JENDL/ImPACT-2018: a new nuclear data library for innovative studies on transmutation of long-lived fission products

ABSTRACT

A new nuclear data library, JENDL/ImPACT-2018, was developed for an innovative study on the transmutation of long-lived fission products. Nuclear reaction cross-sections were newly evaluated for incident neutrons and protons up to 200 MeV for 163 nuclides focusing on long-lived nuclei such as ⁷⁹Se, ⁹³Zr, ¹⁰⁷Pd and ¹³⁵Cs, adopting some parts of JENDL-4.0. Our challenge was an evaluation of cross-sections for a number of unstable nuclei over a wide energy range where the experimental data were very scarce. We estimated cross-sections based on a nuclear model code CCONE by incorporating an advanced knowledge on the nuclear structure theory and a model-parameterization based on new experimental cross-sections measured by the inverse kinematics. Through comparisons with available experimental data on the stable isotopes, it is found that the present data give better agreements with them than those in the existing libraries. In a neutronics simulation by the PHITS code, we also found that the largest impact of the present library was seen on the estimated amount of isotope productions.     

A kinetic model for the direct response matrix method

ABSTRACT

An accurate analysis model for transient reactor behavior is necessary to keep sufficient safety margins of nuclear power plants to prevent cliff edge effects. In this study, the direct response matrix (DRM) method is applied to the kinetic domain and the transient analysis is enabled based on the transport equation. The kinetic DRM model introduces the time delay to four sub-response matrices. The time delay can be evaluated by a Monte Carlo calculation. The model is evaluated in homogeneous and heterogeneous problems. The Doppler feedback is considered in the heterogeneous problem and the calculation results are compared with the experimental data. The calculation results indicate that the calculation step 1.0E-7 s is sufficient for the model and the model provides results in good agreement with the experimental data. It is concluded that the present model with the DRM method can be used for transient analysis.     

Response of nuclear materials in storage and transport to malevolent environments

Abstract

Work related to the assessment of radiological health consequences resulting from a sabotage attack on nuclear fuel storage or transport casks has been continuing since the late 1970s. While the level of effort in this area has been uneven over these three decades due to policy priorities, funding levels and programmatic priorities of the countries funding this type of work, substantial progress has been made. From phenomenology of in cask transport processes to development of aerosol production in high energy attack environments, the analytical and experimental work performed provides substantial justification to consequence assessments that heretofore have had to rely on conservative assumptions in lieu of empirical data. One constant since the late 1990s in addressing this problem has been an international working group whose primary focus has been to develop source term data from experimental simulations of sabotage types of attacks. This working group, titled the International Working Group for Sabotage Concerns of Transport and Storage Casks (WGSTSC) is comprised of experts mainly from the USA, France and Germany. Technical support has also been provided, on an intermittent basis, from the UK and Japan. The WGSTSC has pooled resources and expertise to design and conduct experiments to produce the data needed to perform radiological consequence assessments. In addition to the experimental programme, this group also has coupled modern analytical techniques with experimental results to understand the effects of high energy density devices on nuclear materials. This paper reviews the paradigm that has guided the WGSTSC effort and shows how the results from the experimental programmes of the past three decades have brought us to the current level of understanding of the potential consequences from a malevolent attack on nuclear transport and storage facilities. In addition, the paper provides an update on the status of the work of the WGSTSC and describes what additional experimental and analytical efforts would be most productive in further narrowing of uncertainty in consequence prediction.     

Benchmark experiment of large-angle scattering reaction cross section of iron at 14 MeV using two shadow bars – Comparison of experimental results with ENDF/B-VIII –

ABSTRACT

It is important to perform neutron transport simulations with accurate nuclear data in the neutronics design of a fusion reactor. However, absolute values of large-angle scattering cross sections vary among nuclear data libraries even for well-examined nuclide of iron. Benchmark experiments focusing on large-angle scattering cross sections were thus performed to confirm the correctness of nuclear data libraries. The series benchmark experiments were performed at a DT neutron source facility, OKTAVIAN of Osaka University, Japan, by the unique experimental system established by the authors’ group, which can extract only the contribution of large-angle scattering reactions. This system consists of two shadow bars, target plate (iron), and neutron detector (niobium). Two types of shadow bars were used and four irradiations were conducted for one experiment, so that contribution of room-return neutrons was effectively removed and only large-angle scattering neutrons were extracted from the measured four Nb reaction rates. The obtained experimental results were compared with calculations for five nuclear data libraries including JENDL-4.0, JEFF.-3.3, FENDL-3.1, ENDF/B- VII, and recently released ENDF/B-VIII. It was found from the comparison that ENDF/B-VIII showed the best result, though ENDF/B-VII showed overestimation and others are in large underestimation at 14 MeV.     

