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Vasiliev, A.

Publications and source records attributed to Vasiliev, A..

16 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Precision measurement of the neutral pion lifetime

The explicit breaking of the axial symmetry by quantum fluctuations gives rise to the so-called axial anomaly. This phenomenon is solely responsible for the decay of the neutral pion π0 into two photons (γγ), leading to its unusually short lifetime. Here, we precisely measured the decay width Γ of the π 0 → γγ process. The differential cross sections for π0 photoproduction at forward angles were measured on two targets, carbon-12 and silicon-28, yielding Γ(π 0 → γγ)=7.798±0.056(stat.)±0.109(syst.) eV, where stat. denotes the statistical uncertainty and syst. the systematic uncertainty. We combined the results of this and an earlier experiment to generate a weighted average of Γ(π 0 → γγ)=7.802±0.052(stat.)±0.105(syst.) eV. Our final result has a total uncertainty of 1.50% and confirms the prediction based on the chiral anomaly in quantum chromodynamics.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2019

High precision measurement of Compton scattering in the 5 GeV region

The cross section of atomic electron Compton scattering $\gamma + e \rightarrow \gamma^\prime + e^\prime $ was measured in the 4.40--5.475 GeV photon beam energy region by the {\em PrimEx} collaboration at Jefferson Lab with an accuracy of 2\% and less. The results are consistent with theoretical predictions that include next-to-leading order radiative corrections. The measurements provide the first high precision test of this elementary QED process at beam energies greater than 0.1 GeV.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2018

Erratum: Polarization transfer observables in elastic electron-proton scattering at Q 2 =2.5 , 5.2, 6.8, and 8.5GeV 2 [Phys. Rev. C 96 , 055203 (2017)]

Subsequent to the release of our original paper, we discovered in the context of preparing our technical supplement [1] for journal publication that a typographical error had existed in the text file that the analysis program used to construct the beam polarization "database" for both the original analysis, published in Ref. [2], and our final analysis. The electron-beam polarization P e and the analyzing power A y cancel exactly in the ratio R, which is proportional to the ratio P t / P l of the transferred polarization components. On the other hand, the extraction of the relative e dependence of P l / P Born l relies on knowledge of the beam polarization. As such, data taking was interrupted roughly every two days during the GEp - 2γ experiment to perform invasive measurements of the beam polarization using the Hall C Moller polarimeter [3].

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2017

Polarization transfer observables in elastic electron-proton scattering at Q 2 = 2.5 , 5.2, 6.8, and 8.5 GeV 2

In this paper, interest in the behavior of nucleon electromagnetic form factors at large momentum transfers has steadily increased since the discovery, using polarization observables, of the rapid decrease of the ratio G p E /G p M of the proton's electric and magnetic form factors for momentum transfers Q 2 ≳ 1 GeV 2 , in strong disagreement with previous extractions of this ratio using the traditional Rosenbluth separation technique. The GEp-III and GEp-2$\gamma$ experiments were carried out in Jefferson Lab's (JLab's) Hall C from 2007-2008, to extend the knowledge of $G_E^p/G_M^p$ to the highest practically achievable $Q^2$ and to search for effects beyond the Born approximation in polarization transfer observables of elastic $\vec{e}p$ scattering. This article reports an expanded description of the common experimental apparatus and data analysis procedure, and the results of a final reanalysis of the data from both experiments, including the previously unpublished results of the full-acceptance data of the GEp-2$\gamma$ experiment. The Hall C High Momentum Spectrometer detected and measured the polarization of protons recoiling elastically from collisions of JLab's polarized electron beam with a liquid hydrogen target. A large-acceptance electromagnetic calorimeter detected the elastically scattered electrons in coincidence to suppress inelastic backgrounds. The final GEp-III data are largely unchanged relative to the originally published results. The statistical uncertainties of the final GEp-2$\gamma$ data are significantly reduced at $\epsilon = 0.632$ and $0.783$ relative to the original publication. The decrease with $Q^2$ of $G_E^p/G_M^p$ continues to $Q^2 = 8.5$ GeV$^2$, but at a slowing rate relative to the approximately linear decrease observed in earlier Hall A measurements. At $Q^2 = 2.5$ GeV$^2$, the proton form factor ratio $G_E^p/G_M^p$ shows no statistically significant $\epsilon$-dependence, as expected in the Born approximation. The ratio $P_\ell/P_\ell^{Born}$ of the longitudinal polarization transfer component to its Born value shows an enhancement of roughly 1.7\% at $\epsilon = 0.783$ relative to $\epsilon = 0.149$, with $\approx 2.2\sigma$ significance based on the total uncertainty, implying a similar effect in the transverse component $P_t$ that cancels in the ratio $R$.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2017

