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1.
Monte Carlo calculation methods to estimate the effective delayed neutron fraction βeff are investigated: one is proposed by Meulekamp et al. and the other is by Nauchi et al. It is revealed that both the methods calculate the delayed neutron fraction weighted with the importance functions defined by Kobayashi. The accuracy of the methods are also examined for several simple benchmark systems. Consequently, it is found that Meulekamp’s method causes ∼5% discrepancies in the βeff values for fast systems; Nauchi’s method gives good results for fast bare systems but ∼10% discrepancies for fast reflected systems. Both the methods calculate the βeff values approximately within the accuracy of ∼2% for thermal systems.  相似文献   

2.
The applicability of Monte Carlo techniques, namely the Monte Carlo sensitivity method and the random-sampling method, for uncertainty quantification of the effective delayed neutron fraction βeff is investigated using the continuous-energy Monte Carlo transport code, MCNP, from the perspective of statistical convergence issues. This study focuses on the nuclear data as one of the major sources of βeff uncertainty. For validation of the calculated βeff, a critical configuration of the VENUS-F zero-power reactor was used. It is demonstrated that Chiba's modified k-ratio method is superior to Bretscher's prompt k-ratio method in terms of reducing the statistical uncertainty in calculating not only βeff but also its sensitivities and the uncertainty due to nuclear data. From this result and a comparison of uncertainties obtained by the Monte Carlo sensitivity method and the random-sampling method, it is shown that the Monte Carlo sensitivity method using Chiba's modified k-ratio method is the most practical for uncertainty quantification of βeff. Finally, total βeff uncertainty due to nuclear data for the VENUS-F critical configuration is determined to be approximately 2.7% with JENDL-4.0u, which is dominated by the delayed neutron yield of 235U.  相似文献   

3.
The 10 MW_(th) solid-fueled thorium molten salt reactor(TMSR-SF1) is a FLi Be salt-cooled pebble bed reactor to be deployed in 5–10 years, designed by the TMSR group. Due to a large amount of beryllium in the core, the photoneutrons are produced via(γ , n) reactions.Some of them are generated a long time after the fission event and therefore are considered as delayed neutrons. In this paper, we redefine the effective delayed neutrons into two fractions: the delayed fission neutron fraction and the delayed photoneutron fraction. With some reasonable assumptions, the inner product method and the k-ratio method are adopted for studying the effective delayed photoneutron fraction. In the k-ratio method, the Monte Carlo code MCNP6 is used to evaluate the effective photoneutron fraction as the ratio between the multiplication factors with and without contribution of the delayed neutrons and photoneutrons. In the inner product method, with the Monte Carlo and deterministic codes together, we use the adjoint neutron flux as a weighting function for the neutrons and photoneutrons generated in the core. Results of the two methods agree well with each other, but the k-ratio method requires much more computing time for the same precision.  相似文献   

4.
Today there is no well-established theoretical model to predict the fission delayed neutron yield vd with the required accuracy. In this field the recommended data result from the rare experimental data analysis or from purely phenomenological or semi-phenomenological models. There is another source of valuable information: the related integral data or βeff- data. In this report we demonstrate, via a careful analysis of the experimental methods leading to revisited experimental βeff values and associated uncertainties, that for the major nuclei the vd evaluated data are of acceptable quality. For U-235 U-238 and Pu-239 we recommend vd values for the thermal and the fast reactor ranges which have been obtained from a statistical consistent adjustment to the βeff data. In the course of this study we show that the energy dependence of vd, suspected from a physics point of view, probably exists with a different magnitude according to the nucleus. Concerning the major nuclei it is of negligible importance for the applications. The improvement of the higher Pu isotopes and minor actinides data is the main motivation for developing the theoretical investigations of the delayed neutron generation mechanism at the same level as the necessary experimental activity.  相似文献   

5.
To improve the accuracy of prediction of βeff, an international program of benchmark experiments was planned. This program consisted of two parts; the BERENICE-MASURCA and the FCA XIX series of experiments. The former was carried out in the fast critical facility MASURCA of CEA, FRANCE between 1993 and 1994. The latter one was carried out in the FCA, JAERI between 1995 and 1998. In these benchmark experiments, various experimental techniques were applied to measure the βeff. Through the synthesis of the different results, a recommended value for each core was provided and the accuracy of the measurements was evaluated to be better than 3%. The calculations showed good agreement of the recommended βeff values within 3% for JENDL-3.2 and ENDF/B-VI delayed neutron data sets.  相似文献   

