, K.; Abdullah, R.; Baldock, M.; Saltanov, E.; Pioro, I., E-mail: andrei.vincze@uoit.net, E-mail: khalil.sidawi@uoit.net, E-mail: rand.abdullah@uoit.net, E-mail: matthew.baldock@uoit.net, E-mail: eugene.saltanov@uoit.ca, E-mail: igor.pioro@uoit.ca [Univ. of Ontario Inst. of Tech., Oshawa, ON (Canada)

2014-07-01

Current Nuclear Power Plants (NPPs) play a significant role in energy production around the world. All NPPs operating today employ a Rankine steam cycle for the conversion of thermal power to electricity. This paper will examine the steam cycle arrangement an experimental pressure channel reactor using Nuclear Steam Superheat (NSS) and compare it to two advanced reactor designs, the Advanced CANDU Reactor 1000 (ACR-1000) and the Advanced Boiling Water Reactor (ABWR) designs. The thermodynamic cycle layout and thermal efficiencies of the three reactor types will be discussed. (author)

Searching for full power control rod patterns in a boiling water reactor using genetic algorithms

Energy Technology Data Exchange (ETDEWEB)

Montes, Jose Luis [Departamento Sistemas Nucleares, ININ, Carr. Mexico-Toluca Km. 36.5, Ocoyoacac, Edo. de Mexico (Mexico)]. E-mail: jlmt@nuclear.inin.mx; Ortiz, Juan Jose [Departamento Sistemas Nucleares, ININ, Carr. Mexico-Toluca Km. 36.5, Ocoyoacac, Edo. de Mexico (Mexico)]. E-mail: jjortiz@nuclear.inin.mx; Requena, Ignacio [Departamento Ciencias Computacion e I.A. ETSII, Informatica, Universidad de Granada, C. Daniel Saucedo Aranda s/n. 18071 Granada (Spain)]. E-mail: requena@decsai.ugr.es; Perusquia, Raul [Departamento Sistemas Nucleares, ININ, Carr. Mexico-Toluca Km. 36.5, Ocoyoacac, Edo. de Mexico (Mexico)]. E-mail: rpc@nuclear.inin.mx

2004-11-01

One of the most important questions related to both safety and economic aspects in a nuclear power reactor operation, is without any doubt its reactivity control. During normal operation of a boiling water reactor, the reactivity control of its core is strongly determined by control rods patterns efficiency. In this paper, GACRP system is proposed based on the concepts of genetic algorithms for full power control rod patterns search. This system was carried out using LVNPP transition cycle characteristics, being applied too to an equilibrium cycle. Several operation scenarios, including core water flow variation throughout the cycle and different target axial power distributions, are considered. Genetic algorithm fitness function includes reactor security parameters, such as MLHGR, MCPR, reactor k{sub eff} and axial power density.

Fuel cycle optimization in PWR'S

International Nuclear Information System (INIS)

Castro Lobo, P.D. de; Amorim, E.S. do.

1979-08-01

Neutronics aspects of a reactor core throughout its cycle were investigated in a search for increasing in-core utilization of the residual fissile isotopes content in the cycle discharged disposal. The effects due to design modifications introduced at burnup levels near the end-of-cycle, in an equilibrium cycle condition, have indicated the possibility of a better in-core utilization of the residual fissile isotopes existing in the cycle discharged disposal. The potential benefits are significant to warranty an examination of the mechanical and thermal hydraulic involved. At convenient burnup levels, change in H 2 O/UO 2 volume ratio were introduced allowing an intense depletion of the residual fissile isotopes existing in assemblies with high exposures levels. (Author) [pt

Evaluation of various fuel cycles to control inventories of plutonium and minor in advanced fuel cycles

International Nuclear Information System (INIS)

Miller, L.F.; Anderson, T.; Preston, J.; Humberstone, M.; Hou, J.; McConn, J.; Van Den Durpel, L.

2007-01-01

Inventories of Plutonium and minor actinides are important factors in determination of the risk associated with the use of nuclear energy. This includes the potential of exceeding release limits from a repository and the potential for proliferation. The amount of these materials in any given fleet of reactors is determined in large part by the choice of fuel cycle and by the types of reactors selected for operation. Most of the US reactor fleet will need to be replaced within the next 30 years and additional reactors will need to be added if the contribution of power from nuclear energy is expanded. In order to minimize risk and to make judicious use of repository space, inventories of all radionuclides will need to be effectively managed. Use of hard-spectrum reactors to burn excess Plutonium and other actinides is technologically feasible and is most likely less costly than any other options for minimizing various risks. Calculations for the inventories of several categories of radionuclides indicate that introduction of a modest fraction of fast reactors into the US reactor fleet is effective in stabilizing the growth of problematic radioisotopes. Results are obtained from the DANESS (Dynamic Analysis of Nuclear Energy System Strategies)1,2 Code and from the solution of algebraic equations that define steady state inventories. There are various different possible fuel cycle scenarios to utilize in the implementation of fast, thermal and intermediate spectrum reactors into the U.S. fleet. Results include various combinations of reactor types and fuel with varying times of implementations. Mass flows with uncertainties for equilibrium cycles will also be reported. Time-dependent scenarios are modeled with the DANESS code, and algebraic equations for various fuel cycles are derived. Uncertainties are obtained using Monte Carlo simulations based on estimates of parameters in the models. (authors)

Evaluation of various fuel cycles to control inventories of plutonium and minor in advanced fuel cycles

Energy Technology Data Exchange (ETDEWEB)

Miller, L.F.; Anderson, T.; Preston, J.; Humberstone, M.; Hou, J.; McConn, J. [Tennessee Univ., Nuclear Engineering Dept., Knoxville, TN (United States); Van Den Durpel, L. [Argonne National Laboratory, Argonne, IL (United States)

2007-07-01

Inventories of Plutonium and minor actinides are important factors in determination of the risk associated with the use of nuclear energy. This includes the potential of exceeding release limits from a repository and the potential for proliferation. The amount of these materials in any given fleet of reactors is determined in large part by the choice of fuel cycle and by the types of reactors selected for operation. Most of the US reactor fleet will need to be replaced within the next 30 years and additional reactors will need to be added if the contribution of power from nuclear energy is expanded. In order to minimize risk and to make judicious use of repository space, inventories of all radionuclides will need to be effectively managed. Use of hard-spectrum reactors to burn excess Plutonium and other actinides is technologically feasible and is most likely less costly than any other options for minimizing various risks. Calculations for the inventories of several categories of radionuclides indicate that introduction of a modest fraction of fast reactors into the US reactor fleet is effective in stabilizing the growth of problematic radioisotopes. Results are obtained from the DANESS (Dynamic Analysis of Nuclear Energy System Strategies)1,2 Code and from the solution of algebraic equations that define steady state inventories. There are various different possible fuel cycle scenarios to utilize in the implementation of fast, thermal and intermediate spectrum reactors into the U.S. fleet. Results include various combinations of reactor types and fuel with varying times of implementations. Mass flows with uncertainties for equilibrium cycles will also be reported. Time-dependent scenarios are modeled with the DANESS code, and algebraic equations for various fuel cycles are derived. Uncertainties are obtained using Monte Carlo simulations based on estimates of parameters in the models. (authors)

Sensitivity study on nitrogen Brayton cycle coupled with a small ultra-long cycle fast reactor

International Nuclear Information System (INIS)

Seo, Seok Bin; Seo, Han; Bang, In Cheol

2014-01-01

The main characteristics of UCFR are constant neutron flux and power density. They move their positions every moment at constant speed along with axial position of fuel rod for 60 years. Simultaneously with the development of the reactors, a new power conversion system has been considered. To solve existing issues of vigorous sodium-water reaction in SFR with steam power cycle, many researchers suggested a closed Brayton cycle as an alternative technique for SFR power conversion system. Many inactive gases are selected as a working fluid in Brayton power cycle, mainly supercritical CO 2 (S-CO 2 ). However, S-CO 2 still has potential for reaction with sodium. CO 2 -sodium reaction produces solid product, which has possibility to have an auto ignition reaction around 600 .deg. C. Thus, instead of S-CO 2 , CEA in France has developed nitrogen power cycle for ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration). In addition to inactive characteristic of nitrogen with sodium, its thermal and physical similarity with air enables to easily adopt to existing air Brayton cycle technology. In this study, for an optimized power conversion system for UCFR, a nitrogen Brayton cycle was analyzed in thermodynamic aspect. Based on subchannel analysis data of UCFR-100, a parametric study for thermal performance of nitrogen Brayton cycle was achieved. The system maximum pressure significantly affects to the overall efficiency of cycle, while other parameters show little effects. Little differences of the overall efficiencies for all cases between three stages (BOC, MOC, EOC) indicate that the power cycle of UCFR-100 maintains its performance during the operation

Closed Brayton cycle power conversion systems for nuclear reactors :

Energy Technology Data Exchange (ETDEWEB)

Wright, Steven A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Lipinski, Ronald J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Vernon, Milton E. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sanchez, Travis [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2006-04-01

This report describes the results of a Sandia National Laboratories internally funded research program to study the coupling of nuclear reactors to gas dynamic Brayton power conversion systems. The research focused on developing integrated dynamic system models, fabricating a 10-30 kWe closed loop Brayton cycle, and validating these models by operating the Brayton test-loop. The work tasks were performed in three major areas. First, the system equations and dynamic models for reactors and Closed Brayton Cycle (CBC) systems were developed and implemented in SIMULINKTM. Within this effort, both steady state and dynamic system models for all the components (turbines, compressors, reactors, ducting, alternators, heat exchangers, and space based radiators) were developed and assembled into complete systems for gas cooled reactors, liquid metal reactors, and electrically heated simulators. Various control modules that use proportional-integral-differential (PID) feedback loops for the reactor and the power-conversion shaft speed were also developed and implemented. The simulation code is called RPCSIM (Reactor Power and Control Simulator). In the second task an open cycle commercially available Capstone C30 micro-turbine power generator was modified to provide a small inexpensive closed Brayton cycle test loop called the Sandia Brayton test-Loop (SBL-30). The Capstone gas-turbine unit housing was modified to permit the attachment of an electrical heater and a water cooled chiller to form a closed loop. The Capstone turbine, compressor, and alternator were used without modification. The Capstone systems nominal operating point is 1150 K turbine inlet temperature at 96,000 rpm. The annular recuperator and portions of the Capstone control system (inverter) and starter system also were reused. The rotational speed of the turbo-machinery is controlled by adjusting the alternator load by using the electrical grid as the load bank. The SBL-30 test loop was operated at

Plant accident dynamics of high-temperature reactors with direct gas turbine cycle

International Nuclear Information System (INIS)

Waloch, M.L.

1977-01-01

In the paper submitted, a one-dimensional accident simulation model for high-temperature reactors with direct-cycle gas turbine (single-cycle facilities) is described. The paper assesses the sudden failure of a gas duct caused by the double-ended break of one out of several parallel pipes before and behind the reactor for a non-integrated plant, leading to major loads in the reactor region, as well as the complete loss of vanes of the compressor for an integrated plant. The results of the calculations show especially high loads for the break of a hot-gas pipe immediately behind the flow restrictors of the reactor outlet, because of prolonged effects of pressure gradients in the reactor region and the maximum core differential pressure. A plant accident dynamics calculation therefore allows to find a compromise between the requirements of stable compressor operation, on the one hand, and small loads in the reactor in the course of an accident, on the other, by establishing in a co-ordinated manner the narrowing ratio of the flow restrictors. (GL) [de

Current status of feasibility studies on commercialized fuel cycle system for Fast Breeder Reactor

International Nuclear Information System (INIS)

Ojima, Hisao; Nagaoki, Yoshihiro

2000-01-01

A 'Feasibility Studies on Commercialized Fast Breeder Reactor Cycle System' is underway at the Japan Nuclear Cycle Development Institute (JNC). The study will select the promising concepts with their R and D tasks in order to commercialize the fast breeder reactor (FBR) cycle system. The feasibility studies (F/S) have to present surveyed and screened various relevant technologies, and defined the design requirement of the commercialized fuel cycle system for FBR. The promising technical options are being evaluated and conceptual designs are being examined. At the end of JFY2000, several candidate concepts of the commercialized FBR cycle system will be proposed. (author)

A new paradigm for core design aimed at the sustainability of nuclear energy: The solution of the extended equilibrium state

International Nuclear Information System (INIS)

Artioli, Carlo; Grasso, Giacomo; Petrovich, Carlo

2010-01-01

The future expansion of nuclear energy, a technology identified as one of the main candidates for reducing the world's dependence on fossil fuels, requires a thorough analysis of the sustainability of this energy source for long-term supply. Generation-IV nuclear systems could represent a turning point for energy production by minimizing the environmental footprint of the fuel cycle. A new paradigm is thus required for reactor design, focusing, at the core design level, on both the closure of the fuel cycle and the effective utilization of natural resources. Within this framework, the so-called 'adiabatic core' concept represents a particularly interesting solution. It is based on the idea of ensuring by design a condition of equilibrium in the fuel cycle (i.e., an equilibrium 'fuel vector'), foreseeing nuclear power systems able to maintain a constant total amount of both plutonium and minor actinides (TRU), consuming only uranium (either natural or depleted), while discharging to the environment only fission products and reprocessing losses. Under such a hypothesis, all actinides can be continuously recycled in the same system, reducing both the waste volume and its long-term radiotoxicity, as well as utilizing effectively uranium resources. Two mathematical approaches have been devised to find the 'extended' equilibrium solution for the fuel vector. These methods are compared, validated with the codes MCNPX and FISPACT and applied to the European lead-cooled fast reactor ELSY, confirming the potential of this approach (e.g., a reduction by two orders of magnitude of the TRU mass in the final waste in comparison with the fuel cycle of Light Water Reactors operated in a once-through scenario).

Calculation of the evolution of molten salt breeder reactor

International Nuclear Information System (INIS)

Esteves, Fernando de Avelar

1999-01-01

A forecast for the future electrical consumption in Brazil and forecast of the nuclear electrical generation demand are discussed in this paper, which includes also an analysis on advanced nuclear reactors concept to supply that demand. This paper presents a concise description of the Molten Salt Breeder Reactor, considered the most appropriated to meet that demand. This paper also presents the burnup calculation modeling, including the operation modeling of this type of reactor from an initial load o 233 U up to the equilibrium cycle, the results of these calculations and its analysis. (author)

Non-Proliferative, Thorium-Based, Core and Fuel Cycle for Pressurized Water Reactors

International Nuclear Information System (INIS)

Todosow, M.; Raitses, G.; Galperin, A.

2009-01-01

Two of the major barriers to the expansion of worldwide adoption of nuclear power are related to proliferation potential of the nuclear fuel cycle and issues associated with the final disposal of spent fuel. The Radkowsky Thorium Fuel (RTF) concept proposed by Professor A. Radkowsky offers a partial solution to these problems. The main idea of the concept is the utilization of the seed-blanket unit (SBU) fuel assembly geometry which is a direct replacement for a 'conventional' assembly in either a Russian pressurized water reactor (VVER-1000) or a Western pressurized water reactor (PWR). The seed-blanket fuel assembly consists of a fissile (U) zone, known as seed, and a fertile (Th) zone known as blanket. The separation of fissile and fertile allows separate fuel management schemes for the thorium part of the fuel (a subcritical 'blanket') and the 'driving' part of the core (a supercritical 'seed'). The design objective for the blanket is an efficient generation and in-situ fissioning of the U233 isotope, while the design objective for the seed is to supply neutrons to the blanket in a most economic way, i.e. with minimal investment of natural uranium. The introduction of thorium as a fertile component in the nuclear fuel cycle significantly reduces the quantity of plutonium production and modifies its isotopic composition, reducing the overall proliferation potential of the fuel cycle. Thorium based spent fuel also contains fewer higher actinides, hence reducing the long-term radioactivity of the spent fuel. The analyses show that the RTF core can satisfy the requirements of fuel cycle length, and the safety margins of conventional pressurized water reactors. The coefficients of reactivity are comparable to currently operating VVER's/PWR's. The major feature of the RTF cycle is related to the total amount of spent fuel discharged for each cycle from the reactor core. The fuel management scheme adopted for RTF core designs allows a significant decrease in the

Burn cycle requirements comparison of pulsed and steady-state tokamak reactors

International Nuclear Information System (INIS)

Brooks, J.N.; Ehst, D.A.

1983-12-01

Burn cycle parameters and energy transfer system requirements were analyzed for an 8-m commercial tokamak reactor using four types of cycles: conventional, hybrid, internal transformer, and steady state. Not surprisingly, steady state is the best burn mode if it can be achieved. The hybrid cycle is a promising alternative to the conventional. In contrast, the internal transformer cycle does not appear attractive for the size tokamak in question

Compound process fuel cycle concept

International Nuclear Information System (INIS)

Ikegami, Tetsuo

2005-01-01

Mass flow of light water reactor spent fuel for a newly proposed nuclear fuel cycle concept 'Compound Process Fuel Cycle' has been studied in order to assess the capacity of the concept for accepting light water reactor spent fuels, taking an example for boiling water reactor mixed oxide spent fuel of 60 GWd/t burn-up and for a fast reactor core of 3 GW thermal output. The acceptable heavy metal of boiling water reactor mixed oxide spent fuel is about 3.7 t/y/reactor while the burn-up of the recycled fuel is about 160 GWd/t and about 1.6 t/y reactor with the recycled fuel burn-up of about 300 GWd/t, in the case of 2 times recycle and 4 times recycle respectively. The compound process fuel cycle concept has such flexibility that it can accept so much light water reactor spent fuels as to suppress the light water reactor spent fuel pile-up if not so high fuel burn-up is expected, and can aim at high fuel burn-up if the light water reactor spent fuel pile-up is not so much. Following distinctive features of the concept have also been revealed. A sort of ideal utilization of boiling water reactor mixed oxide spent fuel might be achieved through this concept, since both plutonium and minor actinide reach equilibrium state beyond 2 times recycle. Changes of the reactivity coefficients during recycles are mild, giving roughly same level of reactivity coefficients as the conventional large scale fast breeder core. Both the radio-activity and the heat generation after 4 year cooling and after 4 times recycle are less than 2.5 times of those of the pre recycle fuel. (author)

Actinide recycle potential in the integral fast reactor (IFR) fuel cycle

International Nuclear Information System (INIS)

Chang, Y.I.; Till, C.E.

1991-01-01

In the Integral Fast Reactor (IFR) development program, the entire reactor system -- reactor, fuel cycle, and waste process is being developed and optimized at the same time as a single integral entity. The use of metallic fuel in the IFR allows a radically improved fuel cycle technology. Based on the recent IFR process development, a preliminary assessment has been made to investigate the feasibility of further adapting pyrochemical processes to directly extract actinides from LWR spent fuel. The results of this assessment indicate very promising potential and two most promising flowsheet options have been identified for further research and development. This paper also summarizes current thinking on the rationale for actinide recycle, its ramifications on the geologic repository and the current high-level waste management plans, and the necessary development programs

Fuel management options to extend the IRIS reactor cycle

International Nuclear Information System (INIS)

Petrovic, B.; Franceschini, F.

2004-01-01

To optimize plant operation, reduce scheduled maintenance outage, and increase capacity factor, IRIS is designed to enable extended cycles of up to four years. However, due to the enrichment licensing limitation (less than 5% enriched uranium oxide) there is a tradeoff between the achievable cycle length and fuel utilization, i.e., the average fuel discharge burnup. The longest individual cycle may be achieved with the single-batch straight burn, but at the expense of a lower burnup. Considering the IRIS basic performance requirements, a cycle length in the range of three to four years is deemed desirable. This paper examines different fuel management options, i.e., the influence of the required cycle length on the corresponding reloading strategy, including a two-batch and a three-batch reloading. A reference two-batch core design has been developed for the first cycle, as well as for the transition cycles leading to equilibrium. Main core performance parameters are evaluated. This core design provides the framework for the safety analyses needed to prepare the IRIS safety evaluations. Alternate designs are also considered.(author)

The use of gas based energy conversion cycles for sodium fast reactors

International Nuclear Information System (INIS)

Saez, M.; Haubensack, D.; Alpy, N.; Gerber, A.; Daid, F.

2008-01-01

In the frame of Sodium Fast Reactors, CEA, AREVA and EDF are involved in a substantial effort providing both significant expertise and original work in order to investigate the interest to use a gas based energy conversion cycle as an alternative to the classical steam cycle. These gas cycles consist in different versions of the Brayton cycle, various types of gas being considered (helium, nitrogen, argon, separately or mixed, sub or supercritical carbon dioxide) as well as various cycle arrangements (indirect, indirect / combined cycles). The interest of such cycles is analysed in details by thermodynamic calculations and cycle optimisations. The objective of this paper is to provide a comparison between gas based energy conversion cycles from the viewpoint of the overall plant efficiency. Key factors affecting the Brayton cycle efficiency include the turbine inlet temperature, compressors and turbine efficiencies, recuperator effectiveness and cycle pressure losses. A nitrogen Brayton cycle at high pressure (between 100 and 180 bar) could appear as a potential near-term solution of classical gas power conversion system for maximizing the plant efficiency. At long-term, supercritical carbon dioxide Brayton cycle appears very promising for Sodium Fast Reactors, with a potential of high efficiency using even at a core outlet temperature of 545 deg. C. (authors)

Preliminary neutronic design of high burnup OTTO cycle pebble bed reactor

International Nuclear Information System (INIS)

Setiadipura, T.; Zuhair; Irwanto, D.

