Results of full scale CONSTOR® V/TC prototype 9 m horizontal drop test

Abstract

In the context of the research on the mechanical safety of packages for radioactive material, full scale drop tests with spent fuel and high activity waste transport and storage casks have been performed by the Federal Institute for Materials Research and Testing (BAM). The research reflects national and international interest in acquiring comparative knowledge of full and reduced scale model drop tests as well as in finite element calculations. This paper presents the experimental, analytical and first numerical results of the full scale drop test with the full scale CONSTOR^® V/TC prototype, manufactured by GNS, Gesellschaft für Nuklear-Service mbH, Germany. The prototype was tested by BAM in a 9 m horizontal drop test onto the unyielding target of the BAM drop test facility in Horstwalde, Germany.     

Statistical error estimation of the Feynman-α method using the bootstrap method

Applicability of the bootstrap method is investigated to estimate the statistical error of the Feynman-α method, which is one of the subcritical measurement techniques on the basis of reactor noise analysis. In the Feynman-α method, the statistical error can be simply estimated from multiple measurements of reactor noise, however it requires additional measurement time to repeat the multiple times of measurements. Using a resampling technique called “bootstrap method,” standard deviation and confidence interval of measurement results obtained by the Feynman-α method can be estimated as the statistical error, using only a single measurement of reactor noise. In order to validate our proposed technique, we carried out a passive measurement of reactor noise without any external source, i.e. with only inherent neutron source by spontaneous fission and (α,n) reactions in nuclear fuels at the Kyoto University Criticality Assembly. Through the actual measurement, it is confirmed that the bootstrap method is applicable to approximately estimate the statistical error of measurement results obtained by the Feynman-α method.     

Coal analysis using the pulsed neutron generator

A prototype of elemental analyzer for coal has been developed by using a PFTNA (pulse fast thermal neutron analysis) system. The PFTNA technology is based on the reactions such as (n, γ), (n, n'γ), (n, pγ), etc. by examining the characteristic gamma rays emitted. In our prototype a pulsed neutron generator provides 14 MeV pulse neutrons, which contribute to the separation of spectrum II (the sum of capture and activation spectrum) from spectrum I (the sum of inelastic, capture and activation spectrum), and thus to the measurement of C and O contents in coal. Data management is completed by computer program using the least-square regression method. The experiment in Changshan Power Plant for 3 months showed that the precision of calorific value, whole water, volatile content and ash content is 0.5 kJ/kg, 1.0 wt%, 2.0 wt% and 1.5 wt%, respectively.     

Verification and validation of the flux reconstruction method for CANDU® applications

In this paper, we propose a new method for the cell local flux reconstruction for CANDU® applications. The method is based on the modulation method where the core local mesh flux is reconstructed by modulating the form function with an interpolated cell mesh flux along with an appropriate normalization technique. The reconstruction method was adapted to the diffusion finite difference method. In this paper we present: (a) the flux reconstruction method, (b) the verification of this method using two benchmarking problems and (c) its validation using the ZED-2 data generated from the defueled channel experiments.     

A multigroup finite-element solution of the neutron transport equation—III: Anisotropy of scattering

The extension of the theory of a multigroup finite-element method of solving the neutron transport equation, to include general anisotropy of scattering and an anisotropic spatially-dependent source, is described. The method, based on a variational principle applied to the even-parity transport equation, employs spherical harmonics for the angular basis functions. To illustrate the development in the associated computer code, three test problems, all including energy-dependent anisotropy, are solved, and the results presented in the form of tables and graphs.     

Measurement of cathode surface temperature using the method of CCD imaging in arc discharge

A two-wavelength pyrometry device using ordinary array CCD （charge coupled device） to collect the radiation data in the horizontal and vertical directions has been developed for measuring the cathode surface temperature during the arc discharge. Analyses of experimental results show that the device can make the measurement of the cathode surface temperature feasible. The cathode surface temperatures measured are lower than the melting point of tungsten （3653 K）, and the arc current, cathode diameter, and the cathode length are the main influencing factors of the cathode surface temperature.     

