| 1. |
- Demaziere, Christophe, 1973, et al.
(author)
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Theoretical investigation of the MTC noise estimate in 1-D homogeneous systems
- 2002
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In: Annals of Nuclear Energy. - 0306-4549 .- 1873-2100. ; 29:1, s. 75-100
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Journal article (peer-reviewed)abstract
- In this paper, the accuracy of the noise-based determination of the moderator temperature coefficient (MTC) is investigated theoretically and quantitatively. It is known from earlier work that the noise method systematically underestimates the MTC. In this paper, it is found that the main reason for the underestimation lies with the radial incoherence of the temperature fluctuations. The deviation of the reactor response from point-kinetics is another possible reason, but it was found to play a quite insignificant role. The theory of neutron noise, induced by spatially random perturbations is elaborated and by its help the inaccuracy (bias) of the noise based MTC estimation was quantitatively investigated. It was found that a relatively short correlation length of the temperature fluctuations, which is in agreement with experimental evidence, can explain the observed underestimation of the MTC by the noise method.
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| 2. |
- Arzhanov, Vasily
(author)
-
Monotonicity properties of k(eff) with shape change and with nesting
- 2002
-
In: Annals of Nuclear Energy. - 0306-4549 .- 1873-2100. ; 29:2, s. 137-145
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Journal article (peer-reviewed)abstract
- It was found that, contrary to expectations based on physical intuition, k(eff) can both increase and decrease when changing the shape of an initially regular critical system, while preserving its volume. Physical intuition would only allow for a decrease of k(eff) when the surface/volume ratio increases. The unexpected behaviour of increasing k(eff) was found through numerical investigation. For a convincing demonstration of the possibility of the non-monotonic behaviour, a simple geometrical proof was constructed. This latter proof, in turn, is based on the assumption that k(eff) can only increase (or stay constant) in the case of nesting, i.e. when adding extra volume to a system. Since we found no formal proof of the nesting theorem for the general case, we close the paper by a simple formal proof of the monotonic behaviour of k(eff) by nesting.
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| 3. |
- Arzhanov, Vasily
(author)
-
Multi-group theory of neutron noise induced by vibrating boundaries
- 2002
-
In: Annals of Nuclear Energy. - 0306-4549 .- 1873-2100. ; 29:18, s. 2143-2158
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Journal article (peer-reviewed)abstract
- The paper extends the one-group analysis of the neutron noise induced by fluctuating boundaries [Ann. Nucl. Energy 27(2000)1385] to the general multi-group non-homogeneous model. The full solution is given through the Green's function of the static problem, the static flux, and a quantity describing the boundary movements. A multi-group absorber model is proposed to represent the perturbation. which turns out to be very useful, for instance, to derive the point reactor and adiabatic approximations of the neutron noise arising from the oscillating boundaries. Finally, an equivalent solution is given in terms of the adjoint function.
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| 4. |
- Eriksson, Marcus, et al.
(author)
-
Inherent Shutdown Capabilities in Accelerator-driven Systems
- 2002
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In: Annals of Nuclear Energy. - 0306-4549 .- 1873-2100. ; 29:14, s. 1689-1706
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Journal article (peer-reviewed)abstract
- The applicability for inherent shutdown mechanisms in accelerator-driven systems (ADS) has been investigated. We study the role of reactivity feedbacks. The benefits, in terms of dynamics performance, for enhancing the Doppler effect are examined. Given the performance characteristics of source-driven systems, it is necessary to manage the neutron source in order to achieve inherent shutdown. The shutdown system must be capable of halting the external source before excessive temperatures are obtained. We evaluate methods, based on the analysis of unprotected accidents, to accomplish such means. Pre-concepted designs for self-actuated shutdown of the external source suggested. We investigate time responses and evaluate methods to improve the performance of the safety system. It is shown that maximum beam output must be limited by fundamental means in order to protect against accident initiators that appear to be achievable in source driven systems. Utilizing an appropriate burnup control strategy plays a key role in that effort.
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| 5. |
- Pazsit, I., et al.