2015-01-01

The pebble bed type High Temperature Gas-cooled Reactor (HTGR) is among the interesting nuclear reactor designs in terms of safety and flexibility for co-generation applications. In addition, the strong inherent safety characteristics of the pebble bed reactor (PBR) which is based on natural mechanisms improve the simplicity of the PBR design, in particular for the Once-Through-Then-Out (OTTO) cycle PBR design. One of the important challenges of the OTTO cycle PBR design, and nuclear reactor design in general, is improving the nuclear fuel utilization which is shown by attaining a higher burnup value. This study performed a preliminary neutronic design study of a 200 MWt OTTO cycle PBR with high burnup while fulfilling the safety criteria of the PBR design.The safety criteria of the design was represented by the per-fuel-pebble maximum power generation of 4.5 kW/pebble. The maximum burnup value was also limited by the tested maximum burnup value which maintained the integrity of the pebble fuel. Parametric surveys were performed to obtain the optimized parameters used in this study, which are the fuel enrichment, per-pebble heavy metal (HM) loading, and the average axial speed of the fuel. An optimum design with burnup value of 131.1 MWd/Kg-HM was achieved in this study which is much higher compare to the burnup of the reference design HTR-MODUL and a previously proposed OTTO-cycle PBR design. This optimum design uses 17% U-235 enrichment with 4 g HM-loading per fuel pebble. (author)

Preliminary Neutronic Design of High Burnup OTTO Cycle Pebble Bed Reactor

Directory of Open Access Journals (Sweden)

T. Setiadipura

2015-04-01

Full Text Available The pebble bed type High Temperature Gas-cooled Reactor (HTGR is among the interesting nuclear reactor designs in terms of safety and flexibility for co-generation applications. In addition, the strong inherent safety characteristics of the pebble bed reactor (PBR which is based on natural mechanisms improve the simplicity of the PBR design, in particular for the Once-Through-Then-Out (OTTO cycle PBR design. One of the important challenges of the OTTO cycle PBR design, and nuclear reactor design in general, is improving the nuclear fuel utilization which is shown by attaining a higher burnup value. This study performed a preliminary neutronic design study of a 200 MWt OTTO cycle PBR with high burnup while fulfilling the safety criteria of the PBR design.The safety criteria of the design was represented by the per-fuel-pebble maximum power generation of 4.5 kW/pebble. The maximum burnup value was also limited by the tested maximum burnup value which maintained the integrity of the pebble fuel. Parametric surveys were performed to obtain the optimized parameters used in this study, which are the fuel enrichment, per-pebble heavy metal (HM loading, and the average axial speed of the fuel. An optimum design with burnup value of 131.1 MWd/Kg-HM was achieved in this study which is much higher compare to the burnup of the reference design HTR-MODUL and a previously proposed OTTO-cycle PBR design. This optimum design uses 17% U-235 enrichment with 4 g HM-loading per fuel pebble

Fuel Cycle of Reactor SVBR-100

Energy Technology Data Exchange (ETDEWEB)

Zrodnikov, A.V.; Toshinsky, G.I.; Komlev, O.G. [FSUE State Scientific Center Institute for Physics and Power Engineering, 1, Bondarenko sq., Obninsk, Kaluga rg., 249033 (Russian Federation)

2009-06-15

Modular fast reactor with lead-bismuth heavy liquid-metal coolant in 100 MWe class (SVBR 100) is referred to the IV Generation reactors and shall operate in a closed nuclear fuel cycle (NFC) without consumption of natural uranium. Usually it is considered that launch of fast reactors (FR) is realized using mixed uranium-plutonium fuel. However, such launch of FRs is not economically effective because of the current costs of natural uranium and uranium enrichment servicing. This is conditioned by the fact that the quantity of reprocessing the spent nuclear fuel (SNF) of thermal reactors (TR) calculated for a ton of plutonium that determines the expenditures for construction and operation of the corresponding enterprise is very large due to low content of plutonium in the TR SNF. The economical effectiveness of FRs will be reduced as the enterprises on reprocessing the TR SNF have to be built prior to FRs have been implemented in the nuclear power (NP). Moreover, the pace of putting the FRs in the NP will be constrained by the quantity of the TR SNF. The report grounds an alternative strategy of FRs implementation into the NP, which is considered to be more economically effective. That is conditioned by the fact that in the nearest future use of the mastered uranium oxide fuel for FRs and operation in the open fuel cycle with postponed reprocessing will be most economically expedient. Changeover to the mixed uranium-plutonium fuel and closed NFC will be economically effective when the cost of natural uranium is increased and the expenditures for construction of enterprises on SNF reprocessing, re-fabrication of new fuel with plutonium and their operating becomes lower than the corresponding costs of natural uranium, uranium enrichment servicing, expenditures for fabrication of fresh uranium fuel and long temporary storage of the SNF. As when operating in the open NFC, FRs use much more natural uranium as compared with TRs, and at a planned high pace of NP development

IAEA specialists' meeting on power ramping and cycling behaviour of water reactor fuel. Summary report

Energy Technology Data Exchange (ETDEWEB)

NONE

1983-06-01

At its fourth Annual Meeting, the IAEA International Working Group on Water Reactor Fuel Performance and Technology (IWGFPT) recommended that the Agency should hold a second Specialists' Meeting on 'Power Ramping and Cycling Behaviour of Water Reactor Fuel'. As research activities related to power ramping and cycling of water reactor fuel have been pursued vigorously, it was the objective of this meeting to review and discuss today's State of the Art and current understanding of water reactor fuel behaviour related to this these. Emphasis should be on practical experience and experimental investigations. The meeting was organised in five sessions: Power ramping and power cycling programs in power and and research reactors; Experimental methods; Power ramping and cycling results; Investigations and results of separate effects, especially related to PCI, defect mechanism, mechanical response, fuel design, and specially related to fission gas release; Operational strategies, recommendations and economic implications. The session chairmen, together with the speakers, prepared and presented reports with summary, conclusions and recommendations of the individual sessions. These reports are added to this summary report.

IAEA specialists' meeting on power ramping and cycling behaviour of water reactor fuel. Summary report

International Nuclear Information System (INIS)

1983-06-01

At its fourth Annual Meeting, the IAEA International Working Group on Water Reactor Fuel Performance and Technology (IWGFPT) recommended that the Agency should hold a second Specialists' Meeting on 'Power Ramping and Cycling Behaviour of Water Reactor Fuel'. As research activities related to power ramping and cycling of water reactor fuel have been pursued vigorously, it was the objective of this meeting to review and discuss today's State of the Art and current understanding of water reactor fuel behaviour related to this these. Emphasis should be on practical experience and experimental investigations. The meeting was organised in five sessions: Power ramping and power cycling programs in power and and research reactors; Experimental methods; Power ramping and cycling results; Investigations and results of separate effects, especially related to PCI, defect mechanism, mechanical response, fuel design, and specially related to fission gas release; Operational strategies, recommendations and economic implications. The session chairmen, together with the speakers, prepared and presented reports with summary, conclusions and recommendations of the individual sessions. These reports are added to this summary report

Status of sodium cooled fast reactors with closed fuel cycle in India

International Nuclear Information System (INIS)

Raj, B.

2007-01-01

Fast reactors form the second stage of India's 3-stage nuclear power programme. The seed for India's fast reactor programme was sown through the construction of the Fast Breeder Test Reactor (FBTR) at IGCAR, Kalpakkam, that was commissioned in 1985. FBTR has operated with an unique, indigenously developed plutonium rich mixed carbide fuel, which has reached a burn up as high as 155 GWd/t without any fuel failure in the core. The sodium systems in the reactor have performed excellently. The availability of the reactor has been as high as 92% in the recent campaigns. The fuel discharged from FBTR up to 100 GWd/t has been reprocessed successfully. The experience gained in the construction, commissioning and operation of FBTR has provided the necessary confidence to launch a Prototype FBR of 500 MWe capacity (PFBR). This reactor will be fuelled by uranium, plutonium mixed oxide. The reactor construction started in 2003 and the reactor is scheduled to be commissioned by 2010. The design of the reactor has incorporated the worldwide operating experience from the FBRs and has addressed various safety issues reported in literature, besides introducing a number of innovative features which have reduced the unit energy cost and contributed to its enhanced safety. Simultaneous with the construction of the reactor, the fuel cycle of the reactor has been addressed in a comprehensive manner and construction of a fuel cycle facility has been initiated. Subsequent to the PFBR, 4 more reactors with identical design are proposed to be constructed. Various elements of reactor design are being carefully analysed with the aim of introducing innovative features towards further reduction in unit energy cost and enhancing safety in these reactors

Wastes from the light water reactor fuel cycle

International Nuclear Information System (INIS)

Steindler, M.J.; Trevorrow, L.E.

1976-01-01

The LWR fuel cycle is represented, in the minimum detail necessary to indicate the origin of the wastes, as a system of operations that is typical of those proposed for various commercial fuel cycle ventures. The primary wastes (before any treatment) are described in terms of form, volume, radioactivity, chemical composition, weight, and combustibility (in anticipation of volume reduction treatments). Properties of the wastes expected from the operation of reactors, fuel reprocessing plants, and mixed oxide fuel fabrication plants are expressed in terms of their amounts per unit of nuclear energy produced

Preliminary analysis of combined cycle of modular high-temperature gas cooled reactor

International Nuclear Information System (INIS)

Baogang, Z.; Xiaoyong, Y.; Jie, W.; Gang, Z.; Qian, S.

2015-01-01

Modular high-temperature gas cooled reactor (HTGR) is known as one of the most advanced nuclear reactors because of its inherent safety and high efficiency. The power conversion system of HTGR can be steam turbine based on Rankine cycle or gas turbine based on Brayton cycle respectively. The steam turbine system is mature and the gas turbine system has high efficiency but under development. The Brayton-Rankine combined cycle is an effective way to further promote the efficiency. This paper investigated the performance of combined cycle from the viewpoint of thermodynamics. The effect of non-dimensional parameters on combined cycle’s efficiency, such as temperature ratio, compression ratio, efficiency of compressor, efficiency of turbine, was analyzed. Furthermore, the optimal parameters to achieve highest efficiency was also given by this analysis under engineering constraints. The conclusions could be helpful to the design and development of combined cycle of HTGR. (author)

Waste management in IFR [Integral Fast Reactor] fuel cycle

International Nuclear Information System (INIS)

Johnson, T.R.; Battles, J.E.

1991-01-01

The fuel cycle of the Integral Fast Reactor (IFR) has important potential advantage for the management of high-level wastes. This sodium-cooled, fast reactor will use metal fuels that are reprocessed by pyrochemical methods to recover uranium, plutonium, and the minor actinides from spent core and blanket fuel. More than 99% of all transuranic (TRU) elements will be recovered and returned to the reactor, where they are efficiently burned. The pyrochemical processes being developed to treat the high-level process wastes are capable of producing waste forms with low TRU contents, which should be easier to dispose of. However, the IFR waste forms present new licensing issues because they will contain chloride salts and metal alloys rather than glass or ceramic. These fuel processing and waste treatment methods can also handle TRU-rich materials recovered from light-water reactors and offer the possibility of efficiently and productively consuming these fuel materials in future power reactors

Coupling a Supercritical Carbon Dioxide Brayton Cycle to a Helium-Cooled Reactor.

Energy Technology Data Exchange (ETDEWEB)

Middleton, Bobby [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Pasch, James Jay [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Kruizenga, Alan Michael [Sandia National Lab. (SNL-CA), Livermore, CA (United States); Walker, Matthew [Sandia National Lab. (SNL-CA), Livermore, CA (United States)

2016-01-01

This report outlines the thermodynamics of a supercritical carbon dioxide (sCO₂) recompression closed Brayton cycle (RCBC) coupled to a Helium-cooled nuclear reactor. The baseline reactor design for the study is the AREVA High Temperature Gas-Cooled Reactor (HTGR). Using the AREVA HTGR nominal operating parameters, an initial thermodynamic study was performed using Sandia's deterministic RCBC analysis program. Utilizing the output of the RCBC thermodynamic analysis, preliminary values of reactor power and of Helium flow rate through the reactor were calculated in Sandia's HelCO₂ code. Some research regarding materials requirements was then conducted to determine aspects of corrosion related to both Helium and to sCO₂ , as well as some mechanical considerations for pressures and temperatures that will be seen by the piping and other components. This analysis resulted in a list of materials-related research items that need to be conducted in the future. A short assessment of dry heat rejection advantages of sCO₂> Brayton cycles was also included. This assessment lists some items that should be investigated in the future to better understand how sCO₂ Brayton cycles and nuclear can maximally contribute to optimizing the water efficiency of carbon free power generation

Breeding description for fast reactors and symbiotic reactor systems

International Nuclear Information System (INIS)

Hanan, N.A.

1979-01-01

A mathematical model was developed to provide a breeding description for fast reactors and symbiotic reactor systems by means of figures of merit type quantities. The model was used to investigate the effect of several parameters and different fuel usage strategies on the figures of merit which provide the breeding description. The integrated fuel cycle model for a single-reactor is reviewed. The excess discharge is automatically used to fuel identical reactors. The resulting model describes the accumulation of fuel in a system of identical reactors. Finite burnup and out-of-pile delays and losses are treated in the model. The model is then extended from fast breeder park to symbiotic reactor systems. The asymptotic behavior of the fuel accumulation is analyzed. The asymptotic growth rate appears as the largest eigenvalue in the solution of the characteristic equations of the time dependent differential balance equations for the system. The eigenvector corresponding to the growth rate is the core equilibrium composition. The analogy of the long-term fuel cycle equations, in the framework of this model, and the neutron balance equations is explored. An eigenvalue problem adjoint to the one generated by the characteristic equations of the system is defined. The eigenvector corresponding to the largest eigenvalue, i.e. to the growth rate, represents the ''isotopic breeding worths.'' Analogously to the neutron adjoint flux it is shown that the isotopic breeding worths represent the importance of an isotope for breeding, i.e. for the growth rate of a system

The continuous fuel cycle model and the gas cooled fast reactor

International Nuclear Information System (INIS)

Christie, Stuart; Lathouwers, Danny; Kloosterman, Jan Leen; Hagen, Tim van der

2011-01-01

The gas cooled fast reactor (GFR) is one of the generation IV designs currently being evaluated for future use. It is intended to behave as an isobreeder, producing the same amount of fuel as it consumes during operation. The actinides in the fuel will be recycled repeatedly in order to minimise the waste output to fission products only. Striking the balance of the fissioning of various actinides against transmutation and decay to achieve these goals is a complex problem. This is compounded by the time required for burn-up modelling, which can be considerable for a single cycle, and even longer for studies of fuel evolution over many cycles. The continuous fuel cycle model approximates the discrete steps of loading, operating and unloading a reactor as continuous processes. This simplifies the calculations involved in simulating the behaviour of the fuel, reducing the time needed to model the changes to the fuel composition over many cycles. This method is used to study the behaviour of GFR fuel over many cycles and compared to results obtained from direct calculations. The effects of varying fuel cycle properties such as feed material, recycling of additional actinides and reprocessing losses are also investigated. (author)

Adaptability of Brayton cycle conversion systems to fast, epithermal and thermal spectrum space nuclear reactors

International Nuclear Information System (INIS)

Tilliette, Z.P.

1988-01-01

The two French Government Agencies C.N.E.S. (Centre National d'Etudes Spatiales) and C.E.A. (Commissariat a l'Energie Atomique) are carrying out joint preliminary studies on space nuclear power systems for future ARIANE 5 launch vehicle applications. The Brayton cycle is the reference conversion system, whether the heat source is a liquid metal-cooled (NaK, Na or Li) reactor or a gas-cooled direct cycle concept. The search for an adequate utilization of this energy conversion means has prompted additional evaluations featuring the definition of satisfactory cycle conditions for these various kinds of reactor concepts. In addition to firstly studied fast and epithermal spectrum ones, thermal spectrum reactors can offer an opportunity of bringing out some distinctive features of the Brayton cycle, in particular for the temperature conditioning of the efficient metal hydrides (ZrH, Li/sub 7/H) moderators. One of the purposes of the paper is to confirm the potential of long lifetime ZrH moderated reactors associated with a gas cycle and to assess the thermodynamical consequences for both Nak(Na)-cooled or gas-cooled nuclear heat sources. This investigation is complemented by the definition of appropriate reactor arrangements which could be presented on a further occasion

Thermodynamic Simulation of Equilibrium Composition of Reaction Products at Dehydration of a Technological Channel in a Uranium-Graphite Reactor

Science.gov (United States)

Pavliuk, A. O.; Zagumennov, V. S.; Kotlyarevskiy, S. G.; Bespala, E. V.

2018-01-01

The problems of accumulation of nuclear fuel spills in the graphite stack in the course of operation of uranium-graphite nuclear reactors are considered. The results of thermodynamic analysis of the processes in the graphite stack at dehydration of a technological channel, fuel element shell unsealing and migration of fission products, and activation of stable nuclides in structural elements of the reactor and actinides inside the graphite moderator are given. The main chemical reactions and compounds that are produced in these modes in the reactor channel during its operation and that may be hazardous after its shutdown and decommissioning are presented. Thermodynamic simulation of the equilibrium composition is performed using the specialized code TERRA. The results of thermodynamic simulation of the equilibrium composition in different cases of technological channel dehydration in the course of the reactor operation show that, if the temperature inside the active core of the nuclear reactor increases to the melting temperature of the fuel element, oxides and carbides of nuclear fuel are produced. The mathematical model of the nonstationary heat transfer in a graphite stack of a uranium-graphite reactor in the case of the technological channel dehydration is presented. The results of calculated temperature evolution at the center of the fuel element, the replaceable graphite element, the air gap, and in the surface layer of the block graphite are given. The numerical results show that, in the case of dehydration of the technological channel in the uranium-graphite reactor with metallic uranium, the main reaction product is uranium dioxide UO2 in the condensed phase. Low probability of production of pyrophoric uranium compounds (UH3) in the graphite stack is proven, which allows one to disassemble the graphite stack without the risk of spontaneous graphite ignition in the course of decommissioning of the uranium-graphite nuclear reactor.

Fuel cycle costs for molten-salt reactors

International Nuclear Information System (INIS)

Nagashima, Kikusaburo

1983-01-01

This report describes FCC (fuel cycle cost) estimates for MSCR (molten-salt converter reactor) and MSBR (molten-salt breeder reactor) compared with those for LWRs (PWR and BWR). The calculation is based on the present worth technique with a given discount rate for each cost item, which enables us to make comparison between FCC's for MSCR, MSBR and LWRs. As far as the computational results obtained here are concerned, shown that the FCC's for MSCR and MSBR are 70 -- 60 % lower than the values for LWRs. And it could be said that the FCC for MSCR (Pu-converter) is about 10 % lower than that for MSBR, because of the smaller amount of fissile inventory of MSCR than the inventory of MSBR. (author)

Modeling the nonequilibrium effects in a nonquasi-equilibrium thermodynamic cycle based on steepest entropy ascent and an isothermal-isobaric ensemble

International Nuclear Information System (INIS)

Li, Guanchen; Spakovsky, Michael R. von

2016-01-01

Conventional first principle approaches for studying nonequilibrium or far-from-equilibrium processes depend on the mechanics of individual particles or quantum states. They also require many details of the mechanical features of a system to arrive at a macroscopic property. In contrast, thermodynamics provides an approach for determining macroscopic property values without going into these details, because the overall effect of particle dynamics results, for example, at stable equilibrium in an invariant pattern of the “Maxwellian distribution”, which in turn leads to macroscopic properties. However, such an approach is not generally applicable to a nonequilibrium process except in the near-equilibrium realm. To adequately address these drawbacks, steepest-entropy-ascent quantum thermodynamics (SEAQT) provides a first principle, thermodynamic-ensemble approach applicable to the entire nonequilibrium realm. Based on prior developments by the authors, this paper applies the SEAQT framework to modeling the nonquasi-equilibrium cycle, which a system with variable volume undergoes. Using the concept of hypoequilibrium state and nonequilibrium intensive properties, this framework provides a complete description of the nonequilibrium evolution in state of the system. Results presented here reveal how nonequilibrium effects influence the performance of the cycle. - Highlights: • First-principles nonequilibrium model of thermodynamic cycles. • Study of thermal efficiency losses due to nonequilibrium effects. • Study of systems undergoing nonquasi-equilibrium processes. • Study of the coupling of system relaxation and interaction with a reservoir.

A catalogue of advanced fuel cycles in CANDU-PHW reactors

International Nuclear Information System (INIS)

Veeder, J.; Didsbury, R.

1985-06-01

A catalogue raisonne is presented of various advanced fuel cycle options which have the potential of substantially improving the uranium utilization for CANDU-PHW reactors. Three categories of cycles are: once-through cycles without recovery of fissile materials, cycles that depend on the recovery and recycle of fissile materials in thorium or uranium, cycles that depend primarily on the production of fissile material in a fertile blanket by means of an intense neutron source other than fission, such as an accelerator breeder. Detailed tables are given of the isotopic compositions of the feed and discharge fuels, the logistics of materials and processes required to sustain each of the cycles, and tables of fuel cycle costs based on a method of continuous discounting of cash flow

Thermodynamics of open, nonisothermal chemical systems far from equilibrium

International Nuclear Information System (INIS)

Yoshida, Nobuo

1992-01-01

The thermodynamic behavior of kinetic models based on a continuously stirred tank reactor (CSTR) is studied in an attempt to seek general trends in the thermodynamic properties of open nonlinear systems. The models consist of two reversible reactions, A + nB rightleftharpoons (n + 1) B (n = 0,1,or 2) and B rightleftharpoons C, taking place in an adiabatic CSTR. The heat of reaction is incorporated, and the rate constants are assumed to follow an Arrhenius temperature dependence. The models give rise to multiple stationary states and sustained oscillations (limit cycles). The entropy difference between stationary or oscillatory states and equilibrium and the rate of entropy production in the these states are calculated as a function of the residence time in the reactor. The entropy difference and entropy production may be taken, to some extent, as indicative of the influence of irreversible processes, which disappears at equilibrium. The results of the calculations reveal the following systematic trends: (I) The entropy difference or entropy production for stable states or both always increase as the residence time is shortened, namely, as the system is displaced further from equilibrium. (II) If stable and unstable states (stationary or oscillatory) coexist under identical conditions, then the stable state invariably has a smaller value of the entropy difference or entropy production or both than the corresponding unstable state. 26 refs., 3 figs

Research on nuclear energy in the fields of fuel cycle, PWR reactors and LMFBR reactors

International Nuclear Information System (INIS)

Barre, B.; Camarcat, N.

1995-01-01

In this article we present the CEA research programs to improve the safety of the next generation of reactors, to manage the Plutonium and the wastes of the fuel cycle end and to ameliorate the competitiveness. 6 refs

Operation of CANDU power reactor in thorium self-sufficient fuel cycle

Indian Academy of Sciences (India)

These disadvantages of thorium fuel cycle were seemingly the reasons why that ... According to the data of figure 2, maximum (equilibrium) content of 233U in ..... Self-sufficient mode is related with rather big effort in the extraction of isotopes of.

Direct cycle type nuclear power plant

International Nuclear Information System (INIS)

Tagawa, Hisato; Ibe, Hidefumi.

1990-01-01

In a direct cycle type nuclear power plant such as BWR type reactor, since oxygen atoms in reactor water are actuvated by neutron irradiation in the reactor core, carry over of the thus formed radioactive nitrogen atoms causes increase in the dosage in a turbine system. Since 16 N accompanies in the main steams in the chemical form of 16 NO, it can not effectively be removed in a nitrogen removing device. In view of the above, hydrogen atom concentration is reduced by adding metals having high reaction with hydrogen atoms, for example, silver ions, chromium ions, or ruthenium ions are added to reactor water. Then, equilibrium concentration of 16 NO in water is reduced by suppressing the reaction: 16 NO 2 + H → 16 NO + OH. (T.M.)

Non-equilibrium statistical thermodynamics of neutron gas in reactor

International Nuclear Information System (INIS)

Hayasaka, Hideo

1977-01-01

The thermodynamic structures of non-equilibrium steady states of highly rarefied neutron gas in various media are considered for the irreversible processes owing to creative and destructive reactions of neutrons with nuclei of these media and supply from the external sources. Under the so-called clean and cold condition in reactor, the medium is regarded virtually as offering the different chemical potential fields for each subsystem of a steady neutron gas system. The fluctuations around a steady state are considered in a Markovian-Gaussian process. The generalized Einstein relations are derived for stationary neutron gas systems. The forces and flows of neutron gases in a medium are defined upon the general stationary solution of the Fokker-Planck equation. There exist the symmetry of the kinetic coefficients, and the minimum entropy production upon neutron-nuclear reactions. The distribution functions in various media are determined by each corresponding extremum condition under the vanishing of changes of the respective total entropies in the Gibbs equation. (auth.)

Alternative fuels, fuel cycles, and reactors: are they useful. are they necessary

International Nuclear Information System (INIS)

Spinrad, B.I.

1985-01-01

This chapter discusses reactors, fuel cycles, and fuel production concepts other than those considered conventional in the nuclear community. An attempt is made to look for improvements with the aim of providing cheaper and more durable energy systems, and to contribute toward a solution of the threat of weapons material diversion and weapons proliferation problems. Topics considered include breeding, alternate breeder cycles, alternative reprocessing schemes, symbiotic reactor systems, an interim strategy, and other sources of nuclear fuel. It is determined that the reprocessing of spent fuel is an important safeguard measure in itself

Accelerator-driven systems (ADS) and fast reactors (FR) in advanced nuclear fuel cycles

International Nuclear Information System (INIS)

2002-01-01

The long-term hazard of radioactive waste arising from nuclear energy production is a matter of continued discussion and public concern in many countries. Through partitioning and transmutation (P and T) of the actinides and some of the long-lived fission products, the radiotoxicity of high-level waste (HLW) can be reduced by a factor of 100 compared with the current once-through fuel cycle. This requires very effective reactor and fuel cycle strategies, including fast reactors (FR) and/or accelerator-driven, sub-critical systems (ADS). The present study compares FR- and ADS-based actinide transmutation systems with respect to reactor properties, fuel cycle requirements, safety, economic aspects and (R and D) needs. Several advanced fuel cycle strategies are analysed in a consistent manner to provide insight into the essential differences between the various systems in which the role of ADS is emphasised. The report includes a summary aimed at policy makers and research managers as well as a detailed technical section for experts in this domain. (authors)

World-wide French experience in research reactor fuel cycle transportation

International Nuclear Information System (INIS)

Raisonnier, D.