Elemental PGNAA analysis using gamma–gamma coincidence counting with the library least-squares approach

An accurate method for determining elemental analysis using gamma–gamma coincidence counting is presented. To demonstrate the feasibility of this method for PGNAA, a system of three radioisotopes (Na-24, Co-60 and Cs-134) that emit coincident gamma rays was used. Two HPGe detectors were connected to a system that allowed both singles and coincidences to be collected simultaneously. A known mixture of the three radioisotopes was used and data was deliberately collected at relatively high counting rates to determine the effect of pulse pile-up distortion. The results obtained, with the library least-squares analysis, of both the normal and coincidence counting are presented and compared to the known amounts. The coincidence results are shown to give much better accuracy. It appears that in addition to the expected advantage of reduced background, the coincidence approach is considerably more resistant to pulse pile-up distortion.     

Assembly?oriented reliability analysis method for the top?connection structure of a nuclear fuel assembly

The nuclear fuel assembly is the core component of a nuclear reactor. In a pressurized water reactor fuel assembly, the topconnection structure connects the top nozzle to the guide thimble. Its performance reliability is essential for the stability of the nuclear fuel assembly. In this study, an assembly-oriented reliability analysis method for top-connection structures is presented by establishing an assembly-oriented top-connection structure parameter modeling method and a nonlinear contact ga...     

Huitzoctli: A system to design Control Rod Pattern for BWR’s using a hybrid method

Huitzoctli system was developed to design Control Rod Patterns for Boiling Water Reactors (BWR). The main idea is to obtain a Control Rod Pattern under the following considerations: (a) the critical reactor core state is satisfied, (b) the axial power distribution must be adjusted to a target axial power distribution proposal, and (c) the maximum Fraction of Critical Power Ratio (MFLCPR), the maximum Fraction of Linear Power Density (FLPD) and the maximum Fraction of Average Planar Power Density (MPGR) must be fulfilled. Those parameters were obtained using the 3D CM-PRESTO code. In order to decrease the problem complexity, Control Cell Core load strategy was implemented; in the same way, intermediate axial positions and core eighth symmetry were took into account. In this work, the cycle length was divided in 12 burnup steps. The Fuel Loading Pattern is an input data and it remains without changes during the iterative process. The Huitzoctli system was developed to use the combinatorial heuristics techniques Scatter Search and Tabu Search. The first one was used as a global search method and the second one as a local search method. The Control Rod Patterns obtained with the Huitzoctli system were compared to other Control Rod Patterns designs obtained with other optimization techniques, under the same operating conditions. The results show a good performance of the system. In all cases the thermal limits were satisfied, and the axial power distribution was adjusted to the target axial power distribution almost completely. It is very important to point out that, even though the cycle length improvement was not the main idea of this work, the effective multiplication factor (k_eff) at the end of the cycle was improved in all cases tested. The Huitzoctli system was programmed using Fortran 77 language in an Alpha Workstation with UNIX operating system.     

Spectral analysis method for deriving RBS-like carbon spectra from the C(α,α)C resonant reaction

This article presents a spectral analysis method that detects C using the high-sensitivity of the 4.26 MeV resonance of the ¹²C(α,α)¹²C nuclear reaction while avoiding issues arising from the peaky and asymmetric resonance shape, which complicates depth-sensitive C analysis. By averaging nuclear reaction spectra taken with a set of conveniently chosen He beam energies, we obtain C spectra with amplified intensity, but shape similar to Rutherford backscattering spectrometry (RBS) spectra. The latter fact allows intuitive reading of underlying C depth profiles without employing spectrum simulation software. The method was first applied to simulated samples whose nuclear reaction spectra were generated by SIMNRA, which allowed checking for method accuracy by comparison to corresponding simulated RBS spectra. As real examples of the method application, it was applied to detect depth-sensitive C signals from SiC substrates covered by SiO₂ layers and from 50 nm hafnium-based films deposited on Si by metal-organic chemical vapor deposition.     