SFCOMPO-2.0: An OECD NEA database of spent nuclear fuel isotopic assays, reactor design specifications, and operating data

SFCOMPO-2.0 is the new release of the Organisation for Economic Co-operation and Development (OECD) Nuclear Energy Agency (NEA) database of experimental assay measurements. These measurements are isotopic concentrations from destructive radiochemical analyses of spent nuclear fuel (SNF) samples. We supplement the measurements with design information for the fuel assembly and fuel rod from which each sample was taken, as well as with relevant information on operating conditions and characteristics of the host reactors. These data are necessary for modeling and simulation of the isotopic evolution of the fuel during irradiation. SFCOMPO-2.0 has been developed and is maintained by the OECD NEA under the guidance of the Expert Group on Assay Data of Spent Nuclear Fuel (EGADSNF), which is part of the NEA Working Party on Nuclear Criticality Safety (WPNCS). Significant efforts aimed at establishing a thorough, reliable, publicly available resource for code validation and safety applications have led to the capture and standardization of experimental data from 750 SNF samples from more than 40 reactors. These efforts have resulted in the creation of the SFCOMPO-2.0 database, which is publicly available from the NEA Data Bank. Our paper describes the new database, and applications of SFCOMPO-2.0 for computer code validation, integral nuclear data benchmarking, and uncertainty analysis in nuclear waste package analysis are briefly illustrated.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Complik: A versatile tool for automatizing the representation of material data in MCNP models - 14412

At the Paul Scherrer Institut, the CASMO/SIMULATE code system is used as reference scheme for core analyses of the Swiss reactors. Besides, Monte Carlo (MC) based neutron transport codes are in parallel being established both for out-of-core applications and full core analyses, in order to provide higher fidelity solutions to problems beyond the modelling capabilities of the standard core modelling methods. In this perspective, a necessity is to be able to conduct the MC calculations at any operating conditions, requiring thus proper initialisation of the thermal-hydraulic feedback as well as at any burnup, prompting the need for a detailed modelling of all material zones in the irradiated fuel assemblies. To tackle the later challenge, the development of the COMPLINK tool, aimed at initialising large-scale MC models of burned core configurations, using as basis upstream 2-D assembly/ 3-D core analyses results, was initiated and is presented in this paper. A main feature is the capability to extend in an automatic manner, a user given MC model in order to represent material data with a higher resolution, keeping track of the location of such materials also within nested repeated structures. This capability prevents the user from the burden of developing specific computational framework for a target application. Since the same functionalities are required to analyse out-of-core spent fuel managements systems, a first verification of COMPLINK is illustrated on the basis of a criticality safety evaluation (CSE) with MCNP6.1 of a Swiss PWR wet storage pool loaded with fresh and burned fuel assemblies. Starting from CASMO-5 depletion calculations at each axial layer of the depleted fuel assemblies and reconstructed based on realistic irradiation histories, COMPLINK is then applied to load the isotopic compositions into the material zones of the corresponding MCNP6.1 assembly models. On this basis, the CSE analyses are conducted with focus on investigating axial and radial modelling effects. All in all, the obtained results are found to be well in line with the impact on CSE expected at the considered burnup levels, providing thus confidence in the correct implementation of the COMPLINK functionalities and confirming its effectiveness as well as efficiency for the initialization of all material zones in large-scale MC models of complex and strongly heterogeneous burned configurations. (authors)

42 ENGINEERING↗
DOE OSTI2015

Towards application of a neutron cross-section uncertainty propagation capability in the criticality safety methodology - 14434