1997-01-01

Since 1963 Transnucleaire has safely performed a large number of national and international transports of radioactive material. Transnucleaire has also designed and supplied suitable packagings for all types of nuclear fuel cycle radioactive material from front-end and back-end products and for power or for research reactors. Transportation of the nuclear fuel material for power reactors is made on a regular and industrial basis. The transportation of material for the research reactor fuel cycle is quite different due to the small quantities involved, the categorisation of material and the numerous places of delivery world-wide. Adapted solutions exist, which require a reactive organisation dealing with all the transportation issues for LEU and HEU products as metal, oxide, fresh fuel elements, spent fuel elements including supply of necessary transport packaging and equipment. This presentation will: - explain the choices made by Transnucleaire and its associates to provide and optimise the corresponding services, - demonstrate the capability to achieve, through reliable partnership, transport operations involving new routes, specific equipment and new political constraints while respecting sophisticated safety and security regulations. (author)

Progress and status of the Integral Fast Reactor (IFR) fuel cycle development

International Nuclear Information System (INIS)

Till, C.E.; Chang, Y.I.

1993-01-01

The Integral Fast Reactor (IFR) fuel cycle holds promise for substantial improvements in economics, diversion-resistance, and waste management. This paper discusses technical features of the IFR fuel cycle, its technical progress, the development status, and the future plans and directions

Progress and status of the Integral Fast Reactor (IFR) fuel cycle development

Energy Technology Data Exchange (ETDEWEB)

Till, C.E.; Chang, Y.I.

1993-03-01

The Integral Fast Reactor (IFR) fuel cycle holds promise for substantial improvements in economics, diversion-resistance, and waste management. This paper discusses technical features of the IFR fuel cycle, its technical progress, the development status, and the future plans and directions.

Progress and status of the Integral Fast Reactor (IFR) fuel cycle development

Energy Technology Data Exchange (ETDEWEB)

Till, C.E.; Chang, Y.I.

1993-01-01

The Integral Fast Reactor (IFR) fuel cycle holds promise for substantial improvements in economics, diversion-resistance, and waste management. This paper discusses technical features of the IFR fuel cycle, its technical progress, the development status, and the future plans and directions.

Fuel-Cycle and Nuclear Material Disposition Issues Associated with High-Temperature Gas Reactors

International Nuclear Information System (INIS)

Shropshire, D.E.; Herring, J.S.

2004-01-01

The objective of this paper is to facilitate a better understanding of the fuel-cycle and nuclear material disposition issues associated with high-temperature gas reactors (HTGRs). This paper reviews the nuclear fuel cycles supporting early and present day gas reactors, and identifies challenges for the advanced fuel cycles and waste management systems supporting the next generation of HTGRs, including the Very High Temperature Reactor, which is under development in the Generation IV Program. The earliest gas-cooled reactors were the carbon dioxide (CO2)-cooled reactors. Historical experience is available from over 1,000 reactor-years of operation from 52 electricity-generating, CO2-cooled reactor plants that were placed in operation worldwide. Following the CO2 reactor development, seven HTGR plants were built and operated. The HTGR came about from the combination of helium coolant and graphite moderator. Helium was used instead of air or CO2 as the coolant. The helium gas has a significant technical base due to the experience gained in the United States from the 40-MWe Peach Bottom and 330-MWe Fort St. Vrain reactors designed by General Atomics. Germany also built and operated the 15-MWe Arbeitsgemeinschaft Versuchsreaktor (AVR) and the 300-MWe Thorium High-Temperature Reactor (THTR) power plants. The AVR, THTR, Peach Bottom and Fort St. Vrain all used fuel containing thorium in various forms (i.e., carbides, oxides, thorium particles) and mixtures with highly enriched uranium. The operational experience gained from these early gas reactors can be applied to the next generation of nuclear power systems. HTGR systems are being developed in South Africa, China, Japan, the United States, and Russia. Elements of the HTGR system evaluated included fuel demands on uranium ore mining and milling, conversion, enrichment services, and fuel fabrication; fuel management in-core; spent fuel characteristics affecting fuel recycling and refabrication, fuel handling, interim

Wastes and waste management in the uranium fuel cycle for light water reactors

International Nuclear Information System (INIS)

Costello, J.M.

1975-08-01

The manufacturing processes in the uranium fuel cycle for light water reactors have been described with particular reference to the chemical and radiological wastes produced and the waste management procedures employed. The problems and possible solutions of ultimate disposal of high activity fission products and transuranium elements from reprocessing of irradiated fuel have been reviewed. Quantities of wastes arising in each stage of the fuel cycle have been summarised. Wastes arising from reactor operation have been described briefly. (author)

Status of the Integral Fast Reactor fuel cycle demonstration and waste management practices

International Nuclear Information System (INIS)

Benedict, R.W.; Goff, K.M.; McFarlane, H.F.

1994-01-01

Over the past few years, Argonne National Laboratory has been preparing for the demonstration of the fuel cycle for the Integral Fast Reactor (IFR), an advanced reactor concept that takes advantage of the properties of metallic fuel and liquid metal cooling to offer significant improvements in reactor safety and operations, fuel-cycle economics, environmental protection, and safeguards. The IFR fuel cycle, which will be demonstrated at Argonne-West in Idaho, employs a pyrometallurgical process using molten salts and liquid metals to recover actinides from spent fuel. The required facility modifications and process equipment for the demonstration are nearing completion. Their status and the results from initial fuel fabrication work, including the waste management aspects, are presented. Additionally, estimated compositions of the various process waste streams have been made, and characterization and treatment methods are being developed. The status of advanced waste processing equipment being designed and fabricated is described

Economics of radioactive material transportation in the light-water reactor nuclear fuel cycle

International Nuclear Information System (INIS)

Dupree, S.A.; O'Malley, L.C.

1980-10-01

This report presents estimates of certain transportation costs, in 1979 dollars, associated with Light-Water Reactor (LWR) once-through and recycle fuel cycles. Shipment of fuel, high-level waste and low-level waste was considered. Costs were estimated for existing or planned transportation systems and for recommended alternate systems, based on the assumption of mature fuel cycles. The annual radioactive material transportation costs required to support a nominal 1000-MW(e) LWR in a once-through cycle in which spent fuel is shipped to terminal storage or disposal were found to be approx. $490,000. Analogous costs for an average reactor operating in a fuel cycle with uranium and plutonim recycle were determined to be approx. $770,000. These results assume that certain recommended design changes will occur in radioactive material shipping systems as a mature fuel cycle evolves

The economics of the fuel cycle (light water reactors)

International Nuclear Information System (INIS)

Lepine, J.

1979-01-01

The economical characteristics of the fuel cycle (of light water reactors) as well as the definition and calculation method for the average updated cost of the kWh are recalled. The evolution followed by the unit prices of the different operations of the cycle, their total cost and the part taken by this cost in the overall cost of nuclear kWh are described. The effects on the cost of fuel of certain hypotheses, operating requirements and additional cost factors are considered [fr

Prompt Neutron Lifetime for the NBSR Reactor

Energy Technology Data Exchange (ETDEWEB)

Hanson, A.L.; Diamond, D.

2012-06-24

In preparation for the proposed conversion of the National Institute of Standards and Technology (NIST) research reactor (NBSR) from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel, certain point kinetics parameters must be calculated. We report here values of the prompt neutron lifetime that have been calculated using three independent methods. All three sets of calculations demonstrate that the prompt neutron lifetime is shorter for the LEU fuel when compared to the HEU fuel and longer for the equilibrium end-of-cycle (EOC) condition when compared to the equilibrium startup (SU) condition for both the HEU and LEU fuels.

Progress and status of the Integral Fast Reactor (IFR) fuel cycle development

International Nuclear Information System (INIS)

Till, C.E.; Chang, Y.I.

1991-01-01

The Integral Fast Reactor (IFR) fuel cycle holds promise for substantial improvements in economics, diversion-resistance, and waste management. This paper discusses technical features of the IFR fuel cycle, its technical progress, the development status, and the future plans and directions. 10 refs

Progress and status of the Integral Fast Reactor (IFR) fuel cycle development

Energy Technology Data Exchange (ETDEWEB)

Till, C.E.; Chang, Y.I.

1991-01-01

The Integral Fast Reactor (IFR) fuel cycle holds promise for substantial improvements in economics, diversion-resistance, and waste management. This paper discusses technical features of the IFR fuel cycle, its technical progress, the development status, and the future plans and directions. 10 refs.

Progress and status of the Integral Fast Reactor (IFR) fuel cycle development

International Nuclear Information System (INIS)

Till, C.E.; Chang, Y.I.

1991-01-01

The Integral Fast Reactor (IFR) fuel cycle holds promise for substantial improvements in economics, diversion-resistance, and waste management. This paper discusses technical features of the IFR fuel cycle, its technical progress, the development status, and the future plans and directions. (author)

Plutonium recycle in PWR reactors (Brazilian Nuclear Program)

International Nuclear Information System (INIS)

Rubini, L.A.

1978-02-01

An evaluation is made of the material requirements of the nuclear fuel cycle with plutonium recycle. It starts from the calculation of a reference reactor and allows the evaluation of demand under two alternatives of nuclear fuel cycle for Pressurized Water Reactors (PWR): without plutonium recycle; and with plutonium recycle. Calculations of the reference reactor have been carried out with the CELL-CORE codes. For plutonium recycle, the concept of uranium and plutonium homogeneous mixture has been adopted, using self-produced plutonium at equilibrium, in order to get minimum neutronic perturbations in the reactor core. The refueling model studied in the reference reactor was the 'out-in' scheme with a constant number of changed fuel elements (approximately 1/3 of the core). Variations in the material requirements were studied considering changes in the installed nuclear capacity of PWR reactors, the capacity factor of these reactors, and the introduction of fast breeders. Recycling plutonium produced inside the system can reach economies of about 5%U 3 O 8 and 6% separative work units if recycle is assumed only after the 5th operation cycle of the thermal reactors. The cumulative amount of fissile plutonium obtained by the Brazilian Nuclear Program of PWR reactors by 1991 should be sufficient for a fast breeder with the same capacity as Angra 2. For the proposed fast breeder programs, the fissile plutonium produced by thermal reactors is sufficient to supply fast breeder initial necessities. Howewer, U 3 O 8 and SWU economy with recycle is not significant when the proposed fast breeder program is considered. (Author) [pt

Secondary cycle design considerations for reduction of reactor transients frequency

International Nuclear Information System (INIS)

Bevilacqua, L.; Leal, M.R.L.V.

1980-01-01

The secondary cycle systems should not be considered of secondary importance to the pressurized water reactor safety. The advanced design and analysis techniques used for components related to nuclear safety are suggested. (E.G.) [pt

Synfuels from fusion: producing hydrogen with the Tandem Mirror Reactor and thermochemical cycles

International Nuclear Information System (INIS)

Werner, R.W.; Ribe, F.L.

1981-01-01

This volume contains the following sections: (1) the Tandem Mirror fusion driver, (2) the Cauldron blanket module, (3) the flowing microsphere, (4) coupling the reactor to the process, (5) the thermochemical cycles, and (6) chemical reactors and process units

Extension of cycle 8 of Angra-1 reactor, optimization of electric power generation reduction

International Nuclear Information System (INIS)

Miranda, Anselmo Ferreira; Moreira, Francisco Jose; Valladares, Gastao Lommez

2000-01-01

The main objective of extending fuel cycle length of Angra-1 reactor, is in fact of that each normal refueling are changed about 40 fuel elements of the reactor core. Considering that these elements do not return for the reactor core, this procedure has became possible a more gain of energy of these elements. The extension consists in, after power generation corresponding to a cycle burnup of 13700 MWD/TMU or 363.3 days, to use the reactivity gain by reduction of power and temperature of primary system for power generation in a low energy patamar

Performance analysis of Brayton cycle system for space power reactor

International Nuclear Information System (INIS)

Li Zhi; Yang Xiaoyong; Zhao Gang; Wang Jie; Zhang Zuoyi

2017-01-01

The closed Brayton cycle system now is the potential choice as the power conversion system for High Temperature Gas-cooled Reactors because of its high energy conversion efficiency and compact configuration. The helium is the best working fluid for the system for its chemical stability and small neutron absorption cross section. However, the Helium has small mole mass and big specific volume, which would lead to larger pipes and heat exchanger. What's more, the big compressor enthalpy rise of helium would also lead to an unacceptably large number of compressor's stage. For space use, it's more important to satisfy the limit of the system's volume and mass, instead of the requirement of the system's thermal capacity. So Noble-Gas binary mixture of helium and xenon is presented as the working fluid for space Brayton cycle. This paper makes a mathematical model for space Brayton cycle system by Fortran language, then analyzes the binary mixture of helium and xenon's properties and effects on power conversion units of the space power reactor, which would be helpful to understand and design the space power reactor. The results show that xenon would lead to a worse system's thermodynamic property, the cycle's efficiency and specific power decrease as xenon's mole fraction increasing. On the other hand, proper amount of xenon would decrease the enthalpy changes in turbomachines, which would be good for turbomachines' design. Another optimization method – the specific power optimization is also proposed to make a comparison. (author)

Feasibility study of self sustaining capability on water cooled thorium reactors for different power reactors

International Nuclear Information System (INIS)

Permana, S.; Takaki, N.; Sekimoto, H.

2007-01-01

Thorium fuel cycle can maintain the sustainable system of the reactor for self sustaining system for future sustainable development in the world. Some characteristics of thorium cycle show some advantages in relation to higher breeding capability, higher performance of burn-up and more proliferation resistant. Several investigations was performed to improve the breeding capability which is essential for maintaining the fissile sustainability during reactor operation in thermal reactor such as Shippingport reactor and molten salt breeder reactor (MSBR) project. The preliminary study of breeding capability on water cooled thorium reactor has been investigated for various power output. The iterative calculation system is employed by coupling the equilibrium fuel cycle burn-up calculation and cell calculation of PIJ module of SRAC2000. In this calculation, 1238 fission products and 129 heavy nuclides are employed. In the cell calculation, 26 heavy metals and 66 fission products and 1 pseudo FP are employed. The employed nuclear data library was JENDL 3.2. The reactor is fueled by 2 33U-Th Oxide and it has used the light water coolant as moderator. Some characteristics such as conversion ratio and void reactivity coefficient performances are evaluated for the systems. The moderator to fuel ratio (MFR) values and average burnups are studied for survey parameter. The parametric survey for different power outputs are employed from 10 MWt to 3000 MWt for evaluating the some characteristics of core size and leakage effects to the spectra profile, required enrichment, breeding capability, fissile inventory condition, and void reactivity coefficient. Different power outputs are employed in order to evaluate its effect to the required enrichment for criticality, breeding capability, void reactivity and fissile inventory accumulation. The obtained value of the conversion ratios is evaluated by using the equilibrium atom composition. The conversion ratio is employed based on the

Synfuels from fusion: producing hydrogen with the Tandem Mirror Reactor and thermochemical cycles

Energy Technology Data Exchange (ETDEWEB)

Werner, R.W.; Ribe, F.L.

1981-01-21

This volume contains the following sections: (1) the Tandem Mirror fusion driver, (2) the Cauldron blanket module, (3) the flowing microsphere, (4) coupling the reactor to the process, (5) the thermochemical cycles, and (6) chemical reactors and process units. (MOW)

Neutronics conceptual design of the innovative research reactor core using uranium molybdenum fuel

International Nuclear Information System (INIS)

Tukiran S; Surian Pinem; Tagor MS; Lily S; Jati Susilo

2012-01-01

The multipurpose of research reactor utilization make many countries build the new research reactor. Trend of this reactor for this moment is multipurpose reactor type with a compact core to get high neutron flux at the low or medium level of power. The research newest. Reactor in Indonesia right now is already 25 year old. Therefore, it is needed to design a new research reactor, called innovative research reactor (IRR) and then as an alternative to replace the old research reactor. The aim of this research is to get the optimal configuration of equilibrium core with the acceptance criteria are minimum thermal neutron flux is 2.5E14 n/cm 2 s at the power level of 20 MW (minimum), length of cycle of more than 40 days, and the most efficient of using fuel in the core. Neutronics design has been performed for new fuel of U-9Mo-AI with various fuel density and reflector. Design calculation has been performed using WIMSD-5B and BATAN-FUEL computer codes. The calculation result of the conceptual design shows four core configurations namely 5x5, 5x7, 6x5 and 6x6. The optimalization result for equilibrium core of innovative research reactor is the 5x5 configuration with 450 gU fuel loading, berilium reflector, maximum thermal neutron flux at reflector is 3.33E14 n/cm 2 sand length of cycle is 57 days is the most optimal of IRR. (author)

Analysis of burnable poison in Ford Nuclear Reactor fuel to extend fuel lifetime. Final report, August 1, 1994--September 29, 1996

Energy Technology Data Exchange (ETDEWEB)

Burn, R.R.; Lee, J.C.

1996-12-01

The objective of the project was to establish the feasibility of extending the lifetime of fuel elements for the Ford Nuclear Reactor (FNR) by replacing current aluminide fuel with silicide fuel comprising a heavier uranium loading but with the same fissile enrichment of 19.5 wt% {sup 235}U. The project has focused on fuel designs where burnable absorbers, in the form of B{sub 4}C, are admixed with uranium silicide in fuel plates so that increases in the control reactivity requirements and peak power density, due to the heavier fuel loading, may be minimized. The authors have developed equilibrium cycle models simulating current full-size aluminide core configurations with 43 {approximately} 45 fuel elements. Adequacy of the overall equilibrium cycle approach has been verified through comparison with recent FNR experience in spent fuel discharge rates and simulation of reactor physics characteristics for two representative cycles. Fuel cycle studies have been performed to compare equilibrium cycle characteristics of silicide fuel designs, including burnable absorbers, with current aluminide fuel. These equilibrium cycle studies have established the feasibility of doubling the fuel element lifetime, with minimal perturbations to the control reactivity requirements and peak power density, by judicious additions of burnable absorbers to silicide fuel. Further study will be required to investigate a more practical silicide fuel design, which incorporates burnable absorbers in side plates of each fuel element rather than uniformly mixes them in fuel plates.

Equilíbrio e ciclos Equilibrium and cycles

Directory of Open Access Journals (Sweden)

Matheus Albergaria de Magalhães

2005-12-01

Full Text Available A agenda de pesquisa relacionada a modelos de ciclos reais de negócios (Real-Business-Cycle models - RBC apresentou um crescimento exponencial desde seu surgimento, no início da década de 1980. Os modelos iniciais do gênero partiam de economias simplificadas, sem imperfeições, buscando explicar as oscilações de curto prazo da economia (ciclos a partir de um arcabouço de equilíbrio geral walrasiano. Atualmente, modelos RBC são amplamente usados na macroeconomia, com uma ênfase especial na análise dos aspectos quantitativos das flutuações econômicas. O objetivo do presente artigo é, portanto, realizar uma resenha da primeira fase da agenda RBC. Apesar dos diversos resultados controversos obtidos, é provável que a principal contribuição dessa agenda seja de cunho metodológico, uma vez que a pesquisa daí advinda afetou profundamente as formas de modelagem macroeconômica vigentes hoje em dia.The research agenda related to Real-Business-Cycle models (RBC has shown a remarkable growth since its beginning, in the eighties. The first models of this kind departed from simple economies without imperfections and tried to explain short-run macroeconomic movements (cycles from Walrasian general-equilibrium settings. Nowadays, RBC models are widely used in Macroeconomics, with a special emphasis on the quantitative aspects of economic fluctuations. The main goal of this paper is to provide a survey of the first stage of the RBC research agenda, since there is not a systematic survey of this literature available in Portuguese. Although there are several controversial results related to RBC models, its main contribution seems to be methodological, since these models had a profound impact on current macroeconomic modeling.

A combined gas cooled nuclear reactor and fuel cell cycle

Science.gov (United States)

Palmer, David J.

Rising oil costs, global warming, national security concerns, economic concerns and escalating energy demands are forcing the engineering communities to explore methods to address these concerns. It is the intention of this thesis to offer a proposal for a novel design of a combined cycle, an advanced nuclear helium reactor/solid oxide fuel cell (SOFC) plant that will help to mitigate some of the above concerns. Moreover, the adoption of this proposal may help to reinvigorate the Nuclear Power industry while providing a practical method to foster the development of a hydrogen economy. Specifically, this thesis concentrates on the importance of the U.S. Nuclear Navy adopting this novel design for its nuclear electric vessels of the future with discussion on efficiency and thermodynamic performance characteristics related to the combined cycle. Thus, the goals and objectives are to develop an innovative combined cycle that provides a solution to the stated concerns and show that it provides superior performance. In order to show performance, it is necessary to develop a rigorous thermodynamic model and computer program to analyze the SOFC in relation with the overall cycle. A large increase in efficiency over the conventional pressurized water reactor cycle is realized. Both sides of the cycle achieve higher efficiencies at partial loads which is extremely important as most naval vessels operate at partial loads as well as the fact that traditional gas turbines operating alone have poor performance at reduced speeds. Furthermore, each side of the cycle provides important benefits to the other side. The high temperature exhaust from the overall exothermic reaction of the fuel cell provides heat for the reheater allowing for an overall increase in power on the nuclear side of the cycle. Likewise, the high temperature helium exiting the nuclear reactor provides a controllable method to stabilize the fuel cell at an optimal temperature band even during transients helping

Progress and status of the integral fast reactor (IFR) fuel cycle development

International Nuclear Information System (INIS)

Till, C.E.; Chang, Y.I.

1993-01-01

The Integral Fast Reactor (IFR) fuel cycle holds promise for substantial improvements in economics, diversion-resistance, and waste management. This paper discusses technical features of the IFR fuel cycle, its technical progress, the development status, and the future plans and directions. The Integral Fast Reactor (IFR) fuel cycle, is based on the use of a metallic fuel alloy (U-Pu-Zr) that permits use of an innovative method for processing of spent fuel. This method, a combination of pyrometallurgical and electrochemical processes, has been termed pyroprocessing. It offers the advantages of a simple, compact processing system and limited volumes of stabilized high-level wastes. This translates to an economically viable system that is likely to receive favorable public response, particularly when combined with the other attributes of the Integral Fast Reactor. Substantial progress has been made in the development of the IFR pyroprocessing method. A comprehensive demonstration of the process will soon begin at the Argonne National Laboratory Idaho site, using spent fuel from the EBR-II reactor. An important advantage of the IFR is its ability to recycle fuel in the process of power generation, extending fuel resources by a considerable amount and assuring the continued viability of nuclear power stations by reducing dependence on external fuel supplies. Pyroprocessing is the means whereby the recycle process is accomplished. It can also be applied to the recovery of fuel constituents from spent fuel generated in the process of operation of conventional light water reactor power plants, offering the means to recover the valuable fuel resources remaining in that material

Transition Analysis of Promising U.S. Future Fuel Cycles Using ORION

International Nuclear Information System (INIS)

Sunny, Eva E.; Worrall, Andrew; Peterson, Joshua L.; Powers, Jeffrey J.; Gehin, Jess C.; Gregg, Robert

2015-01-01

The US Department of Energy Office of Fuel Cycle Technologies performed an evaluation and screening (E&S) study of nuclear fuel cycle options to help prioritize future research and development decisions. Previous work for this E&S study focused on establishing equilibrium conditions for analysis examples of 40 nuclear fuel cycle evaluation groups (EGs) and evaluating their performance according to a set of 22 standardized metrics. Following the E&S study, additional studies are being conducted to assess transitioning from the current US fuel cycle to future fuel cycle options identified by the E&S study as being most promising. These studies help inform decisions on how to effectively achieve full transition, estimate the length of time needed to undergo transition from the current fuel cycle, and evaluate performance of nuclear systems and facilities in place during the transition. These studies also help identify any barriers to achieve transition. Oak Ridge National Laboratory (ORNL) Fuel Cycle Options Campaign team used ORION to analyze the transition pathway from the existing US nuclear fuel cycle—the once-through use of low-enriched-uranium (LEU) fuel in thermal-spectrum light water reactors (LWRs)—to a new fuel cycle with continuous recycling of plutonium and uranium in sodium fast reactors (SFRs). This paper discusses the analysis of the transition from an LWR to an SFR fleet using ORION, highlights the role of lifetime extensions of existing LWRs to aid transition, and discusses how a slight delay in SFR deployment can actually reduce the time to achieve an equilibrium fuel cycle.

