| 1. |
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| 2. |
- Dufek, Jan, et al.
(författare)
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An efficient parallel computing scheme for Monte Carlo criticality calculations
- 2009
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Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 36:8, s. 1276-1279
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Tidskriftsartikel (refereegranskat)abstract
- The existing parallel computing schemes for Monte Carlo criticality calculations suffer from a low efficiency when applied on many processors. We suggest a new fission matrix based scheme for efficient parallel computing. The results are derived from the fission matrix that is combined from all parallel simulations. The scheme allows for a practically ideal parallel scaling as no communication among the parallel simulations is required, and inactive cycles are not needed. (C) 2009 Elsevier Ltd. All rights reserved.
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| 3. |
- Dufek, Jan, 1978-
(författare)
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Development of New Monte Carlo Methods in Reactor Physics : Criticality, Non-Linear Steady-State and Burnup Problems
- 2009
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Doktorsavhandling (övrigt vetenskapligt/konstnärligt)abstract
- The Monte Carlo method is, practically, the only approach capable of giving detail insight into complex neutron transport problems. In reactor physics, the method has been used mainly for determining the keff in criticality calculations. In the last decade, the continuously growing computer performance has allowed to apply the Monte Carlo method also on simple burnup simulations of nuclear systems. Nevertheless, due to its extensive computational demands the Monte Carlo method is still not used as commonly as deterministic methods. One of the reasons for the large computational demands of Monte Carlo criticality calculations is the necessity to carry out a number of inactive cycles to converge the fission source. This thesis presents a new concept of fission matrix based Monte Carlo criticality calculations where inactive cycles are not required. It is shown that the fission matrix is not sensitive to the errors in the fission source, and can be thus calculated by a Monte Carlo calculation without inactive cycles. All required results, including keff, are then derived via the final fission matrix. The confidence interval for the estimated keff can be conservatively derived from the variance in the fission matrix. This was confirmed by numerical test calculations of Whitesides's ``keff of the world problem'' model where other Monte Carlo methods fail to estimate the confidence interval correctly unless a large number of inactive cycles is simulated. Another problem is that the existing Monte Carlo criticality codes are not well shaped for parallel computations; they cannot fully utilise the processing power of modern multi-processor computers and computer clusters. This thesis presents a new parallel computing scheme for Monte Carlo criticality calculations based on the fission matrix. The fission matrix is combined over a number of independent parallel simulations, and the final results are derived by means of the fission matrix. This scheme allows for a practically ideal parallel scaling since no communication among the parallel simulations is required, and no inactive cycles need to be simulated. When the Monte Carlo criticality calculations are sufficiently fast, they will be more commonly applied on complex reactor physics problems, like non-linear steady-state calculations and fuel cycle calculations. This thesis develops an efficient method that introduces thermal-hydraulic and other feedbacks into the numerical model of a power reactor, allowing to carry out a non-linear Monte Carlo analysis of the reactor with steady-state core conditions. The thesis also shows that the major existing Monte Carlo burnup codes use unstable algorithms for coupling the neutronic and burnup calculations; therefore, they cannot be used for fuel cycle calculations. Nevertheless, stable coupling algorithms are known and can be implemented into the future Monte Carlo burnup codes.
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| 4. |
- Dufek, Jan, et al.
(författare)
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Fission matrix based Monte Carlo criticality calculations
- 2009
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Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 36:8, s. 1270-1275
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Tidskriftsartikel (refereegranskat)abstract
- We have described a fission matrix based method that allows to cancel the inactive cycles in Monte Carlo criticality calculations. The fission matrix must be sampled in the course of the Monte Carlo calculation using a space mesh with sufficiently small zones as it causes the fission matrix be insensitive to errors in the initial fission source. The k(eff) and other quantities can be derived by means of the final fission matrix. The confidence interval for the k(eff) estimate can be conservatively determined via the variance in the fission matrix.
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| 5. |
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| 6. |
- Dufek, Jan, et al.
(författare)
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Stability and convergence problems of the Monte Carlo fission matrix acceleration methods
- 2009
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Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 36:10, s. 1648-1651
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Tidskriftsartikel (refereegranskat)abstract
- The Monte Carlo fission matrix acceleration methods aim at accelerating the convergence of the fission source in inactive cycles of Monte Carlo criticality calculations. In practice, however, these methods may corrupt the fission source, or slow down its convergence. These phenomena have not been completely understood so far. We demonstrate the convergence problems, and explain their reasons.
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| 7. |
- Dufek, Jan, et al.
(författare)
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Stochastic Approximation for Monte Carlo Calculation of Steady-State Conditions in Thermal Reactors
- 2006
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Ingår i: Nuclear science and engineering. - 0029-5639 .- 1943-748X. ; 152, s. 274-283
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Tidskriftsartikel (refereegranskat)abstract
- A new adaptive stochastic approximation method for an efficient Monte Carlo calculation of steady-state conditions in thermal reactor cores is described The core conditions that we consider are spatial distributions of power, neutron flux, coolant density, and strongly absorbing fission products like Xe-135. These distributions relate to each other; thus, the steady-state conditions are described by a system of nonlinear equations. When a Monte Carlo method is used to evaluate the power or neutron flux, then the task turns to a nonlinear stochastic root-finding problem that is usually solved in the iterative manner by stochastic optimization methods. One of those methods is stochastic approximation where efficiency depends on a sequence of stepsize and sample size parameters. The stepsize generation is often based on the well-known Robbins-Monro algorithm; however, the efficient generation of the sample size (number of neutrons simulated at each iteration step) was not published yet. The proposed method controls both the stepsize and the sample size in an efficient way; according to the results, the method reaches the highest possible convergence rate.
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| 8. |
- Eriksson, Marcus, et al.
(författare)
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Safety Analysis of Na and Pb-Bi Coolants in Response to Beam Instabilities
- 2003
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Ingår i: UTILISATION AND RELIABILITY OF HIGH POWER PROTON ACCELERATORS, WORKSHOP PROCEEDINGS. - 9264102116 ; , s. 227-236
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Konferensbidrag (refereegranskat)abstract
- A comparative safety study has been performed on sodium vs. lead/bismuth as coolant for accelerator-driven systems. Transient studies are performed for a beam overpower event. We examine a fuel type of recent interest in the research on minor actinide burners, i.e. uranium-free oxide fuel. A strong positive void coefficient is calculated for both sodium and lead/bismuth. This is attributed to the high fraction of americium in the fuel. It is shown that the lead/bismuth-cooled reactor features twice the grace time with respect to fuel or cladding damage compared to the sodium-cooled reactor of comparable core size and power rating. This accounts to the difference in void reactivity contribution and to the low boiling point of sodium. For improved safety features the general objective is to reduce the coolant void reactivity effect. An important safety issue is the high void worth that could possibly drive the system to prompt criticality.
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| 9. |
- Gottlieb, C., et al.