6.
An optimization approach to establish an appropriate multi-group energy structure for boiling water reactor (BWR) pin-by-pin fine mesh core analysis is proposed. In the present approach, the number of energy groups of cross sections is successively reduced or increased. In order to select an energy group boundary that is removed or added, performances of all possible candidates of energy group structures are tested in multi-assembly geometries. Then, the energy group boundary, which provides the minimum difference of the k-infinity or the pin-by-pin fission rate distribution, is finally removed or added. This procedure is repeated until the number of energy groups reaches to the target value. In order to confirm the applicability of the present approach, the accuracies of the k-infinity and the pin-by-pin fission rate distribution are investigated in various 2 × 2 multi-assembly geometries with the established energy group structure. From the verification results, the differences of the k-infinity and the pin-by-pin fission rate distribution between the reference (fine) and the established (coarse) energy group structure are small in the various 2 × 2 multi-assembly geometries. Therefore, we can conclude that the present approach is efficient to establish an appropriate energy group structure for BWR pin-by-pin fine mesh core analysis.  相似文献   

7.
ABSTRACT

An effective dose calculation method is important in the design of efficient shields in radiation facilities. Some analytical methods have been shown to provide a simple and quick design analysis; however, no suitable method exists that can be applied to a room located directly under an X-ray irradiation room. We propose a new analytical method that uses the multiple reflection ratio predetermined by a Monte Carlo simulation and the differential dose albedo given by the Chilton–Huddleston semi-empirical equation. Our method is verified by comparison with the Monte Carlo simulation, performed for the case of an electron linac facility with an accelerated energy of 10 MeV, where the shielding floor has a thickness of 1.6–2.0 m and the downstairs room has a height of 0.5–1.5 m. The difference between the effective X-ray doses in the downstairs room calculated via the proposed analytical method and the Monte Carlo simulation is less than 25% when the horizontal distance from the X-ray beam to the evaluation point exceeds 3 m and the evaluation point is set at half of the height of the room. The new analytical method can be efficiently and accurately applied to the calculation of the effective dose.  相似文献   

8.
ABSTRACT

For recent boron neutron capture therapy (BNCT), accelerator-based neutron sources have been actively developed in place of reactor-based neutron sources. In this study, a novel neutron energy spectrometer for the daily quality assurance (QA) of BNCT was designed on the basis of a CsI self-activation method for accelerator-based neutron sources. The spectrometer design was optimized in terms of its energy resolution. To verify its applicability to high-intensity BNCT neutron fields, some practical simulations were performed. It was shown that the designed spectrometer was able to evaluate a neutron energy spectrum in approximately 900 s after an instantaneous neutron irradiation. In addition, its energy resolution was sufficient for detecting an unexpected distortion in the spectrum. The results confirm that the designed spectrometer can be employed for the daily QA of BNCT to check that the expected spectrum remains unchanged.  相似文献   

9.
In this study, we developed a 45 MeV neutron fluence rate standard of Japan. Quasi-monoenergetic neutrons with a peak energy of 45 MeV in the neutron standard field were produced by the 7Li(p,n)7Be reaction using a 50-MeV proton beam from an azimuthally varying field (AVF) cyclotron of the Takasaki Ion Accelerators for Advanced Radiation Application (TIARA). The neutron energy spectrum was measured using an organic liquid scintillation detector and a 6Li-glass scintillation detector by the time-of-flight method, and using a Bonner sphere spectrometer by the unfolding method. The absolute neutron fluence was determined using a proton recoil telescope (PRT) composed of the liquid scintillation detector and a Si(Li) detector that was newly developed in the present study. The detection efficiency of the PRT was obtained using the MCNPX code. The peak neutron production cross section for the 7Li(p,n)7Be reaction was also derived from the neutron fluence in order to confirm the neutron fluence of the TIARA high-energy neutron field. The peak neutron production cross section obtained in the present study was in good agreement with those of previous studies. The characteristics of the 45-MeV neutron field in TIARA were successfully evaluated in order to calibrate high-energy neutron detectors and high-energy neutron dosimeters.  相似文献   