Radiotoxicity study of a boiling water reactor core design based on a thorium-uranium fuel concept

International Nuclear Information System (INIS)

Nunez C, A.; Espinosa P, G.

2007-01-01

Full text: The innovative design of a Boiling Water Reactor (BWR) equilibrium core using the thorium-uranium (blanket-seed) concept in the same integrated fuel assembly is presented in this paper. The lattice design uses the thorium conversion capability to 233 U in a BWR spectrum. A core design was developed to achieve an equilibrium cycle of one effective full power year in a standard BWR. A comparison of the toxicity of the spent fuel showed that toxicity is lower in the thorium cycle than other commercial fuels as UO 2 and MOX (uranium and plutonium) in case of the one-through cycle for LWR. (Author)

A strategy analysis of the fast breeder reactor introduction and nuclear fuel cycle systems deployment

International Nuclear Information System (INIS)

Wajima, Tsunetaka; Kawashima, Katsuyuki; Yamashita, Takashi

1996-01-01

A study is made on a strategy analysis of the long term nuclear fuel cycle systems deployment in accordance with the nuclear power growth projection and fast breeder reactor (FBR) introduction. In the analysis, the reprocessed plutonium (Pu) is charged into the reactor in such a way that the reprocessed Pu is not stored outside the reactor, i.e., there is no excess Pu outside the reactor. The analysis characterized the fuel cycle systems, and showed the usefulness of the present method to determine future directions for the FBR introduction and nuclear fuel cycle systems deployment. Concerning an intermediate-term strategy, the time of introduction and required capacities of a second commercial LWR reprocessing plant, Pu-thermal, and the first FBR reprocessing plant deployment are evaluated. A long term strategy analysis shows that the two or three large plants are run in parallel for each fuel cycle facility and that FBR related facilities deal with a markedly large amount of Pu. It is concluded that the early stage introduction of FBRs of significant capacities seems necessary to materialize a consistent total FBR/fuel cycle system where Pu balance becomes feasible through its flexible operation of, for instance, adjusting breeding ratio, in order to keep the transparency of the Pu utilization. (author)

Driver options and burn cycle selection based on power reactor considerations

International Nuclear Information System (INIS)

Ehst, D.A.

1983-01-01

Reactor implications for noninductive current drive are presented based on a number of studies. First, the lower hybrid driver for the STARFIRE reactor is discussed and the disadvantages of this driver are reviewed. Next, the results of an extensive search for a better current driver are presented. A large number of alternatives were compared in a common context, the DEMO reactor, in order to examine their suitability on a standard basis. Finally, the methodology of a study, currently in progress, is described. The goals of this last study are to compare tokamak reactor designs optimized for operation under different burn cycles, in order to assess the actual benefits and costs of pulsed versus steady-state operation. (author)

Driver options and burn-cycle selection based on power-reactor considerations

International Nuclear Information System (INIS)

Ehst, D.A.

1983-04-01

Reactor implications for noninductive current drive are presented based on a number of studies. First, the lower hybrid driver for the STARFIRE reactor is discussed and the disadvantages of this driver are reviewed. Next, the results of an extensive search for a better current driver are presented. A large number of alternatives were compared in a common context, the DEMO reactor, in order to examine their suitability on a standard basis. Finally, the methodology of a study, currently in progress, is described. The goals of this last study are to compare tokamak reactor designs optimized for operation under different burn cycles, in order to assess the actual benefits and costs of pulsed versus steady-state operation

Reduction of repository heat load using advanced fuel cycles

International Nuclear Information System (INIS)

Preston, Jeff; Miller, L.F.

2008-01-01

With the geologic repository at Yucca Mountain already nearing capacity full before opening, advanced fuel cycles that introduce reprocessing, fast reactors, and temporary storage sites have the potential to allow the repository to support the current reactor fleet and future expansion. An uncertainty analysis methodology that combines Monte Carlo distribution sampling, reactor physics data simulation, and neural network interpolation methods enable investigation into the factor reduction of heat capacity by using the hybrid fuel cycle. Using a Super PRISM fast reactor with a conversion ratio of 0.75, burn ups reach up to 200 MWd/t that decrease the plutonium inventory by about 5 metric tons every 12 years. Using the long burn up allows the footprint of 1 single core loading of FR fuel to have an integral decay heat of about 2.5x10 5 MW*yr over a 1500 year period that replaces the footprint of about 6 full core loadings of LWR fuel for the number of years required to fuel the FR, which have an integral decay heat of about.3 MW*yr for the same time integral. This results in an increase of a factor of 4 in repository support capacity from implementing a single fast reactor in an equilibrium cycle. (authors)

A prospective study of power cycles based on the expected sodium fast reactor parameters

International Nuclear Information System (INIS)

Herranz, L. E.; Linares, J. I.; Moratilla, B. Y.; Perez, G. D.

2010-01-01

One of the main issues that has not been solved yet in the frame of Sodium Fast Reactors (SFR) is to choose the most appropriate power conversion system. This paper explores the performance of different power cycles, from traditional to innovative layouts trying to find the optimized solution. Based on the expected reactor parameters (i.e., inlet and outlet coolant temperatures, 395 deg.C and 545 deg.C, respectively), a subcritical Rankine similar to those of fossil power plant cycles has been proposed as a reference layout. Then, alternative layouts based on innovative Rankine and Brayton cycles have been investigated. Two Rankine supercritical layouts have been modeled and analyzed: one of them, adopted from the Supercritical Water Reactor of GIV (one reheater, nine pre-heaters and one moisture separator) and the other similar to some fossil plants (two reheaters, nine pre-heaters with no moisture separator). Simple Brayton cycle configurations based on Helium has been also studied. Several layouts have been modeled to study the effects of: inter-cooling between compression stages, absence of an intermediate loop and coupling of an organic Rankine cycle (ORC). (authors)

Computational analysis of supercritical CO2 Brayton cycle power conversion system for fusion reactor

International Nuclear Information System (INIS)

Halimi, Burhanuddin; Suh, Kune Y.

2012-01-01

Highlights: ► Computational analysis of S-CO 2 Brayton cycle power conversion system. ► Validation of numerical model with literature data. ► Recompression S-CO 2 Brayton cycle thermal efficiency of 42.44%. ► Reheating concept to enhance the cycle thermal efficiency. ► Higher efficiency achieved by the proposed concept. - Abstract: The Optimized Supercritical Cycle Analysis (OSCA) code is being developed to analyze the design of a supercritical carbon dioxide (S-CO 2 ) driven Brayton cycle for a fusion reactor as part of the Modular Optimal Balance Integral System (MOBIS). This system is based on a recompression Brayton cycle. S-CO 2 is adopted as the working fluid for MOBIS because of its easy availability, high density and low chemical reactivity. The reheating concept is introduced to enhance the cycle thermal efficiency. The helium-cooled lithium lead model AB of DEMO fusion reactor is used as reference in this paper.

Study of potential of nuclear waste transmutation and safety characteristics of an hybrid system: sub critical accelerator reactor; Etude du potentiel de transmutation et des caracteristiques de surete d`un systeme hybride: accelerateur reacteur sous critique

Energy Technology Data Exchange (ETDEWEB)

Tchistiakov, A

1998-04-01

The study of potential of nuclear waste transmutation for the new reactor systems - hybrid reactors - was the object of this work. Global review of different projects is presented. The basic physical parameters definitions, as neutron surplus and relative importance of external source neutrons, are introduced and explained. For these parameters, numerical values are obtained. The advantage in neutron surplus of fast system is noted. Equilibrium model and corresponding toxicities of different isotopes nd nuclear cycles are presented. Numerical analysis for equilibrium model converge validation are performed also. The study of neutron consumption by `transmutable` Long-Lived Fission Products (Tc, I and Cs) show the possibility of their incineration in dedicated fast hybrid reactors. Equilibrium model shown the influence of reprocessing losses level to cycle toxicity level. Relations between specific fuel inventories (mass normalised by power unit) for thermal and fast spectra are examined. The differences are relatively small. Finally, few hybrid reactor concepts with different objects were analysed. These studies confirm that in frameworks of certain Nuclear Energy scenarios the fast hybrid systems can reduce significantly the radio-toxicity of fuel cycle. Preliminary analyses of sub-critical reactor behaviour show big potential of this reactor type in `Transient of Power` kind of accident, even if more detailed study is necessary. (author)

Improving Fuel Cycle Design and Safety Characteristics of a Gas Cooled Fast Reactor

International Nuclear Information System (INIS)

Rooijen, W.F.G. van

2006-01-01

The Gas Cooled Fast Reactor (GCFR)is one of the Generation IV reactor concepts. This concept specifically targets sustainability of nuclear power generation. In nuclear reactors fertile material is converted to fissile fuel. If the neutrons inducing fission are highly energetic, the opportunity exists to convert more than one fertile nucleus per fission, thereby effectively breeding new nuclear fuel. Reactors operating on this principle are called ‘Fast Breeder Reactor’. Since natural uranium contains 99.3%of the fertile isotope 238 U, breeding increases the energy harvested from the nuclear fuel. If nuclear energy is to play an important role as a source of energy in the future, fast breeder reactors are essential for breeding nuclear fuel. Fast neutrons are also more efficient to destruct heavy (Minor Actinide, MA) isotopes, such as Np, Am and Cm isotopes, which dominate the long-term radioactivity of nuclear waste. So the waste life-time can be shortened if the MA nuclei are destroyed. An important prerequisite of sustainable nuclear energy is the closed fuel cycle, where only fission products are discharged to a final repository, and all Heavy Metal (HM) are recycled. The reactor should breed just enough fissile material to allow refueling of the same reactor, adding only fertile material to the recycled material. Other key design choices are highly efficient power conversion using a direct cycle gas turbine, and better safety through the use of helium, a chemically inert coolant which cannot have phase changes in the reactor core. Because the envisaged core temperatures and operating conditions are similar to thermal-spectrum High Temperature Reactor (HTR) concepts, the research for this thesis initially focused on a design based on existing HTR fuel technology: coated particle fuel, assembled into fuel assemblies. It was found that such a fuel concept could not meet the Generation IV criteria set for GCFR: self-breeding is difficult, the temperature

Transition analysis of promising U.S. future fuel cycles using ORION - 5114

International Nuclear Information System (INIS)

Sunny, E.; Worrall, A.; Peterson, J.; Powers, J.; Gehin, J.

2015-01-01

The US Department of Energy Office of Fuel Cycle Technologies performed an evaluation and screening (E/S) study of nuclear fuel cycle options to help prioritize future research and development decisions. Previous work for this E/S study focused on establishing equilibrium conditions for analysis examples of 40 nuclear fuel cycle evaluation groups and evaluating their performance according to a set of 22 standardized metrics. Following the E/S study, additional studies are being conducted to assess transition period from the current US fuel cycle to future fuel cycle options identified by the E/S study as being most promising. These studies help inform decisions on how to effectively achieve full transition, estimate the length of time needed to undergo transition from the current fuel cycle, and evaluate performance of nuclear systems and facilities in place during the transition. These studies also help identify any barriers to achieve transition. Oak Ridge National Laboratory (ORNL) Fuel Cycle Options Campaign team used ORION to analyze the transition pathway from the existing US nuclear fuel cycle - the once-through use of low-enriched-uranium (LEU) fuel in thermal-spectrum light water reactors (LWRs) - to a new fuel cycle with continuous recycling of plutonium and uranium in sodium fast reactors (SFRs). This paper discusses the analysis of the transition from an LWR to an SFR fleet using ORION, highlights the role of lifetime extensions of existing LWRs to aid transition, and discusses how a slight delay in SFR deployment can actually reduce the time to achieve an equilibrium fuel cycle. (authors)

An integrated multicriteria decision-making approach for evaluating nuclear fuel cycle systems for long-term sustainability on the basis of an equilibrium model: Technique for order of preference by similarity to ideal solution, preference ranking organization method for enrichment evaluation, and multiattribute utility theory combined with analytic hierarchy process

Energy Technology Data Exchange (ETDEWEB)

Yoon, Sae Rom [Dept of Quantum Energy Chemical Engineering, Korea University of Science and Technology (KUST), Daejeon (Korea, Republic of); Choi, Sung Yeol [Ulsan National Institute of Science and Technology, Ulju (Korea, Republic of); Ko, Wonil [Nonproliferation System Development Division, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2017-02-15

The focus on the issues surrounding spent nuclear fuel and lifetime extension of old nuclear power plants continues to grow nowadays. A transparent decision-making process to identify the best suitable nuclear fuel cycle (NFC) is considered to be the key task in the current situation. Through this study, an attempt is made to develop an equilibrium model for the NFC to calculate the material flows based on 1 TWh of electricity production, and to perform integrated multicriteria decision-making method analyses via the analytic hierarchy process technique for order of preference by similarity to ideal solution, preference ranking organization method for enrichment evaluation, and multiattribute utility theory methods. This comparative study is aimed at screening and ranking the three selected NFC options against five aspects: sustainability, environmental friendliness, economics, proliferation resistance, and technical feasibility. The selected fuel cycle options include pressurized water reactor (PWR) once-through cycle, PWR mixed oxide cycle, or pyroprocessing sodium-cooled fast reactor cycle. A sensitivity analysis was performed to prove the robustness of the results and explore the influence of criteria on the obtained ranking. As a result of the comparative analysis, the pyroprocessing sodium-cooled fast reactor cycle is determined to be the most competitive option among the NFC scenarios.

An integrated multicriteria decision-making approach for evaluating nuclear fuel cycle systems for long-term sustainability on the basis of an equilibrium model: Technique for order of preference by similarity to ideal solution, preference ranking organization method for enrichment evaluation, and multiattribute utility theory combined with analytic hierarchy process

International Nuclear Information System (INIS)

Yoon, Sae Rom; Choi, Sung Yeol; Ko, Wonil

2017-01-01

The focus on the issues surrounding spent nuclear fuel and lifetime extension of old nuclear power plants continues to grow nowadays. A transparent decision-making process to identify the best suitable nuclear fuel cycle (NFC) is considered to be the key task in the current situation. Through this study, an attempt is made to develop an equilibrium model for the NFC to calculate the material flows based on 1 TWh of electricity production, and to perform integrated multicriteria decision-making method analyses via the analytic hierarchy process technique for order of preference by similarity to ideal solution, preference ranking organization method for enrichment evaluation, and multiattribute utility theory methods. This comparative study is aimed at screening and ranking the three selected NFC options against five aspects: sustainability, environmental friendliness, economics, proliferation resistance, and technical feasibility. The selected fuel cycle options include pressurized water reactor (PWR) once-through cycle, PWR mixed oxide cycle, or pyroprocessing sodium-cooled fast reactor cycle. A sensitivity analysis was performed to prove the robustness of the results and explore the influence of criteria on the obtained ranking. As a result of the comparative analysis, the pyroprocessing sodium-cooled fast reactor cycle is determined to be the most competitive option among the NFC scenarios

An Integrated Multicriteria Decision-Making Approach for Evaluating Nuclear Fuel Cycle Systems for Long-term Sustainability on the Basis of an Equilibrium Model: Technique for Order of Preference by Similarity to Ideal Solution, Preference Ranking Organization Method for Enrichment Evaluation, and Multiattribute Utility Theory Combined with Analytic Hierarchy Process

Directory of Open Access Journals (Sweden)

Saerom Yoon

2017-02-01

Full Text Available The focus on the issues surrounding spent nuclear fuel and lifetime extension of old nuclear power plants continues to grow nowadays. A transparent decision-making process to identify the best suitable nuclear fuel cycle (NFC is considered to be the key task in the current situation. Through this study, an attempt is made to develop an equilibrium model for the NFC to calculate the material flows based on 1 TWh of electricity production, and to perform integrated multicriteria decision-making method analyses via the analytic hierarchy process technique for order of preference by similarity to ideal solution, preference ranking organization method for enrichment evaluation, and multiattribute utility theory methods. This comparative study is aimed at screening and ranking the three selected NFC options against five aspects: sustainability, environmental friendliness, economics, proliferation resistance, and technical feasibility. The selected fuel cycle options include pressurized water reactor (PWR once-through cycle, PWR mixed oxide cycle, or pyroprocessing sodium-cooled fast reactor cycle. A sensitivity analysis was performed to prove the robustness of the results and explore the influence of criteria on the obtained ranking. As a result of the comparative analysis, the pyroprocessing sodium-cooled fast reactor cycle is determined to be the most competitive option among the NFC scenarios.

Neutronic Core Performance of CAREM-25 Reactor

International Nuclear Information System (INIS)

Villarino, Eduardo; Hergenreder, Daniel; Matzkin, S

2000-01-01

The actual design state of core of CAREM-25 reactor is presented.It is shown that the core design complains with the safety and operation established requirements.It is analyzed the behavior of the reactor safety and control systems (single failure of the fast shut down system, single failure of the shut down system, single failure of the second shut down system, reactivity worth of the adjust and control system in normal operation and hot shut down, reactivity worth of the adjust and control system and the scheme of movement of the control rod during the operation cycle).It is shown the burnup profile of fuel elements with the proposed scheme of refueling and the burnup and power density distribution at different moments of the operation cycle.The power peaking factor of the equilibrium core is 2.56, the minimum DNBR is 1.90 and its average is 2.09 during the operation cycle

Steam generators in indirect-cycle water-cooled reactors

International Nuclear Information System (INIS)

Fajeau, M.

1976-01-01

In the indirect cycle water-cooled nuclear reactors, the steam generators are placed between the primary circuit and the turbine. They act both as an energy transmitter and as a leaktigh barrier against fission or corrosion products. Their study is thus very important from a performance and reliability point of view. Two main types are presented here: the U-tube and the once-through steam generators [fr

The slightly-enriched spectral shift control reactor

Energy Technology Data Exchange (ETDEWEB)

Martin, W.R.; Lee, J.C.; Larsen, E.W. (Michigan Univ., Ann Arbor, MI (United States). Dept. of Nuclear Engineering); Edlund, M.C. (Virginia Polytechnic Inst. and State Univ., Blacksburg, VA (United States). Dept. of Mechanical and Nuclear Engineering)

1991-11-01

An advanced converter reactor design utilizing mechanical spectral shift control rods in a conventional pressurized water reactor configuration is under investigation. The design is based on the principle that a harder spectrum during the early part of the fuel cycle will result in large neutron captures in fertile {sup 238}U, which can then be burned in situ in a softer spectrum later in the cycle. Preliminary design calculations performed during FY 89 showed that the slightly-enriched spectral shift reactor design offers the benefit of substantially increased fuel resource utilization with the proven safety characteristics of the pressurized water reactor technology retained. Optimization of the fuel design and development of fuel management strategies were carried out in FY 90, along with effort to develop and validate neutronic methodology for tight-lattice configurations with hard spectra. During FY 91, the final year of the grant, the final Slightly-Enriched Spectral Shift Reactor (SESSR) design was determined, and reference design analyses were performed for the assemblies as well as the global core configuration, both at the beginning of cycle (BOC) and with depletion. The final SESSR design results in approximately a 20% increase in the utilization of uranium resources, based on equilibrium fuel cycle analyses. Acceptable pin power peaking is obtained with the final core design, with assembly peaking factors equal to less than 1.04 for spectral shift control rods both inserted and withdrawn, and global peaking factors at BOC predicted to be 1.4. In addition, a negative Moderation Temperature Coefficient (MTC) is maintained for BOC, which is difficult to achieve with conventional advanced converter designs based on a closed fuel cycle. The SESSR design avoids the need for burnable poison absorber, although they could be added if desired to increase the cycle length while maintaining a negative MTC.

The slightly-enriched spectral shift control reactor

International Nuclear Information System (INIS)

Martin, W.R.; Lee, J.C.; Larsen, E.W.; Edlund, M.C.

1991-11-01

An advanced converter reactor design utilizing mechanical spectral shift control rods in a conventional pressurized water reactor configuration is under investigation. The design is based on the principle that a harder spectrum during the early part of the fuel cycle will result in large neutron captures in fertile 238 U, which can then be burned in situ in a softer spectrum later in the cycle. Preliminary design calculations performed during FY 89 showed that the slightly-enriched spectral shift reactor design offers the benefit of substantially increased fuel resource utilization with the proven safety characteristics of the pressurized water reactor technology retained. Optimization of the fuel design and development of fuel management strategies were carried out in FY 90, along with effort to develop and validate neutronic methodology for tight-lattice configurations with hard spectra. During FY 91, the final year of the grant, the final Slightly-Enriched Spectral Shift Reactor (SESSR) design was determined, and reference design analyses were performed for the assemblies as well as the global core configuration, both at the beginning of cycle (BOC) and with depletion. The final SESSR design results in approximately a 20% increase in the utilization of uranium resources, based on equilibrium fuel cycle analyses. Acceptable pin power peaking is obtained with the final core design, with assembly peaking factors equal to less than 1.04 for spectral shift control rods both inserted and withdrawn, and global peaking factors at BOC predicted to be 1.4. In addition, a negative Moderation Temperature Coefficient (MTC) is maintained for BOC, which is difficult to achieve with conventional advanced converter designs based on a closed fuel cycle. The SESSR design avoids the need for burnable poison absorber, although they could be added if desired to increase the cycle length while maintaining a negative MTC

A sensitivity analysis and assessment on the reactivity, economics and resorce implications of reactor systems and cycles with respect to uncertainity in nuclear data and other reactor parameters

International Nuclear Information System (INIS)

Quan, B.L.

1980-01-01

A general sensitivity analysis system for analyzing the effects of uncertainity in nuclear data and reactor parameters on fuel cycle economics, resources and physics has been developed. The sensitivity analysis has been performed on various reactor systems and cycles such as the thorium cycles, plutonium cycles, CANDU reactor fuel cycles and alternate once-through LWR cycles such as the 18 month cycle. Sensitivity coefficients were generated for a variety of materials pertinent to the LWR fuel cycle using a series of fast running codes developed for this purpose and running on a local PDP-15 computer. Their relative order of importance were assessed and the reasons explaining this difference were examined. This work is a result of EPRI project in determining the data needs for the LWR industry and should be valuable in identifying areas in which data improvements are worthwhile

One-group constant libraries for nuclear equilibrium state

Energy Technology Data Exchange (ETDEWEB)

Mizutani, Akihiko; Sekimoto, Hiroshi [Tokyo Inst. of Tech. (Japan). Research Lab. for Nuclear Reactors

1997-03-01

One-group constant libraries for the nuclear equilibrium state were generated for both liquid sodium cooled MOX fuel type fast reactor and PWR type thermal reactor with Equilibrium Cell Iterative Calculation System (ECICS) using JENDL-3.2, -3, -2 and ENDF/B-VI nuclear data libraries. ECICS produced one-group constant sets for 129 heavy metal nuclides and 1238 fission products. (author)

Evaluation of the radial design of fuel cells in an operation cycle of a BWR reactor

International Nuclear Information System (INIS)

Gonzalez C, J.; Martin del Campo M, C.