Development of a “best representativity” method for experimental data analysis and an application to the critical experiments at the Toshiba NCA facility

For nuclear critical experiments, it is essential to certify similarities of the experiment with the objective of the actual reactor conditions or actual reactor equipment. To judge the applicability of the experimental data, the concept of a “representativity factor” has recently been adopted in the critical experiment field, particularly for fast breeder reactors and future reactor studies. In this study, we extended this concept to the design of a light water reactor system. We developed a new numerical evaluation method and a calculation system. The method is based on a linear combination of the sensitivity coefficient vector of an experiment in which the representativity factor to the target system is maximized to utilize experimental data effectively. Simultaneously, using the measurement data of critical experiments, the method enables us to evaluate calculation errors caused by errors or uncertainties of physical parameters. The derivation of the new calculation method is explained first. We then qualify it with a sample calculation, presenting numerical results for three kinds of critical experiments conducted at the Toshiba Nuclear Critical Assembly facility. Finally, the results are compared with those of an extended bias factor method to clarify the performance of the new method.     

Development of a method to lower recontamination after chemical decontamination by depositing Pt nano particles: (I) Consideration of the suppression mechanism using a quantum dynamics calculation

The Pt coating (Pt-C) process has been developed to lower the recontamination by radioactive elements after chemical decontamination of piping surfaces. In this process, a layer of fine Pt nano particles is formed in an aqueous solution on the base metal of the piping following the chemical decontamination. In this study, we confirmed that the suppression effect by the Pt-C toward ⁶⁰Co deposition on type 316 stainless steel using a ⁶⁰Co deposition test under hydrogen water chemistry. Furthermore, we investigated the suppression mechanism of deposition of radioactive elements by a quantum molecular simulation. The deposition amounts of ⁶⁰Co which were incorporated in oxides after 1000 h with and without the Pt-C process were about 90 and 10.2 Bq/cm², respectively. The amount of ⁶⁰Co deposition with Pt-C is about 10% that of non-coated specimens. The ⁶⁰Co incorporation for the Pt-C specimen was suppressed by decreasing the formation of oxides. We considered this phenomenon using a quantum dynamics calculation and concluded that the Fe–O bonds in oxides were weakened by the effect of Pt and hydrogen radicals which were produced in the reaction between H₂ and Pt, and then oxides were dissolved into the water.     

A simple method for estimating the structure temperatures and the cesium revaporization inside the reactor pressure vessel – II: Feasibility study for the Fukushima Daiichi Nuclear Power Plant

The present article is the second part of a two-part article, examining the feasibility of the method proposed in the preceding part. In this method, the ratio of the core uncovered from water x _u, the ratio of the core flooded by water x _f, the ratio of the core slumped into the pedestal area of the drywell x _s, and the ratio of the injected water leaking before reaching the core x _wl are the four important uncertain parameters. The base case study shows that water injection via the core spray line is more effective to cool the uncovered core and to reduce the amount of cesium hydroxide (CsOH) released from the drywell. The sensitivity study is conducted by introducing the dimensionless decay heat N _qd, which combines the effects of x _f, x _s, and x _wl on the steam generation rate associated with the forced convection cooling in the reactor pressure vessel (RPV). The results show that the temperatures of the uncovered core and the other structures increase with N _qd. Consequently, the release rate of CsOH also increases with N _qd. The relationships of the measurable RPV wall temperature with the temperatures of the uncovered core and the structures as well as the release characteristics of CsOH are also examined.     

NAPTH: Neutronics analysis code system for the fusion–fission hybrid reactor with pressure tube type blanket

The fusion–fission hybrid reactor is considered as a potential path to the early application of fusion energy. A new concept with pressure tube type blanket has recently been proposed for a feasible hybrid reactor. In this paper, a code system for the neutronics analysis of the pressure tube type hybrid reactor is developed based on the two-step calculation scheme: the few-group homogeneous constant calculation and the full blanket calculation. The few-group homogeneous constants are calculated using the lattice code DRAGON4. The blanket transport calculation is performed by the multigroup Monte Carlo code. A link procedure for fitting the cross sections is developed between these two steps. An additional procedure is developed to calculate the burnup, power distribution, energy multiplication factor, tritium breeding ratio and neutron multiplication factor. From some numerical results, it is found that the code system NAPTH is reliable and exhibits good calculation efficiency. It can be used for the conceptual design of the pressure tube type hybrid reactor with precise geometry.     