(författare)
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Feasibility study on transient identification in nuclear power plants using support vector machines
- 2006
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Ingår i: Nuclear Technology. - 0029-5450 .- 1943-7471. ; 155:1, s. 67-77
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Tidskriftsartikel (refereegranskat)abstract
- Support vector machines (SVMs), a relatively new paradigm in statistical learning theory, are studied for their potential to recognize transient behavior of detector signals corresponding to various accident events at nuclear power plants (NPPs). Transient classification is a major task for any computer-aided system for recognition of various malfunctions. The ability to identify the state of operation or events occurring at an NPP is crucial so that personnel can select adequate response actions. The Modular Accident Analysis Program, version 4 (MAAP4) is a program that can be used to model various normal and abnormal events in an NPP. This study uses MAAP signals describing various loss-of-coolant accidents in boiling water reactors. The simulated sensor readings corresponding to these events have been used to train and test SVM classifiers. SVM calculations have demonstrated that they can produce classifiers with good generalization ability for our data. This in, turn indicates that SVMs show promise as classifiers for the learning problem of identifying transients.
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| 10. |
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| 11. |
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| 12. |
- Gudowski, Waclaw, et al.
(författare)
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Review of the European project - Impact of Accelerator-Based Technologies on Nuclear Fission Safety (IABAT)
- 2001
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Ingår i: Progress in nuclear energy (New series). - 0149-1970 .- 1878-4224. ; 38:1-2, s. 135-151
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Tidskriftsartikel (refereegranskat)abstract
- The IABAT project - Impact of Accelerator Based Technologies on Nuclear Fission Safety - started in 1996 in the frame of 4(th) Framework Programme of the European Union, R&D specific programme Nuclear fission safety 1994-1998, area A.2 Exploring innovative approaches/Fuel cycle concepts, as one of the first common European activities in ADS. The project was completed October 31, 1999. The overall objective of the IABAT project has been a preliminary assessment of the potential of Accelerator-Driven Systems (ADS) for transmutation of nuclear waste and for nuclear energy production with minimum waste generation. Moreover, more specific topics related to nuclear data and code development for ADS have been studied in more detail. Four ADSs have been studied for different fuel/coolant combinations: liquid metal coolant and solid fuel, liquid metal coolant and dispersed fuel, and fast and thermal molten salt systems. Target studies comprised multiple target solutions and radiation damage problems in a target environment. In a tool development part of the project a methodology of subcriticality monitoring has been developed based on Feynman-alpha and Rossi-alpha methods. Moreover, a new Monte-Carlo burnup code taking full advantage of continuous neutron cross-section data has been developed and benchmarked. Impact on the risk from high-level waste repositories fi om radiotoxicity reduction using ADS has been assessed giving no crystal-clear benefits of ADS for repository radiotoxicity reduction but concluding some important prerequisites for effective transmutation. In proliferation studies important differences between critical reactors and ADS have been underlined and non-proliferation measures have been proposed. In assessment of accelerator technology costing models have been created that allow the circular and linear accelerator options to be compared and the effect of parameter variations examined. The calculations reported show that cyclotron systems would be more economical, due mainly to the advantage of the cost of RF power supplies. However, the accelerator community regards with skepticism the possibility of transporting and extracting more than a 10mA beam current from a 1GeV cyclotron and therefore technical factors may limit the application of cyclotrons. Finally, this review summarizes development of nuclear data in the energy region between 20 Mev and 150 MeV. Neutron and proton transport data files for Fe, Ni, Pb, Th, U-238 and Pu-239 have been created. The high-energy part of the data files consists completely of results from model calculations, which are benchmarked against the available experimental data. Although there is obviously future work left regarding fine-tuning of several parts of the data files, the representation of nuclear reaction information up to 150 MeV is already better than can be attained with intranuclear cascade codes.
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| 13. |
- Gudowski, Waclaw, et al.
(författare)
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The Subcritical Assembly in Dubna (SAD) - Part II : Research program for ADS-demo experiment
- 2006
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Ingår i: Nuclear Instruments and Methods in Physics Research Section A. - : Elsevier BV. - 0168-9002 .- 1872-9576. ; 562:2, s. 887-891
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Tidskriftsartikel (refereegranskat)abstract
- Subcritical Assembly in Dubna (SAD), a project funded by the International Science and Technology Centre, driven in collaboration with many European partners, may become the first Accelerator Driven Subcritical experiment coupling an existing proton accelerator of 660 MeV with a compact MQX-fuelled subcritical core. The main objective of the SAD experiment is to study physics of Accelerator Driven System ranging from a very deep subcriticality up to k(eff) of 0.98. All experiences with subcriticality monitoring from previous subcritical experiments like MUSE, Yalina and IBR-30 booster mode will be verified in order to select the most reliable subcriticality monitoring technique. Particular attention will be given to validation of the core power-beam current relation. Moreover, some studies have been done to assess possibility of power upgrade for SAD.
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| 14. |
- Gudowski, Waclaw
(författare)
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Transmutation of isotopes - Ecological and energy production aspects
- 2000
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Ingår i: Acta Physica Polonica B. - 0587-4254 .- 1509-5770. ; 31:1, s. 107-122
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Tidskriftsartikel (refereegranskat)abstract
- This paper describes principles of Accelerator-Driven Transmutation of Nuclear Wastes (ATW) and gives some flavour of the most important topics which are today under investigations in many countries. Pin assessment of the potential impact of ATW on a future of nuclear energy is also given. Nuclear reactors based on self-sustained fission reactions - after spectacular development in fifties and sixties, that resulted in deployment of over 400 power reactors - are wrestling today more with public acceptance than with irresolvable technological problems. In a whole spectrum of reasons which resulted in today's opposition against nuclear power few of them are very relevant for the nuclear physics community and they arose from the fact that development of nuclear power had been handed over to the nuclear engineers and technicians with some generically unresolved problems, which should have been solved properly by nuclear scientists. In a certain degree of simplification one can say, that most of the problems originate from very specific features of a fission phenomenon: self-sustained chain reaction in fissile materials and very strong radioactivity of fission products and very long half-life of some of the fission and activation products. And just this enormous concentration of radioactive fission products in the reactor core is the main problem of managing nuclear reactors: it requires unconditional guarantee for the reactor core integrity in order to avoid radioactive contamination of the environment; it creates problems to handle decay heat in the reactor core and finally it makes handling and/or disposal of spent fuel almost a philosophical issue, due to unimaginable long time scales of radioactive decay of some isotopes. A lot can be done to improve the design of conventional nuclear reactors (like Light Water Reactors); new, better reactors can be designed but it seems today very improbable to expect any radical change in the public perception of conventional nuclear power. In this context a lot of hopes and expectations have been expressed for novel systems called Accelerator-Driven Systems, Accelerator-Driven Transmutation of Waste or just Hybrid Reactors. All these names are used for description of the same nuclear system combining a powerful particle accelerator with a subcritical reactor. A careful analysis of possible environmental impact of ATW together with limitation of this technology is presented also in this paper.