(author)
-
Linear reactor kinetics and neutron noise in systems with fluctuating boundaries
- 2000
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In: Annals of Nuclear Energy. - 0306-4549 .- 1873-2100. ; 27:15, s. 1385-1398
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Journal article (peer-reviewed)abstract
- The general theory of linear reactor kinetics and that of the induced neutron noise is developed for systems with varying size, i.e. in which the position of the boundary fluctuates around a stationary value. The point kinetic and adiabatic approximations are defined by a generalisation of the flux factorisation, and the full solution of the general problem with an arbitrarily fluctuating boundary is given by the Green's function technique. The correctness of the general solution is proven both generally and also by considering the simple case of a 2-D cylindrical reactor with a fluctuating radius, in which case a direct compact solution is possible.
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| 6. |
- Demaziere, Christophe, 1973
(author)
-
Development of a 2-D 2-group neutron noise simulator
- 2004
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In: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 31:6, s. 647-680
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Journal article (peer-reviewed)abstract
- In this paper, the development of a so-called neutron noise simulator is reported. This simulator calculates both the direct and the adjoint reactor transfer function between a stationary noise source and its induced neutron noise for any 2-dimensional heterogeneous critical system. The main advantage of this neutron noise simulator is that any realistic core can be modelled, since the simulator is designed to rely on a set of material constants corresponding to the actual reactor operating conditions. The calculations are performed in the 2-group diffusion approximation and in the frequency domain. The spatial discretisation is carried out with respect to the finite difference scheme. The noise source, expressed as an "absorber of variable strength" type, is defined directly from the fluctuations of the macroscopic cross-sections and can be spatially distributed over the core or concentrated in a few discrete nodes. If the noise source is a point-source, the simulator actually estimates the 2-dimensional 2-group discretised Green's function of the system. From the calculated Green's function, the neutron noise induced by a "vibrating absorber" type of noise source can also be determined. Different benchmark cases show that this neutron noise simulator works satisfactorily.
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| 7. |
- Talamo, Alberto, et al.
(author)
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Key Physical Parameters and Temperature Reactivity Coefficients of the Deep Burn Modular Helium Reactor Fueled with LWRs Waste
- 2004
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In: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 31, s. 1913-1931
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Journal article (peer-reviewed)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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| 8. |
- Tucek, Kamil, et al.
(author)
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Coolant void worth in fast breeder reactors and accelerator-driven transuranium and minor-actinide burners
- 2004
-
In: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 31:15, s. 1783-1801
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Journal article (peer-reviewed)abstract
- Liquid metal coolant void worth have been calculated as a function of fuel composition and core geometry for several model fast breeder reactors and accelerator-driven systems (ADSs). The Monte Carlo transport code MCNP with continuous energy cross-section libraries was used for this study. With respect to the core void worth, lead/bismuth cooled FBR, appear to be inferior to those employing sodium for pitch-to-diameter ratios exceeding 1.4. It is shown that in reactor systems cooled by lead/bismuth eutectic radial steel pin reflector significantly lowers the void worth. The void worth proves to be a strong function of the fuel composition, reactor cores with high content of minor actinides in fuel exhibiting larger void reactivities than systems with plutonium based fuel. Enlarging the lattice pitch in ADS burners operating on Pu rich fuel decreases the void worth while the opposite fact is true for ADSs employing americium based fuels.
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| 9. |
- Arul, A.J., et al.
(author)
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The power law character of off-site power failures
- 2003
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In: Annals of Nuclear Energy. - 0306-4549. ; 30:14, s. 1401-1408
-
Journal article (peer-reviewed)abstract
- A study on the behavior of off-site AC power failure recovery times at three nuclear plant sites is presented. It is shown, that power law is appropriate for the representation of failure frequency–duration correlation function of off-site power failure events, based on simple assumptions about component failure and repair rates. It is also found that the annual maxima of power failure duration follow Frechet distribution, which is a type II asymptotic distribution, strengthening our assumption of power law for the parent distribution. The extreme value distributions obtained are used for extrapolation beyond the region of observation.
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