2003-01-01

This work is continuation of one previous in the one that the application of the optimization technique called Tabu search to the radial design of fuel cells of boiling water reactors (BWR, Boiling Water Reactor) is presented. The objective function used in the optimization process only include neutron parameters (k-infinite and peak of radial power) considering the cell at infinite media. It was obtained to reduce the cell average enrichment completing the characteristics of reactivity of an original cell. The objective of the present work is to validate the objective function that was used for the radial design of the fuel cell (test cell), analyzing the operation of a one cycle of the reactor in which fuels have been fresh recharged that contain an axial area with the nuclear database of the cell designed instead of the original cell. For it is simulated it with Cm-Presto the cycle 10 of the reactor operation of the Unit 1 of the Nuclear Power station of Laguna Verde (U1-CNLV). For the cycle evaluation its were applied so much the simulation with the Haling strategy, as the simulation of the one cycle with control rod patterns and they were evaluated the energy generation and several power limits and reactivity that are used as design parameters in fuel reloads of BWR reactors. The results at level of an operation cycle of the reactor, show that the objective function used in the optimization and radial design of the cell is adequate and that it can induce to one good use of the fuel. (Author)

Design and analysis of helium Brayton power cycles for HiPER reactor

Energy Technology Data Exchange (ETDEWEB)

Sánchez, Consuelo, E-mail: csanchez@ind.uned.es [Dpto. Ingeniería Energética UNED, Madrid (Spain); Juárez, Rafael; Sanz, Javier [Dpto. Ingeniería Energética UNED, Madrid (Spain); Instituto de Fusión Nuclear/UPM, Madrid (Spain); Perlado, Manuel [Instituto de Fusión Nuclear/UPM, Madrid (Spain)

2013-10-15

Highlights: ► A helium Brayton cycle has been designed integrating the two energy sources of HiPER. ► The Brayton cycle has intercooling stages and a recovery process. ► The low temperature of HiPER heat sources results in low cycle efficiency (35.2%). ► Two inter-cooling stages and a reheating process increases efficiency to over 37%. ► Helium Brayton cycles are to be considered as candidates for HiPER power cycles. -- Abstract: Helium Brayton cycles have been studied as power cycles for both fission and fusion reactors obtaining high thermal efficiency. This paper studies several technological schemes of helium Brayton cycles applied for the HiPER reactor proposal. Since HiPER integrates technologies available at short term, its working conditions results in a very low maximum temperature of the energy sources, something that limits the thermal performance of the cycle. The aim of this work is to analyze the potential of the helium Brayton cycles as power cycles for HiPER. Several helium Brayton cycle configurations have been investigated with the purpose of raising the cycle thermal efficiency under the working conditions of HiPER. The effects of inter-cooling and reheating have specifically been studied. Sensitivity analyses of the key cycle parameters and component performances on the maximum thermal efficiency have also been carried out. The addition of several inter-cooling stages in a helium Brayton cycle has allowed obtaining a maximum thermal efficiency of over 36%, and the inclusion of a reheating process may also yield an added increase of nearly 1 percentage point to reach 37%. These results confirm that helium Brayton cycles are to be considered among the power cycle candidates for HiPER.

Design and analysis of helium Brayton power cycles for HiPER reactor

International Nuclear Information System (INIS)

Sánchez, Consuelo; Juárez, Rafael; Sanz, Javier; Perlado, Manuel

2013-01-01

Highlights: ► A helium Brayton cycle has been designed integrating the two energy sources of HiPER. ► The Brayton cycle has intercooling stages and a recovery process. ► The low temperature of HiPER heat sources results in low cycle efficiency (35.2%). ► Two inter-cooling stages and a reheating process increases efficiency to over 37%. ► Helium Brayton cycles are to be considered as candidates for HiPER power cycles. -- Abstract: Helium Brayton cycles have been studied as power cycles for both fission and fusion reactors obtaining high thermal efficiency. This paper studies several technological schemes of helium Brayton cycles applied for the HiPER reactor proposal. Since HiPER integrates technologies available at short term, its working conditions results in a very low maximum temperature of the energy sources, something that limits the thermal performance of the cycle. The aim of this work is to analyze the potential of the helium Brayton cycles as power cycles for HiPER. Several helium Brayton cycle configurations have been investigated with the purpose of raising the cycle thermal efficiency under the working conditions of HiPER. The effects of inter-cooling and reheating have specifically been studied. Sensitivity analyses of the key cycle parameters and component performances on the maximum thermal efficiency have also been carried out. The addition of several inter-cooling stages in a helium Brayton cycle has allowed obtaining a maximum thermal efficiency of over 36%, and the inclusion of a reheating process may also yield an added increase of nearly 1 percentage point to reach 37%. These results confirm that helium Brayton cycles are to be considered among the power cycle candidates for HiPER

Development of the fuel-cycle costs in nuclear power stations with light-water reactors

International Nuclear Information System (INIS)

Brosch, R.; Moraw, G.; Musil, G.; Schneeberger, M.

1976-01-01

The authors investigate the fuel-cycle costs in nuclear power stations with light-water reactors in the Federal Republic of Germany in the years 1966 to 1976. They determine the effect of the price development for the individual components of the nuclear fuel cycle on the fuel-cycle costs averaged over the whole power station life. Here account is taken also of inflation rates and the change in the DM/US $ parity. In addition they give the percentage apportionment of the fuel-cycle costs. The authors show that real fuel-cycle costs for nuclear power stations with light-water reactors in the Federal Republic of Germany have risen by 11% between 1966 and 1976. This contradicts the often repeated reproach that fuel costs in nuclear power stations are rising very steeply and are no longer competitive. (orig.) [de

The benefits of a fast reactor closed fuel cycle in the UK

International Nuclear Information System (INIS)

Gregg, R.; Hesketh, K.

2013-01-01

The work has shown that starting a fast reactor closed fuel cycle in the UK, requires virtually all of Britain's existing and future PWR spent fuel to be reprocessed, in order to obtain the plutonium needed. The existing UK Pu stockpile is sufficient to initially support only a modest SFR 'closed' fleet assuming spent fuel can be reprocessed shortly after discharge (i.e. after two years cooling). For a substantial fast reactor fleet, most Pu will have to originate from reprocessing future spent PWR fuel. Therefore, the maximum fast reactor fleet size will be limited by the preceding PWR fleet size, so scenarios involving fast reactors still require significant quantities of uranium ore indirectly. However, once a fast reactor fuel cycle has been established, the very substantial quantities of uranium tails in the UK would ensure there is sufficient material for several centuries. Both the short and long term impacts on a repository have been considered in this work. Over the short term, the decay heat emanating from the HLW and spent fuel will limit the density of waste within a repository. For scenarios involving fast reactors, the only significant heat bearing actinide content will be present in the final cores, resulting in a 50% overall reduction in decay energy deposited within the repository when compared with an equivalent open fuel cycle. Over the longer term, radiological dose becomes more important. Total radiotoxicity (normalised by electricity generated) is lower for scenarios with Pu recycle after 2000 years. Scenarios involving fast reactors have the lowest radiotoxicity since the quantities of certain actinides (Np, Pu and Am) eventually stabilise. However, total radiotoxicity as a measure of radiological risk does not account for differences in radionuclide mobility once in repository. Radiological dose is dominated by a small number of fission products so is therefore not affected significantly by reactor type or recycling strategy (since the

Chemical vapour deposition of silicon under reduced pressure in a hot-wall reactor: Equilibrium and kinetics

International Nuclear Information System (INIS)

Langlais, F.; Hottier, F.; Cadoret, R.

1982-01-01

Silicon chemical vapour deposition (SiH 2 Cl 2 /H 2 system), under reduced pressure conditions, in a hot-wall reactor, is presented. The vapour phase composition is assessed by evaluating two distinct equilibria. The homogeneous equilibrium , which assumes that the vapour phase is not in equilibrium with solid silicon, is thought to give an adequate description of the vapour phase in the case of low pressure, high gas velocities, good temperature homogeneity conditions. A comparison with heterogeneous equilibrium enables us to calculate the supersaturation so evidencing a highly irreversible growth system. The experimental determination of the growth rates reveals two distinct temperature ranges: below 1000 0 C, polycrystalline films are usually obtained with a thermally activated growth rate (+40 kcal mole -1 ) and a reaction order, with respect to the predominant species SiCl 2 , close to one; above 1000 0 C, the films are always monocrystalline and their growth rate exhibits a much lower or even negative activation energy, the reaction order in SiCl 2 remaining about one. (orig.)

Analysis of transition to fuel cycle system with continuous recycling in fast and thermal reactors - 5060

International Nuclear Information System (INIS)

Passereini, S.; Feng, B.; Fei, T.; Kim, T.K.; Taiwo, T.A.; Brown, N.R.; Cuadra, A.

2015-01-01

A recent Evaluation and Screening study of nuclear fuel cycle options identified a few groups of options as most promising. One of these most promising Evaluation Groups (EGs) is characterized by the continuous recycling of uranium (U) and transuranics (TRU) with natural uranium feed in both fast and thermal critical reactors. This evaluation group, designated as EG30, is represented by an example fuel cycle option that employs a two-technology, two-stage fuel cycle system. The first stage involves the continuous recycling of co-extracted U/TRU in Sodium-cooled Fast Reactors (SFRs) with metallic fuel and breeding ratio greater than 1. The second stage involves the use of the surplus TRU in Mixed Oxide (MOX) fuel in Pressurized Water Reactors that are MOX-capable (MOX-PWRs). This paper presents and discusses preliminary fuel cycle analysis results from the fuel cycle codes VISION and DYMOND for the transition to this fuel cycle option from the current once-through cycle in the United States (U.S.) that consists of Light Water Reactors (LWRs) that only use conventional UO 2 fuel. The analyses in this paper are applicable for a constant 100 GWe capacity, roughly the size of the U.S. nuclear fleet. Two main strategies for the transition to EG30 were analyzed: 1) deploying both SFRs and MOX-PWRs in parallel or 2) deploying them in series with the SFR fleet first. With an estimated retirement schedule for the existing LWRs, an assumed reactor lifetime of 60 years, and no growth, the nuclear system fully transitions to the new fuel cycle within 100 years for both strategies without SFR fuel shortages. Compared to the once-through cycle, transition to the SFR/MOX-PWR fleet with continuous recycle was shown to offer significant reductions in uranium consumption and waste disposal requirements. In addition, these initial calculations revealed a few notable modeling and strategy questions regarding how recycled resources are allocated, reactors that can switch between

BN800: The advanced sodium cooled fast reactor plant based on close fuel cycle

International Nuclear Information System (INIS)

Wu Xingman

2011-01-01

As one of the advanced countries with actually fastest reactor technology, Russia has always taken a leading role in the forefront of the development of fast reactor technology. After successful operation of BN600 fast reactor nuclear power station with a capacity of six hundred thousand kilowatts of electric power for nearly 30 years, and after a few decades of several design optimization improved and completed on its basis, it is finally decided to build Unit 4 of Beloyarsk nuclear power station (BN800 fast reactor power station). The BN800 fast reactor nuclear power station is considered to be the project of the world's most advanced fast reactor nuclear power being put into implementation. The fast reactor technology in China has been developed for decades. With the Chinese pilot fast reactor to be put into operation soon, the Chinese model fast reactor power station has been put on the agenda. Meanwhile, the closed fuel cycle development strategy with fast reactor as key aspect has given rise to the concern of experts and decision-making level in relevant areas. Based on the experiences accumulated in many years in dealing the Sino-Russian cooperation in fast reactor technology, with reference to the latest Russian published and authoritative literatures regarding BN800 fast reactor nuclear power station, the author compiled this article into a comprehensive introduction for reference by leaders and experts dealing in the related fields of nuclear fuel cycle strategy and fast reactor technology development researches, etc. (authors)

Assessment of the dry process fuel sodium-cooled fast reactors

Energy Technology Data Exchange (ETDEWEB)

Roh, Gyu Hong; Choi, Hang Bok

2004-04-01

The feasibility of using dry-processed oxide fuel in a Sodium-cooled Fast Reactor (SFR) was analyzed for the equilibrium fuel cycle of two reference cores: Hybrid BN-600 benchmark core with a enlarged lattice pitch and modified BN-600 core. The dry process technology assumed in this study based on the molten-salt process, which was developed by Russian scientists for recycling oxide fuels. The core calculation was performed by the REBUS-3 code and the reactor characteristics such as the transuranic enrichment, breeding ratio, peak linear power, burnup reactivity swing, etc. were calculated for the equilibrium core under a fixed fuel management scheme. The results showed that a self-sustainable breakeven core was achievable without blanket fuels when the fuel volume fraction was {approx}50% and most of the fission products were removed.

Assessment of the dry process fuel sodium-cooled fast reactors

International Nuclear Information System (INIS)

Roh, Gyu Hong; Choi, Hang Bok

2004-04-01

The feasibility of using dry-processed oxide fuel in a Sodium-cooled Fast Reactor (SFR) was analyzed for the equilibrium fuel cycle of two reference cores: Hybrid BN-600 benchmark core with a enlarged lattice pitch and modified BN-600 core. The dry process technology assumed in this study based on the molten-salt process, which was developed by Russian scientists for recycling oxide fuels. The core calculation was performed by the REBUS-3 code and the reactor characteristics such as the transuranic enrichment, breeding ratio, peak linear power, burnup reactivity swing, etc. were calculated for the equilibrium core under a fixed fuel management scheme. The results showed that a self-sustainable breakeven core was achievable without blanket fuels when the fuel volume fraction was ∼50% and most of the fission products were removed

Feasibility study on commercialized fast reactor cycle systems. (1) Current status of the phase-II study

International Nuclear Information System (INIS)

Sagayama, Yutaka

2005-01-01

A feasibility study on commercialized fast reactors including related nuclear fuel cycle systems has been started from Japanese fiscal year 1999 by a Japanese joint project team of Japan Nuclear Cycle Development Institute and the Japan Atomic Power Company. This project aims at elucidating prominent fast reactor cycle systems that will respond to various needs of society in the future, together with economic competitiveness as future electricity supply systems. Challenging technology goals for the fast reactor cycle systems were defined in five targets: safety, economic competitiveness, reduction of environmental burden, efficient utilization of nuclear fuel resources and enhancement of nuclear non-proliferation. As the results of the feasibility study up to now, it is confirmed as the interim results that the combination of sodium-cooled fast reactors with oxide fuels, advanced aqueous reprocessing and simplified pellet fuel fabrication is highly suited to the development targets. The cost would be highly reduced by the adoption of innovative technologies, which feasibility is relatively clear and some R and D issues are now under progress. (author)

Technical Meeting on Fast Reactors and Related Fuel Cycle Facilities with Improved Economic Characteristics. Working Material

International Nuclear Information System (INIS)

2013-01-01

In recent years, engineering oriented work, rather than basic research and development (R&D), has led to significant progress in improving the economics of innovative fast reactors and associated fuel cycle facilities, while maintaining and even enhancing the safety features of these systems. Optimization of plant size and layout, more compact designs, reduction of the amount of plant materials and the building volumes, higher operating temperatures to attain higher generating efficiencies, improvement of load factor, extended core lifetimes, high fuel burnup, etc. are good examples of achievements to date that have improved the economics of fast neutron systems. The IAEA, through its Technical Working Group on Fast Reactors (TWG-FR) and Technical Working Group on Nuclear Fuel Cycle Options and Spent Fuel Management (TWG-NFCO), devotes many of its initiatives to encouraging technical cooperation and promoting common research and technology development projects among Member States with fast reactor and advanced fuel cycle development programmes, with the general aim of catalysing and accelerating technology advances in these fields. In particular the theme of fast reactor deployment, scenarios and economics has been largely debated during the recent IAEA International Conference on Fast Reactors and Related Fuel Cycles: Safe Technologies and Sustainable Scenarios, held in Paris in March 2013. Several papers presented at this conference discussed the economics of fast reactors from different national and regional perspectives, including business cases, investment scenarios, funding mechanisms and design options that offer significant capital and energy production cost reductions. This Technical Meeting on Fast Reactors and Related Fuel Cycle Facilities with Improved Economic Characteristics addresses Member States’ expressed need for information exchange in the field, with the aim of identifying the main open issues and launching possible initiatives to help and

Integrated gasification gas combined cycle plant with membrane reactors: Technological and economical analysis

International Nuclear Information System (INIS)

Amelio, Mario; Morrone, Pietropaolo; Gallucci, Fausto; Basile, Angelo

2007-01-01

In the present work, the capture and storage of carbon dioxide from the fossil fuel power plant have been considered. The main objective was to analyze the thermodynamic performances and the technological aspects of two integrated gasification gas combined cycle plants (IGCC), as well as to give a forecast of the investment costs for the plants and the resulting energy consumptions. The first plant considered is an IGCC* plant (integrated gasification gas combined cycle plant with traditional shift reactors) characterized by the traditional water gas shift reactors and a CO 2 physical adsorption system followed by the power section. The second one is an IGCC M plant (integrated gasification gas combined cycle plant with membrane reactor) where the coal thermal input is the same as the first one, but the traditional shift reactors and the physical adsorption unit are replaced by catalytic palladium membrane reactors (CMR). In the present work, a mono-dimensional computational model of the membrane reactor was proposed to simulate and evaluate the capability of the IGCC M plant to capture carbon dioxide. The energetic performances, efficiency and net power of the IGCC* and IGCC M plants were, thus, compared, assuming as standard a traditional IGCC plant without carbon dioxide capture. The economical aspects of the three plants were compared through an economical analysis. Since the IGCC* and IGCC M plants have additional costs related to the capture and disposal of the carbon dioxide, a Carbon Tax (adopted in some countries like Sweden) proportional to the number of kilograms of carbon dioxide released in the environment was assumed. According to the economical analysis, the IGCC M plant proved to be more convenient than the IGCC* one

Burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1

Directory of Open Access Journals (Sweden)

Muhammad Atta

2011-01-01

Full Text Available The burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1, reference operating core, has been carried out utilizing standard computer codes WIMS/D4, CITATION, and RELAP5/MOD3.4. Reactor codes WIMS/D4 and CITATION have been used for the calculations of neutronic parameters including peaking factors and power profiles at different burn-up considering a xenon free core and also the equilibrium xenon values. RELAP5/MOD3.4 code was utilized for the determination of peak fuel centerline, clad and coolant temperatures to ensure the safety of the reactor throughout the cycle. The calculations reveal that the reactor is safe and no nucleate boiling will commence at any part of the core throughout the cycle and that the safety margin increases with burnup as peaking factors decrease.

Fuel cycle model and the cost of a recycling thorium in the CANDU reactor

International Nuclear Information System (INIS)

Choi, Hangbok; Park, Chang Je

2005-01-01

The dry process fuel technology has a high proliferation-resistance, which allows applications not only to the existing but also to the future nuclear fuel cycle systems. In this study, the homogeneous ThO 2 -UO 2 recycling fuel cycle in a Canada deuterium uranium (CANDU) reactor was assessed for a fuel cycle cost evaluation. A series of parametric calculations were performed for the uranium fraction, enrichment of the initial uranium fuel, and the fission product removal rated of the recycled fuel. The fuel cycle cost was estimated by the levelized lifetime cost model provided by the Organization for Economic Cooperation and Development/Nuclear Energy Agency. Though it is feasible to recycle the homogeneous ThO 2 -UO 2 fuel in the CANDU reactor from the viewpoint of a mass balance, the recycling fuel cycle cost is much higher than the conventional natural uranium fuel cycle cost for most cases due to the high fuel fabrication cost. (author)

Hydrogen production with fully integrated fuel cycle gas and vapour core reactors

International Nuclear Information System (INIS)

Anghaie, S.; Smith, B.

2004-01-01

This paper presents results of a conceptual design study involving gas and vapour core reactors (G/VCR) with a combined scheme to generate hydrogen and power. The hydrogen production schemes include high temperature electrolysis as well as two dominant thermochemical hydrogen production processes. Thermochemical hydrogen production processes considered in this study included the calcium-bromine process and the sulphur-iodine processes. G/VCR systems are externally reflected and moderated nuclear energy systems fuelled by stable uranium compounds in gaseous or vapour phase that are usually operated at temperatures above 1500 K. A gas core reactor with a condensable fuel such as uranium tetrafluoride (UF 4 ) or a mixture of UF 4 and other metallic fluorides (BeF 2 , LiF, KF, etc.) is commonly known as a vapour core reactor (VCR). The single most relevant and unique feature of gas/vapour core reactors is that the functions of fuel and coolant are combined into one. The reactor outlet temperature is not constrained by solid fuel-cladding temperature limits. The maximum fuel/working fluid temperature in G/VCR is only constrained by the reactor vessel material limits, which is far less restrictive than the fuel clad. Therefore, G/VCRs can potentially provide the highest reactor and cycle temperature among all existing or proposed fission reactor designs. Gas and vapour fuel reactors feature very low fuel inventory and fully integrated fuel cycle that provide for exceptional sustainability and safety characteristics. With respect to fuel utilisation, there is no fuel burn-up limit for gas core reactors due to continuous recycling of the fuel. Owing to the flexibility in nuclear design characteristics of cavity reactors, a wide range of conversion ratio from completely burner to breeder is achievable. The continuous recycling of fuel in G/VCR systems allow for complete burning of actinides without removing and reprocessing of the fuel. The only waste products at the back

Monte Carlo reactor calculation with substantially reduced number of cycles

International Nuclear Information System (INIS)

Lee, M. J.; Joo, H. G.; Lee, D.; Smith, K.

2012-01-01

A new Monte Carlo (MC) eigenvalue calculation scheme that substantially reduces the number of cycles is introduced with the aid of coarse mesh finite difference (CMFD) formulation. First, it is confirmed in terms of pin power errors that using extremely many particles resulting in short active cycles is beneficial even in the conventional MC scheme although wasted operations in inactive cycles cannot be reduced with more particles. A CMFD-assisted MC scheme is introduced as an effort to reduce the number of inactive cycles and the fast convergence behavior and reduced inter-cycle effect of the CMFD assisted MC calculation is investigated in detail. As a practical means of providing a good initial fission source distribution, an assembly based few-group condensation and homogenization scheme is introduced and it is shown that efficient MC eigenvalue calculations with fewer than 20 total cycles (including inactive cycles) are possible for large power reactor problems. (authors)

Fuel cycle options for light water reactors in Germany

International Nuclear Information System (INIS)

Broecking, D.; Mester, W.

1999-01-01

In Germany 19 nuclear power plants with an electrical output of 22 GWe are in operation. Annually about 450 t of spent fuel are unloaded from the reactors. Currently most of the spent fuel elements are shipped to France and the United Kingdom for reprocessing according to contracts which have been signed since the late 70es. By the amendment of the Atomic Energy Act in 1994 the previous priority for reprocessing of spent nuclear fuel was substituted by a legal equivalency of the reprocessing and direct disposal option. As a consequence some utilities take into consideration the direct disposal of their spent fuel for economical reasons. The separated plutonium will be recycled as MOX fuel in light water reactors. About 30 tons of fissile plutonium will be available to German utilities for recycling by the year 2000. Twelve German reactors are already licensed for the use of MOX fuel, five others have applied for MOX use. Eight reactors are currently using MOX fuel or used it in the past. The spent fuel elements which shall be disposed of without reprocessing will be stored in two interim dry storage facilities at Gorleben and Ahaus. The storage capacities are 3800 and 4200 tHM, respectively. The Gorleben salt dome is currently investigated to prove its suitability as a repository for high level radioactive waste, either in a vitrified form or as conditioned spent fuel. The future development of the nuclear fuel cycle and radioactive waste management depends on the future role of nuclear energy in Germany. According to estimations of the German utilities no additional nuclear power plants are needed in the near future. Around the middle of the next decade it will have to be decided whether existing plants should be substituted by new ones. For the foreseeable time German utilities are interested in a highly flexible approach to the nuclear fuel cycle and waste management keeping open both spent fuel management options: the closed fuel cycle and direct disposal of

Small particle bed reactors: Sensitivity to Brayton cycle parameters

Science.gov (United States)

Coiner, John R.; Short, Barry J.

Relatively simple particle bed reactor (PBR) algorithms were developed for optimizing low power closed Brayton cycle (CBC) systems. These algorithms allow the system designer to understand the relationship among key system parameters as well as the sensitivity of the PBR size and mass (a major system component) to variations in these parameters. Thus, system optimization can be achieved.