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| 15. |
- Gudowski, Waclaw
(författare)
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Transmutation of nuclear waste
- 2000
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Ingår i: Nuclear Physics A. - 0375-9474 .- 1873-1554. ; 663, s. 169C-182C
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Tidskriftsartikel (refereegranskat)
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| 16. |
- Henriksson, Hans, et al.
(författare)
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Long-term operation of the Swedish Centre for Nuclear Technology (SKC): New challenges and solutions in competence building
- 2016
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Ingår i: Proc. European Conf. Nuclear Education and Training (NESTet 2016), Berlin, Germany, May 22-26, 2016.
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Konferensbidrag (refereegranskat)abstract
- The Swedish Centre for Nuclear Technology (Svenskt Kärntekniskt Centrum, SKC) is a national initiative to perform industry relevant research at Swedish universities, and to support dedicated education of direct use to the Swedish nuclear industry.SKC has been the meeting point between industry and academia for 25 years, and has coped with varying needs from industry and political situations. The present situation in the Nordic countries is split: Sweden plans to shut down four out of ten reactors by 2020, while Finland is planning and constructing new reactors. Even without a strong signal to construct new reactors in Sweden, the need for nuclear competence will stay, as we have challenges in front of us to operate and dismantle power plants, operate the intermediate storage facility CLAB in Oskarshamn, and to build and fill the final repositories in Forsmark.The education supported by the SKC at the selected universities will facilitate possible recruitment for nuclear installations. The funding body of SKC consists of all the Swedish Nuclear Power Plants (NPP) situated in Forsmark (three BWRs), Oskarshamn (three BWRs) and Ringhals (one BWR and three PWRs), and the nuclear fuel manufacturer (Westinghouse), while the main research and education is carried out at Chalmers, KTH Royal Institute of Technology and Uppsala University with corresponding in-kind contribution. The research activities cover highly requested studies for today’s nuclear fleet: material embrittlement, stress-corrosion cracking, accident-tolerant fuel development and ageing management for long-term operation (LTO), while the educational part consists of Bachelor and Master programmes as well as elective courses for students outside the main nuclear programmes and contract education.In the master programmes, focus is on e-learning platforms for courses and examination. Examples from such development are to be presented in the full contribution to the conference. Another success story is project based courses in industry, especially within the Bachelor programmes. This is highly appreciated by students, providing a direct contact with future employers, The success of SKC originates from close contact between the funding bodies and academia on many levels: base funding for course preparation, project support in annual calls, and dedicated long-term research funding. But also dedicated industrial experts, and an enthusiasm from academia to enlighten present research issues, as well as strong presence at universities during student fairs and career days. That is LTO of SKC!
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| 17. |
- Ignatyuk, A. V., et al.
(författare)
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Neutron and proton cross-section evaluations for Th-232 up to 150 MeV
- 2002
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Ingår i: Nuclear science and engineering. - 0029-5639 .- 1943-748X. ; 142:2, s. 177-194
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Tidskriftsartikel (refereegranskat)abstract
- Investigations aimed at the development of neutron and proton cross-section evaluations for Th-232 at intermediate energies in the range of 0 to 200 MeV are described The coupled-channel optical model has been used to calculate the neutron total, elastic, and reaction cross sections and the elastic scattering angular distributions. Evaluations of the neutron and charged particle emission cross sections and of the fission cross sections have been obtained on the basis of the statistical description that includes direct, preequilibrium, and equilibrium mechanisms of nuclear reactions. The Kalbach parameterization of angular distributions has been used to describe the double-differential cross sections of emitted neutrons and charged particles in ENDF/B-VI format.
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| 18. |
- Ignatyuk, A. V., et al.
(författare)
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Neutron cross-section evaluations for U-238 up to 150 MeV
- 2000
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Ingår i: Nuclear science and engineering. - 0029-5639 .- 1943-748X. ; 136:3, s. 340-356
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Tidskriftsartikel (refereegranskat)abstract
- Investigations aimed at the development of neutron cross-section evaluations for U-238 at intermediate energies are briefly described. The coupled-channels optical model is used to calculate the neutron total, the elastic and reaction cross sections, and the elastic-scattering angular distributions. Evaluations of the neutron and charged particle emission cross sections and of the fission cross sections are obtained on the basis of the statistical description that includes direct, preequilibrium, and equilibrium mechanisms of nuclear reactions. The Kalbach parameterization of angular distributions is used to describe the double-differential cross sections of emitted neutrons and charged particles in ENDF/B-VI format.
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| 19. |
- Persson, Carl Magnus, et al.
(författare)
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Analysis of reactivity determination methods in the subcritical experiment Yalina
- 2005
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Ingår i: Nuclear Instruments and Methods in Physics Research Section A. - : Elsevier BV. - 0168-9002 .- 1872-9576. ; 554:1-3, s. 374-383
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Tidskriftsartikel (refereegranskat)abstract
- Different reactivity determination methods have been investigated, based on experiments performed at the subcritical assembly Yalina in Minsk, Belarus. The development of techniques for on-line monitoring of the reactivity level in a future accelerator-driven system (ADS) is of major importance for safe operation. Since an ADS is operating in a subcritical mode, the safety margin to criticality must be sufficiently large. The investigated methods are the Slope Fit Method, the Sjostrand Method and the Source Jerk Method. The results are compared with Monte Carlo simulations performed with different nuclear data libraries. The results of the Slope Fit Method are in good agreement with the Monte Carlo simulation results, whereas the Sjostrand Method appears to underestimate the criticality somewhat. The Source Jerk Method is subject to inadequate statistical accuracy.
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| 20. |
- Persson, C. -M, et al.
(författare)
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Comparison of neutron kinetic parameters of the subcritical ads experiments Yalina and Yalina Booster
- 2005
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Konferensbidrag (refereegranskat)abstract
- Subcritical experiments, devoted to investigation of characteristics of accelerator-driven systems, have been constructed at the Joint Institute for Power and Nuclear Research - Sosny in Minsk, Belarus. Different methods for reactivity determination have previously been investigated in the thermal ADS experiment "Yalina", and recently, a coupled fast-thermal facility "Yalina Booster" was launched. This study presents the neutron kinetic characteristics of the Yalina and the new Yalina Booster setups, and points out some important differences. For the Yalina setup, neutron kinetic parameters, such as keff, α, βeff and Λ have been determined by Monte Carlo simulations and they have previously been verified experimentally. For Yalina Booster, these parameters have been estimated by Monte Carlo simulations in a preliminary study, and they will be verified in upcoming experiments.
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| 21. |
- Persson, Carl-Magnus, et al.