At the Paul Scherrer Institute (PSI) the methodology for criticality safety evaluations (CSE) used for Swiss applications is focused, following common practice, on establishing an upper subcriticality limit (USL) for the effective neutron multiplication factor (k{sub eff}), based on a statistical evaluation of the obtained validation results in terms of the calculated-to-benchmark k{sub eff} values. Recently, a new computational tool referred to as NUSS (Nuclear data Uncertainty Stochastic Sampling) was developed in order to propagate neutron-nuclear data (ND) uncertainties in calculations with Monte Carlo transport codes in combination with general-purpose ND libraries in ACE format. Thereby, this tool should now allow for a quantification of ND data uncertainties in CSE and to introduce these into the CSE criteria definition. This paper concerns an application of the NUSS methodology for criticality validation studies. Calculated k{sub eff} uncertainty estimations were obtained for the current PSI validation suite including in total 149 critical benchmark configurations for which in-house MCNP(X) models were developed, covering both uranium and mixed uranium-plutonium oxide fuels. The MCNPX-2.7.0 code was used together with the ENDF/B-VII.1 ND library while the ND variance-covariance matrixes (VCM) were taken from both SCALE-6.0 44-group libraries and also from ENDF/B-VII.1 files. Based on the newly obtained information on the uncertainties, a re-assessment of the lower tolerance bound (LTB) for k{sub eff} using the total uncertainties as weighting factors is proposed as means for the present PSI CSE methodology upgrade and some possible directions of further methodology enhancements are discussed next. (authors)

42 ENGINEERING↗
DOE OSTI2015

Computational scheme for burnup credit applied to long term waste disposal - 14448

The application of burnup credit (BUC) to criticality safety evaluations (CSE) of geological repositories for long term disposal of spent nuclear fuel (SNF) is under consideration in Switzerland. Therefore, in collaboration with the national waste management organization, PSI is conducting research towards the development of a BUC methodology tailored for Swiss specific applications. As part of this research, a computational sequence from the in-core burnup phase, through the interim storage, until final disposal is being established. The tools selected for this methodology, referred to as BUCSS-R (Burnup Credit System for Swiss reactors and Repositories), are CASMO for the in-core fuel depletion, the decay module of SERPENT for the decay calculations, and MCNP for the criticality calculations. In connection to BUCSS-R, a computational interface to validated reference core models of the Swiss reactors is being developed. The purpose is to apply reconstructed depletion histories based on real core operation conditions with CASMO assembly calculations, to transfer the pin wise nuclide compositions at discharge to the subsequent SERPENT decay calculations, and afterwards to the final MCNP criticality calculations. An objective of this paper is to present the elements of the BUCSS-R methodology, and to show a first application to so-called bounding analyses of a preliminary Swiss disposal canister design loaded with fuel assembly types operated in a Swiss PWR reactor. The rationale for these bounding analyses is that before starting the integration of methodologies for bias and uncertainty quantifications, the span of the k-eff variation between fresh and burnt fuel loadings should first be quantified in a realistic manner. The obtained results are reported in this paper together with studies performed in order to evaluate the effects from burnup levels and nuclides credit at flooded conditions on the k-eff values. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2014

High precision photon flux determination for photon tagging experiments

The Jefferson Laboratory PrimEx Collaboration has developed and implemented a method to control the tagged photon flux in photoproduction experiments at the 1% level over the photon energy range from 4.9 to 5.5 GeV. This method has been successfully implemented in a high precision measurement of the neutral pion lifetime. Here, we outline the experimental equipment and the analysis techniques used to accomplish this. These include the use of a total absorption counter for absolute flux calibration, a pair spectrometer for online relative flux monitoring, and a new method for post-bremsstrahlung electron counting.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2013

Comparison of SERPENT and CASMO-5M for pressurized water reactors models

The objective of this work is to perform a preliminary assessment of the capability of SERPENT to generate cross sections for a PWR Beginning-of-Life (BOL) isothermal mini-core by comparing a SERPENT/PARCS solution with the results obtained using a CASMO-5M/PARCS approach. The PARCS code is used instead of the usual SIMULATE-3 to analyze the Swiss Reactors, because interfaces with PARCS already exist to obtain neutronic data from SERPENT. For the PWR configurations, the differences between CASMO-5M and SERPENT solutions are within 200 pcm at the assembly level and thus rather small when considering the deterministic transport method (energy/angular/space discretization) in CASMO-5M versus the stochastic treatment of SERPENT, the statistical uncertainties in the Monte-Carlo approach as well as the eventual differences in nuclear data used by both codes. At the 2D mini-core level, no major difference is observed when comparing PARCS run with CASMO-5M versus SERPENT cross sections. For the generation of kinetic parameters, non trivial differences are observed due both to the methods and the data used. For the relatively limited number of configurations considered, it is hard to make any definitive statement on the benefits of using Monte Carlo codes in terms of nuclear data generation. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Stochastic sampling method with MCNPX for nuclear data uncertainty propagation in criticality safety applications