p+~(107,109)Ag核反应数据的计算和分析

本工作通过理论计算的方法获得了一套适合入射能量为从阈值到200 MeV的p+107,109 Ag核反应全套微观数据。首先,使用光学模型理论进行调参计算,得到了一套适合入射能量为从阈值到340MeV的p+107,109 Ag核反应Becchetti-Greenlees光学势参数,这套参数与实验数据符合很好。其次,在这套光学势参数的基础上用扭曲波玻恩近似对入射能量从阈值到200MeV的p+107,109 Ag直接非弹性散射截面进行了计算。最后,使用核反应统计理论计算了入射能量从阈值到200MeV的p+107,109 Ag核反应各反应道的截面和出射粒子能谱,得到了该能区p+107,109 Ag核反应全套微观数据。将所有计算值与实验数据进行比较,结果表明,所得到的全套微观数据与实验数据符合很好。     

Defect inspection system of nuclear fuel pellet end faces based on machine vision

ABSTRACT

In order to effectively conduct the defect detection of nuclear fuel pellets end face and avoid the leakage of nuclear radiation, a defect detection system for the nuclear fuel pellets end face based on machine vision is proposed. Firstly, aiming at the complexity of the defect detection of nuclear fuel pellets, a set of image acquisition system lighted by left-right symmetric grating is designed. Then, after fusing the images of left-right structured light those cross points are extracted which classified based on the Gaussian mixture model (GMM). Finally, a series of morphological operations such as dilation operation are conducted for the classified points to obtain the defect area of nuclear fuel pellets end face. The experimental results show that this method reduces the influence of complex characteristics of form, texture, and color of the sample end face on the defect detection and relatively good detection results are gained for various defects with 99.5% accuracy. It takes less than 0.4 s to fully meet the requirements of industrial automation testing.     

Optimization of Fuel Assembly Allocation for Boiling Water Reactors

Abstract

A direct search algorithm is applied to the optimization of fuel assembly allocation of BWR with particular consideration given to the nuclear model and the treatment of operating constraints. A simple expression is derived for evaluating the stuck rod margin, based on regression analysis of data obtained by three-dimensional full core analysis, and the expression is applied to optimization procedure.

The practical applicability of the method is confirmed through trial computations for the second and equilibrium cycles of a medium-sized commercial BWR, with an examination based on various initial guesses and objective functions for radial power peaking.     

NSE abundance data

A novel method of calculating nuclear statistical equilibrium (NSE) is presented. Basic equations are carefully solved using arbitrary precision arithmetic. A special interpolation procedure is then used to retrieve all abundances using tabulated results for neutrons and protons, together with basic nuclear data. Proton and neutron abundance tables, basic nuclear data, and partition functions for nuclides used in the calculations are provided. A simple interpolation algorithm using pre-calculated p and n abundances tabulated as functions of kT, ρ and Y_e is outlined. Unique properties of this method are: (1) ability to pick up out of NSE selected nuclei only, (2) computational time scaling linearly with number of re-calculated abundances, (3) relatively small amount of stored data: only two large tables, (4) slightly faster than solving the NSE equations using traditional Newton-Raphson methods for small networks (few tens of species); superior for huge (800-3000) networks, (5) does not require initial guess; works well on random input, (6) can be tailored to specific application, (7) ability to use third-party NSE solvers to obtain fully compatible tables, and (8) encapsulation of the NSE code for bug-free calculations. A range of applications for this approach is possible: covering tests of traditional NSE Newton-Raphson codes, generating starting values, code-to-code verification, and possible replacement of the old legacy procedures in supernova simulations.     

Analysis of the content of global 90Sr in the water and bottom deposits of reservoirs

A computational scheme, based on the systematic use of two models and permitting detailed (with different degrees of accuracy) estimation of the volume and specific activity of global-fallout ⁹⁰Sr in the water and bottom deposits of reservoirs, is proposed. As an example of the application of this approach, a comparison is made between the computational results and measurement data for reconstruction of the background contamination with global ⁹⁰Sr of Lakes Pes’vo and Udomlya-reservoirs for the Kalinin nuclear power plant. The computational results agree satisfactorily with the experimental data. __________ Translated from Atomnaya énergiya,Vol. 100, No. 6, pp. 471–478, June, 2006.     