(författare)
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Pulsed neutron source measurements in the subcritical ADS experiment YALINA-Booster
- 2008
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Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 35:12, s. 2357-2364
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Tidskriftsartikel (refereegranskat)abstract
- A subcritical zero-power source-driven coupled core, the YALINA-Booster. has been constructed for experimental investigations of neutron kinetics of source-driven systems. In this study, the reactivity of two subcritical configurations has been determined by the area ratio method. The prompt neutron decay constants have been evaluated through slope fitting of the prompt neutron decay as well as through the pulsed Rossi-alpha method. It is shown that the slope fitting method and the pulsed Rossi-alpha method give stable results whereas the area ratio method results show spatial dependence. The reasons for the spatial spread are addressed.
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| 22. |
- Persson, Carl-Magnus, 1980-
(författare)
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Reactivity Assessment in Subcritical Systems
- 2007
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Licentiatavhandling (övrigt vetenskapligt/konstnärligt)abstract
- Accelerator-driven systems have been proposed for incineration of transuranic elements from spent nuclear fuel. For safe operation of such facilities, a robust method for reactivity monitoring is required. In this thesis, the most important existing reactivity determination methods have been evaluated experimentally in the subcritical YALINA-experiments in Belarus. It is concluded that the existing methods are sufficient for calibration purposes, but not for reactivity monitoring during regular operation of an accelerator-driven system. Conditions for successful utilization of the various methods are presented, based on the experimental experience.
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| 23. |
- Persson, Carl-Magnus, et al.
(författare)
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Results from pulsed neutron source measurements in the YALINA-Booster ads experiment
- 2007
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Ingår i: 8th International Topical Meeting on Nuclear Applications and Utilization of Accelerators, ACCAPP'07. - 0894480545 - 9780894480546 ; , s. 924-929
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Konferensbidrag (refereegranskat)abstract
- Two subcritical configurations of the zero-power coupled subcritical core YALINA-Booster have been identified through pulsed neutron source measurements. The area ratio and the slope fitting reactivity estimation methods have been utilized as well as the pulsed Rossi-a noise method. The measurements showed that despite the inhomogeneous two-zone core composition a clear single exponential prompt neutron decay was obtained. Spatial spread of the results and converegence issues related to the area ratio method are addressed.
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| 24. |
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| 25. |
- Plukienė, R., et al.
(författare)
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Transmutation considerations of LWR and RBMK spent nuclear fuel by the fusion–fission hybrid system
- 2018
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Ingår i: Nuclear Engineering and Design. - : Elsevier. - 0029-5493 .- 1872-759X. ; 330, s. 241-249
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Tidskriftsartikel (refereegranskat)abstract
- The performance of the fusion–fission hybrid system based on the molten salt (flibe) blanket, driven by a plasma based fusion device, was analyzed by comparing transmutation scenarios of actinides extracted from the LWR (Sweden) and RBMK (Lithuania) spent nuclear fuel in the scope of the EURATOM project BRILLIANT. The IAEA nuclear fuel cycle simulation system (NFCSS) has been applied for the estimation of the approximate amount of heavy metals of the spent nuclear fuel in Sweden reactors and the SCALE 6 code package has been used for the determination of the RBMK-1500 spent nuclear fuel composition. The total amount of trans-uranium elements has been estimated in both countries by 2015. Major parameters of the hybrid system performance (e.g., kscr, keff, Φn(E), equilibrium conditions, etc.) have been investigated for LWR and RBMK trans-uranium transmutation cases. Detailed burn-up calculations with continuous feeding to replenish the incinerated trans-uranium material and partial treatment of fission products were done using the Monteburns (MCNP + ORIGEN) code system. About 1.1 tons of spent fuel trans-uranium elements could be burned annually with an output of the 3 GWth fission power, but the equilibrium stage is reached differently depending on the initial trans-uranium composition. The radiotoxicity of the remaining LWR and RBMK transmuted waste after the hybrid system operation time has been estimated.
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| 26. |
- Polanski, A., et al.
(författare)
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Power upgrade of the subcritical assembly in Dubna (SAD) to 100 kW
- 2006
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Ingår i: Nuclear Instruments and Methods in Physics Research Section A. - : Elsevier BV. - 0168-9002 .- 1872-9576. ; 562:2, s. 879-882
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Tidskriftsartikel (refereegranskat)abstract
- The paper present results of Monte Carlo modeling of an Experimental Accelerator Driven System (ADS), which employs a subcritical assembly and a 660 MeV proton accelerator operating at the Laboratory of Nuclear Problems of the Joint Institute for Nuclear Research in Dubna. The mix of oxides PuO2 + UO2 MOX fuel designed for the reactor will be adopted for the core of the assembly. The design of the experimental subcritical assembly in Dubna (SAD) is based on the core with a nominal unit capacity of 30 kW (thermal). This corresponds to the multiplication coefficient K-eff = 0.95 and the accelerator beam power of I kW. A subcritical assembly has been modeled in order to increase power of this experimental set up. Different options for the target and fuel elements have been considered.
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| 27. |
- Seltborg, Per, et al.
(författare)
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Definition and application of proton source efficiency in accelerator driven systems
- 2003
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Ingår i: Nuclear science and engineering. - 0029-5639 .- 1943-748X. ; 145:3, s. 390-399
-
Tidskriftsartikel (refereegranskat)abstract
- In order to study the beam power amplification of an accelerator-driven system (ADS), a new parameter, the proton source efficiency psi* is introduced. psi* represents the average importance of the external proton source, relative to the average importance of the eigenmode production, and is closely related to the neutron source efficiency rho*, which is frequently used in the ADS field. rho* is commonly used in the physics of subcritical systems driven by any external source (spallation source, (d,d), (d, t), Cf-252 spontaneous fissions, etc.). On the contrary, psi* has been defined in this paper exclusively for ADS studies where the system is driven by a spallation source. The main advantage with using psi* instead of rho* for ADS is that the way of defining the external source is unique and that it is proportional to the core power divided by the proton beam power, independent of the neutron source distribution. Numerical simulations have been performed with the Monte Carlo code MCNPX in order to study psi* as a function of different design parameters. It was found that, in order to maximize psi* and therefore minimize the proton current needs, a target radius as small as possible should be chosen. For target radii smaller than similar to30 cm, lead-bismuth is a better choice of coolant material than sodium, regarding the proton source efficiency, while for larger target radii the two materials are equally good. The optimal axial proton beam impact was found to be located similar to 20 cm above the core center. Varying the proton energy, psi*/E-p was found to have a maximum for proton energies between 1200 and 1400 MeV Increasing the americium content in the fuel decreases psi* considerably, in particular when the target radius is large.
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| 28. |
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| 29. |
- Seltborg, Per, et al.