Rossi-α and Feynmann Y functions for non-Poissonian pulsed sources of neutrons in the stochastic pulsing method: Application to subcriticality monitoring in ADS and comparison with the results of Poissonian pulsed neutron sources

In this paper, analytical expressions for the Rossi-α and the Feynmann Y functions are deduced for the case of Poissonian and non-Poissonian neutron sources when the stochastic pulsing method is used. These analytical expressions are used to fit the experimental data and to obtain the prompt neutron time constant. Also we perform in this paper a comparison of the results obtained for the Rossi-α and Feynmann Y functions with Poissonian and non-Poissonian neutron sources, and we study how much change the shape of these functions when the fission probability decreases and the capture probability increases due to the depletion with time of the fuel, and the increase of the fission products. Some comparisons with experimental data and with the results of other authors have been performed. Another important question analyzed in this paper and that it is interesting from an academic point of view is that the average number of detected counts induced by one single neutron injected in the system at an arbitrary time t′, should obey in point kinetics theory an adjoint equation in the time domain. Also the cross-factorial moment of the number of counts induced by one neutron in two counting intervals should obey also an adjoint equation in the time domain with a source term that depends on the first moments. These results are a consequence of more general results that have been obtained using stochastic transport theory for the one particle probability generating function or Kernel generating function.     

Fast-sweeping Langmuir probes: what happens to the I–V trace when sweeping frequency is higher than the ion plasma frequency?

Limited particle transit time is one of several limiting factors which determine the maximum temporal resolution of a Langmuir probe. In this work, we have revisited the known fast sweep Langmuir probe techniques in a uniform, quiescent multi-dipole confined hot cathode discharge with two operation scenarios: one in which the probe sweeping frequency fsweep is much lower than the ion plasma frequency fpi, another one where fsweep is much greater than fpi, respectively. This allows investigation into the effect of limited ion-motion on I–V traces. Serious distortions of I–V traces at high frequencies, previously claimed to be an ion-motion limitation effect, were not found unless shunt resistance is sufficiently high, despite a fsweep/fpi ratio of ∼3. On the other hand, evidences of sheath capacitance on the I–V traces have been observed. Distortions of I–V traces qualitatively agree with predictions of sheath capacitance response to the sweeping voltage. Additionally, techniques in fast sweep Langmuir probe are briefly discussed. The comparison between the high-speed dual Langmuir probe (HDLP) and the single probe setup shows that the capacitive response can be removed via subtracting a leakage current for the single probe setup almost as effectively as using the HDLP setup, but the HDLP setup does remain advantageous in its facilitation of better recovery of weak current signal commonly in low density plasma.     

Calculation and analysis of n+Zr reactions in the En ⩽ 250 MeV energy range

All of reaction cross sections, angular distributions, energy spectra, γ-ray production cross sections, and the double differential cross section for neutron, proton, deuteron, triton, helium and alpha emission are calculated and analyzed for n+^{90,91,92,94,96,nat}Zr at incident neutron energies from 0.1 to 250 MeV. The optical model, intranuclear cascade model, the unified Hauser–Feshbach theory and the exciton model which included the improved Iwamoto–Harada model are used. Theoretical calculated results are compared with existing experimental data and other evaluated data from ENDF/B-VI.8, ENDF/B-VII.0 and JENDL-3.3. The optical model potential parameters are obtained according to the experimental data of total, nonelastic cross sections and elastic scattering angular distributions.     

Energy calibration of HPGe detector using the high-energy characteristic γ rays in 13C formed in 6Li + 12C reaction

An ⁶ Li+⁸⁹Y experiment was conducted at the Laboratori Nazinali di Legnaro,INFN,Italy.The 550 μg/cm² thick ⁸⁹Y target was backed on a 340 μg/cm² thick ¹²C foil.The several γ rays in the experiment with energies higher than 3000 keV can most likely be ascribed to the transitions in the ¹³C nuclei,which can be formed through various interactions between the ⁶ Li beam and the ¹²C foil.The high-energy ...