Transition period fuel cycle from current to next generation reactors for Japan

International Nuclear Information System (INIS)

Yamashita, Junichi; Fukasawa, Tetsuo; Hoshino, Kuniyoshi; Kawamura, Fumio; Shiina, Kouji; Sasahira, Akira

2007-01-01

Long-term energy security and global warming prevention can be achieved by a sustainable electricity supply with next generation fast breeder reactors (FBRs). Current light water reactors (LWRs) will be replaced by FBRs and FBR cycle will be established in the future considering the limited amount of uranium (U) resource. The introduction of FBRs requires plutonium (Pu) recovered from LWR spent fuel. The authors propose advanced system named Flexible Fuel Cycle Initiative (FFCI)' which can supply enough Pu and hold no surplus Pu, can respond flexibly the future technical and social uncertainties, and can achieve an economical FBR cycle. FFCI can simplify the 2nd LWR reprocessing facility for Japan (after Rokkasho Reprocessing Plant) which only carries out U removal from LWR spent fuel. Residual 'Recycle Material' is, according to FBRs introduction status, immediately treated in the FBR reprocessing to fabricate FBR fuel or temporarily stored for the utilization in FBRs at necessary timing. FFCI has high flexibility by having several options for future uncertainties by the introduction of Recycle Material as a buffer material between LWR and FBR cycles. (author)

DIRECT-CYCLE, BOILING-WATER NUCLEAR REACTOR

Science.gov (United States)

Harrer, J.M.; Fromm, L.W. Jr.; Kolba, V.M.

1962-08-14

A direct-cycle boiling-water nuclear reactor is described that employs a closed vessel and a plurality of fuel assemblies, each comprising an outer tube closed at its lower end, an inner tube, fuel rods in the space between the tubes and within the inner tube. A body of water lying within the pressure vessel and outside the fuel assemblies is converted to saturated steam, which enters each fuel assembly at the top and is converted to superheated steam in the fuel assembly while it is passing therethrough first downward through the space between the inner and outer tubes of the fuel assembly and then upward through the inner tube. (AEC)

Physics studies of weapons plutonium disposition in the Integral Fast Reactor closed fuel cycle

International Nuclear Information System (INIS)

Hill, R.N.; Wade, D.C.; Liaw, J.R.; Fujita, E.K.

1995-01-01

The core performance impact of weapons plutonium introduction into the Integral Fast Reactor (IFR) closed fuel cycle is investigated by comparing three disposition scenarios: a power production mode, a moderate destruction mode, and a maximum destruction mode, all at a constant heat rating of 840 MW(thermal). For each scenario, two fuel cycle models are evaluated: cores using weapons material as the sole source of transuranics in a once-through mode and recycle cores using weapons material only as required for a makeup feed. In addition, the impact of alternative feeds (recycled light water reactor or liquid-metal reactor transuranics) on burner core performance is assessed. Calculated results include mass flows, detailed isotopic distributions, neutronic performance characteristics, and reactivity feedback coefficients. In general, it is shown that weapons plutonium does not have an adverse effect on IFR core performance characteristics; also, favorable performance can be maintained for a wide variety of feed materials and fuel cycle strategies

An integral metallic-fueled and lead-cooled reactor concept for the 4th generation reactor

International Nuclear Information System (INIS)

Santos, Adimir dos; Nascimento, Jamil Alves do

2002-01-01

An Integral Lead Reactor (ILR) concept is proposed for the 4th generation reactor to be used in the future. The ILR is loaded with metallic fuel and cooled by lead. It was evaluated in the 300-1500 MWe power range with the Japanese Fast Set 2 cross sections library. This set was tested against several fast benchmarks and the criticality uncertainty was found to be 0.51 %Δk. The reactor is started with U-Zr and changes to the U-TRU-Zr-RE fuel in a stepwise way. In the equilibrium cycle, the burnup reactivity is less than β eff for a core of the order of 300 MWe, pin diameter of 10.4 mm and a pin-pinch to diameter ratio of 1.308. The lead void reactivity is negative for reactor power less than 750 MWe. There is a need to improve the nuclear data for the major actinides. (author)

An integral metallic-fueled and lead-cooled reactor concept for the 4th generation reactor

International Nuclear Information System (INIS)

Santos, A. dos; Nascimento, J.A. do

2002-01-01

An Integral Lead Reactor (ILR) concept is proposed for the 4th generation reactor to be used in the future. The ILR is loaded with metallic fuel and cooled by lead. It was evaluated in the 300-1500 MWe power range with the Japanese Fast Set 2 cross sections library. This set was tested against several fast benchmarks and the criticality uncertainty was found to be 0.51 % Δk. The reactor is started with U-Zr and changes to the U-TRU-Zr-RE fuel in a stepwise way. In the equilibrium cycle, the burnup reactivity is less than β eff for a core of the order of 300 MWe, pin diameter of 10.4 mm and a pin-pitch to diameter ratio of 1.308. The lead void reactivity is negative for reactor power less than 750 MWe. There is a need to improve the nuclear data for the major actinides. (author)

Fuel and fuel cycles with high burnup for WWER reactors

International Nuclear Information System (INIS)

Chernushev, V.; Sokolov, F.

2002-01-01

The paper discusses the status and trends in development of nuclear fuel and fuel cycles for WWER reactors. Parameters and main stages of implementation of new fuel cycles will be presented. At present, these new fuel cycles are offered to NPPs. Development of new fuel and fuel cycles based on the following principles: profiling fuel enrichment in a cross section of fuel assemblies; increase of average fuel enrichment in fuel assemblies; use of refuelling schemes with lower neutron leakage ('in-in-out'); use of integrated fuel gadolinium-based burnable absorber (for a five-year fuel cycle); increase of fuel burnup in fuel assemblies; improving the neutron balance by using structural materials with low neutron absorption; use of zirconium alloy claddings which are highly resistant to irradiation and corrosion. The paper also presents the results of fuel operation. (author)

Non-Equilibrium Thermodynamics of Self-Replicating Protocells

DEFF Research Database (Denmark)

Fellermann, Harold; Corominas-Murtra, Bernat; Hansen, Per Lyngs

2018-01-01

We provide a non-equilibrium thermodynamic description of the life-cycle of a droplet based, chemically feasible, system of protocells. By coupling the protocells metabolic kinetics with its thermodynamics, we demonstrate how the system can be driven out of equilibrium to ensure protocell growth...... and replication. This coupling allows us to derive the equations of evolution and to rigorously demonstrate how growth and replication life-cycle can be understood as a non-equilibrium thermodynamic cycle. The process does not appeal to genetic information or inheritance, and is based only on non......-equilibrium physics considerations. Our non-equilibrium thermodynamic description of simple, yet realistic, processes of protocell growth and replication, represents an advance in our physical understanding of a central biological phenomenon both in connection to the origin of life and for modern biology....

Core Designs and Economic Analyses of Homogeneous Thoria-Urania Fuel in Light Water Reactors

International Nuclear Information System (INIS)

Saglam, Mehmet; Sapyta, Joe J.; Spetz, Stewart W.; Hassler, Lawrence A.

2004-01-01

The objective is to develop equilibrium fuel cycle designs for a typical pressurized water reactor (PWR) loaded with homogeneously mixed uranium-thorium dioxide (ThO 2 -UO 2 ) fuel and compare those designs with more conventional UO 2 designs.The fuel cycle analyses indicate that ThO 2 -UO 2 fuel cycles are technically feasible in modern PWRs. Both power peaking and soluble boron concentrations tend to be lower than in conventional UO 2 fuel cycles, and the burnable poison requirements are less.However, the additional costs associated with the use of homogeneous ThO 2 -UO 2 fuel in a PWR are significant, and extrapolation of the results gives no indication that further increases in burnup will make thoria-urania fuel economically competitive with the current UO 2 fuel used in light water reactors

Feasibility study on commercialized fast reactor cycle systems. Phase II final report

International Nuclear Information System (INIS)

Ieda, Yoshiaki; Uchikawa, Sadao; Okubo, Tsutomu; Ono, Kiyoshi; Kato, Atsushi; Kurisaka, Kenichi; Sakamoto, Yoshihiko; Sato, Kazujiro; Sato, Koji; Chikazawa, Yoshitaka; Nakai, Ryodai; Nakabayashi, Hiroki; Nakamura, Hirofumi; Namekawa, Takashi; Niwa, Hajime; Nomura, Kazunori; Hayashi, Hideyuki; Hayafune, Hiroki; Hirao, Kazunori; Mizuno, Tomoyasu; Muramatsu, Toshiharu; Ando, Masato; Ono, Katsumi; Ogata, Takanari; Kubo, Shigenobu; Kotake, Shoji; Sagayama, Yutaka; Takakuma, Katsuyuki; Tanaka, Toshihiko; Namba, Takashi; Fujii, Sumio; Muramatsu, Kazuyoshi

2006-06-01

A joint project team of Japan Atomic Energy Agency and the Japan Atomic Power Company (as the representative of the electric utilities) started the feasibility study on commercialized fast reactor cycle systems (F/S) in July 1999 in cooperation with Central Research Institute of Electric Power Industry and vendors. On the major premise of safety assurance, F/S aims to present an appropriate picture of commercialization of fast reactor (FR) cycle system which has economic competitiveness with light water reactor cycle systems and other electricity base load systems, and to establish FR cycle technologies for the future major energy supply. In the period from Japanese fiscal year (JFY) 1999 to 2000, the phase-I of F/S was carried out to screen our representative FR, reprocessing and fuel fabrication technologies. In the phase-II (JFY 2001-2005), the design study of FR cycle concepts, the development of significant technologies necessary for the feasibility evaluation, and the confirmation of key technical issues were performed to clarify the promising candidate concepts toward the commercialization. In this final phase-II report clarified the most promising concept, the R and D plan until around 2015, and the key issues for the commercialization. Based on the comprehensive evaluation in F/S, the combination of the sodium-cooled FR with MOX fuel core, the advanced-aqueous reprocessing process and the simplified-pelletizing fuel fabrication process was recommended as the mainline choice for the most promising concept. The concept exceeds in technical advancement, and the conformity to the development targets was higher compared with that of the others. Alternative technologies are prepared to be decrease the development risk of innovative technologies in the mainline choice. (author)

V.S.O.P.-computer code system for reactor physics and fuel cycle simulation

International Nuclear Information System (INIS)

Teuchert, E.; Hansen, U.; Haas, K.A.

1980-03-01

V.S.O.P. (Very Superior Old Programs) is a system of codes linked together for the simulation of reactor life histories. It comprises neutron cross section libraries and processing routines, repeated neutron spectrum evaluation, 2-D diffusion calculation based on neutron flux synthesis with depletion and shutdown features, incore and out-of-pile fuel management, fuel cycle cost analysis, and thermal hydraulics (at present restricted to Pebble Bed HTRs). Various techniques have been employed to accelerate the iterative processes and to optimize the internal data transfer. A limitation of the storage requirement to 360 K-bites is achieved by an overlay structure. The code system has been used extensively for comparison studies of reactors, their fuel cycles, and related detailed features. Beside its use in research and development work for the high temperature reactor the system has been applied successfully to LWR and Heavy Water Reactors. (orig.) [de

Introduction of long term cycle of reactor operation

International Nuclear Information System (INIS)

Aoyati, M.; Tanaka, T.

2004-01-01

Introduction of long term cycle of LWR reactor operation at NPP in Japan is considered, and problems of technical, legislative and economical character, increase of power coefficient are discussed. More long term operation period provides decreasing frequency of periodic inspections and reduction of personnel radiation doses. Reliability of fuel, energetic equipment, mechanisms and devices must be taken into account for the decision of technical problems. Consumptions for electric power generation are studied [ru

Nuclear proliferation and civilian nuclear power: report of the Nonproliferation Alternative Systems Assessment Program. Volume IX. Reactor and fuel cycle descriptions

Energy Technology Data Exchange (ETDEWEB)

1979-12-01

The Nonproliferation Alternative Systems Assessment Program (NASAP) has characterized and assessed various reactor/fuel-cycle systems. Volume IX provides, in summary form, the technical descriptions of the reactor/fuel-cycle systems studied. This includes the status of the system technology, as well as a discussion of the safety, environmental, and licensing needs from a technical perspective. This information was then used in developing the research, development, and demonstration (RD and D) program, including its cost and time frame, to advance the existing technology to the level needed for commercial use. Wherever possible, the cost data are given as ranges to reflect the uncertainties in the estimates. Volume IX is divided into three sections: Chapter 1, Reactor Systems; Chapter 2, Fuel-Cycle Systems; and the Appendixes. Chapter 1 contains the characterizations of the following 12 reactor types: light-water reactor; heavy-water reactor; water-cooled breeder reactor; high-temperature gas-cooled reactor; gas-cooled fast reactor; liquid-metal fast breeder reactor; spectral-shift-controlled reactor; accelerator-driven reactor; molten-salt reactor; gaseous-core reactor; tokamak fusion-fisson hybrid reactor; and fast mixed-spectrum reactor. Chapter 2 contains similar information developed for fuel-cycle facilities in the following categories: mining and milling; conversion and enrichment; fuel fabrication; spent fuel reprocessing; waste handling and disposal; and transportation of nuclear materials.

Nuclear proliferation and civilian nuclear power: report of the Nonproliferation Alternative Systems Assessment Program. Volume IX. Reactor and fuel cycle descriptions

International Nuclear Information System (INIS)

1979-12-01

The Nonproliferation Alternative Systems Assessment Program (NASAP) has characterized and assessed various reactor/fuel-cycle systems. Volume IX provides, in summary form, the technical descriptions of the reactor/fuel-cycle systems studied. This includes the status of the system technology, as well as a discussion of the safety, environmental, and licensing needs from a technical perspective. This information was then used in developing the research, development, and demonstration (RD and D) program, including its cost and time frame, to advance the existing technology to the level needed for commercial use. Wherever possible, the cost data are given as ranges to reflect the uncertainties in the estimates. Volume IX is divided into three sections: Chapter 1, Reactor Systems; Chapter 2, Fuel-Cycle Systems; and the Appendixes. Chapter 1 contains the characterizations of the following 12 reactor types: light-water reactor; heavy-water reactor; water-cooled breeder reactor; high-temperature gas-cooled reactor; gas-cooled fast reactor; liquid-metal fast breeder reactor; spectral-shift-controlled reactor; accelerator-driven reactor; molten-salt reactor; gaseous-core reactor; tokamak fusion-fisson hybrid reactor; and fast mixed-spectrum reactor. Chapter 2 contains similar information developed for fuel-cycle facilities in the following categories: mining and milling; conversion and enrichment; fuel fabrication; spent fuel reprocessing; waste handling and disposal; and transportation of nuclear materials

Control rod studies for alternative fuel cycles in the GA 1160 MW(e) high temperature reactor

Energy Technology Data Exchange (ETDEWEB)

Neef, H. J.

1975-06-15

The control system, which is investigated in this paper for both the low enriched uranium high enriched uranium/thorium fuel cycles, has been developed to control the General Atomics (GA) thorium fuel cycle 1160 MW(e) reactor. It has been shown in this investigation that its effectiveness in the low enriched and subsequent thorium cycle switch-over reactor is equivalent to the effectiveness in the thorium cycle. The shutdown margin in the low enriched core is even higher compared to the thorium core, mainly due to the presence of Pa-233 in the thorium cycle. As long as the fuel cycle for the thorium cycle is not closed with the recycling of U-233, the low enriched cycle will offer an attractive alternative. It was found that the GA 1160 MW(e) control system has enough built-in control rod capacity to accommodate the low enriched uranium cycle and to perform a later switch-over to a thorium-based fuel cycle.

Application of S-CO_2 Cycle for Small Modular Reactor coupled with Desalination System

International Nuclear Information System (INIS)

Lee, Won Woong; Bae, Seong Jun; Lee, Jeong Ik

2016-01-01

The Korean small modular reactor, SMART (System-integrated Modular Advanced ReacTor, 100MWe), is designed to achieve enhanced safety and improved economics through reliable passive safety systems, a system simplification and component modularization. SMART can generate electricity and provide water by seawater desalination. However, due to the desalination aspect of SMART, the total amount of net electricity generation is decreased from 100MWe to 90MWe. The authors suggest in this presentation that the reduction of electricity generation can be replenished by applying S-CO_2 power cycle technology. The S-CO_2 Brayton cycle, which is recently receiving significant attention as the next generation power conversion system, has some benefits such as high cycle efficiency, simple configuration, compactness and so on. In this study, the cycle performance analysis of the S-CO_2 cycles for SMART with desalination system is conducted. The simple recuperated S-CO_2 cycle is revised for coupling with desalination system. The three revised layout are proposed for the cycle performance comparison. In this results of the 3rd revised layout, the cycle efficiency reached 37.8%, which is higher than the efficiency of current SMART with the conventional power conversion system 30%

Nuclear Systems Enhanced Performance Program, Maintenance Cycle Extension in Advanced Light Water Reactor Design

Energy Technology Data Exchange (ETDEWEB)

Professor Neill Todreas

2001-10-01

A renewed interest in new nuclear power generation in the US has spurred interest in developing advanced reactors with features which will address the public's concerns regarding nuclear generation. However, it is economic performance which will dictate whether any new orders for these plants will materialize. Economic performance is, to a great extent, improved by maximizing the time that the plant is on-line generating electricity relative to the time spent off-line conducting maintenance and refueling. Indeed, the strategy for the advanced light water reactor plant IRIS (International Reactor, Innovative and Secure) is to utilize an eight year operating cycle. This report describes a formalized strategy to address, during the design phase, the maintenance-related barriers to an extended operating cycle. The top-level objective of this investigation was to develop a methodology for injecting component and system maintainability issues into the reactor plant design process to overcome these barriers. A primary goal was to demonstrate the applicability and utility of the methodology in the context of the IRIS design. The first step in meeting the top-level objective was to determine the types of operating cycle length barriers that the IRIS design team is likely to face. Evaluation of previously identified regulatory and investment protection surveillance program barriers preventing a candidate operating PWR from achieving an extended (48 month) cycle was conducted in the context of the IRIS design. From this analysis, 54 known IRIS operating cycle length barriers were identified. The resolution methodology was applied to each of these barriers to generate design solution alternatives for consideration in the IRIS design. The methodology developed has been demonstrated to narrow the design space to feasible design solutions which enable a desired operating cycle length, yet is general enough to have broad applicability. Feedback from the IRIS design team

Reloading pattern optimization of VVER-1000 reactors in transient cycles using genetic algorithm

International Nuclear Information System (INIS)

Rahmani, Yashar

2017-01-01

Highlights: • The genetic algorithm (GA) and the innovative weighting factors method were used. • The coupling of WIMSD5-B and CITATION-LDI2 neutronic codes with the thermohydraulic WERL code was employed. • Optimization of reloading patterns was carried out in two states. • First an arrangement with satisfactory excess reactivity and the flattest power distribution was searched. • Second, it is tried to obtain an arrangement with satisfactory safety threshold and the maximum K_e_f_f. - Abstract: The present paper proposes application of the genetic algorithm (GA) and the innovative weighting factor method to optimize the reloading pattern of Bushehr VVER-1000 reactor in the second cycle. To estimate the composition of fuel assemblies remaining from the first cycle and precisely calculate the objective parameters of each reloading pattern in the second cycle, coupling of WIMSD5-B and CITATION-LDI2 codes in the neutronic section and the WERL code in the thermo-hydraulic section was employed. Optimization of the reloading patterns was carried out in two states. To meet the mentioned objective, with application of the weighting factor method in the first state, the type and quantity of the loadable fresh assemblies were determined to enable the reactor core to maintain the core criticality over the entire cycle length. Afterwards, the genetic algorithm was used to optimize the reloading pattern of the reactor to obtain an arrangement with flat radial power distribution. In the second state, the optimization algorithm was free to select the type and number of fresh fuel assemblies to be able to search for an arrangement with the maximum effective multiplication factor and the safe power peaking factor. In addition, in order to ensure the safety and desirability of the proposed patterns in both states, a time-dependent examination of the thermo-neutronic behavior of the reactor core was carried out during the second cycle. With consideration of the new

The electronuclear cycle: from fission to new reactor systems

International Nuclear Information System (INIS)

Belier, G.; Cugnon, J.; Lapoux, V.; Liatard, E.; Porquet, Marie-Genevieve; Rudolf, G.

2006-09-01

The Joliot Curie School trains each year, and since 1981, PhD students, post-Doctorates and researchers on scientific breakthroughs performed in a topic related to nuclear physics, in a broad range. These proceedings brings together the 11 lectures given at the 2006 session of Joliot Curie School on the topic of the electronuclear cycle: - Fission: from phenomenology to theory (Berger, J.F.); - Physics of nuclear reactors (Baeten, P.); - Data modeling and evaluation (Bauge, E.; Hilaire, S.); - Measurement of cross sections of interest for minor actinides incineration (Jurado, B.); - Spallation data and modelling for hybrid reactors (Boudard, A.); - Nuclear wastes: overview (Billard, I.); - Long living nuclear wastes transmutation processes and feasibility (Varaine, F.); - Hybrid reactors: recent advances for a demonstrator (Billebaud, A.); - Systems of the future and strategy (David, S.); - Non-nuclear energies (Nifenecker, H.); - Fundamental physics with ultracold neutrons (Protasov, K). The last section is a compilation of abstracts of presentations given by Young searchers' (Young searchers' seminars)

Cooling of nuclear power stations with high temperature reactors and helium turbine cycles

International Nuclear Information System (INIS)

Foerster, S.; Hewing, G.

1977-01-01

On nuclear power stations with high temperature reactors and helium turbine cycles (HTR-single circuits) the residual heat from the energy conversion process in the primary and intermediate coolers is removed from cycled gas, helium. Water, which is circulated for safety reasons through a closed circuit, is used for cooling. The primary and intermediate coolers as well as other cooling equipment of the power plant are installed within the reactor building. The heat from the helium turbine cycle is removed to the environment most effectively by natural draught cooling towers. In this way a net plant efficiency of about 40% is attainable. The low quantities of residual heat thereby produced and the high (in comparison with power stations with steam turbine cycles) cooling agent pressure and cooling water reheat pressure in the circulating coolers enable an economically favourable design of the overall 'cold end' to be expected. In the so-called unit range it is possible to make do with one or two cooling towers. Known techniques and existing operating experience can be used for these dry cooling towers. After-heat removal reactor shutdown is effected by a separate, redundant cooling system with forced air dry coolers. The heat from the cooling process at such locations in the power station is removed to the environment either by a forced air dry cooling installation or by a wet cooling system. (orig.) [de

Evaluation of Indirect Combined Cycle in Very High Temperature Gas--Cooled Reactor

International Nuclear Information System (INIS)

Chang Oh; Robert Barner; Cliff Davis; Steven Sherman; Paul Pickard

2006-01-01

The U.S. Department of Energy and Idaho National Laboratory are developing a very high temperature reactor to serve as a demonstration of state-of-the-art nuclear technology. The purpose of the demonstration is twofold: (a) efficient, low-cost energy generation and (b) hydrogen production. Although a next-generation plant could be developed as a single-purpose facility, early designs are expected to be dual purpose, as assumed here. A dual-purpose design with a combined cycle of a Brayton top cycle and a bottom Rankine cycle was investigated. An intermediate heat transport loop for transporting heat to a hydrogen production plant was used. Helium, CO2, and a helium-nitrogen mixture were studied to determine the best working fluid in terms of the cycle efficiency. The relative component sizes were estimated for the different working fluids to provide an indication of the relative capital costs. The relative size of the turbomachinery was measured by comparing the power input/output of the component. For heat exchangers the volume was computed and compared. Parametric studies away from the baseline values of the cycle were performed to determine the effects of varying conditions in the cycle. This gives some insight into the sensitivity of the cycle to various operating conditions as well as trade-offs between efficiency and component size. Parametric studies were carried out on reactor outlet temperature, mass flow, pressure, and turbine cooling

Power ramping, cycling and load following behaviour of water reactor fuel

International Nuclear Information System (INIS)

1988-05-01

The present meeting was scheduled by the International Atomic Energy Agency upon proposal of the Members of the International Working Group on Water Reactor Fuel Performance and Technology. Sixty-three participants representing 15 countries and one international organization attended the meeting. Twenty papers were presented during three technical sessions, followed by panel discussions which allowed to formulate the conclusions of the meeting and recommendations to the Agency. The objective of this Technical Committee Meeting is to review the ''State-of-the-Art'', make critical comments and recommendations with the aim of improving fuel reliability and assure integrity of the cladding and core materials when subjected to ramping and cycling sequences. The Meeting was organized in three sessions: Session 1. ''Mechanical Behaviour and Fission Gas Release'' (7 papers); Session 2. ''Power Ramping and Power Cycling Demonstration Programmes in Research Reactors'' (5 papers); Session 3. ''Fuel Behaviour in Power Reactors'' (9 papers). Between the sessions, the session chairmen, together with the speakers, prepared and presented reports with summary, conclusions and recommendations of the individual sessions. These reports are added to this summary report. A separate abstract was prepared for each of these 21 presentations. Refs, figs and tabs

Exploratory Design of a Reactor/Fuel Cycle Using Spent Nuclear Fuel Without Conventional Reprocessing - 13579

International Nuclear Information System (INIS)

Bertch, Timothy C.; Schleicher, Robert W.; Rawls, John D.