Fission time for the U + α reaction measured by the crystal blocking technique

The crystal blocking technique has been used to measure the total time of the induced fission process for the ²³⁵U + α reaction in the energy range of bombarding α-particles from 25.9 to 31.2 MeV. Experimental fission times observed in this reaction vary from 10⁻¹⁷ to 10⁻¹⁶ s, depending on the projectile energy. Together with the corresponding experimental data on angular anisotropy in the same reaction they were analyzed within the dynamic-statistical approach with allowance for the nuclear dissipation phenomenon and the double-humped fission barrier model. It was demonstrated that the time of induced fission at low excitation energies is sensitive to the nuclear dissipation magnitude.     

Establishing and Maintaining Compliant Flask Fleet Operations

Abstract

The transport of irradiated nuclear fuel from the BNFL Magnox Generation nuclear power stations is a necessary requirement in the continued safe generation of electricity. As such the CEGB (and its successor companies) have invested heavily in developing an irradiated fuel transport flask, the M2 flask design, that meets the twin aims of maintaining regulatory compliance and securing the fuel delivery requirements of the generating stations to the end of their economic lives. This paper describes the steps that have been taken in establishing a compliant flask design and how that design has been maintained in terms of compliance, quality assurance, operation and maintenance.     

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.     

Optimization of energy-group structure for LWR high-fidelity neutronics calculation based on the contributon theory

ABSTRACT

Deterministic high-fidelity neutronics calculation is to solve the neutron transport equation using the multi-group (MG) nuclear data libraries. The energy-group structure (ES) in MG nuclear data libraries has a significant impact on the precision and efficiency of neutronics calculation. Therefore, to meet the requirement of high precision and efficiency for high-fidelity neutronics calculation, the contributon theory is adopted to select the optimal ESs for the high-fidelity neutronics code NECP-X which is developed by Nuclear Engineering Computational Physics (NECP) lab. at Xi’an Jiaotong University (XJTU). By combining the contributon theory with the exhaustive searching method, the optimal ESs can be selected effectively. Two optimal ESs named NECP-69 and NECP-47 are obtained and the nuclear data processing code NECP-Atlas developed by NECP lab. is utilized to generate the corresponding MG nuclear data libraries for NECP-X. The neutronics parameters of the MOX pin cell problems and the VERA benchmarks are evaluated. The numerical results show that more precise neutronics parameters are obtained based on the optimal ESs compared with those based on the conventional WIMS-69 and HELIOS-47 for NECP-X.     

An experimental procedure for measuring accelerations and strains from a tie down system of a heavy nuclear transport package during a rail journey

Abstract

The transportation of nuclear waste and new nuclear fuel is an important aspect in sustaining the generation of electricity by nuclear power. The design of packages that satisfy regulatory requirements for normal operating and accident conditions is a complex engineering challenge. The ancillary equipment used to constrain the packages to their conveyance, a tie down system, is part of a multicomponent system used to transport packages. Traditionally, the individual components of the transport system have been designed in isolation. This approach does not account for the interaction between components of the system such as the conveyance, tie down system and package. The current design process for tie down systems is well established but, due to its heuristic development, suffers from uncertainties over which loading conditions should be applied. This paper presents a method for collecting measured acceleration and strain data that can be used to derive customised load cases for the design of tie down systems during rail transportation. The data was collected from a tie down system that restrained an empty TN81 package, weighing 99·7 tonnes during a routine rail journey from Barrow-in-Furness to Sellafield. Furthermore, the data can be used to validate modern computer models, allowing for the development of the previously described holistic approach to tie down system design. The results are unique because an ensemble of acceleration and strain time histories from a transport system laden with a nuclear package is unprecedented. A visual examination indicates that the loading a tie down system incurs during a rail journey consists of low magnitude accelerations. The measurement points also show that the general trend of acceleration levels is highest nearest the track and is attenuated by the package. The implications for the design of tie down systems are that two potential failure modes, fatigue and static strength, have been identified. The data provides scope for customising accurate static strength and fatigue calculations using modern computational techniques. This allows for the safety margins inherent in new designs to be determined and optimised design solutions made possible.

INS makes no representations or warranties or any kind concerning this article, express or implied, statutory or otherwise, including without limitation, warranties of accuracy or the absence of errors.     

Analysis of the Rossi-α Experiment through Pál-Bell's Equation

Pál -Bell's equation for the probability generating function of neutron counts has been analytically solved in the case of three time states, using two-forked approximation. From this solution it is found that all experimental data on neutron fluctuation consist individually of only three basic parameters. The average counting rate C, the decay constant α and the chain register rate C_r are in this instance chosen for the three fundamental measures of correlation. The original observation is presented that C_r can be obtained precisely by determining the waiting time distribution for the triggering of the time analyzer.

The correlated and uncorrelated parts of the Rossi-α data in a thermal system have been analyzed by this three-parameter scheme, and a consistent explanation is given of the results obtained.     