(författare)
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Impact of heterogeneous Cm-distribution on proton source efficiency in accelerator-driven systems
- 2004
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Ingår i: Proceedings of the PHYSOR 2004. - 0894486837 ; , s. 489-497
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Konferensbidrag (refereegranskat)abstract
- The proton source efficiency (ψ*) was studied for homogeneous and heterogeneous distributions of minor actinides in a nitride-fuelled and lead-bismuth-cooled accelerator-driven system. The findings from the MCNPX simulations indicate that, compared to a homogeneous configuration, a gain in ψ* by up to 16% can be obtained by distributing the minor actinides heterogeneously, Cm being placed in the inner zone of the active core and Am in the outer zone. The reason for this is the higher fission probability for neutrons for Cm than for Am in the energy range below 1.0 MeV. Moreover, a comparative study of two different physics packages available in MCNPX, the Bertini and the CEM models, has been performed, focusing on the production of neutrons in the spallation target and on the proton source efficiency. The Bertini model was found to produce a higher number of neutrons in the low-energy range (below ∼15 MeV) than the CEM model. Consequently, the Bertini model also over-estimates ψ* by about 10%, compared to the CEM model.
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| 30. |
- Seltborg, P., et al.
(författare)
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Investigation of radiation fields outside the sub-critical assembly in Dubna
- 2005
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Ingår i: Radiation Protection Dosimetry. - : Oxford University Press (OUP). - 0144-8420 .- 1742-3406. ; 116:1-4, s. 449-453
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Tidskriftsartikel (refereegranskat)abstract
- The radiation fields outside the planned experimental Sub-critical Assembly in Dubna (SAD) have been studied in order to provide a basis for the design of the concrete shielding that cover the reactor core. The effective doses around the reactor, induced by leakage of neutrons and photons through the shielding, have been determined for a shielding thickness varying from 100 to 200 cm. It was shown that the neutron flux and the effective dose is higher above the shielding than at the side of it, owing to the higher fraction of high-energy spallation neutrons emitted in the direction of the incident beam protons. At the top, the effective dose was found to be similar to 150 mu Sv s(-1) for a concrete thickness of 100 cm, while similar to 2.5 mu Sv s(-1) for a concrete thickness of 200 cm. It was also shown that the high-energy neutrons (> 10 MeV), which are created in the proton-induced spallation interactions in the target, contribute for the major part of the effective doses outside the reactor.
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| 31. |
- Seltborg, Per, et al.
(författare)
-
Radiation shielding of high-energy neutrons in SAD
- 2005
-
Ingår i: Nuclear Instruments and Methods in Physics Research Section A. - : Elsevier BV. - 0168-9002 .- 1872-9576. ; 550, s. 313-328
-
Tidskriftsartikel (refereegranskat)abstract
- The radiation fields and the effective dose at the Sub-critical Assembly in Dubna (SAD) have been studied with the Monte Carlo code MCNPX. The effective dose above the shielding, i.e. in the direction of the incident proton beam of 3.0 mu A, was found to be about 190 mu Sv h(-1). This value meets the dose limits according to Russian radiation protection regulations, provided that access to the rooms in this area is not allowed for working personnel during operation.By separating the radiation fields into a spallation- and a fission-induced part, it was shown that the neutrons with energy higher than 10MeV, originating exclusively from the proton-induced spallation reactions in the target, contribute for the entire part of the radiation fields and the effective dose at the top of the shielding. Consequently, the effective dose above the SAD reactor system is merely dependent on the proton beam properties and not on the reactivity of the core.
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| 32. |
- Seltborg, Per, 1973-
(författare)
-
Source efficiency and high-energy neutronics in accelerator-driven systems
- 2005
-
Doktorsavhandling (övrigt vetenskapligt/konstnärligt)abstract
- Transmutation of plutonium and minor actinides in accelerator-driven systems (ADS) is being envisaged for the purpose of reducing the long-term radiotoxic inventory of spent nuclear reactor fuel. For this reason, the physics of sub-critical systems are being studied in several different experimental programs across the world. Three of these experiments have been studied within the scope of the present thesis; the MUSE experiments in France, the Yalina experiments in Belarus and the SAD experiments in Russia. The investigations of the MUSE experiments have focused on three different neutronic parameters; the neutron energy spectrum, the external neutron source efficiency and the dynamic neutron source response. It has been shown that the choice of external neutron source has negligible effect on the neutron energy spectrum in the core. Therefore, from this point of view, the MUSE experiments can be considered representative of an ADS. From the analyses of different reactivity determination methods in the Yalina experiments, it can be concluded that the slope fit method gives results in good agreement with the results obtained by the Monte Carlo method MCNP. Moreover, it was found that the Sjöstrand method underestimates keff slightly, in comparison with MCNP and the other investigated methods. In the radiation shielding studies of the SAD experiments, it was shown that the entire part of the effective dose detected at the top of the biological shielding originates from the proton-induced spallation reactions in the target. Thus, it can be concluded that the effective dose is directly proportional to the proton beam power, but independent of the reactivity of the sub-critical core. In order to study the energy gain of an ADS, i.e., the core power divided by the proton beam power, the proton source efficiency, ψ*, has been studied for various ADS models. ψ* is defined in analogy with the neutron source efficiency, φ*, but relates the core power directly to the source protons instead of to the source neutrons. φ* is commonly used in the physics of sub-critical systems, driven by any external neutron source (spallation source, (D,D), (D,T), 252Cf spontaneous fission etc.). On the contrary, ψ* has been defined only for ADS studies, where the system is driven by a proton-induced spallation source. The main advantages of using ψ* instead of φ* are that the way of defining the external source is unique and that ψ* is proportional to the energy gain. An important part of this thesis has been devoted to studies of ψ* as a function of different system parameters, thereby providing a basis for an ADS design with optimal properties for obtaining a high core power over beam power ratio. For instance, ψ* was found to decrease considerably with increasing spallation target radius.
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| 33. |
- Shvetsov, V., et al.
(författare)
-
The Subcritical Assembly in Dubna (SAD) - Part I : Coupling all major components of an Accelerator Driven System (ADS)
- 2006
-
Ingår i: Nuclear Instruments and Methods in Physics Research Section A. - : Elsevier BV. - 0168-9002 .- 1872-9576. ; 562:2, s. 883-886
-
Tidskriftsartikel (refereegranskat)abstract
- A demonstration facility for Accelerator Driven Systems has been proposed to be constructed at the Joint Institute of Nuclear Research in Dubna. The Subcritical Assembly in Dubna project proposes to couple an existing proton accelerator of 660 MeV and 1 mu A current with a specially designed U-Pu MOX subcritical core. Project objectives, technical description and current status of the project are presented in this paper.
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| 34. |
- Talamo, Alberto, et al.