In the domain of criticality safety, the efficient propagation of uncertainty in nuclear data to uncertainty in k{sub eff} is an important area of current research. In this paper, a method based on stochastic sampling is presented for uncertainty propagation in MCNPX calculations. To that aim, the nuclear data (i.e. cross sections) are assumed to have a multivariate normal distribution and simple random sampling is performed following this presumed probability distribution. A verification of the developed stochastic sampling procedure with MCNPX is then conducted using the {sup 239}Pu Jezebel experiment as well as the PB-2 BWR and TMI-1 PWR pin cell models from the Uncertainty Analysis in Modeling (UAM) exercises. For the Jezebel case, it is found that the developed stochastic sampling approach predicts similar k{sub eff} uncertainties compared to conventional sensitivity and uncertainty methods. For the UAM models, slightly lower uncertainties are obtained when comparing to existing preliminary results. Further details of these verification studies are discussed and directions for future work are outlined. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Towards a reference numerical scheme using MCNPX for PWR control rod tip fluence estimations

Recent occurrences of cracks and fissures on the cladding tubes of PWR control rod (CR) fingers employed in the Swiss reactors prompted the need to develop more reliable analytical methods for CR tip fluence estimations. To partly address this need, a deterministic methodology based on SIMULATE-3/CASMO-4 was in recent years developed at PSI. Although this methodology has already been applied for independent support to licensing issues related to CR lifetime, two main questions are currently being the center of attention for further enhancements. First, the methodology relies on several assumptions that have so far not been verified. Secondly, an assessment of the achieved accuracy has not been addressed. In an attempt to answer both these open questions, it was considered appropriate to develop an alternative computational scheme based on the stochastic MCNPX code with the objective to provide reference numerical solutions. This paper presents the first steps undertaken in that direction. To start, a methodology for a volumetric neutron source transfer to full core MCNPX models with detailed CR as well as axial reflector representations is established. On this basis, the assumptions of the deterministic methodology are studied for selected CR configurations for two Beginning-of-Life cores by comparing the spatial neutron flux distributions obtained with the two approaches for the entire spectrum. Finally, for the high-energy range (E> 1 MeV) and for a few CRs, the new MCNPX scheme is applied to estimate the accumulated fluence over one real operated cycle and the results are compared with the deterministic approach. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Reactor physics studies for assessment of tramp uranium methods

This paper presents calculation studies towards validation of a methodology for estimations of the tramp uranium mass from water chemistry measurements. Particular emphasis is given to verify, from a reactor physics point of view, the justification basis for the so-called 'Pu-based model' versus the 'U-based model' as a key assumption for the methodology. The computational studies are carried out for a typical BWR fuel assembly with CASMO-5M and MCNPX. By approximating the evolution of fissile nuclides and the fraction of {sup 235}U fissions to total fissions in different zones of a fuel rod, including tramp uranium on the clad surface, it is found that Pu gives the dominant contribution to fissions for tramp uranium after an irradiation on the outer clad surface of at least one cycle in a BWR. Thus, the use of the so-called Pu model for the determination of the tramp uranium mass (this means in particular using the yields for {sup 239}Pu fission) appears justified in the cases considered. On that basis, replacing the older U model by a Pu model is recommended. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Nuclear data uncertainty propagation in a lattice physics code using stochastic sampling

A methodology is presented for 'black box' nuclear data uncertainty propagation in a lattice physics code using stochastic sampling. The methodology has 4 components: i) processing nuclear data variance/covariance matrices including converting the native group structure to a group structure 'compatible' with the lattice physics code, ii) generating (relative) random samples of nuclear data, iii) perturbing the lattice physics code nuclear data according to the random samples, and iv) analyzing the distribution of outputs to estimate the uncertainty. The scheme is described as implemented at PSI, in a modified version of the lattice physics code CASMO-5M, including all relevant practical details. Uncertainty results are presented for a BWR pin-cell at hot zero power conditions and a PWR assembly at hot full power conditions with depletion. Results are presented for uncertainties in eigenvalue, 1-group microscopic cross sections, 2-group macroscopic cross sections, and isotopics. Interesting behavior is observed with burnup, including a minimum uncertainty due to the presence of fertile U-238 and a global effect described as 'synergy', observed when comparing the uncertainty resulting from simultaneous and one-at-a-time variations of nuclear data. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