Thinking outside the box: options for transport of high burnup spent fuel

Abstract

The current uncertainty surrounding the licensing and eventual opening of a long term geologic repository for the nation’s civilian and defense spent nuclear fuel and high level radioactive waste has shifted the window for the length of time spent fuel could be stored to periods of time significantly longer than the current licensing period of 40 years for dry storage. An alternative approach may be needed to the licensing of high burnup fuel for storage and transportation based on the assumption that spent fuel cladding may not always remain intact. The approach would permit spent fuel to be retrieved on a canister basis and could lessen the need for repackaging of spent fuel. This approach is being presented as a possible engineering solution to address the uncertainties and lack of data availability for cladding properties for high burnup fuel and extended storage time frames. The proposed approach does not involve relaxing current safety standards for criticality safety, containment, or permissible external dose rates.     

Modeling the thermochemical behavior of AmO2−x

A thermochemical representation of the fluorite structure AmO_2−x phase was developed using the compound energy formalism approach assuming constituents of (Am⁴⁺)₁(O²⁻)₂, (Am⁴⁺)₁(Va)₂, (Am³⁺)₁(O²⁻)₂, and (Am³⁺)₁(Va)₂. The Gibbs free energies for the constituents and a set of interaction parameters were determined using reported oxygen potential-temperature-composition data. A good fit to the experimental information was obtained which well-reproduces the behavior. The representation is also in a format that will allow incorporation of other dissolved metals and thus will be useful in generating multi-component compound energy formalism representations for complex oxide nuclear fuel and waste systems. A full assessment relating the fluorite structure phase to the phase equilibria for Am-O, however, must await adequate data for the remainder of the system.     

A critical survey of experimental cross section data, comparison with nuclear model calculations and estimation of production yields of Br and Kr in no-carrier-added form via various nuclear processes

Experimentally investigated nuclear reactions for production of no-carrier-added ⁷⁷Br and ⁷⁷Kr were critically surveyed. The survey covered nine reactions for the formation of ⁷⁷Br and six reactions for ⁷⁷Kr. Both radionuclides are simultaneously produced in many of the studied nuclear processes. The experimental data were compared with the results of nuclear model calculations based on the computer code ALICE-IPPE and the third version of TALYS-based Evaluated Nuclear Data Library, TENDL-2010. Good agreement was found over extended energy regions for the p-, ³He- and α-particle induced reactions on several target materials. In case of d-induced reactions, however, considerable discrepancies were noted between the experimental and theoretical data. The concordant sets of experimental cross section data for each reaction were fitted by a polynomial function to obtain a trend curve. From the thus obtained trend curves the yields of both ⁷⁷Br and ⁷⁷Kr were calculated. A discussion of suitable production routes is given.     

Theory of Third Order Moments-Related Reactor Noise Experiments with Delayed Neutrons

Abstract

A general theory of third order moments-related reactor noise pulse experiments is presented. Detailed computation of a two-intervals moment of counts for a zero power one velocity point reactor with delayed neutrons is performed. The neutronic process is analyzed up to third order covariances via an extension of the usual mobility and diffusion matrices' approach. For this purpose, a stochastic diffusion matrix and a third order stochastic diffusion tensor are introduced. Detection effects are taken into account through a non-homogeneous Poisson distribution. No a priori approximation for the probability generating function of counts is assumed.     

Validation of Pb nuclear data by Monte Carlo analyses of sample reactivity experiments at Kyoto University Critical Assembly

Sample reactivity experiments on the uncertainty analyses of Pb nuclear data are carried out by substituting Al plates for Pb ones at the Kyoto University Critical Assembly, as part of basic research on Pb–Bi for the coolant. Numerical simulations of sample reactivity experiments are performed with the Monte Carlo calculation code MCNP6.1 together with four nuclear data libraries JENDL-3.3, JENDL-4.0, ENDF/B-VII.0 and JEFF-3.1, to examine the accuracy of cross-section uncertainties of Pb isotopes by comparing measured and calculated sample reactivities. A library update from JENDL-3.3 to JENDL-4.0 is demonstrated by the fact that the difference between Pb isotopes of the two JENDL libraries is dominant in the comparative study, through the experimental analyses of sample reactivity by the MCNP approach. In addition, JENDL-4.0 reveals a slight difference from ENDF/B-VII.0 in all Pb isotopes and ²⁷Al, and from JEFF-3.1 in ²³⁸U and ²⁷Al. Based on these results, further experiments are needed to investigate the uncertainties of Bi isotopes with the use of the Pb–Bi and Bi plates.