(författare)
-
A deep burn fuel management strategy for the incineration of military plutonium in the gas turbine-modular helium reactor modeled in a detailed three-dimensional geometry by the Monte Carlo continuous energy burnup code
- 2006
-
Ingår i: Nuclear science and engineering. - 0029-5639 .- 1943-748X. ; 153:2, s. 172-183
-
Tidskriftsartikel (refereegranskat)abstract
- In the future development of nuclear energy, the graphite-moderated helium-cooled reactors may play an important role because of their valuable technical advantages: passive safety, low cost, flexibility in the choice of fuel, high conversion energy efficiency, high burnup, more resistant fuel cladding, and low power density. General Atomics possesses a long experience with this type of reactor, and it has recently developed the gas turbine-modular helium reactor (GT-MHR), a design where the nuclear power plant is structured into four reactor modules of 600 MW(thermal). Amid its benefits, the GT-MHR offers a rather large flexibility in the choice of fuel type; Th, U, and Pu may be used in the manufacture of fuel with some degrees of freedom. As a consequence, the fuel management may be designed for different objectives aside from energy production, e.g., the reduction of actinide waste production through a fuel based on thorium. In our previous studies we analyzed the behavior of the GT-MHR with a plutonium fuel based on light water reactor (LWR) waste; in the present study we focused on the incineration of military Pu. This choice of fuel requires a detailed numerical modeling of the reactor since a high value of keff at the beginning of the reactor operation requires the modeling both of control rods and of burnable poison; by contrast, when the GT-MHR is fueled with LWR waste, at the equilibrium of the fuel composition, the reactivity swing is small.
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| 35. |
- Talamo, Alberto, et al.
(författare)
-
Adapting the Deep Burn In-Core Fuel Management Strategy for the Gas Turbine - Modular Helium Reactor to a Uranium-Thorium Fue
- 2005
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 32:16, s. 1750-1781
-
Tidskriftsartikel (refereegranskat)abstract
- In 1966, Philadelphia Electric has put into operation the Peach Bottom I nuclear reactor, it was the first high temperature gas reactor (HTGR); the pioneering of the helium-cooled and graphite-moderated power reactors continued with the Fort St. Vrain and THTR reactors, which operated until 1989. The experience on HTGRs lead General Atomics to design the gas turbine - modular helium reactor (GT-MHR), which adapts the previous HTGRs to the generation IV of nuclear reactors. One of the major benefits of the GT-MHR is the ability to work on the most different types of fuels: light water reactors waste, military plutonium, MOX and thorium. In this work, we focused on the last type of fuel and we propose a mixture of 40% thorium and 60% uranium. In a uranium-thorium fuel, three fissile isotopes mainly sustain the criticality of the reactor: U-235, which represents the 20% of the fresh uranium, U-233, which is produced by the transmutation of fertile Th-212, and Pu-239, which is produced by the transmutation of fertile U-238. In order to compensate the depletion of U-235 with the breeding of U-233 and Pu-239, the quantity of fertile nuclides must be much larger than that one of 235 U because of the small capture cross-section of the fertile nuclides, in the thermal neutron energy range, compared to that one of U-235. At the same time, the amount of U-235 must be large enough to set the criticality condition of the reactor. The simultaneous satisfaction of the two above constrains induces the necessity to load the reactor with a huge mass of fuel; that is accomplished by equipping the fuel pins with the JAERI TRISO particles. We start the operation of the reactor with loading fresh fuel into all the three rings of the GT-MHR and after 810 days we initiate a refueling and shuffling schedule that, in 9 irradiation periods, approaches the equilibrium of the fuel composition. The analysis of the k(eff) and mass evolution, reaction rates, neutron flux and spectrum at the equilibrium of the fuel composition, highlights the features of a deep burn in-core fuel management strategy for a uranium-thorium fuel.
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| 36. |
|
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| 37. |
- Talamo, Alberto, 1971-
(författare)
-
Advanced In-Core Fuel Cycles for the Gas Turbine-Modular Helium Reactor
- 2006
-
Doktorsavhandling (övrigt vetenskapligt/konstnärligt)abstract
- In 1789 a German chemist, Martin Heinrich Klaproth, announced the discovery of a new element: uranium; few years later, the head of father of the modern chemistry, Antoine Lavoisier, was swept away by guillotine: a new era was destined to be opened, either where energy would have been produced in large scale by nuclear processes delivering hundreds of times the energy of chemical processes or where a mass of people, revolutionary or not, would have been melted down into a couple of seconds. After a quite long time, on the 2nd December 1942, the first nuclear reactor has been put into operation by Enrico Fermi in Chicago; few years later, came also the dark side utilization of fissile materials in Hiroshima and Nagasaki. Since those moments, three power plants generations succeeded, until the current one which is the generation IV of nuclear reactors. The latter has the goal of generating electricity in a safe manner, for the core is designed to provide an effective passive cooling of the decay heat. Amid generation IV of nuclear power plants, the Gas Turbine – Modular Helium Reactor, designed by General Atomics, is the only core with an energy conversion efficiency of 50%; the above consideration, coupled to construction and operation costs lower than ordinary Light Water Reactors, renders the Gas Turbine – Modular Helium reactor rather unequaled.In the present studies we investigated the possibility to operate the GT-MHR with two types of fuels: LWRs waste and thorium; since thorium is made of only fertile 232Th, we tried to mix it with pure 233U, 235U or 239Pu; ex post facto, only uranium isotopes allow the reactor operation, that induced us to examine the possibility to use a mixture of uranium, enriched 20% in 235U, and thorium. We performed all calculations by the MCNP and MCB codes, which allowed to model the reactor in a very detailed threedimensional geometry and to describe the nuclides transmutation in a continuous energy approach; finally, we completed our studies by verifying the influence of the major nuclear data libraries, JEFF, JENDL and ENDF/B, on the obtained results.
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| 38. |
- Talamo, Alberto, et al.
(författare)
-
Comparative Studies of ENDF/B-6.8, JEF-2.2 and JENDL-3.2 Data Libraries by Monte Carlo Modeling of High Temperature Reactors on Plutonium Based Fuel Cycles
- 2004
-
Ingår i: Journal of Nuclear Science and Technology. - 0022-3131 .- 1881-1248. ; 41:12, s. 1228-1236
-
Tidskriftsartikel (refereegranskat)abstract
- We performed a numerical comparative analysis of the burnup capability of the Gas Turbine-Modular Helium Reactor (GT-MHR) by the Monte Carlo Continuous Energy Burnup Code (MCB). The MCB code is an extension of MCNP that includes the burnup implementation; it adopts continuous energy cross sections and it evaluates the transmutation trajectories for over 2,400 decaying nuclides. We equipped the MCB code with three different nuclear data libraries: JENDL-3.2, JEF-2.2 and ENDF/B-6.8 processed for temperatures from 300 to 1,800 K. The GT-MHR model studied in this paper is fueled by actinides coming from the Light Water Reactors waste, converted into two different types of fuel: Driver Fuel and Transmutation Fuel. The Driver Fuel supplies the fissile nuclides needed to maintain the criticality of the reactor, whereas the Transmutation Fuel depletes non-fissile isotopes and controls reactivity excess. We set the refueling and shuffling period to one year and the in-core fuel residency time to three years. The comparative analysis of the MCB code consists of accuracy and precision studies. In the accuracy studies, we performed the burnup calculation with different nuclear data libraries during the year at which the refueling and shuffling schedule set the equilibrium of the fuel composition. In the precision studies, we repeated the same simulations 20 times with a different pseudorandom number stride and the same nuclear data library.