10.
Heterogeneous nuclear reactors require numerical methods to solve the neutron diffusion equation (NDE) to obtain the neutron flux distribution inside them, by discretizing the heterogeneous geometry in a set of homogeneous regions. This discretization requires additional equations at the inner faces of two adjacent cells: neutron flux and current continuity, which imply an excess of equations. The finite volume method (FVM) is suitable to be applied to NDE, because it can be easily applied to any mesh and it is typically used in the transport equations due to the conservation of the transported quantity within the volume. However, the gradient and face-averaged values in the FVM are typically calculated as a function of the cell-averaged values of adjacent cells. So, if the materials of the adjacent cells are different, the neutron current condition could not be accomplished. Therefore, a polynomial expansion of the neutron flux is developed in each cell for assuring the accomplishment of the flux and current continuity and calculating both analytically. In this polynomial expansion, the polynomial terms for each cell were assigned previously and the constant coefficients are determined by solving the eigenvalue problem with SLEPc. A sensitivity analysis for determining the best set of polynomial terms is performed.  相似文献   

11.
12.
The effective delayed neutron fraction βeff for a light water moderated low-enriched UO2 core has been re-evaluated to obtain benchmark data for the validation of calculation codes and nuclear data. Originally, the βeff value was measured by the substitution method. In that method, the βeff value was obtained from measured reactivity change by substituting a Sb-Cd-Pb absorber rod for a 2.6 wt% UO2 rod for all core regions. In the present evaluation, we have employed the latest value for the buckling coefficient of reactivity to re-evaluate the substitution reactivity with high accuracy. In addition, the correction factor, which was ignored in the previous measurement, has been calculated to compensate the difference in the absorption cross sections of fuel and absorber rods. Consequently, the obtained βeff value in the present evaluation was 0.00771±0.00017, and it is more credible than the previous one. The present result is available as benchmark data for the verification of delayed neutron data for light water reactors.

For comparison, we have calculated the βeff value using a transport code TWODANT with the JENDL-3.2 nuclear data library. The calculated βeff value overestimated the experiments; the difference slightly exceeded the experimental error.  相似文献   

13.
A correction technique to capture the spectral interference effect on collapsed cross sections is combined with the superhomogenization (SPH) factor or the discontinuity factor (DF) and is applied to the pin-by-pin core analysis for boiling water reactors (BWRs). The spectral interference effect has relationship with variations of neutron leakage in each pin-cell from the viewpoint of neutron balance. In order to correct collapsed cross sections, a new correction technique, in which the neutron leakage in each pin-cell is used as a correction index, was proposed in the previous study. By this correction technique, the reference coarse group cross sections are well reproduced and the calculation accuracies are improved. However, the reference fine group calculation results could not be reproduced since the correction technique cannot reduce energy collapsing errors. Thus, we combine the correction technique with the SPH factor or the DF to reduce energy collapsing errors. In order to verify and discuss the applicability of the correction technique with the SPH factor or the DF, two-dimensional benchmark calculations considering typical characteristics of BWR cores are carried out. The correction technique with the DF more accurately reproduces the reference fine group calculation results than that with the SPH factor.  相似文献   

14.
This paper presents the architecture for upgrading the instrumentation and control (I&C) systems of a Korean standard nuclear power plant (KSNP) as an operating nuclear power plant. This paper uses the analysis results of KSNP's I&C systems performed in a previous study. This paper proposes a Preparation–Decision–Design–Assessment (PDDA) process that focuses on quality oriented development, as a cyclical process to develop the architecture. The PDDA was motivated from the practice of architecture-based development used in software engineering fields. In the preparation step of the PDDA, the architecture of digital-based I&C systems was setup for an architectural goal. Single failure criterion and determinism were setup for architectural drivers. In the decision step, defense-in-depth, diversity, redundancy, and independence were determined as architectural tactics to satisfy the single failure criterion, and sequential execution was determined as a tactic to satisfy the determinism. After determining the tactics, the primitive digital-based I&C architecture was determined. In the design step, 17 systems were selected from the KSNP's I&C systems for the upgrade and functionally grouped based on the primitive architecture. The overall architecture was developed to show the deployment of the systems. The detailed architecture of the safety systems was developed by applying a 2-out-of-3 voting logic, and the detailed architecture of the non-safety systems was developed by hot-standby redundancy. While developing the detailed architecture, three ways of signal transmission were determined with proper rationales: hardwire, datalink, and network. In the assessment step, the required network performance, considering the worst-case of data transmission was calculated: the datalink was required by 120 kbps, the safety network by 5 Mbps, and the non-safety network by 60 Mbps. The architecture covered 17 systems out of 22 KSNP's I&C systems. The architecture is implementable with the equipment developed in South Korea. The architecture can be used as a model to upgrade the existing I&C systems in a planned, large-scale, and one-shot manner. A more detailed architecture down to software level will be developed in the future.  相似文献   

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