2013-01-01

General Atomics has started design of a waste to energy nuclear reactor (EM2) that can use light water reactor (LWR) spent nuclear fuel (SNF). This effort addresses two problems: using an advanced small reactor with long core life to reduce nuclear energy overnight cost and providing a disposal path for LWR SNF. LWR SNF is re-fabricated into new EM2 fuel using a dry voloxidation process modeled on AIROX/ OREOX processes which remove some of the fission products but no heavy metals. By not removing all of the fission products the fuel remains self-protecting. By not separating heavy metals, the process remains proliferation resistant. Implementation of Energy Multiplier Module (EM2) fuel cycle will provide low cost nuclear energy while providing a long term LWR SNF disposition path which is important for LWR waste confidence. With LWR waste confidence recent impacts on reactor licensing, an alternate disposition path is highly relevant. Centered on a reactor operating at 250 MWe, the compact electricity generating system design maximizes site flexibility with truck transport of all system components and available dry cooling features that removes the need to be located near a body of water. A high temperature system using helium coolant, electricity is efficiently produced using an asynchronous high-speed gas turbine while the LWR SNF is converted to fission products. Reactor design features such as vented fuel and silicon carbide cladding support reactor operation for decades between refueling, with improved fuel utilization. Beyond the reactor, the fuel cycle is designed so that subsequent generations of EM2 reactor fuel will use the previous EM2 discharge, providing its own waste confidence plus eliminating the need for enrichment after the first generation. Additional LWR SNF is added at each re-fabrication to replace the removed fission products. The fuel cycle uses a dry voloxidation process for both the initial LWR SNF re-fabrication and later for EM2

Exploratory Design of a Reactor/Fuel Cycle Using Spent Nuclear Fuel Without Conventional Reprocessing - 13579

Energy Technology Data Exchange (ETDEWEB)

Bertch, Timothy C.; Schleicher, Robert W.; Rawls, John D. [General Atomics 3550 General Atomics Court San Diego, CA 92130 (United States)

2013-07-01

General Atomics has started design of a waste to energy nuclear reactor (EM2) that can use light water reactor (LWR) spent nuclear fuel (SNF). This effort addresses two problems: using an advanced small reactor with long core life to reduce nuclear energy overnight cost and providing a disposal path for LWR SNF. LWR SNF is re-fabricated into new EM2 fuel using a dry voloxidation process modeled on AIROX/ OREOX processes which remove some of the fission products but no heavy metals. By not removing all of the fission products the fuel remains self-protecting. By not separating heavy metals, the process remains proliferation resistant. Implementation of Energy Multiplier Module (EM2) fuel cycle will provide low cost nuclear energy while providing a long term LWR SNF disposition path which is important for LWR waste confidence. With LWR waste confidence recent impacts on reactor licensing, an alternate disposition path is highly relevant. Centered on a reactor operating at 250 MWe, the compact electricity generating system design maximizes site flexibility with truck transport of all system components and available dry cooling features that removes the need to be located near a body of water. A high temperature system using helium coolant, electricity is efficiently produced using an asynchronous high-speed gas turbine while the LWR SNF is converted to fission products. Reactor design features such as vented fuel and silicon carbide cladding support reactor operation for decades between refueling, with improved fuel utilization. Beyond the reactor, the fuel cycle is designed so that subsequent generations of EM2 reactor fuel will use the previous EM2 discharge, providing its own waste confidence plus eliminating the need for enrichment after the first generation. Additional LWR SNF is added at each re-fabrication to replace the removed fission products. The fuel cycle uses a dry voloxidation process for both the initial LWR SNF re-fabrication and later for EM2

An advanced conceptual Tokamak fusion power reactor utilizing closed cycle helium gas turbines

International Nuclear Information System (INIS)

Conn, R.W.

1976-01-01

UWMAK-III is a conceptual Tokamak reactor designed to study the potential and the problems associated with an advanced version of Tokamaks as power reactors. Design choices have been made which represent reasonable extrapolations of present technology. The major features are the noncircular plasma cross section, the use of TZM, a molybdenum based alloy, as the primary structural material, and the incorporation of a closed-cycle helium gas turbine power conversion system. A conceptual design of the turbomachinery is given together with a preliminary heat exchanger analysis that results in relatively compact designs for the generator, precooler, and intercooler. This paper contains a general description of the UWMAK-III system and a discussion of those aspects of the reactor, such as the burn cycle, the blanket design and the heat transfer analysis, which are required to form the basis for discussing the power conversion system. The authors concentrate on the power conversion system and include a parametric performance analysis, an interface and trade-off study and a description of the reference conceptual design of the closed-cycle helium gas turbine power conversion system. (Auth.)

Feasibility study for fast reactor and related fuel cycle. Preliminary studies in 1998

International Nuclear Information System (INIS)

Hayafune, Hiroki; Enuma, Yasuhiro; Kubota, Kenichi; Yoshida, Masashi; Uno, Osamu; Ishikawa, Hiroyasu; Kobayashi, Jun; Umetsu, Youichiro; Ichimiya, Masakazu

1999-10-01

Prior to the feasibility study for fast reactors (FRs) starting from the 1999 fiscal year, planned in the medium and long-term program of JNC, preliminarily studies were performed on 'FR systems except sodium cooled MOX fueled reactors'. Small scale or module type reactors, heavy metal (Pb or Pb-Bi) cooled reactors, gas cooled reactors, light water cooled reactors, and molten salt reactors were studied on the basis of literature. They were evaluated from the viewpoint of the technical possibility (the structure integrity, earthquake resistance, safety, productivity, operability, maintenance repair, difficulty of the development), the long-term targets (market competitiveness as an energy system, utilization of uranium resources, reduction of radioactive waste, security of the non-proliferation), and developmental risk. As the result, the following concepts should be studied for future commercialized FRs. Small scale and module type reactor: Middle-sized reactor with an excellent economical efficiency. Small power reactor with a multipurpose design concept. Gas cooled reactor: CO2 gas cooled reactor, He gas cooled reactor. Heavy metal cooled reactor: Russian type lead cooled reactor. Light water cooled reactor: Light water cooled high converter reactor and super critical pressure light water cooled reactor. Molten salt reactor: Trichloride molten salt reactor which matches the U-Pu cycle. (author)

Systems design of direct-cycle supercritical-water-cooled fast reactors

International Nuclear Information System (INIS)

Oka, Yoshiaki; Koshizuka, Seiichi; Jevremovic, Tatjana; Okano, Yashushi

1995-01-01

The system design of a direct-cycle supercritical-water-cooled fast reactor is presented. The supercritical water does not exhibit a change of phase. the recirculation system, steam separator, and dryer of a boiling water reactor (BWR) are unnecessary. Roughly speaking, the reactor pressure vessel and control rods are similar to those of a pressurized water reactor, the containment and emergency core cooling system are similar to a BWR, and the balance of plant is similar to a supercritical-pressure fossil-fired power plant (FPP). the electric power of the fast converter is 1,508 MW(electric). The number of coolant loops is only two because of the high coolant enthalpy. Containment volume is much reduced. The thermal efficiency is improved 24% over a BWR. The coolant void reactivity is negative by placing thin zirconium-hydride layers between seeds and blankets. The power costs would be much reduced compared with those of a light water reactor (LWR) and a liquid-metal fast breeder reactor. The concept is based on the huge amount of experience with the water coolant technology of LWRs and FPPs. The oxidation of stainless steel cladding is avoided by adopting a much lower coolant temperature than that of the FPP

Prospects of power ramping and cycling supervision in Finnish power reactors

Energy Technology Data Exchange (ETDEWEB)

Antila, M; Kaikkonen, H T [Imatran Voima Oy, Helsinki (Finland); Mannola, E [Teollisuuden Voima Oy Industries Kraft Ab, Helsinki (Finland)

1983-06-01

Since 1977 2x440 MWe PWR and 2x660 MWe BWR nuclear power has been taken in operation in Finland, which until the middle of 1982 has given favourable fuel operating experiences from 10 reactor years. This paper describes the core supervision systems of the plants especially from the viewpoint of ramp surveillance and the potentials and needs to improve the supervision capability to meet the future needs in case more load follow operation is required. As a special feature for Imatran Voima is the demand of general basic understanding of the behaviour of Loviisa reactors' fuel in different operating conditions. A possibility to investigate the fuel seem to be power cycling tests in Loviisa reactors. (author)

Prospects of power ramping and cycling supervision in Finnish power reactors

International Nuclear Information System (INIS)

Antila, M.; Kaikkonen, H.T.; Mannola, E.

1983-01-01

Since 1977 2x440 MWe PWR and 2x660 MWe BWR nuclear power has been taken in operation in Finland, which until the middle of 1982 has given favourable fuel operating experiences from 10 reactor years. This paper describes the core supervision systems of the plants especially from the viewpoint of ramp surveillance and the potentials and needs to improve the supervision capability to meet the future needs in case more load follow operation is required. As a special feature for Imatran Voima is the demand of general basic understanding of the behaviour of Loviisa reactors' fuel in different operating conditions. A possibility to investigate the fuel seem to be power cycling tests in Loviisa reactors. (author)

Advanced plutonium management in PWR - complementarity of thorium and uranium cycles

International Nuclear Information System (INIS)

Ernoult, Marc

2014-01-01

In order to study the possibility of advanced management of plutonium in existing reactors, 8 strategies for plutonium multi-recycling in PWRs are studied. Following equilibrium studies, it was shown that, by using homogeneous assemblies, the use of thorium cannot reduce the plutonium inventory of equilibrium cycle or production of americium. By distributing the different fuel types within the same assembly, some thoriated strategies allow however lower inventories and lower production americium best strategies using only the uranium cycle. However, in all cases, low fuel conversion theories in PWRs makes it impossible to lower resource consumption more than a few percent compared to strategies without thorium. To study the transition, active participation in development of the scenario code CLASS has been taken. It led to the two simulation scenarios among those studied in equilibrium with CLASS. These simulations have shown discrepancies with previously simulated scenarios. The major causes of these differences were identified and quantified. (author)

10-75-kWe-reactor-powered organic Rankine-cycle electric power systems (ORCEPS) study. Final technical report

Energy Technology Data Exchange (ETDEWEB)

1977-03-30

This 10-75 kW(e) Reactor-ORCEPS study was concerned with the evaluation of several organic Rankine cycle energy conversion systems which utilized a /sup 235/U-ZrH reactor as a heat source. A liquid metal (NaK) loop employing a thermoelectric converter-powered EM pump was used to transfer the reactor energy to the organic working fluid. At moderate peak cycle temperatures (750/sup 0/F), power conversion unit cycle efficiencies of up to 25% and overall efficiencies of 20% can be obtained. The required operating life of seven years should be readily achievable. The CP-25 (toluene) working fluid cycle was found to provide the highest performance levels at the lowest system weights. Specific weights varies from 100 to 50 lb/kW(e) over the power level range 10 to 75 kW(e). (DLC)

High Temperature Fusion Reactor Cooling Using Brayton Cycle Based Partial Energy Conversion

Science.gov (United States)

Juhasz, Albert J.; Sawicki, Jerzy T.

2003-01-01

For some future space power systems using high temperature nuclear heat sources most of the output energy will be used in other than electrical form, and only a fraction of the total thermal energy generated will need to be converted to electrical work. The paper describes the conceptual design of such a partial energy conversion system, consisting of a high temperature fusion reactor operating in series with a high temperature radiator and in parallel with dual closed cycle gas turbine (CCGT) power systems, also referred to as closed Brayton cycle (CBC) systems, which are supplied with a fraction of the reactor thermal energy for conversion to electric power. Most of the fusion reactor's output is in the form of charged plasma which is expanded through a magnetic nozzle of the interplanetary propulsion system. Reactor heat energy is ducted to the high temperature series radiator utilizing the electric power generated to drive a helium gas circulation fan. In addition to discussing the thermodynamic aspects of the system design the authors include a brief overview of the gas turbine and fan rotor-dynamics and proposed bearing support technology along with performance characteristics of the three phase AC electric power generator and fan drive motor.

Evaluation of fuel fabrication and the back end of the fuel cycle for light-water- and heavy-water-cooled nuclear power reactors

International Nuclear Information System (INIS)

Carter, W.L.; Olsen, A.R.

1979-06-01

The classification of water-cooled nuclear reactors offers a number of fuel cycles that present inherently low risk of weapons proliferation while making power available to the international community. Eight fuel cycles in light water reactor (LWR), heavy water reactor (HWR), and the spectral shift controlled reactor (SSCR) systems have been proposed to promote these objectives in the International Fuel Cycle Evaluation (INFCE) program. Each was examined in an effort to provide technical and economic data to INFCE on fuel fabrication, refabrication, and reprocessing for an initial comparison of alternate cycles. The fuel cycles include three once-through cycles that require only fresh fuel fabrication, shipping, and spent fuel storage; four cycles that utilize denatured uranium--thorium and require all recycle operations; and one cycle that considers the LWR--HWR tandem operation requiring refabrication but no reprocessing

Evaluation of fuel fabrication and the back end of the fuel cycle for light-water- and heavy-water-cooled nuclear power reactors

Energy Technology Data Exchange (ETDEWEB)

Carter, W.L.; Olsen, A.R.

1979-06-01

The classification of water-cooled nuclear reactors offers a number of fuel cycles that present inherently low risk of weapons proliferation while making power available to the international community. Eight fuel cycles in light water reactor (LWR), heavy water reactor (HWR), and the spectral shift controlled reactor (SSCR) systems have been proposed to promote these objectives in the International Fuel Cycle Evaluation (INFCE) program. Each was examined in an effort to provide technical and economic data to INFCE on fuel fabrication, refabrication, and reprocessing for an initial comparison of alternate cycles. The fuel cycles include three once-through cycles that require only fresh fuel fabrication, shipping, and spent fuel storage; four cycles that utilize denatured uranium--thorium and require all recycle operations; and one cycle that considers the LWR--HWR tandem operation requiring refabrication but no reprocessing.

Development of Monte Carlo-based pebble bed reactor fuel management code

International Nuclear Information System (INIS)

Setiadipura, Topan; Obara, Toru

2014-01-01

Highlights: • A new Monte Carlo-based fuel management code for OTTO cycle pebble bed reactor was developed. • The double-heterogeneity was modeled using statistical method in MVP-BURN code. • The code can perform analysis of equilibrium and non-equilibrium phase. • Code-to-code comparisons for Once-Through-Then-Out case were investigated. • Ability of the code to accommodate the void cavity was confirmed. - Abstract: A fuel management code for pebble bed reactors (PBRs) based on the Monte Carlo method has been developed in this study. The code, named Monte Carlo burnup analysis code for PBR (MCPBR), enables a simulation of the Once-Through-Then-Out (OTTO) cycle of a PBR from the running-in phase to the equilibrium condition. In MCPBR, a burnup calculation based on a continuous-energy Monte Carlo code, MVP-BURN, is coupled with an additional utility code to be able to simulate the OTTO cycle of PBR. MCPBR has several advantages in modeling PBRs, namely its Monte Carlo neutron transport modeling, its capability of explicitly modeling the double heterogeneity of the PBR core, and its ability to model different axial fuel speeds in the PBR core. Analysis at the equilibrium condition of the simplified PBR was used as the validation test of MCPBR. The calculation results of the code were compared with the results of diffusion-based fuel management PBR codes, namely the VSOP and PEBBED codes. Using JENDL-4.0 nuclide library, MCPBR gave a 4.15% and 3.32% lower k eff value compared to VSOP and PEBBED, respectively. While using JENDL-3.3, MCPBR gave a 2.22% and 3.11% higher k eff value compared to VSOP and PEBBED, respectively. The ability of MCPBR to analyze neutron transport in the top void of the PBR core and its effects was also confirmed

Advanced fuel cycles of WWER-1000 reactors

International Nuclear Information System (INIS)

Lunin, G.; Novikov, A.; Pavlov, V.; Pavlovichev, A.

2003-01-01

The present paper considers characteristics of fuel cycles for the WWER-1000 reactor satisfying the following conditions: duration of the campaign at the nominal power is extended from 250 EFPD up to 470 and more ones; fuel enrichment does not exceed 5 wt.%; fuel assemblies maximum burnup does not exceed 55 MWd/kgHM. Along with uranium fuel, the use of mixed Uranium-Plutonium fuel is considered. Calculations were conducted by codes TVS-M, BIPR-7A and PERMAK-A developed in the RRC Kurchatov Institute, verified for the calculations of uranium fuel and certified by GAN RF

Thorium fuel for light water reactors - reducing proliferation potential of nuclear power fuel cycle

Energy Technology Data Exchange (ETDEWEB)

Galperin, A; Radkowski, A [Ben-Gurion Univ. of the Negev, Beersheba (Israel)

1996-12-01

The proliferation potential of the light water reactor fuel cycle may be significantly reduced by utilization of thorium as a fertile component of the nuclear fuel. The main challenge of Th utilization is to design a core and a fuel cycle, which would be proliferation-resistant and economically feasible. This challenge is met by the Radkowsky Thorium Reactor (RTR) concept. So far the concept has been applied to a Russian design of a 1,000 MWe pressurized water reactor, known as a WWER-1000, and designated as VVERT. The following are the main results of the preliminary reference design: * The amount of Pu contained in the RTR spent fuel stockpile is reduced by 80% in comparison with a VVER of a current design. * The isotopic composition of the RTR-Pu greatly increases the probability of pre-initiation and yield degradation of a nuclear explosion. An extremely large Pu-238 content causes correspondingly large heat emission, which would complicate the design of an explosive device based on RTR-Pu. The economic incentive to reprocess and reuse the fissile component of the RTR spent fuel is decreased. The once-through cycle is economically optimal for the RTR core and cycle. To summarize all the items above: the replacement of a standard (U-based) fuel for nuclear reactors of current generation by the RTR fuel will provide an inherent barrier for nuclear weapon proliferation. This inherent barrier, in combination with existing safeguard measures and procedures is adequate to unambiguously disassociate civilian nuclear power from military nuclear power. * The RTR concept is applied to existing power plants to assure its economic feasibility. Reductions in waste disposal requirements, as well as in natural U and fabrication expenses, as compared to a standard WWER fuel, provide approximately 20% reduction in fuel cycle (authors).

Thorium cycle and molten salt reactors: field parameters and field constraints investigations toward 'thorium molten salt reactor' definition

International Nuclear Information System (INIS)

Mathieu, L.

2005-09-01

Producing nuclear energy in order to reduce the anthropic CO 2 emission requires major technological advances. Nuclear plants of 4. generation have to respond to several constraints, as safety improvements, fuel breeding and radioactive waste minimization. For this purpose, it seems promising to use Thorium Cycle in Molten Salt Reactors. Studies on this domain have already been carried out. However, the final concept suffered from serious issues and was discontinued. A new reflection on this topic is being led in order to find acceptable solutions, and to design the Thorium Molten Salt Reactor concept. A nuclear reactor is simulated by the coupling of a neutron transport code with a materials evolution code. This allows us to reproduce the reactor behavior and its evolution all along its operation. Thanks to this method, we have studied a large number of reactor configurations. We have evaluated their efficiency through a group of constraints they have to satisfy. This work leads us to a better understanding of many physical phenomena controlling the reactor behavior. As a consequence, several efficient configurations have been discovered, allowing the emergence of new points of view in the research of Molten Salt Reactors. (author)

Extension of the supercritical carbon dioxide Brayton cycle for application to the Very High Temperature Reactor

International Nuclear Information System (INIS)

Moisseytsev, A.; Sienicki, J. J.

2010-01-01

An investigation has been carried out of the feasibility of applying the supercritical carbon dioxide (S-CO 2 ) Brayton cycle to the Very High Temperature Reactor (VHTR). Direct application of the standard S-CO 2 recompression cycle to the VHTR was found to be challenging because of the mismatch in the inherent temperature drops across the He and CO 2 sides of the reactor heat exchanger resulting in a relatively low cycle efficiency of 45 % compared to 48 % for a direct helium cycle. Two approaches consisting of either a cascaded cycle arrangement with three separate cascaded S-CO 2 cycles or, alternately, operation of a single S-CO 2 cycle with the minimum pressure below the critical pressure and the minimum temperature above the critical temperature have been identified and shown to successfully enable the S-CO 2 Brayton cycle to be adapted to the VHTR such that the benefits of the higher S-CO 2 cycle efficiency can be realized. For both approaches, S-CO 2 cycle efficiencies in excess of 49 % are calculated. (authors)

A comparative study on recycling spent fuels in gas-cooled fast reactors

International Nuclear Information System (INIS)

Choi, Hangbok; Baxter, Alan

2010-01-01

This study evaluates advanced Gas-cooled Fast Reactor (GFR) fuel cycle scenarios which are based on recycling spent nuclear fuel for the sustainability of nuclear energy. A 600 MWth GFR was used for the fuel cycle analysis, and the equilibrium core was searched with different fuel-to-matrix volume ratios such as 70/30 and 60/40. Two fuel cycle scenarios, i.e., a one-tier case combining a Light Water Reactor (LWR) and a GFR, and a two-tier case using an LWR, a Very High Temperature Reactor (VHTR), and a GFR, were evaluated for mass flow and fuel cycle cost, and the results were compared to those of LWR once-through fuel cycle. The mass flow calculations showed that the natural uranium consumption can be reduced by more than 57% and 27% for the one-tier and two-tier cycles, respectively, when compared to the once-through fuel cycle. The transuranics (TRU) which pose a long-term problem in a high-level waste repository, can be significantly reduced in the multiple recycle operation of these options, resulting in more than 110 and 220 times reduction of TRU inventory to be geologically disposed for the one-tier and two-tier fuel cycles, respectively. The fuel cycle costs were estimated to be 9.4 and 8.6 USD/MWh for the one-tier fuel cycle when the GFR fuel-to-matrix volume ratio was 70/30 and 60/40, respectively. However the fuel cycle cost is reduced to 7.3 and 7.1 USD/MWh for the two-tier fuel cycle, which is even smaller than that of the once-through fuel cycle. In conclusion the GFR can provide alternative fuel cycle options to the once-through and other fast reactor fuel cycle options, by increasing the natural uranium utilization and reducing the fuel cycle cost.

Nonlinear dynamics and stability of boiling water reactors: qualitative and quantitative analyses

International Nuclear Information System (INIS)

March-Leuba, J.; Cacuci, D.G.; Perez, R.B.