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| 39. |
- Talamo, Alberto, et al.
(författare)
-
Comparative Studies of JENDL-3.3, JENDL-3.2, JEFF-3, JEF-2.2 and ENDF/B-6.8 Data Libraries on the Monte Carlo Continuous Energy Modeling of the Gas Turbine - Modular Helium Reactor Operating with Thorium Fuel
- 2005
-
Ingår i: Journal of Nuclear Science and Technology. - 0022-3131 .- 1881-1248. ; 42:12, s. 1040-1053
-
Tidskriftsartikel (refereegranskat)abstract
- One of the major benefits of the Gas Turbine-Modular Helium Reactor is the capability to operate with several different types of fuel; either Light Water Reactors waste, military plutonium or thorium represent valid candidates as possible types of fuel. In the present studies, we performed a comparison of various nuclear data libraries by the Monte Carlo Continuous Energy Burnup Code MCB applied to the Gas Turbine-Modular Helium Reactor operating on a thorium fuel. A thorium fuel offers valuable attractive advantages: low fuel cost, high reduction of actinides production and the possibility to enable the reactor to act as a breeder of fuel by the neutron capture of fertile Th-232. We evaluated the possibility to mix thorium with small quantities, about 3% in atomic composition, of Pu-239, U-233 and U-235. The mass of thorium must be much larger than that one of plutonium or uranium because of the low capture cross section of thorium compared to the fission one of the fissile nuclides; at the same time, the quantity of the fissile isotopes must grant the criticality condition. These two simultaneous constraints force to load a huge mass of fuel in the reactor; consequently, we propose to allocate the fuel in TRISO particles with a large radius of the kernel. For each of the three different fuels we calculated the evolution of the fuel composition by the MCB code equipped with five different nuclear data libraries: JENDL-3.3, JENDL-3.2, JEFF-3, JEF-2.2 and ENDF/B.
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| 40. |
- Talamo, Alberto, et al.
(författare)
-
Comparative studies of jendl-3.3, jendl-3.2, jeff-3, jef-2.2 and endf/b-6.8 data libraries on the monte carlo continuous energy modeling of the gas turbine-modular helium reactor operating with thorium fuels
- 2005
-
Ingår i: Journal of Nuclear Science and Technology. - : Informa UK Limited. - 0022-3131 .- 1881-1248. ; 42:12, s. 1040-1053
-
Tidskriftsartikel (refereegranskat)abstract
- One of the major benefits of the Gas Turbine-Modular Helium Reactor is the capability to operate with several different types of fuel; either Light Water Reactors waste, military plutonium or thorium represent valid candidates as possible types of fuel. In the present studies, we performed a comparison of various nuclear data libraries by the Monte Carlo Continuous Energy Burnup Code MCB applied to the Gas Turbine-Modular Helium Reactor operating on a thorium fuel. A thorium fuel offers valuable attractive advantages: low fuel cost, high reduction of actinides production and the possibility to enable the reactor to act as a breeder of fuel by the neutron capture of fertile 232Th. We evaluated the possibility to mix thorium with small quantities, about 3% in atomic composition, of 239Pu, 233U and 235U. The mass of thorium must be much larger than that one of plutonium or uranium because of the low capture cross section of thorium compared to the fission one of the fissile nuclides; at the same time, the quantity of the fissile isotopes must grant the criticality condition. These two simultaneous constraints force to load a huge mass of fuel in the reactor; consequently, we propose to allocate the fuel in TRISO particles with a large radius of the kernel. For each of the three different fuels we calculated the evolution of the fuel composition by the MCB code equipped with five different nuclear data libraries: JENDL-3.3, JENDL-3.2, JEFF-3, JEF-2.2 and ENDF/B.
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| 41. |
- Talamo, Alberto, et al.
(författare)
-
Comparison of MCB and MONTEBURNS Monte Carlo burnup codes on a one-pass deep burn
- 2006
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 33:14-15, s. 1176-1188
-
Tidskriftsartikel (refereegranskat)abstract
- Numerical applications implemented on the Monte Carlo method have developed in line with the increase of computer power; nowadays, in the field of nuclear reactor physics, it is possible to perform burnup simulations in a detailed 3D geometry and a continuous energy description by the Monte Carlo method; moreover, the required computing time can be abundantly reduced by taking advantage of a computer cluster. In this paper we focused on comparing the results of the two major Monte Carlo burnup codes, MONTEBURNS and MCB, when they share the same MCNP geometry, nuclear data library, core thermal power, and they apply the same refueling and shuffling schedule. While simulating a total operation time of the Gas Turbine-Modular Helium Reactor of 2100 effective full power days and a one-pass deep burn in-core fuel management schedule, we have found that the two Monte Carlo codes produce very similar results both on the criticality value of the core and the transmutation of the key actinides.
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| 42. |
|
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| 43. |
- Talamo, Alberto, et al.
(författare)
-
Incineration of light water reactor waste in high-temperature gas reactors : Axial fuel management and efficiency of americium and curium transmutation
- 2007
-
Ingår i: Nuclear science and engineering. - 0029-5639 .- 1943-748X. ; 156:2, s. 244-266
-
Tidskriftsartikel (refereegranskat)abstract
- In the present study we investigate the influence of the fuel axial shuffling and the operational control rod maneuvering on the performances of the one-pass (no reprocessing) deep-burn incineration of light water reactor waste in the gas turbine-modular helium reactor. After an irradiation period, the fuel axial shuffling schedule has to take into account the fuel depletion profile generated by the adjustments of the position of the operational control rods, because the insertion of the rods strongly alters the neutron flux shape. We aimed at implementing a numerical simulation as close as possible to a real scenario and therefore took advantage of the powerful geometrical modeling capability of the MCB code to describe the reactor in a detailed three-dimensional geometry model in which we simulated over 120 different burnable materials, each of them undergoing a different neutron flux intensity. We adjusted the position of the control rods every 90 effective full-power days of irradiation to maintain the core as close as possible to the critical condition; thereafter, we recalculated the neutron flux and cross sections by a new MCNP/ MCB run. At the present time, this sophisticated approach can be realized only by a computer cluster of ten 64-bit processors working in parallel mode. The fuel axial shuffling adds from 3 to 5% to the transmutation rates of 239Pu, plutonium, and all actinides, which range from 80 to 86, 50 to 53, and 46 to 48%, respectively; the present results are 5 to 14% less compared to the case of a two-pass (reprocessing) deep burn. The efficiency of transmuting minor actinides has been estimated by comparing the long-term radio-toxicity of the fresh and irradiated americium and curium fuel; this comparison revealed that it is not worthwhile to transmute americium and curium in the current design of the gas turbine-modular helium reactor by a one-pass deep burn.
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| 44. |
- Talamo, Alberto, et al.
(författare)
-
Key Physical Parameters and Temperature Reactivity Coefficients of the Deep Burn Modular Helium Reactor Fueled with LWRs Waste
- 2004
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 31, s. 1913-1931
-
Tidskriftsartikel (refereegranskat)abstract
- We investigated some important neutronic features of the deep burn modular helium reactor (DB-MHR) using the MCNP/MCB codes. Our attention was focused on the neutron flux and its spectrum, capture to fission ratio of Pu-239 and the temperature coefficient of fuel and moderator. The DB-MHR is a graphite-moderated helium-cooled reactor proposed by General Atomic to address the need for a fast and efficient incineration of plutonium for nonproliferation purposes as well as the management of light water reactors (LWRs) waste. In fact, recent studies have shown that the use of the DB-MHR coupled to ordinary LWRs would keep constant the world inventory of plutonium for a reactor fleet producing 400 TWe/y. In the present studies, the DB-MHR is loaded with Np-Pu driver fuel (DF) with an isotopic composition corresponding to LWRs spent fuel waste. DF uses fissile isotopes (e.g. Pu-239 and Pu-241), previously generated in the LWRs, and maintains criticality conditions in the DB-MHR. After an irradiation of three years, the spent DF is reprocessed and its remaining actinides are manufactured into fresh transmutation fuel (TF). TF mainly contains non-fissile actinides which undergo neutron capture and transmutation during the subsequent three-year irradiation in the DB-MHR. At the same time, TF provides control and negative reactivity feedback to the reactor. After extraction of the spent TF, irradiated for three years, over 94% of Pu-239 and 53% of all actinides coming from LWRs waste will have been destroyed in the DB-MHR. In this paper we look at the operation conditions at equilibrium for the DB-MHR and evaluate fluxes and reactivity responses using state of the art 3-D Monte Carlo simulations.
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| 45. |
- Talamo, A., et al.
(författare)
-
MCB1C2 bug on thermal reactors
- 2006
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 33:7, s. 653-654
-
Tidskriftsartikel (refereegranskat)
|
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| 46. |
- Talamo, Alberto, et al.
(författare)
-
Performance of the Gas Turbine – Modular Helium Reactor fuelled with different types of fertile TRISO particles
- 2005
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 32:16, s. 1719-1749
-
Tidskriftsartikel (refereegranskat)abstract
- Preliminary studies have been performed on operation of the gas turbine-modular helium reactor (GT-MHR) with a thorium based fuel. The major options for a thorium fuel are a mixture with light water reactors spent fuel, mixture with military plutonium or with with fissile isotopes of uranium. Consequently, we assumed three models of the fuel containing a mixture of thorium with 239Pu, 233U or 235U in TRISO particles with a different kernel radius keeping constant the packing fraction at the level of 37.5%, which corresponds to the current compacting process limit. In order to allow thorium to act as a breeder of fissile uranium and ensure conditions for a self-sustaining fission chain, the fresh fuel must contain a certain quantity of fissile isotope at beginning of life; we refer to the initial fissile nuclide as triggering isotope. The small capture cross-section of 232Th in the thermal neutron energy range, compared to the fission one of the common fissile isotopes (239Pu, 233U and 235U), requires a quantity of thorium 25-30 times greater than that one of the triggering isotope in order to equilibrate the reaction rates. At the same time, the amount of the triggering isotope must be enough to set the criticality condition of the reactor. These two conditions must be simultaneously satisfied. The necessity of a large mass of fuel forces to utilize TRISO particles with a large radius of the kernel, 300 μm. Moreover, in order to improve the neutron economics, a fuel cycle based on thorium requires a low capture to fission ratio of the triggering isotope. Amid the common fissile isotopes, 233U, 235U and 239Pu, we have found that only the uranium nuclides have shown to have the suitable neutronic features to enable the GT-MHR to work on a fuel based on thorium.
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| 47. |
|
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| 48. |
- Talamo, Alberto, et al.
(författare)
-
Studies of a deep burn fuel cycle for the incineration of military plutonium in the GT-MHR using the Monte-Carlo burnup code
- 2004
-
Ingår i: Proceedings of the PHYSOR 2004: The Physics of Fuel Cycles and Advanced Nuclear Systems - Global Developments. - 0894486837 ; , s. 433-442
-
Konferensbidrag (refereegranskat)abstract
- The deep burn fuel cycle for the incineration of military plutonium in the GT-MHR is studied using the Monte-Carlo burnup code. The irradiation is DF is so rich in fissile isotopes that the TF cannot guarantee a negative reactive feedback, and the presence of erbium as burnable poison is absolutely necessary for the reactivity safety reasons. At beginning of life (BOL) the fuel composed of DF, consisting of fresh military plutonium, after an irradiation period of three years the fuel is reprocessed into post driver fuel (PDF). The mass flow of the GT-MHR fuelled by military plutonium at the equilibrium of the fuel composition shows that 66% of 239Pu is burned in three years and 92% in six years.
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| 49. |
|
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| 50. |
- Talamo, Alberto, et al.
(författare)
-
The burnup capabilities of the deep burn modular helium reactor analyzed by the Monte Carlo continuous energy code MCB
- 2004
-
Ingår i: Annals of nuclear energy. - 0029-5639. ; 31:2, s. 173-196
-
Tidskriftsartikel (refereegranskat)abstract
- In the future development of nuclear energy, the graphite-moderated helium-cooled reactors may play an important role because of their valuable technical advantages: passive safety, low cost, flexibility in the choice of fuel, high conversion energy efficiency, high burnup, more resistant fuel cladding, and low power density. General Atomics possesses a long experience with this type of reactor, and it has recently developed the gas turbine-modular helium reactor (GT-MHR), a design where the nuclear power plant is structured into four reactor modules of 600 MW(thermal). Amid its benefits, the GT-MHR offers a rather large flexibility in the choice of fuel type; Th, U, and Pu may be used in the manufacture of fuel with some degrees of freedom. As a consequence, the fuel management may be designed for different objectives aside from energy production, e.g., the reduction of actinide waste production through a fuel based on thorium. In our previous studies we analyzed the behavior of the GT-MHR with a plutonium fuel based on light water reactor (LWR) waste; in the present study we focused on the incineration of military Pu. This choice of fuel requires a detailed numerical modeling of the reactor since a high value of keff at the beginning of the reactor operation requires the modeling both of control rods and of burnable poison; by contrast, when the GT-MHR is fueled with LWR waste, at the equilibrium of the fuel composition, the reactivity swing is small.
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