1985-01-01

A phenomenological model has been developed to simulate the qualitative behavior of boiling water reactors (BWRs) in the nonlinear regime under deterministic and stochastic excitations. After the linear stability threshold is crossed, limit cycle oscillations appear due to interactions between two unstable equilibrium points and the phase-space trajectories. This limit cycle becomes unstable when the feedback gain exceeds a certain critical value. Subsequent limit cycle instabilities produce a cascade of period-doubling bifurcations that leads to a periodic pulsed behavior. Under stochastic excitations, BWRs exhibit a single characteristic resonance, at approx.0.5 Hz, in the linear regime. By contrast, this work shows that harmonics of this characteristic frequency appear in the nonlinear regime. Furthermore, this work also demonstrates that amplitudes of the limit cycle oscillations do not depend on the variance of the stochastic excitation and remain bounded at all times. A physical model of nonlinear BWR dynamics has also been developed and employed to calculate the amplitude of limit cycle oscillations and their effects on fuel integrity over a wide range of operating conditions in the Vermont Yankee reactor. These calculations have confirmed that, beyond the threshold for linear stability, the reactor's state variable undergo limit cycle oscillations

POWER CYCLE AND STRESS ANALYSES FOR HIGH TEMPERATURE GAS-COOLED REACTOR

International Nuclear Information System (INIS)

Oh, Chang H; Davis, Cliff; Hawkes, Brian D; Sherman, Steven R

2007-01-01

The Department of Energy and the Idaho National Laboratory are developing a Next Generation Nuclear Plant (NGNP) to serve as a demonstration of state-of-the-art nuclear technology. The purpose of the demonstration is two fold (1) efficient low cost energy generation and (2) hydrogen production. Although a next generation plant could be developed as a single-purpose facility, early designs are expected to be dual-purpose. While hydrogen production and advanced energy cycles are still in its early stages of development, research towards coupling a high temperature reactor, electrical generation and hydrogen production is under way. Many aspects of the NGNP must be researched and developed in order to make recommendations on the final design of the plant. Parameters such as working conditions, cycle components, working fluids, and power conversion unit configurations must be understood. Three configurations of the power conversion unit were demonstrated in this study. A three-shaft design with three turbines and four compressors, a combined cycle with a Brayton top cycle and a Rankine bottoming cycle, and a reheated cycle with three stages of reheat were investigated. An intermediate heat transport loop for transporting process heat to a High Temperature Steam Electrolysis (HTSE) hydrogen production plant was used. Helium, CO2, and a 80% nitrogen, 20% helium mixture (by weight) were studied to determine the best working fluid in terms cycle efficiency and development cost. In each of these configurations the relative component size were estimated for the different working fluids. The relative size of the turbomachinery was measured by comparing the power input/output of the component. For heat exchangers the volume was computed and compared. Parametric studies away from the baseline values of the three-shaft and combined cycles were performed to determine the effect of varying conditions in the cycle. This gives some insight into the sensitivity of these cycles to

Nuclear fuel cycle and reactor strategies: Adjusting to new realities. Key issue papers

International Nuclear Information System (INIS)

1997-01-01

The international symposium ''Nuclear Fuel Cycle and Reactor Strategy: Adjusting to new Realities'' was organized to face the new realities in the nuclear fuel cycle and to consider options on how these new realities could be addressed. The Key Issue Papers treat the various subjects from both short and long term perspectives. In so doing, they address the likely development of all aspects concerning the nuclear fuel cycle up to the year 2050

Fuel cycle: the transition between the third and the fourth generation of reactors

International Nuclear Information System (INIS)

2008-01-01

Many challenges arrive today for the french research and development on the fuel cycle: promote the industrial technologies, improve the world increase of the nuclear and adapt the fuel cycle technologies to the future reactors. In this framework the report presents after a recall on the fuel cycle, the researches on the fuel, the optimization of the recycling, the wastes management, the simulation and Phenix an experimentation tool for the fuel. (A.L.B.)

The uncertainty analysis of a liquid metal reactor for burning minor actinides from light water reactors

Energy Technology Data Exchange (ETDEWEB)

Choi, Hang Bok [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1999-12-31

The neutronics analysis of a liquid metal reactor for burning minor actinides has shown that uncertainties in the nuclear data of several key minor actinide isotopes can introduce large uncertainties in the predicted performance of the core. A comprehensive sensitivity and uncertainty analysis was performed on a 1200 MWth actinide burner designed for a low burnup reactivity swing, negative doppler coefficient, and low sodium void worth. Sensitivities were generated using depletion perturbation methods for the equilibrium cycle of the reactor and covariance data was taken ENDF-B/V and other published sources. The relative uncertainties in the burnup swing, doppler coefficient, and void worth were conservatively estimated to be 180%, 97%, and 46%, respectively. 5 refs., 1 fig., 3 tabs. (Author)

The uncertainty analysis of a liquid metal reactor for burning minor actinides from light water reactors

Energy Technology Data Exchange (ETDEWEB)

Choi, Hang Bok [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1998-12-31

The neutronics analysis of a liquid metal reactor for burning minor actinides has shown that uncertainties in the nuclear data of several key minor actinide isotopes can introduce large uncertainties in the predicted performance of the core. A comprehensive sensitivity and uncertainty analysis was performed on a 1200 MWth actinide burner designed for a low burnup reactivity swing, negative doppler coefficient, and low sodium void worth. Sensitivities were generated using depletion perturbation methods for the equilibrium cycle of the reactor and covariance data was taken ENDF-B/V and other published sources. The relative uncertainties in the burnup swing, doppler coefficient, and void worth were conservatively estimated to be 180%, 97%, and 46%, respectively. 5 refs., 1 fig., 3 tabs. (Author)

Definition of breeding gain for the closed fuel cycle and application to a gas cooled fast reactor

International Nuclear Information System (INIS)

Van Rooijen, W. F. G.; Kloosterman, J. L.; Van Der Hagen, T. H. J. J.; Van Dam, H.

2006-01-01

In this paper a definition is given for the Breeding Gain (BG) of a nuclear reactor, taking into account compositional changes of the fuel during irradiation, cool down and reprocessing. A definition is given for the reactivity weights required to calculate BG. To calculate the effects of changes in the initial fuel composition on BG, first order nuclide perturbation theory is used. The theory is applied to the fuel cycle of GFR600, a 600 MWth Generation IV Gas Cooled Fast Reactor. This reactor should have a closed fuel cycle, with a BG equal to zero, breeding just enough new fuel during irradiation to allow refueling by only adding fertile material. All Heavy Metal is recycled in the closed fuel cycle. The result is that a closed fuel cycle is possible if the reprocessing has low losses ( 238 U, 15% Pu, and low amounts of the Minor Actinides. (authors)

A computer program for calculation of the fuel cycle in pressurized water reactors

International Nuclear Information System (INIS)

Solanilla, R.

1976-01-01

The purpose of the FUCEFURE program is two-fold: first, it is designed to solve the problem of nuclear fuel cycle cost in one pressurized light water reactor calculation. The code was developed primarily for comparative and sensitivity studies. The program contains simple correlations between exposure and available depletion data used to predict the uranium and plutonium content of the fuel as a function of the fuel initial enrichment. Second, it has been devised to evaluate the nuclear fuel demand associated with an expanding nuclear power system. Evaluation can be carried out at any time and stage in the fuel cycle. The program can calculate the natural uranium and separate work requirements of any final and tails enrichment. It also can determine the nuclear power share of each reactor in the system when a decision has been made about the long-term nuclear power installations to be used and the types of PWR and fast breeder reactor characteristics to be involved in them. (author)

Uranium and thorium cycles for sodium fast reactors: Neutronic aspects and associated wastes

International Nuclear Information System (INIS)

Brizi, J.

2010-10-01

Sodium fast reactors (SFR-Na) with uranium 238/plutonium 239(U/Pu) cycle, its technical feasibility has already proven, allow to overcome the problem of natural uranium resources in achieving the regeneration of the fuel fissile element. In addition, a waste management can be performed to reduce the radiotoxicity of actinides produced by the reactor in transmuting the AM in the core (homogeneous transmutation). Another alternative to minimize waste is to use another couple fertile-fissile: the thorium 232 and the uranium 233 (Th/U). The comparison is performed on neutronic and safety aspects and on waste production, in using an evolutive Monte Carlo. Although one does not disclose real clear advantages concerning the radiotoxicity of wastes for a particular cycle, the Th/U cycle reduces the radiotoxicity during periods when it is the highest. The homogeneous transmutation minimizes significantly for both cycles, radiotoxicity of wastes, with different factors depending on the considered time period. However, it is done to the detriment of an important increase of AM in the core. If we consider the nuclear stop, the inventory of the reactor core becomes a waste. The gain provided by the transmutation, taking into account both the core and accumulated waste radio-toxicities, will be quantified, and shows the transmutation does not provide a significant gain if the burning of main fissile elements is not considered when the nuclear is stopped. (author)

Design of a power conversion system for an indirect cycle, helium cooled pebble bed reactor system

International Nuclear Information System (INIS)

Wang, C.; Ballinger, R.G.; Stahle, P.W.; Demetri, E.; Koronowski, M.

2002-01-01

A design is presented for the turbomachinery for an indirect cycle, closed, helium cooled modular pebble bed reactor system. The design makes use of current technology and will operate with an overall efficiency of 45%. The design uses an intermediate heat exchanger which isolated the reactor cycle from the turbomachinery. This design excludes radioactive fission products from the turbomachinery. This minimizes the probability of an air ingress accident and greatly simplifies maintenance. (author)

Status of fast breeder reactors and associated fuel cycle in India

International Nuclear Information System (INIS)

Chellapandi, P.

2009-01-01

Full text: India is the largest democracy with the current population of about 1.05 billion and is on a road to rapid growth in economy. An impressive average domestic product (GDP) growth rate of about 8 % per year has been achieved in 2006-07 and it is targeted to touch 10 % per year for the next 10 years. Towards realizing this targeted growth, development activities are planned based on well-conceived road map and clear vision. Like elsewhere, the energy and electricity growth in India are also closely linked to growth in economy. Indices of socio-economic development like literacy, longevity, GDP and human development are directly dependent upon the per capita energy consumption of a country. India is aiming to reach at least per capita energy consumption equal to the present world average (2200 kWh/a) by 2030 from the current value of (660 kWh/a). The current installed capacity of ∼138 GW(e) needs to be increased to about 600 GWe by 2030 assuming the population of about 1.4 billion. Energy strategists in the country have realized the importance of judicious mix of energy resources to meet this challenge. A large share of nuclear energy is an inevitable choice in this judicious energy mix from resources, sustainability and environment considerations. The nuclear is expected to contribute about 63 GWe by 2030, which will be steadily increased to 275 GWe by 2052, against the total projected capacity of 1344 GWe. The three stage visionary programme of India envisages Water Reactors (first stage), Fast Breeders with high breeding (second stage) and Thorium based Reactors as third stage. Closed fuel cycle in all stages is an essential ingredient. The success of each stage depends upon expeditious maturity of the earlier stage as India has limited indigenous resources of uranium, but vast resources of thorium. India ranks high in nuclear technology scale with strong R and D, high quality human resources, sound infrastructure, unwavering Government support and

Evaluation of spectral shift controlled reactors operating on the uranium fuel cycle. Final report

International Nuclear Information System (INIS)

Matzie, R.A.; Sider, F.M.

1979-08-01

The performance of the spectral shift controlled reactor (SSCR) operating on uranium fuel cycles was evaluated and compared with the conventional pressurized water reactor (PWR). In order to analyze the SSCR, the PSR design methodology was extended to include systems moderated by mixtures of light water and heavy water and these methods were validated by comparison with experimental results. Once the design methods had been formulated, the resouce requirements and power costs were determined for the uranium-fueled SSCR. The ore requirements of the UO 2 once-through fuel cycle and the UO 2 fuel cycle with self-generated recycle (SGR) of plutonium were found to be 10% and 19% less than those of similarly fueled PWRs, respectively. A fuel cycle optimization study was performed for the UO 2 once-through SSCR and the SGR SSCR. By individually altering lattice parameters, discharge exposure or number of in-core batches, savings of less than 8% in resource requirements and less than 1% in power costs were obtained

Comparative study on plutonium and MA recycling in equilibrium burnup and standard burnup of PWR

International Nuclear Information System (INIS)

Waris, Abdul; Kurniadi, Rizal; Su'ud, Zaki; Permana, Sidik

2005-01-01

The equilibrium burnup model is a powerful method since its can handle all possible generated nuclides in any nuclear system. Moreover, this method is a simple time independent method. Hence the equilibrium burnup method could be very useful for evaluating and forecasting the characteristics of any nuclear fuel cycle, even the strange one, e.g. all nuclides are confined in the reactor. However, this method needs to be verified since the method is not a standard tool. The present study aimed to compare the characteristics of plutonium recycling and plutonium and minor actinides (MA) recycling in PWR with the equilibrium burnup and the standard burnup. In order to become more comprehensive study, an influence of moderator-to-fuel volume ratio (MFR) changes by changing the pin-pitch of fuel cell has been evaluated. The MFR ranges from 0.5 to 4.0. For the equilibrium burnup we used equilibrium cell-burnup code. We have employed 1368 nuclides in the equilibrium calculation with 129 of them are heavy metals (HMs). For standard burnup, SRAC2002 code has been utilized with 26 HMs and 66 fission products (FPs). The JENDL 3.2 library has been employed for both burnup schemes. The uranium, plutonium and MA vector, which resulted from the equilibrium burnup are directly used as fuel input composition for the standard burnup calculation. Both burnup results demonstrate that plutonium recycling and plutonium and MA recycling can be conducted safer in tight lattice core. They are also show the similar trend in neutron spectrum, which become harder with the increasing number of recycled heavy nuclides as well as the decreasing of the MFR values. However, there are some discrepancy on the effective multiplication factor and the conversion ratio, especially for the reactor core for MFR ≥ 2.0. (author)

Development of dynamic simulation code for fuel cycle fusion reactor

Energy Technology Data Exchange (ETDEWEB)

Aoki, Isao; Seki, Yasushi [Department of Fusion Engineering Research, Naka Fusion Research Establishment, Japan Atomic Energy Research Institute, Naka, Ibaraki (Japan); Sasaki, Makoto; Shintani, Kiyonori; Kim, Yeong-Chan

1999-02-01

A dynamic simulation code for fuel cycle of a fusion experimental reactor has been developed. The code follows the fuel inventory change with time in the plasma chamber and the fuel cycle system during 2 days pulse operation cycles. The time dependence of the fuel inventory distribution is evaluated considering the fuel burn and exhaust in the plasma chamber, purification and supply functions. For each subsystem of the plasma chamber and the fuel cycle system, the fuel inventory equation is written based on the equation of state considering the fuel burn and the function of exhaust, purification, and supply. The processing constants of subsystem for steady states were taken from the values in the ITER Conceptual Design Activity (CDA) report. Using this code, the time dependence of the fuel supply and inventory depending on the burn state and subsystem processing functions are shown. (author)

Radiological impact of plutonium recycle in the fuel cycle of LWR type reactors: professional exposure during mormal operation

International Nuclear Information System (INIS)

White, I.F.; Kelly, G.N.

1983-01-01

The radiological impact of the fuel cycle of light water type reactors using enriched uranium may be changed by plutonium recycle. The impact on human population and on the persons professionally exposed may be different according to the different steps of the fuel cycle. This report analyses the differential radiological impact on the different types of personnel involed in the fuel cycle. Each step of the fuel cycle is separately studied (fuel fabrication, reactor operation, fuel reprocessing), as also the transport of the radioactive materials between the different steps. For the whole fuel cycle, one estimates that, with regard to the fuel cycle using enriched uranium, the plutonium recycle involves a small increase of the professional exposure

Nuclear reactors and technology in the next stage

International Nuclear Information System (INIS)

Orlov, V.

2000-01-01

Author deals with the perspectives of development of nuclear power. It is possible to create in a fairly short time reactors and fuel technology that would meet the main requirements for large-scale power production, i.e.: (a) to afford a 100-fold reduction in the specific consumption of uranium, by utilizing thousands of tonnes of Pu accumulated in the spent fuel from the reactors of the fl t stage; .to rule out nuclear disasters, by taking advantage of the intrinsic properties and behavior of reactor, coolant, fuel, etc., with the plants made simpler and cheaper; (b) to hit a balance between the radiotoxicity of waste and that of feed uranium, by providing neutron transmutation; (c) to create power reactors and fuel cycle technology that would not afford extraction of weapon-grade materials. To fulfil all these requirements, it is necessary to provide substantial neutron excess in a chain reaction for Pu breeding, to use fuel with an equilibrium composition, to bum actinides and LLFPs. All this can be done only in fast reactors. Fast reactors can also provide fuel for thermal reactors that might still be used for some applications, operating in a Th/U cycle, which is the best option for such facilities. Novel engineering solutions will be necessary: high-density heat-conductive fuel (UPuN), chemically inert high-boiling coolant (Pb), dry reprocessing. These issues have been studied well enough to allow embarking on the development of advanced fast reactors. Minatom institutions are finalizing a detailed design of a demonstration BREST-300 plant, complete with an on-site fuel cycle that will meet the requirements of large-scale nuclear power. Hopefully, construction of this plant at Beloyarsk site with its subsequent trial operation would open a door to the next stage in nuclear power development. (author)

Adjusting to new realities. IAEO meeting on fuel cycle and reactor strategies, June 3-6, 1997 in Vienna

International Nuclear Information System (INIS)

Jelinek-Fink, P.

1997-01-01

The IAEA, in co-operation with the European Commission, the Nuclear Energy Agency of the OECD and the Uranium Institute, organized an international symposium to discuss the 'new realities' in the nuclear fuel cycle with special emphasis on plutonium management. The symposium covered all aspects of the fuel cycle in six sessions: Global Energy Outlook; Present Status and Immediate Prospects of Plutonium Management; Future Fuel Cycle and Reactor Strategies; Safety, Health and Environmental Implications of the Different Fuel Cycles; Non-Proliferation and Safeguards Aspects; and International Co-operation. The Symposium took place in Vienna, 3-6 June 1997. About 300 participants from 41 countries and 4 international organizations attended. The Symposium gave a comprehensive overview on the present situation of the fuel cycle and the expected developments up to 2050. During this period nuclear energy will continue to be dominated by Thermal Reactors; Fast Reactors will be introduced slowly after 2030. In general, an evolutionary and not a revolutionary process is expected. (orig.) [de

Evaluation of core compositions for use in breed and burn reactors and limited-separations fuel cycles

International Nuclear Information System (INIS)

Petroski, Robert; Forget, Benoit; Forsberg, Charles

2013-01-01

Highlights: ► Calculated minimum burnup and irradiation damage for B and B reactor compositions. ► Computed doubling time of fuel cycles using B and B reactors and no chemical separations. ► Determined sensitivity of doubling time to using melt refining vs. direct reuse. ► Examined tradeoff between power density and neutronics for different coolants. - Abstract: Previously developed methods for analyzing breed-and-burn (B and B) reactors are applied to a wide range of core compositions. The compositions studied include different fuel types, steel and silicon carbide structure, and sodium, lead/lead bismuth eutectic (LBE), and gas coolants. These compositions are evaluated for use in “minimum burnup” B and B reactors in which it is assumed that blocks comprising the core can be shuffled in all three dimensions to flatten out non-uniformities in burnup. The two figures of merit evaluated are the minimum irradiation damage requirement and reactor fleet doubling time. To minimize irradiation damage, gas coolants perform best, followed by lead/LBE then sodium. High uranium-content metal fuel outperforms compound fuels, and different types of steel are similar and perform slightly better than silicon carbide. Once-through irradiation damage requirements can be surprisingly modest in minimum burnup B and B reactors, with a wide range of compositions viable at irradiation damage levels 50% higher than existing materials data. Doubling times were calculated for a reactor fleet consisting of B and B reactors operating in a limited-separations fuel cycle; i.e., a fuel cycle with no chemical separation of actinides. The effects of different cooling times and removal of fission products using a melt refining process are evaluated. To minimize doubling time, sodium cooled compositions perform best because they are able to achieve core power densities several times larger than compositions using other coolants. A hypothetical sodium-cooled core composition with high

Subchannel analysis of a small ultra-long cycle fast reactor core

International Nuclear Information System (INIS)

Seo, Han; Kim, Ji Hyun; Bang, In Cheol

2014-01-01

Highlights: • The UCFR-100 is small-sized one of 60 years long-life nuclear reactors without refueling. • The design safety limits of the UCFR-100 are evaluated using MATRA-LMR. • The subchannel results are below the safety limits of general SFR design criteria. - Abstract: Thermal-hydraulic evaluation of a small ultra-long cycle fast reactor (UCFR) core is performed based on existing safety regulations. The UCFR is an innovative reactor newly designed with long-life core based on the breed-and-burn strategy and has a target electric power of 100 MWe (UCFR-100). Low enriched uranium (LEU) located at the bottom region of the core play the role of igniter to operate the UCFR for 60 years without refueling. A metallic form is selected as a burning fuel region material after the LEU location. HT-9 and sodium are used as cladding and coolant materials, respectively. In the present study, MATRA-LMR, subchannel analysis code, is used for evaluating the safety design limit of the UCFR-100 in terms of fuel, cladding, and coolant temperature distributions in the core as design criteria of a general fast reactor. The start-up period (0 year of operation), the middle of operating period (30 years of operation), and the end of operating cycle (60 years of operation) are analyzed and evaluated. The maximum cladding surface temperature (MCST) at the BOC (beginning of core life) is 498 °C on average and 551 °C when considering peaking factor, while the MCST at the MOC (middle of core life) is 498 °C on average and 548 °C in the hot channel, respectively, and the MCST at the EOC (end of core life) is 499 °C on average and 538 °C in the hot channel, respectively. The maximum cladding surface temperature over the long cycle is found at the BOC due to its high peaking factor. It is found that all results including fuel rods, cladding, and coolant exit temperature are below the safety limit of general SFR design criteria

Coupling of Modular High-Temperature Gas-Cooled Reactor with Supercritical Rankine Cycle

Directory of Open Access Journals (Sweden)

Shutang Zhu

2008-01-01

Full Text Available This paper presents investigations on the possible combination of modular high-temperature gas-cooled reactor (MHTGR technology with the supercritical (SC steam turbine technology and the prospective deployments of the MHTGR SC power plant. Energy conversion efficiency of steam turbine cycle can be improved by increasing the main steam pressure and temperature. Investigations on SC water reactor (SCWR reveal that the development of SCWR power plants still needs further research and development. The MHTGR SC plant coupling the existing technologies of current MHTGR module design with operation experiences of SC FPP will achieve high cycle efficiency in addition to its inherent safety. The standard once-reheat SC steam turbine cycle and the once-reheat steam cycle with life-steam have been studied and corresponding parameters were computed. Efficiencies of thermodynamic processes of MHTGR SC plants were analyzed, while comparisons were made between an MHTGR SC plant and a designed advanced passive PWR - AP1000. It was shown that the net plant efficiency of an MHTGR SC plant can reach 45% or above, 30% higher than that of AP1000 (35% net efficiency. Furthermore, an MHTGR SC plant has higher environmental competitiveness without emission of greenhouse gases and other pollutants.

Report on the 9th workshop on the innovative water reactor for flexible fuel cycle

International Nuclear Information System (INIS)

Ishikawa, Nobuyuki; Kobayashi, Noboru; Okubo, Tsutomu; Uchikawa, Sadao

2006-07-01

The research on Innovative Water Reactor for Flexible Fuel Cycle (FLWR) has been performed in JAEA for the development of future innovative reactor. The workshop on FLWRs has been held every year since 1998 aiming at information exchange with other organizations such as universities, laboratories, utilities and vendors. The 9th workshop was held on March 1, 2006 under the joint auspices of JAEA and North Kanto and Kanto-Koetsu branches of Atomic Energy Society of Japan with 64 participants. The workshop began with presentation entitled 'Activities on Nuclear Science and Engineering Research and Collaboration with Industry in JAEA', followed by presentations entitled 'Progress of Research and Development on FLWR' and 'On Final Report of Feasibility Study (phase 2) on Commercialized FBR Cycle Systems'. Then two lectures followed: 'Core and Fuel Design on Super Light Water Reactor' by Tokyo University and 'Recent trends on the Development of Next Generation Nuclear Reactor' by Institute of Applied Energy. This report summarizes the lectures of the workshop. (author)

Potential advantages of coupling supercritical CO2 Brayton cycle to water cooled small and medium size reactor